CONF-821215-DE83 017117
Proceedings Of The
Fourth DOE Environmental Protectjon Information Meeting Held At
Denver, Colorado December 7-9,1982 Published: August 1983
DISCLAIMER This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.
NOTICE PORTIONS OF THIS REPORT ARE IlLEfllBU.' It has been reproduced from the best available copy to permit the broadest possible availability.
Prepared for:
U.S. Department of Energy Assistant Secretary, Environmental Protection, Safety, and Emergency Preparedness Washington, D.C. 20545 OF mis DOCUMENT IS
PROCEEDINGS OF THE FOURTH DOE ENVIRONMENTAL PROTECTION INFORMATION MEETING
Held at Denver, Colorado December 7-9,1982
ISSUE DATE: June, 1983
General Chairman
D.E. Patterson
Program Chairman
C.G. Welty, Jr.
Meeting Coordinator
J.P. Corley
Publications
E M Toomey Program Committee
L.J. Deal
J.F. Wing
O.R. Elle
S.R. Wright
G J . Werkema
Sponsored By The OFFICE OF OPERATIONAL SAFETY U.S. DEPARTMENT OF ENERGY WASHINGTON. D.C. 20645
May 13, 1983 FOREWORD The Department of Energy (DOE) conducts both nuclear and non-nuclear programs involving both research and production activities. These programs are conducted at more than 30 sites with a variety of missions and located throughout the U.S. Consistent with national environmental policy, the Department of Energy conducts its operations not only in conformance with applicable laws and regulations but with a due regard for protection of the environment and of public health. To accomplish these objectives, the Department conducts a comprehensive environmental program to establish policy and requirements, provide technical support, and maintain program overview for all of its facilities and contractors. The Department of Energy and its predecessor organizations, the Atomic Energy Commission and the Energy Research and Development Administration, have sponsored three previous meetings. In addition, there have been less formal interchanges of information among DOE and contractor staff on environmental protection matters. The stated purpose of this 4th DOE Environmental Protection Information Meeting was to enhance the Department's efforts to assure environmental compliance and public protection by promoting the continued exchange of technical information and operational experience. Thanks to the quality and content of the presentations and ensuing discussions, this purpose was realized. Notable in these Proceedings are the numbers of papers concerned with the quality and reliability of environmental data, the management of hazardous wastes, improvements in pollution control, and groundwater monitoring and protection. Expressed concerns primarily centered around difficulties anticipated in implementing DOE's hazardous waste requirements, the need for continued assurance of the quality of environmental data, and the availability of funding for pollution control projects. The Office of Operational Safety expects that the positive environmental protection efforts discussed in these Proceedings will be increased in those areas where environmental concerns persist. These Proceedings contain the papers presented through the three days of meetings. The Office of Operational Safety acknowedges with thanks the participation and contributions of the members of the program committee, the session chairmen, the members of the hazardous waste management discussion panel, and all authors of papers. Their efforts made this a most successful meeting.
c
t ? D. E. Patterson, Director Office of Operational Safety Meeting General Chairman
TABLE OF CONTENTS FOREWORD
v
SESSION ONE: OPENING 1A OPENING ADDRESS R. W. Davies IB OVERVIEW OF CURRENT AND EXPECTED CHANGES IN LEGISLATION AND REGULATIONS W. J. Dennison
1
5
1C OVERVIEW OF CURRENT PROGRAMS AND TRENDS IN ASSESSMENT OF HEALTH EFFECTS Dr. J. W. Thiessen
13
ID OFFICE OF OPERATIONAL SAFETY, ENVIRONMENTAL PROGRAM MISSIONS AND OBJECTIVES C. J. Welty, Jr.
25
IE PNL'S ENVIRONMENTAL PROTECTION SUPPORT PROJECT FOR THE OFFICE OF OPERATIONAL SAFETY J. P. Corley and R. E. Jaquish
31
SESSION TWO: HAZARDOUS WASTE MANAGEMENT 2A THE ARGONNE HIGH SULPHER DRY SCRUBBER P. S. Farber 2B TRAPPING OF GASEOUS FLUORIDE EMISSICMS ON SOLID SORBENTS N. F, Windt
45 .
2C FERMILAB'S APPROACH TO PCBS S. I. Baker 2D RAFFINATE TREATMENT AT THE PORTSMOUTH GASEOUS DIFFUSION PLANT T. A. Acox 2E ASSESSMENT OF DOE INACTIVE CHEMICAL WASTE SITES: ARE ADEQUATE INVENTORY DATA AVAILABLE? C. J. English and D. H. Denham PANEL DISCUSSION V. DeCarlo, J. E. Bresson, N. J. Stas, J. F. Wing, and S. R. Wright
vii
.
55 69
.
77
.
87
95
SESSION THREE: RADIOACTIVE WASTE MANAGEMENT AND QUALITY ASSESSMENT 3A QUALITY ASSURANCE FOR ENVIRONMENTAL ANALYTICAL CHEMISTRY AT LOS ALAMOS E. S. Gladney, D. R. Perrin, and W. E. Goode
107
)
3B QUALITY ASSURANCE IN ENVIRONMENTAL MEASUREMENTS T. W. Oakes
119.
. . . .
3C QUALITY CONTROL ACTIVITIES OF THE HANFORD ENVIRONMENTAL SURVEILLANCE PROGRAM K. R. Price and R. E. Jaquish
135
3D DESTRUCTION OF URANIUM-CONTAMINATED WASTE OIL L. F. Hary
141
. . . .
3E THE WATER MONITORING PROGRAM AT THE ROCKY FLATS PLANT R. L. Henry
. . .
153
3F DETAILED SITE STUDY PHASE: AN ENVIRONMENTAL APPROACH TO PLANNING AND DESIGN M. L. Brown SESSION FOUR: 4A
163
REGULATORY IMPACTS
ENVIRONMENTAL PROTECTION ANALYSIS AND PLANNING FOR PROPOSED SAVANNAH RIVER PLANT ACTIONS T. V. Crawford and J. C. Tseng
173
4B NUCLEAR WASTE DISPOSAL AND NEPA COMPLIANCE B. H. Smith and M. L. Brown 4C ROCKY FLATS PLANT INTERACTION WITH THE COLORADO DEPARTMENT OF HEALTH D. D. Hornbacher
181
189
40 THE INFLUENCE OF PCB AND HAZARDOUS WASTE REGULATIONS ON THE OPERATIONS OF THE WESTERN AREA POWER ADMINISTRATION . . . K. E. Mathias and J. J. Kirby
195
4E COAL CONVERSION: THE CH EXPERIENCE V. L. Alspaugh 4F ENVIRONMENTAL REGULATORY ASPECTS OF ALTERNATE LIQUID FUEL USE AT BROOKHAVEN NATIONAL LABORATORY M. J. Bebon, J. R. Naidu, and L. C. Emma
201
211
SESSION FIVE: EFFLUENT MONITORING 5A NUCLEAR AIR CLEANING: E. H. Carbaugh
THE NEED FOR A CHANGE IN EMPHASIS
viii
.
.
227
5B OPERATIONAL EXPERIENCE WITH TWO TRITIUM EFFLUENT MONITORING SYSTEMS J. S. Haynie and J. A. Gutierrez 5C ON-LINE LIQUID EFFLUENT MONITORING OF SEWAGE AT LAWRENCE LIVERMORE NATIONAL LABORATORY . . . M. Dreicer, et al. 5D GENERIC PARTICULATE MONITORING SYSTEM FOR RETROFIT TO HANFORD EXHAUST STACKS J. W. Cammann and E. H. Carbaugh SESSION SIX: ENVIRONMENTAL MONITORING I
237
.
.
.
247
257
6A SOME PROBLEMS IN EVALUATING RADIATION DOSE FOR EFFECTIVE ENVIRONMENTAL ASSESSMENT E. C. Watson, et al.
279
6B STANDARDS FOR CONCENTRATIONS OF RADIONUCLIDES IN SOIL: BASIS AND IMPLEMENTATION R. C. Brown and G. S. Kephart
291
6C SCREENING LEVELS FOR RADIONUCLIDES IN SOIL: APPLICATION TO DECONTAMINATION AND DECOMMISSIONING (D&D) CRITERIA . . . S. K. Rope and S. R. Adams
301
6D THE TEXAS PANHANDLE SOIL-CROP-BEEF FOOD CHAIN FOR URANIUM: A DYNAMIC MODEL VALIDATED BY EXPERIMENTAL DATA . . . . W. J. Wenzel, et al.
311
6E RISK ASSESSMENT AS A MEANS OF EVALUATING ENVIRONMENTAL CONTROLS S. A. Dover 6F FEASIBILITY OF AN EIS FOLLOW-UP PROGRAM I. C. Nelson, R. E. Jaquish, and D. G. Watson SESSION SEVEN: ENVIRONMENTAL MONITORING II 7A A METHODOLOGY FOR MAKING ENVIRONMENTAL AS LOW AS REASONABLY ACHIEVABLE (ALARA) DETERMINATIONS R. C. Brown and D. R. Speer 7B ALARA BEYOND DOLLARS PER PERSON-REM D. Waite and W. Harper 7C IMPROVEMENTS TO ENVIRONMENTAL SURVEILLANCE AT THE RADIOACTIVE WASTE MANAGEMENT COMPLEX (RWMC) OF THE IDAHO NATIONAL ENGINEERING LABORATORY K. S. Moor, et al.
IX
.
323 333
345 367
387
Id INTERNAL REVIEW SYSTEM FOR ENVIRONMENTAL PROTECTION PROGRAMS OF THE DEPARTMENT OF ENERGY FACILITIES OPERATED BY UNION CARBIDE CORPORATION NUCLEAR DIVISION . . . . H. H. Abee and M. E. Mitchell
399
7E SITE MONITORING FROM SOIL SAMPLE ANALYSIS . . . . C. T. Illsley 7F SUMMARY AND RESULTS OF THE COMPREHENSIVE ENVIRONMENTAL MONITORING PROGRAM AT THE INEL'S RAFT RIVER GEOTHERMAL SITE . R. A. Mayes, T. L. Thurow, and L. S. Cahn 7G ENVIRONMENTAL PROGRAM AT THE MHD COAL FIRED FLOW FACILITY T. P. Lynch, et al.
.
.
409
.
419
.
431
7H MEASUREMENTS AND MODELING OF GAMMA ABSORBED DOSES DUE TO RELEASES FROM A LINEAR PROTON ACCELERATOR: EXPERIMENTAL DESIGN AND PRELIMINARY RESULTS B. M. Bowen, et al. SESSION EIGHT: GROUND-WATER MONITORING AND ASSESSMENTS 8A MONITORING FOR ORGANIC CONTAMINANTS IN GROUND WATER . . . D. A. Myers and J. M. Msuser 8B GROUNDWATER MONITORING PROGRAM AT THE ROCKY FLATS PLANT . N. D. Hoffman
447
459 .
471
8C OAK RIDGE Y-12 PLANT, GROUNDWATER MONITORING AND ASSESSMENT PROGRAM P. M. Pritz
481
8D GROUND-WATER MONITORING PROGRAMS AT THE HANFORD SITE, WASHINGTON STATE J. S. Wilbur, L. S. Prater, and P. A. Eddy
489
SESSION NINE: EMERGENCY ASSESSMENTS 9A ACCIDENT ANALYSIS AND DOE CRITERIA J. M. Graf and J. C. Elder 9B EMERGENCY PREPAREDNESS FOR NONRADIOLOGICAL INCIDENTS AT HANFORD R. D. Gilmore, B. J. McMurray, and K. R. Heid
507 /
519
9C DOSE PROJECTION CONSIDERATIONS FOR EMERGENCY CONDITIONS AT NUCLEAR POWER PLANTS G. A. Stoetzel, et al.
527
9D IRDAM-INTERACTIVE RAPID DOSE ASSESSMENT MODEL R. W. Poeton, et al.
541
. . . .
/
9E EMERGENCY EFFLUENT MONITORING AND ASSESSMENT C. D. Corbit 9F POST ACCIDENT RADIATION MONITORS G. J. Laugh!in and R. L, Kathren 9G CAPABILITIES OF THE LOS ALAMOS NATIONAL LABORATORY'S ENVIRONMENTAL EMERGENCY RESPONSE VEHICLE D. Van Etten, et al. ALPHABETICAL INDEX BY AUTHOR
555 563
573 Index.1
SESSION ONE OPENING
1A
OPENING ADDRESS R. W. Davies Deputy Assistant Secretary, Environment, Safety and Health U.S. Department of Energy Washington, D.C.
Let me speak about the Office of Environment, Safety and Health (ES&H). When I arrived in Washington, I came with 10 to 15 years experience as a licensee under the Atomic Energy Commission (later the Nuclear Regulatory Commission) and my perspective was based on that experience. When I was told that the DOE had an internal self-regulatory system, I said "Well, where is the counterpart of the Nuclear Regulatory Commission or of OSHA or of EPA?" I looked but I couldn't find it. It was surprising that nobody on my immediate staff could explain it to me, at least not in simple terms that I could understand. Later on, as I got into discussions about the process, I began to see how it worked, and began to realize that there is no direct counterpart to the NRC in the DOE. The ES&H safety offices run a different kind of shop than the NRC. There is no strong centralized agency, like the NRC is with its relationship with the private sector. The office that I represent sets standards like the NRC, but people use these standards more as if they were criteria or suggested guidelines, rather than as mandated standards as with the NRC. I often wonder about that, because the NRC system requires one to follow their standards exactly, In DOE, we establish standards and say, "Here's a set of criteria you ought to look at and think about. We'll be back later to see if you are doing an adequate job." It would be very difficult to apply specific criteria to all parts of DOE because its operation is so diverse. As an example, consider toxic chemicals, if there were 100 of them known to man, then the DOE would have 110. DOE is a very widely ranging operation; it is fascinating. My office also provides technical support in the areas of environment, safety, &nd health. If we can't provide the support ourselves, we have contractors who can. By setting our standards and criteria and by providing technical assistance, we approach safety and environmental protection in a cooperative and supportive manner rather than by mandating requirements as from an outside agency. If training is needed, we sponsor training programs, again to help develop and build a safety and environmental program. The one area that looks somewhat like the NRC is the area where we conduct appraisals. Our system of self-regulation is a tiered system. The individual on the very bottom, the working person in the contractor's organization, is looked at by the contractors' safety or QA organization, and Federal people in the Field Office look at the contractor. The program offices from Headquarters look at the Field Office. Later on, my office looks at both the program and Field Offices to see that they are doing their job. This system has been developed over the yearb and it is a good system; the proof of the pudding is in the eating. The safety record for DOE is excellent. In every category, the DOE record is considerably better than that of the comparable industry record in the same field. 1
You might say, well, that's what it's all about, but there is another thing. We should be proud of our good record, but we must not get complacent. The Three Mile Island plant had a very good operating record as a commercial facility before its accident. I have often said that a good operating record is a record of the past, it is not a predictor of the future. So we must not get complacent with our good record. We've got to always keep on our toes. Also, with the cooperative arrangement we have, I do sense a lack of discipline in parts of the safety program. This concerns me because of the potential that others may decide that we should have a more tightly controlled program dominated by a strong central organization. If we only issue guidelines or criteria, there is a chance that somebody in DOE might say, "Well, I don't want to do that" or "I don't have the manpower." Here we can apply our professional judgment to a problem. When you work with the NRC system , you can't apply judgment. You must perform according to a set of rules. The theory is that when one is permitted to use their judgment there is always the possibility of a slip up. However, I am confident that no one bypasses safety regulations, something involving safety of employees, or the public. On the other hand, there appears to be more discretion exercised in the environmental protection area. I believe the budget process has an effect. A decision may be made to clean up a particular waste stream, but when it comes down to doing the work on that waste stream and there is a tight budget, you might be forced to say "Hey, we can do that next year." I am aware of a number of instances where certain environmental protection projects are being delayed because of the budget process. I often think my office should get a little hard-nosed about these kind of things. We should call attention to decisions which might not be best in terms of environmental protection. This is something we need to work out. A couple of things of interest are taking place right now relative to our ES&H program. As you know after the Three Mile Island accident, the Crawford Committee looked at DOE's nuclear safety operation. One of the points made by the Crawford Committee and the resulting Action Plan was the recommendation that we carefully review our manpower resources, considering both the technical talent and numbers of people in the ES&H operation. The Office of Management and Administration in Washington has recently completed a DOE-wide manpower study on environmental, safety and health personnel and identified a shortage. I was encouraged by the fact that they determined that the requirements of the program were measurable and identifiable. Secretary Hodel has accepted the M&A report and its recommendations for manpower and he is in the process right now of passing that word back to OMB. As an example, in my office we were budgeted for FY 1983 at 95 FTE's. The OMB pass-back for FY 1984 dropped us to 90 because they were told to drop everybody 5 percent. The M&A study concluded that we should really be at 117 instead of 95. He is also recommending that our 1983 manpower be increased by 19 more people than it is today. This personnel problem affects you in the field as well as it does us in Headquarters. Secretary Hodel is supporting this activity and that is very encouraging. The last point to mention is recent activity of the General Accounting Office related to ES&H. A little over a year ago they reached the conclusions that better oversight was needed of DOE's nuclear safety program.
We replied to that study and in effect said that, "Hey, GAO you don't understand how we operate our safety program." They had recommended a much stronger centralized system. They did not recognize that DOE has developed a good system and that it does work. GAO then came out with a supplement to their original report and said, "We didn't like your reply, you didn't really answer our questions, we still think you need better oversight." So it came down to a stand off, but the official position within the Department of Energy was that additional discussions with GAO were not warranted. We had a difference in philosophy; the point of view was so totally different, we couldn't communicate any more. GAO is presently back in our offices for a re-audit. They are going to try to determine if we have made improvements since the Three Mile Island accident, and if we have complied with the Crawford Committee recommendations. Are we doing our job better today than we were when GAO was in here before? If they reach the conclusion that we are, then they may let it drop. If they don't like what they see, they're going to probe some more. I mention this because the DOE system of self-regulation is being questioned. Some members of Congress are pushing this investigation and we are concerned about what might happen from this particular thrust. I believe we must do something if we're going to successfully ward off such threats to the way we operate. When I say do something, I mean that we must elevate the consciousness of the Department of Energy that we really know what we are doing, that we intend to do it and do it well. The process starts at Headquarters and it flows down through the whole field organization. We all need to do our job and do it professionally, and we will then be permitted to remain as our own self-regulator.
IB
OVERVIEW OF CURRENT AND EXPECTED CHANGES IN LEGISLATION AND REGULATIONS VI. J. Dennison Special Assistant to the Deputy General Counsel for Programs U.S. Department of Energy Washington, D.C. 20585
I would like to focus upon two or three issues which Carl Welty and I agreed might be of some general interest to this group: an old issue in a sense, and a new issue. The old issue is the question of the applicability of EPA's hazardous waste regulations to DOE's Atomic Energy Act facilities. After discussing that, I would then like to talk about the prospective development by EPA of regulations for radionuclide emissions from DOE facilities. Turning first to the hazardous waste matter, as most of you already know, EPA has prescribed regulations for the management of hazardous waste under the Resource Conservation and Recovery Act, or RCRA for short. The last time this group met (in Germantown in November 1980) we were at a stage where these EPA regulations were just beginning to be phased in. A question arose at that time, posed most notably by the Office of Defense Programs, as to whether these regulations were applicable to DOE facilities. There was concern among some program managers that the EPA regulations might in some instances be overly stringent, by requiring protective measures which would not necessarily enhance the protection of the environment, but which might interfere significantly w'th the accomplishment of important DOE program objectives. The Office of Defense Programs asked the DOE General Counsel's office to analyze this issue, and that analysis revealed the presence of some interesting statutory language in RCRA which, to the best of my knowledge, is not duplicated in any other environmental statute. The language says (I am paraphrasing) that RCRA does not apply to activities regulated under the Atomic Energy Act, to the extent that the applicability of RCRA woulo be inconsistent with the requirements of the Atomic Energy Act. The practical significance of that language was not immediately clear, nor did legislative history shed any light upon what the Congress may have intended those words to mean. We did conclude, however, in light of the fact that the Atomic Energy Act gives DOE the authority to regulate its own AEA activities (including any necessary environmental protection measures), that at the very least the application by the EPA of RCRA regulations would be duplicative of DOE's authority under the AEA. Moreover, in light of the fact that DOE is authorized and required under the AEA to protect the national security aspects of those AEA activities, we concluded that it would not be unreasonable for DOE to assert the position that the application of the EPA regulations would be "inconsistent" with the AEA. In a letter dated November 14, 1980, DOE notified EPA that DOE did not consider its AEA facilities subject to regulation under RCRA. The letter arrived at a time of considerable confusion at EPA; they were faced with
phasing in an enormously complex set of regulations and, to complicate matters further, the government was in the middle of a transition from the Carter Administration to the Reagan administration. For these reasons, perhaps not surprisingly,DOE did not receive an immediate response to its letter. DOE's Acting Undersecretary nonetheless informed the DOE Field Offices of DOE's position, and directed them not to submit to the EPA regulations except where necessary to keep DOE programs running, and even then to make clear that DOE's actions were being undertaken as a matter of comity rather than regulatory compliance. The emphasis was thus on flexibility. Indeed, our concern throughout the history of this matter has been that our programs remain free to operate as they should, without undue and unproductive constraints. As the months passed, DOE continued to wait for an EPA response to the November 1980 letter. The issue gradually became more pressing as EPA began to delegate enforcement responsibility to the states. Accordingly, DOE has taken a number of occasions to remind E°A of the urgent need to resolve this issue. Among other initiatives, we agreed to prepare a draft Memorandum of Understanding (MOLI) between DOE and EPA that when signed would ratify the DOE position. Such a draft MOU was sent to EPA on September 2 of this year. It simply said that, because of the inconsistency between RCRA and the Atomic Energy Act, DOE's Atomic Energy Act facilities were not subject to EPA's RCRA regulations. The MOU indicated that DOE facilities, to the extent possible, would cooperate with the states and with EPA in a spirit of comity in order to demonstrate to them that we were handling our hazardous waste responsibilities sensibly. The MOU also indicated that DOE would prescribe a formal DOE Order which would regulate hazardous waste at the AEA facilities. And finally, the MOU for the first time provided EPA with a comprehensive list of the specific facilities that DOE was claiming not to be subject to RCRA. Although that MOU was sent to EPA on September 2, we still haven't had a response from EPA. Before concluding my remarks on this issue, I would like to expand upon the idea that Bob Davies mentioned in his remarks, the idea that many issues which are ostensibly technical or legal issues eventually come down to a political or a policy decision. Nowhere is Bob's point better taken than with regard to this RCRA issue. I can't stand before you today and say that, as a matter of law, it is clear that DOE facilities are not subject to these EPA regulations. What I can do is to assure you that there exists a plausible legal basis upon which policy makers can base a decision that these regulations should not be applicable to DOE facilities. As far as DOE is concerned, we can certainly try very strongly to make it clear that we think our position ought to be adopted by EPA, but in the final analysis, they or OMB are going tohave to make that policy decision. It's apparently a difficult decision for them; they're having trouble grappling with it. All we can do is to continue to press our point as strongly as we can. Finally, I think it appropriate at this point for me to offer some advice to any of you that are dealing with EPA regional offices or states that want to come in and perform monitoring enforcement-type audits of your hazardous waste operations. My advice today is pretty much what it was two years ago. In the first place, do not under any circumstances admit EPA jurisdiction. We don't want to suggest, either by our actions or by our words, that
we consider EPA's regulations applicable to us. We have to make it consistently clear to the states and to the EPA regional offices that our position is to the contrary. On the other hand, I think it very important for us not to antagonize needlessly state officials or EPA regional officials. We should work with them to the extent that working with them in a spirit of comity would not compromise the continuance of our programs. Some of them are quite willing to accept our cooperation on those terms; others are less willing. There are no hard and fast rules in this area; we have to look at each situation separately. Our paramount objective, as always, is to keep our programs operating, but we should also strive to keep tempers as cool as possible under the circumstances without compromising the basic DOE position. Sooner or later we will come to a resolution of this matter, but it may yet take some time. Until then, we have no choice but to continue to act in the way that wa have been acting for the last two years. One last bit of advice that I might give you would be to urge that you use the talents of the DOE field office attorneys who are working on this problem. I can't mention them all, but certainly Bob Carosino at Richland, Karen Hoewing in Idaho, Jim Randall at Albuquerque, Vicki Alspaugh at Chicago, and Joan Shands and Mike Bebon at Brookhaven have proven themselves most resourceful in working out these problems as they arose. The are all sensitive to the need to be tactful, the need to be flexible in order to protect their facilities' interests without compressing DOE's overall position. When problem? arise, when you hear from the states or from EPA, I urge you to talk to your field office counsel, who in turn talk to me. Of course, you may also call me if you wish. That concludes my remarks on the RCRA/AEA issue. I am sorry that so much of what I have said has been merely repetitive of my remarks two years ago. In another two years, when this group meets again, I hope that I will be able to talk about something different. I'd like now to turn to a more recent issue, which is the prospective development by EPA of radionuclide emission standards under Section 112 of the Clean Air Act. Over the course of the past few months, various DOE managers at Headquarters have become increasingly aware that EPA is going to be developing radionuclide emission standards for DOE facilities. EPA was required to do this under the 1977 amendments to the CLean Air Act. Those amendments set out a regulatory timetable which is long since past. The Sierra Club filed suit in Federal Court in San Francisco to compel EPA to develop the regulations. As a consequence of that suit, EPA is now under obligation to prescribe these regulations for a number of different categories of emitters, one of which is DOE facilities. We know from talking to EPA that the DOE facilities are at the top of EPA's list, primarily because they think that they have better information on our facilities, which in turn is due largely to the fact that we have provided them with better information. EPA is under court order to act; we can't really do anything about that. We have some concerns, however, not the least of which is the fact that EPA itself admits that their own scientific expertise is not terribly good in this area. Our second cause for concern is that there is no proven
level of exposure below which we can consider that there are no risks regardless of how small. The statute requires that the EPA standards for hazardous air pollutants provide for an ample margin of safety. Thus, when you consider that there is no generally acceptable threshold below which there are no possible minimal risks, we're concerned that EPA could propose a standard which would have an emission limitation, if not of zero, then close to zero. A final concern, similar to our concern in the RCRA issue, is that once again the states are free to regulate our facilities more stringently than might be justified on environmental grounds alone. So far we don't know of any states who are interested in doing that, but once EPA makes the first cut it's not unreasonable to assume that some states may decide to refine or to secondguess the EPA standard. The choice that we were faced with is whether we should get involved now, to try to influence what EPA is going to do, or whether we should sit back and take our chances, to see what they come up with, and then get involved at that point. The decision was made last summer that we should get involved now, do what we can to influence the standards before they are formally proposed. As with the RCRA issue, the suggestion was made by EPA that we enter into a Memorandum of Understanding with them in order to define the two agencies' respective responsibilities. We were fortunate in having at EPA, in the person of Glen Sjoblom, someone who understood DOE's concerns from his own experience in the Naval Reactors program, and was frank to admit EPA's need for cooperation from DOE. Accordingly, the MOU which was prepared by my office was formally signed in October of this year. The MOU commits DOE to assist bPA anci to provide available information on an as-needed basis. It further commits DOE to participate in any special studies which the agencies agree needs to be done, and, most importantly from our point of view, it obligates EPA to discuss with us the direction in which their standard is heading before they formalize anything in a Federal Register proposal. We think this is of enormous significance to DOE. We will be able to see what they're doing, and we have every reason to believe that we'll have an opportunity to provide input that will make for better standards. Before concluding, I might just mention the problem with state regulation under the Clean Air Act. As I am sure most of you are aware, the Clean Air Act is now before the Congress for reauthorization. It's not clear when action is going to take place, or what the nature of that action will be, but we are prepared at DOE to take advantage of any opportunity that presents itself to propose an amendment to the Clean Air Act language dealing with Presidental authority to exempt facilities by regulation from requirements of the act. Under the present provisions, the President can grant an exemption by regulation, effective for three years only for facilities which are military in character, basically facilities owned by the armed forces. We are prepared to offer an amendment, which we think is reasonably subtle and would therefore have a good change of passage, that would grant the President the authority to issue exemptions by regulation, for any government facilities which are essential to the national defense. We think that under that language most if not all of the DOE facilities would be eligible for a Presidental exemption. 8
RESPONSES TO QUESTIONS The meaning of the Executive Order 12088 is not altogether clear. I believe I talked about that two years ago to this same group. Executive Order 12088 requires federal agencies to comply with applicable state environmental laws. The question is, what is an "applicable" state law; the Executive Order doesn't say. OMB has never said formally what the Executive Order means; they have said informally, and we have accepted this as legitimate guidance, that Executive Order 12088 applies only to environmental requirements undertaken pursuant to a federal statute which requires federal facilities to comply with state regulations. The argument may begin to seem circular at this point. If our position is accepted that RCRA does not apply to these DOE facilities, then Executive Order 12088 wouldn't apply; the South Carolina regulations would not be an "applicable'Yequirement; and we would not be required to comply with them under that Executive Order. That is the basic position that we would take. I haven't talked to .John Cumbee lately; I'm not sure what he is doing on this, but that's the advice that I would give him and you. I am surprised that South Carolina hasn't seen the draft MOU by this point. It's been sent to the EPA regional offices, and I might have expected that it would have filtered down to soma of the more concerned states by this point. The issue is considerably unsettled as you can easily gather, but we're not without a plausible argument. I think that we can maintain our position on that ground. We know what we want the Executive Order to mean; we're not sure who else agrees with us. I see two different issues here; one dealing with legal requirements and another dealing with what we ought to be doing as a matter of policy. I'll speak first to the legal issue. Obviously, where the law says federal facilities must comply with state regulations, we have to comply. Where the law appears not to impose such a requirement, however, and where at the same time we foresee the possibility of state regulations which will not materially contribute to improved environmental protection, but which could burden our facilities considerably, then as a matter of law I don't see any reed for us to comply with the state regulations. We have independent authority under the Atomic Energy Act to regulate the environmental effluents from DOE facilities ourselves. Regarding the second issue, I think we have a serious obligation (and perhaps Bob Davies may wish to speak to this) to step into the state regulators' or the EPA regulators' shoes and do an unquestionably excellent job of managing our activities in such a manner that the degree of environmental protection will be comparable to what that protection would have been if the activities were subject to state laws or the EPA regulations. Bob, would you like to say something? (Davies—the only thing that I would really comment on is that we have to be equivalent to either; if we were regulated by another federal agency or if we were regulated by the state. If we're not equivalent to that, then we're in jeopardy). That's right, that is our policy as we have stated it to EPA. Talking specifically about the RCRA issue, we have stated repeatedly to EPA that our policy is to develop a DOE
Order which will provide a degree of environmental protection that is comparable to their regulations. We have to take that responsibility very seriously, it seems to me. (Davies—I'd like to comment on the two conditions you brought up; I don't think either one of those is really acceptable. That's the lack of discipline that I was talking about. I think we need to clean our act up in these kinds of areas. If things are optional, you're in trouble; I think you're in trouble if somebody begins to press. It may be perfectly okay as far as environmental protection is concerned, but procedurally you're not in order and I think we're in jeopardy if anybody presses on that point). Yes, and we're seeing that not only with regard to Congresswoman Schroeder's concerns about the regulation of our nuclear safety activities; we're also seeing it with EPA in the RCRA area. Attached to the RCRA draft MOU that we sent to EPA on September 2, was a draft of the DOE Order on hazardous waste management. We felt it was important to attach that so that EPA would be able to see for itself that v»e do have this commitment, that we are developing an Order, and that we do intend to policy ourselves. Informally, I've been told by EPA staff that the draft Order appeared to be vague in places, it appeared to provide only criteria rather than requirements, and there was a certain sensitivity on the part of EPA officials to turning over to the Department of Energy a responsibility which would not necessarily be managed as strictly or as carefully as they would manage it themselves. We may very well have to alleviate that type of concern before we are able to settle that issue with hPA. I'm not sure if I understand what you're saying. Are you suggesting that we go ahead and comply with requirements which we don't believe to be necessary from an environmental protection standpoint, so long as we can comply without interfering with our programs? Certainly the advice of my office over the course of the last two years has been to do whatever is necessary to keep DOE programs running. I don't know what kinds of problems are going to present themselves when we have an official DOE Order on hazardous waste management. I think it not unreasonable to expect that you will be presented with some difficult requirements, something which you would not, of yourselves, choose to do. I think one of the important aspects of the Order is that, for purposes of your budgetary planning, you will finally know what it is that you are expected to do and you can start to plan accordingly. I don't think it at all likely that you are going to find the DOE Order perfectly simple'to comply with and undemanding. I don't know just what the problems are going to be, but I think you can expect some. As you say, it would be a lot easier if you simply know that any time the state or locality comes out with a regulation, your responsibility is to comply with it. This brings me back a second time to the point that Bob Dayies made earlier about all decisions ending up as political decisions or policy decisions. The legal issues here are not always clear. The legal issue with RCRA, as I have already indicated, isn't clear. The interpretation of Executive Order 12088 is not crystal clear. I can't stand here and tell you as a matter of law that it definitely means this or that it definitely means something else. DOE program managers, particularly in the Office of Defense Programs, have taken the position that the kind of easy interpretation that you are talking about—that we simply comply with all state and local 10
requirements—is something that they can't live with. In serving them as lawyers, we have shown them whatever plausible bases exist for taking positions other than automatic submission to state or local regulation. It's a policy decision; they feel that their tasks as managers are made easier by the very sort of ambiguity of which you're complaining. As you know, the cost of compliance with environmental requirements can be enormous. In view of such considerations, and particularly in view of the fact that much of this cost may be for compliance with requirements that may not materially enhance the degree of actual environmental protection, I think it would be a reasonable policy decision for DOE to choose not to comply with requirements which as a matter of law do not clearly apply to us. An area which I haven't worked on directly. I'm not sure I know the answer, but Carl may. (Welty—yes, I would like to speak to that. We have been keeping in close touch with the staff of the radiation programs.at EPA for the last several years and it's the position of the staff, as told to us over this long period of time, that their intent is that we will have generic standards applied to us and they will be in the form of dose limitations. They would like to have a single number that applies generically to all of our operations. Now they're testing this out on EPA general counsel. Apparently there are positive indications back to technical staff, but the policy decision has not been made and we're just hopeful that the staff of Radiations Programs will win out in this. There are some legal questions too to be addressed because the implications under Section 112 would be that the limitations have tc be on the effluents, but there are precedents under Section 112. For example, asbestos is controlled according to procedural requirements laid on the operators rather than emission limits). This is something that I haven't worked on directly; I don't know a great deal about it myself, but as Ed Patterson said in his opening remarks, to the extent that we here can't help you, we can tell you who can. There is an attorney at DOE Headquarters by the name of Tom Wolfe, who is involved in some detail in this matter, and I suggest you get in touch with him if you want to pursue it further. (Welty—I might add that we keep in touch with Bill Dennison and Steve Greenleigh and Attorney Wolfe). I drafted the MOU that was later signed by Secretary Edwards and Mrs. Gorsuch, and I find that it doesn't take a great deal of talent to put something on paper whereby two people agree to work with each other,but when their staffs actually get into the details of the subject, it gets a little more complicated. As I tried to indicate earlier, regardless of this MOU, we see the need to amend the Clean Air Act at the first opportunity to give us some further protection, through Presidential exemptions, from unduly burdensome state regulations. I think it is \/ery important to get that amendment into the Clean Air Act if we possibly can. We're watching legislative developments very closely to take advantage of any opportunity to do that. Our amendment would not be a cure-all,, but at least it would give us some assurance that a state would not be able to come in and shut one of our facilities down, if the President were authorized to give us an exemption by regulation for three years. The language in RCRA, as I indicated earlier, is not duplicated in any other environmental statute. Although the significance of that language 11
is ambiguous, as I've said, it seems plausible to suggest that we can avoid submitting to state hazardous waste regulations altogether. We don't have that luxury in the Clean Air Act. We have to live with state regulations under that Act. We may need the umbrella of Presidential exemptions to get us through that. State regulation is, I think, the biggest problem that we face in the area of environmental regulation generally. DOE managers seem to be fairly consistent in their policy view that we should avoid state regulation wherever possible, so as to avoid the imposition of any unreasonable state requirements on our facilities. The strategy by which we go about that varies from statute to statute because the statutory language varies. Ir. RCRA, we feel we have a strong argument for avoiding state regulation altogether. In other cases it's going to be more difficult. In those cases, perhaps the only thing we can do is to ask for EPA's good offices to talk to the states on our behalf to try to make them see reason in the instances where we think that they're not seeing reason, and frankly, I am not sure how much of a help that's going to be. Once a state has been delegated enforcement authority, that authority is pretty broad—there's not a whole lot that EPA can do to rein in overenthusiastic state regulators. So it is a difficult matter. A ( lot of these problems are not readily resolved. I dor't know what the answers are to a lot of them, and we will have to work together to find those answers.
12
1C OVERVIEW OF CURRENT PROGRAMS AND TRENDS IN ASSESSMENT OF HEALTH EFFECTS Dr. J. W. Thiessen Office of Health and Environmental Research U.S. Department of Energy Washington, D.C. 20545 I am glad to be here, because what we do in the Office of Health and Environmental Research is meant to be of use to you. Our name reflects our intentions to contribute to the protection of human health and the environment to the extent possible. Two years ago we were part of what is now called EP. Since then, our office was moved into the Office of Energy Research, which represented a change in emphasis, an essential change which I would like to stress at the outset. We are now bound to the goals of the nation's Energy Research, i.e., performing long-range research. In other words, we are aiming our research at resolving the problems of the future. In the old EV we were very much also involved in solving the problems of today, and if you can say anything about a difference between "regulatory" and "nonregulatory" activities, as in the case of EP and ER, it is that the regulatory agencies try to solve the problems of today, and very often, as you know, the problems of yesterday. We are aiming at tomorrow. Ostensibly, in our "ivory tower" we can look down and further away than you all can, and so we are trying to identify the problems of tomorrow and work on them now, so that we have the answers when they are necessary. Of course: yesterday's tomorrow is today. It is certainly true that we haven't solved a number of problems that you are dealing with today, and both you and we should identify those remaining problems and work towards their solution now. From my personal point of view, there is not much real practical advantage in making too fine a distinction between today and tomorrow in this sense. Also in EV, we definitely had a mission to support the operations of the Agency. It has been mentioned before this morning, if DOE is its own EPA and OSHA, there is a point to be made that OHER should be its NIOSH and NIEHS. So there is a need and that need has not been resolved, and from that point of view also we may have to look at a framework that somehow addresses the need for health and environmental reseanch now, and for your purposes. For example, if work is being done in Utah to determine population exposure downwind from the Nevada Test Site, then clearly that appears to be a problem left over from yesterday, and yet it has consequences way beyond that in the future. If there is time, I would like to say a few things about our nonnuclear, energy-related research which is quite extensive also, but I don't want to try to do too much in too short a time, so I will emphasize our radiation research. Furthermore, I will primarily talk about health effects research in humans. I will not talk too much, if at all, about environmental research. Again, in order to be able to say something that is worthwhile listening to rather than go through a whole lot in a very short time. If we could know everything there is to know by examining human populations or individuals, we would of course do that. There is no question that direct human observations, even though they may be incomplete, are to be 13
preferred above observations in animals, or further down in cells or, on the subcellular level, in DNA. But as you all know, suitable human populations are not always available and there are no such human individuals, or, if there are, they are too few. Consequently, we need a rather extensive underpinning in animal research, i.e., research on animal organs, complete animals, or organs or cellular systems, cell tissues, or molecules of biological importance, i.e., DNA and other proteins in the cell. On the question of the effects of external ionizing radiation, we will discuss the information on what we know by direct observation. As far as the internal emitters are concerned, the situation is the same. In human health effects research, we have a few groups—the miners, the plutonium workers, the radium dial painters, the medical users of radium--where we have a rather substantial data base. But in most cases we are still very much dependent upon experiments on animals to determine these effects in terms of internal dosimetry and metabolism of internal emitters. I must say also that, in the old AEC days, the miners were not considered to be under the jurisdiction of AEC. Consequently, all the health effects research on miners or most of it has been done by the Public Health Service. We now intend doing an epidemiological study on miners in New Hexico. In our minds there are still quite a few unresolved questions with respect to the health effects of radon and radon daughters, and you all know of course that has become more and more important in the context of conservation measures, which tend to increase the exposure to radon in the indoor environment. As far as the role of the other government agencies is concerned, about 70 percent of the federal radiation research is performed by DOE, about 15 percent by the National Institutes of Health, and about 15 percent by all others. Of course, this is 1982; five years ago, we did even more than 70 percent. It is also worthwhile looking at the categories of research. Human effects research in DOE is about one-fifth of the total budget in 1982. There is quite a bit that has to do with determining source characteristics, physics, chemistry, etc. Our program is a multidisciplinary program, it is not just a program that concerns biologists and physicians, it spans the entire range from characterization of the source in physical and chemical terms through pathway studies and environmental studies, to, as I explained, the animal research. Let's look at the epidemiological studies, the ones that I spoke about before (Figure 1). You see that we have quite a few rather large groups that, though not exposed to very high levels of radiation, seem to or are beginning to meet the statistical requirement of large numbers. For small exposures, the effects are small, of course, so that there is only one way to get any sort of reliable data, and that is to look at large numbers. That is what we are doing. Now as far as the ABCC (the old Atomic Bomb Casualty Commission) survivor studies are concerned (now carried out by the Radiation Effects
14
STUDY POPULATION
TYPE OF EXPOSURE
NUMBER OF SUBJECTS
ATOMIC BOMB SURVIVORS (RERF)
HIGH DOSE, ACUTE EXTERNAL
170,000
DOE/DOE CONTRACTOR WORKERS
LOW DOSE, CHRONIC EXTERNAL
600,000
NUCLEAR SHIPYARD WORKERS
LOW DOSE, CHRONIC EXTERNAL
250.000
PLUTONIUM WORKERS
CHRONIC, INTERNAL
15,000
RADIUM DIAL PAINTERS
CHRONIC, INTERNAL
10,000
CYCLOTRON WORKERS
MAGNETIC FIELDS
U.S. OIL SHALE WORKERS
SHALE OIL AND RELATED PRODUCTS
SCOTTISH OIL SHALE WORKERS
SHALE OIL AND RELATED PRODUCTS
U.S. GENERAL POPULATION
CARBON MONOXIDE
750
87+ 10,003 9,000
FIGURE 1. Department of Energy Epidemiologic Studies Research Foundation in Hiroshima and Nagasaki), that is still an on-going study. It represents about 50 percent of the budget in human health. So it is an enormous study. It is a $10 million a year study as far as the United Jtates is concerned, with about the same amount contributed by the Japanese government. So there is no question that it is one of our more important studies, and I would like to talk a little bit about that because to a large extent, a lot of what we know about carcinogenesis in man rests rather heavily on what we have observed (and are still observing) in Hiroshima and Nagasaki. As you know, it was very clear, pretty soon after the detonations in 1945, that leukemia went up rather rapidly among the A-bomb survivors. Since then, incontrovertible evidence has been collected on relationships between radiation and lung cancer, breast cancer, stomach cancer, urinary tract cancer, and esophageal cancer. For all of those, based on the data of Hiroshima and Nagasaki, there is solid evidence that cannot really be questioned as to the solidity of the relationship between cause and effect. This whole data base is still being added to. Every four or five years the numbers are being looked at and statistically analyzed. In the latest review, in 1978, colon cancer and multiple myeloma were added to the list of cancers for which we have incontrovertible evidence of their relationship with radiation exposure. As yet we have no significant evidence of such relationships with respect to malignant lymphoma, rectal cancer, pancreas cancer, and uterine cancer. Furthermore, from the data that we have on Hiroshima and Nagasaki, we are not really able to determine whether those relationships are linear or nonlinear. All the discussions about linearity, and they are very important of course with respect to standard setting, rest on information that is taken mostly from animal experiments, information that, as you know, is not supportive of linearity down to very low doses. 15
For the RBE of neutrons, all we knew in man was derived years ago from the Hiroshima data. It appeared at that time that the RBE for neutrons, at least with respect to induction of leukemia, breast cancer and stomach cancer, was between two and seven. Of course, you know that Rossi and Mays came up with an entirely different approach which resulted in the conclusion that the RBE of neutrons at very low doses might be considerably higher, even as high as 100. And that upset quite a lot of people. About that time though, and there had been some rumblings of things to come, about that time Livermore came out with the results of a study by Loewe and Mendelsohn that indicated that the neutron doses at Hiroshima were considerably lower than we had previously thought, about an order of magnitude, ar)d that the gamma doses were probably somewhat higher. In Nagasaki, the neutron doses were also lower, but as they had been inconsequential anyway, that did not make much of a difference. The gamma doses were somewhat lower, but then again, the differences were not all that important (Figure 2). The attention was redirected to the whole situation in Hiroshima. And when that happened, it was very clear that the whole neutron issue, at least the issue of very high RBEs seemed to sort of resolve itself and disappear. However, it wes also clear that the whole matter of the relationship between exposures or doses on th6 one hand, and carcinogenesis on the other hand, had to be reconsidered on the basis of a reassessment of the doses in Hiroshima and, to a lesser extent, in Nagasaki. In September 1981, we had a workshop in Germantown, with all the investigators and everybody who had an interest, regulatory or otherwise, and it was evident that we would have to start or at least coordinate a rather extensive program. There clearly was a need from the ep^emiological point of view for the people working on Hiroshima and Nagasaki who needed a firm data base to redo tlieir dose-effect analyses. In December 1981, we found Dr. Christy at Cal Tech willing to coordinate this program. His professional career goes back to the Manhattan Project. He is very familiar with weaponsrelated radiation, and so forth. And above ail, he was accepted by all the investigators. We are extremely fortunate to have him. At the same time we made it clear that the entire effort would be done in the open, with the cooperation of the Japanese scientists as much as possible, and with some type of overview and review authority to be given to the National Academy of Sciences. So we asked the Academy to set up a panel for overview purposes. Dr. Seitz, ex-president of the Academy, became the chairman, again somebody who has extensive experience and insights into the particular problem. They have met twice now, and in October 1982, we had a joint meeting. In February 1983, we will have a workshop together with the Japanese in Nagasaki. It is very clear that to them this is a very sensitive issue. They are heavily involved with this, even though, of course, a large portion of the research is done in the United States. The role of the Japanese scientists is to, as much as possible, confirm our.calculated doses by inducedradioactivity studies, and so on. By late 1983, we hope to have at least the free air dose calculations finished and sometime in 1984 practically 95 percent of the dosimetry reassessment; that is, the individual organ and whole body doses of each individual in the study group will, we hope, be completed sometime in late 1984. It is very clear that everybody is waiting for the conclusion of the dosimetry reassessment, because it may result in a revision 16
MODEL
MODEL
"PRELIMINARY"
"PRELIMINARY"
100
Y - LLNL
o <
3
0.1
CO
0.01
0.01 -
1
2D "AIR" DOSE HIROSHIMA 12.5 kt (MOISTURE INCLUDED) I
I
1.2
1.4
\~
2D "AIR" DOSE NAGASAKI 22 kt (MOISTURE INCLUDED)
I
I
1.6
1.8
1
2.0
DISTANCE FROM GROUND ZERO-km
1.2
1.4
1.6
I
I
1.8
2.0
DISTANCE FROM GROUND ZERO-km
FIGURE 2. Air Doses in Hiroshima and Nagasaki; T65D and LLNL Data of our risk factors. For example, it looks now that if there is a change, it is likely to be no more than a factor of two upwards, that is that radiation may be twice as effective as we thought before, at least on the basis of Hiroshima and Nagasaki data. That, however, is not that important, because other information that we have from clinical practice data, therapeutic radiation, etc., appears to indicate that risk factors derived from Hiroshima and Nagasaki data were a bit low anyway. I don't think we can expect a complete revision, but there certainly will be some. Frankly, if we look at what we have now, things appear tc look much better with respect to, for example, dose-response relationships, and especially with respect to the discrepancies between the two cities. There were inexplicable differences between the Hiroshima and Nagasaki data; most of these differences seem, statistically, to disappear. Our data base is improving, and I think we will eventually have a much better dosimetric basis for our epidemiologic studies in Japan. The Hiroshima analogue of internal dosimetry (on a smaller scale, of course) is the radium dial painter situation. In this photograph (Figure 3 ) , taken in 1925, 68 women are pictured of which 55 could be traced. Among those 55, four developed bone cancer, and seven carcinomas of the sinus or mastoid. That is 20 percent, a tremendously large fraction. 17
00
FIGURE 3 -
Radium Dial Painters: Radium-Dial Painting Plant in I l l i n o i s (1925). Of these 68 women, 55 located, 4 bone cancers and 7 carcinomas of the sinus or mastoids have been diagnosed.
Argonne National Laboratory, in the Center for Human Radiobiology, has been examining radium dial painters on a periodic basis and still occasionally finds "new" ones. Of all the cases.foilowed, we have not been able to find a case of bone cancer or sinus tumor at total uptakes to bone below 30 microcuries. This case was found quite a long time ago, and it is on this single datapoint that the maximum permissible bone burden of radium, and through it, of all other bone seekers, was based, until the recent complete revisions by ICRP. Figure 4 indicates the epidemiological study population included in the DOE "Health and Mortality Study." What I would like to point out here is that the DOE-contractor employees included go back to the Manhattan Project. These are subpopulations that are being studied on the basis of priority. Many of the uranium miners, for example, are getting pretty old. The time actually is to a large extent past to look into these. So we have mortality studies covering all DOE, contractor, and past employees. The plutonium worker study is a subpopulation containing a number of groups. The U.S. Radiation Accident Registry is an interesting project because we have so few real cases of overexposures in humans that every accident case that occurs becomes a research subject. That is why we have a contract with Dr. Lushbaugh at Oak Ridge Associated Universities, to identify and evaluate every accident case, wherever it occurs. The "5 Rem Study" is also part of the Oak Ridge study. I will not go into much detail on this and the other studies, but want you to understand that we are talking about a large group of people that is extremely diverse and extremely complicated. This is not something for which we can give an expected date of completion. Research of this kind by its very nature takes decades, and there is no way to expedite that appreciably. The next research area I would like to discuss is related to the residents living downwind from the Nevada Test Site. These are some of the isodose contours downwind of the Nevada Test Site (Figure 5 ) , with the Harry and Nancy Shot as two of the more contaminating events. There is a lawsuit going on now concerning individuals in southern Utah alleging that they have been exposed to radiation from these tests, and that because of that they have been developing cancer. This matter started to develop when a Utah University researcher, Dr. Jos. Lyon, evaluated leukemia rates for what he called the high fallout and low fallout counties, some three or four years ago. This study indicated that for the so-called high fallout counties, i.e., the southern counties of Utah, and especially for the period of high exposures in Utah that appeared to be between 1951 and 1958, there was a significant increase of leukemias, compared to leukemia rates in the so-called low fallout counties (Figure 6). These results can be interpreted in quite different ways. For instance, another study done by Dr. Charles Land, of the National Cancer Institute, concluded that, if you look at other childhood cancers, you see the opposite happening, i.e., in the s-called high fallout counties, the number of other childhood cancers is actually lower than in the so-called low fallout counties (Figure 7). That didn't make sense. At the same time, DOE's Environmental Measurements Laboratory was involved in a study of the population exposure downwind from the Nevada Test Site, and much to everybody's surprise, they found that the exposures in northern Utah were not lower than those in 19
RADIATION ACCIDENT REGISTRY
FIGURE 4. Department of Energy Epidemiologic Studies Populations 20
Between 1951 and 1962, the governmant conducted 84 above-ground nuclear tests. Since 1 9 6 2 , 1 8 belowground blasts have vented radioactivity into the atmosphere.
I Bryce Canyon National Park
Parowan ,-r; ^'CedarCity , / / 2ion i_ National
\\
I iPark
Washington Fredonia*
Mesquite
Arizona
Grand Canyon National Park San Francisco
\NIV CAL.
Las Vegas
Lake Mead National Recreation Area
40 Miles
Cal. FIGURE 5. Isodose Contours, Downwind from Nevada Test Site
AREA
LOW-EXPOSURE COHORT, 1944-50
HIGH-EXPOSURE COHORT, 1951-58
LOW-EXPOSURE COHORT, 1959-75
UTAH
3.45 (2.50+4.40)
4.23 (3.60±4.87)
3.11 (2,54±3.69)
HIGH-FALLOUT COUNTIES
2.10 (0.54±3.66)
4.39 (2.81 ±5.98)
1.96 <0.73±3.19>
LOW-FALLOUT COUNTIES
3.84 (2.70+4.97)
4.21 (3.51 ±4.90)
3.28 (2.64±3.92)
FIGURE 6 .
Leukemia Mortality Rates per 100,000 Population
of Both Sexes Adjusted for Age and Sex 21
OTHER CHILDHOOD CANCER DEATHS
LEUKEMIA DEATHS FALLOUT LEVEL
EXPOSURE COHORT HIGH
HIGH
32
LOW
152
ODDS RATIO
FALLOUT LEVEL
LOW |
EXPOSURE COHORT HIGH
LOW
17
HIGH
21
36
156
LOW
165
156
1.93
0.55
3.68 (P = 0.055)
3.53 (P = 0.060)
FIGURE 7. Comparison of Deaths from Leukemia and Other Childhood Cancers in the Two Cohorts According to Fallout Level southern Utah. As a matter of fact they were higher, and to cite from the EML's latest reports, in the northern counties the average dose is about 1.3 R, and in the southern counties, something like 0.86 R. So you have the opposite of what was assumed initially in Lyon's study. Well, to make a long story short, there was heavy political pressure on HHS, and on us, and on Defense to do a more definitive study to resolve this and other issues. So after long discussions, a proposal was developed by the University of Utah to study both thyroid cancers and leukemias in the downwind population. This is something like a six million dollar study over five years that is funded by NCI, DOE, and by DOD together, and that will look into this with much more detail than both earlier studies could afford. I would like to talk a little more about the relationship between EP and ER. We have a Health and Environmental Risk Analysis Program (HERAP), which, in my opinion, is developing quite nicely. Its mission is to evaluate energy technologies and related problems with respect to their health and environmental uncertainties. That is, HERAP is to indicate on a recurring basis, annually or bi-annually, what is the state-of-the-art with respect to health and environmental uncertainties. So that we can establish research programs to improve these. They have been producing so-called Health and Environmental Effects Documents, HEEDS, and here you see a list of these HEEDS (Figure 8 ) . The Health and Environmental Risk Analysis Program also involves the preparation of more generic analyses and methodology development. It is very clear that there is still a lot that can be done to analyze problems such as this one. These problems definitely are the ones that are \/ery much with us today, and in some of those, I am sure, you are heavily involved. For that matter a lot of the DOE activities, even not directly energy related, could require, in my feeling, a lot more extensive contact between EP and ER. Some of these projects or subject areas have been mentioned already. We are talking about the Clean Air Act radionuclide standards for example. 22
1
CY1979 |
TECHNOLOGY SPECIFIC ANALYSES ELECTRIC STORAGE SYSTEMS DIESEL ENGINES PHOTOVOLTAICS
BNL
OIL SHALE
IWG
COAL LIQUEFACTION H-COAL GEOTHERMAL
ORNL LLNL
COAL GASIFICATION HYGAS
ANL
FLUIDIZED BED COMBUSTION
ITRI
REFUSE DERIVED FUEL SPACE NUCLEAR SYSTEMS
FY 1980
FY 1979 AML ITRI
.
BREEDER REACTORS
CY1981
FY1981
A
.
CY 1982 FY 1982
| CY 1983 FY1983
A
o
i \
A j
L
/
A»
\
J L
A,
AMES
o o
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A,
BUILDING CONSERVATION COAL LIQUEFACTION SRC II
CY1980 |
PNL ORNL
A_
A
FUSION A—
LOW BTU GASIFICATION
ANEGOTIATION A S T A R T
^REPORTSCHEDULED O
REF>
1
O R T SUBMITTED
FIGURE 8. Health and Environmental Risk Analysis Program There is no question that we should have the data base in hand, and if not, if we need additional information we better find out what it is that we need and when we need it—radon mill tailings, Marshall Islands rehabilitation, routine and nonroutine tritium and plutonium releases or what have you. A lot of these subjects have their own uncertainties with respect to effects, and we better find out what they are because, if we don't, we will have to rely on very cost-ineffective command-and-control regulatory policies. In many cases regulatory agencies, and I include ourselves, have a tendency to add on a safety margin. If we keep doing this, we add conservatism upon conservatism until the final levels or the final criteria, or the final standard is really so conservative as to be laughable. Then because we don't have any other more reasonable figure or procedure or standard, we have to accept it. So we try to reduce these uncertainties, that is to increase our knowledge, in order not to get in that sort of situation, and some of these situations have been discussed this morning. If anybody is interested in a particular subject, I have quite a bit of information with me, and will be willing to discuss this with you personally Thank you.
ID
OFFICE OF OPERATIONAL SAFETY ENVIRONMENTAL PROGRAM MISSIONS AND OBJECTIVES C. G. Nelty, Jr. Office of Operational Safety U.S. Department of Energy Washington, D.C.
I want to take a few minutes this morning to discuss the 00S Environmental Protection Program, how it functions and who is involved in carrying it out. I want to also emphasize the importance of the program and to encourage each of you in your efforts to assure that all elements of the program
particularly important to tsiose function areas, radiological and hazardous waste, where DOE is self regulating. Note that the overview subprogram makes use of information system, appraisals, and periodic reports. The Outlay Subprogram is shown in figure number 7. This subprogram is supportive of the other three subprograms and is primarily geared to assuring that program requirements and guidance are adequate and up-to-date. This subprogram has expanded significantly in the last five years, with the greatest increase taking place in the hazardous waste management programs. Figure 8 indicates the present staffing and functional assignments in the Environmental Protection Group. FIGURE 1. Environmental Protection Program PROGRAM MISSION TO ASSURE THAT DOE OPERATIONS COMPLY WITH APPLICABLE FEDERAL, STATE, AND LOCAL ENVIRONMENTAL PROTECTION LAWS, RULES AND REGULATIONS AND PROTECTION OF THE PUBLIC HEALTH AND WELFARE.
FIGURE 2. Objectives •
HAVE DEPARTMENTAL POLICY AND REQUIREMENTS IN PLACE
•
PROVIDE A COHERENT DOE ENVIRONMENTAL PROTECTION PROGRAM
•
ASSURE ADEQUATE OVERVIEW OF THE DOE PROGRAM AT ALL LEVELS
•
BE RESPONSIVE TO LINE ORGANIZATION NEEDS (HQ & FO)
•
PROVIDE A CENTRAL POINT OF COORDINATION WITHIN AND OUTSIDE DOE
FIGURE 3. Subprograms •
POLICY, REQUIREMENTS AND GUIDANCE
•
COMMUNICATION AND COORDINATION
•
OVERVIEW
•
TECHNICAL SUPPORT
26
FIGURE 4. Policy, Requirements and Guidance Subprogram •
PROMULGATE FEDERAL & DEPARTMENTAL REQUIREMENTS + ORDER 5480.1A, ENVIRONMENTAL PROTECTION, SAFETY AND HEALTH PROTECTION CHAPTER XI - REQUIREMENTS FOR RADIATION PROTECTION CHAPTER XII - PROVENTION CONTROL AND ABATEMENT OF ENVIRONMENTAL POLLUTION + ORDER 5484.1, ENVIRONMENTAL PROTECTION, SAFETY AND HEALTH REPORTING REQUIREMENTS CHAPTER III - EFFLUENT AND ENVIRONMENTAL MONITORING PROGRAM REQUIREMENTS CHAPTER IV - ENVIRONMENTAL PROTECTION, SAFETY AND HEALTH REPORTS
•
INTERPRET AND UPGRADE REQUIREMENTS (ORDERS) + REPORTING (PERIODIC REPORTS & INFORMATION SYSTEMS) + QUALITY ASSURANCE + RADIATION DOSE MODELING + GROUNDWATER MONITORING + HAZARDOUS WASTE MANAGEMENT
o
DEVELOP GUIDANCE IN SUPPORT OF ORDERS + ENVIRONMENTAL MONITORING GUIDE + EFFLUENT MONITORING GUIDE :• PREPARATION OF HAZARDOUS WASTE MANAGEMENT PLANS
27
FIGURE 5. Communication and Coordination Subprogram • WITHIN ASEP
)
•
OTHER FEDERAL AGENCIES AND THE PUBLIC
•
PROGRAM OFFICES + PROBLEMS AND INQUIRIES (HQ & FOs) + CONFERENCES AND WORKSHOPS + SPECIAL PROBLEMS + EARLY WARNING ON LAWS AND REGULATIONS + BUDGET CROSSCUT + WASTE MANAGEMENT POLICY AND REQUIREMENTS
FIGURE 6. Overview Subprogram • OVERVIEW VIA INFORMATION SYSTEMS + EFFLUENT INFORMATION SYSTEMS (EIS/ODIS) + SITE ENVIRONMENTAL MONITORING REPORTS + REPORTING OF LIMITS EXCEEDED AND VIOLATIONS + POLLUTION ABATEMENT REPORT (OMB CIRCULAR A-106) + GRAPHIC OVERVIEW + AERIAL MEASURING SYSTEM (AMS) •
OVERVIEW VIA PROGRAM APPRAISAL
•
ANNUAL REPORTS
28
FIGURE 7. Outlay Subprogram LEAD LABORATORY PROJECT (PNL) (MULTI TASK SUPPORT PROGRAM) RADIOLOGICAL CRITERIA DEVELOPMENT (LANL) (HUMAN EXPOSURE PARAMETERS) ANALYTICAL LABORATORY QUALITY ASSESSMENT PROGRAM (EML) HAZARDOUS WASTE (LANL, ORNL, & BNL) WASTE MANAGEMENT PLANNING GUIDANCE (ORNL) REFUSE MANAGEMENT AND DISPOSAL CRITERIA (BNL)
FIGURE 8. Environmental Protection Staff CARL WELTY - GROUP LEADER VINCE DECARLO - HAZARDOUS WASTES ABDUL DASTI - REGULATIONS, INFO. SYSTEMS, ABATEMENT JOHN MATHUR - RADIOLOGICAL, GROUNDWATER, QA DAN HENNINGER - NEPA RELATED
29J3-)
IE
PNL'S ENVIRONMENTAL PROTECTION SUPPORT PROJECT FOR THE OFFICE OF OPERATIONAL SAFETY* J. P. Corley ani R. E. Jaquish Pacific Northwfst Laboratory Richland, Washington 99352
ABSTRACT The objectives and a brief history of the PNL project entitled "Environmental Protection Support and Assistance to DOE Office of Operational Safety" are presented, including the major tasks undertaken to date, and the current year's efforts. The major products of this effort have been: "A Guide for Environmental Radiological Surveillance at U.S. Department of Energy Installations," DOE/EP-0023 (ERDA 77-24 revised); "A Guide for Effluent Radiological Measurements at DOE Installations," in process of publication, and summaries of the annual environmental reports from DOE sites. Support documentation is mentioned. A major effort is currently underway to proyide additional guidance on environmental data reporting, including non-radiological pollutants and witn emphasis on non-compliance events.
Those of you who have any responsibilities for environmental surveillance at the Department of Energy's nuclear sites w i l l have been affected to some degree by the Pacific Northwest Laboratory's continuing support project for the Office of Operational Safety. I trust you will be interested then in a summary of the project's history, its objectives, and its current direction. Although the project as such is only in its third year, i t had its inception in two tasks undertaken for Art Schoen's Environmental Protection Branch of the Atomic Energy Commission. The f i r s t of these was to identify, *This paper is based on work performed under United States Department of Energy Contract DE-AC06-76RL0-1830.
31
on the basis of the annual environmental reports submitted to the Branch, any areas of technical inconsistency and potential weaknesses in site environmental programs, and especially to identify and recommend needed supplemental guidance to the requirements of AEC Manual Chapter 0513. The latter, you may recall, covered Effluent and Environmental Monitoring and Reporting, and after one revision as ERDA Manual Chapter 0513, was largely incorporated in DOE Order 5484.1. The outcome of our early reviews, followed by a questionnaire, site visits, and extensive reviews by other contractors and ERDA staff, was the 1977 editionjof the Environmental Guide for ERDA Installations.(D
T
In preparing for this presentation, I came across a set of transparencies I haJ used in discussing the preparation of the Environmental Guide. Figures 1 and 2 show what we believed to be the problem areas, and Figure 3 shows our intended emphasis. The 1977 edition did, I believe, provide useful guidance on each of these items. As you may know, the Environmental Guide was updated and reissued in 1981 as D0E/EP-Q023.(2) \ WO u1d like to emphasize several concepts to which we have attempted to hold in our subsequent efforts for the Office of Operational Safety. First, the product was a Guide, not a manual of prescribed procedures. Second, we tried to avail ourselves of the expertise and insights available in other contractor organizations and DOE offices. Although there have certainly been disagreements as well as external constraints, we have tried to consider objectively all comments and suggestions. Third, we have attempted to reflect in our recommendations not necessarily the most current technological capabilities, but what seemed to be good practices, operationally achievable, taking into consideration the variety of problem areas and concerns at the various sites. Recognizing that further study of capabilities was warranted on several aspects of environmental surveillance, and that both technology and the regulatory limits were changeable, agreement was reached with the Office of Operational Safety in FY 1981 on a more generalized program of technical support and assistance. The general objectives of this program are shown in Figure 4. Obviously our efforts can only assist 00S in achieving these objectives. Generalized task area descriptions are shown in Figure 5 and a generalized approach described in Figure 6. Figure 7 lists major publications originating to date in this project. The Effluent Guide, intended as a companion document to the Environmental Guide, is in the final review process by 00S before publication. The 1980 Annual Environmental Summary^3) is the first in this series to reach publication. In the process the format and content have gone through extensive changes in an attempt to provide a product that would be useful
32
PATHWAY ANALYSIS • • • •
CRITICAL PATHWAY IDENTIFICATION QUANTIFICATION OF LOCAL PARAMETERS SUMMATION OF DOSES TO CRITICAL ORGAN(S) 80 km RADIUS POPULATION DOSE
MEASUREMENT METHODOLOGY •
ISOTOPIC VERSUS GROSS ACTIVITY MEASUREMENTS o DIRECT VERSUS INDIRECT -AVAILABILITY OF APPLICABLE METEOROLOGICAL DATA • MINIMUM DETECTION LEVELS Figure 1. Population Dose Assessment Comparability Problem Areas-I
SEPARATION OF OPERATIONAL EFFECTS • • • •
USE OF CONTROL (BACKGROUND) LOCATIONS STATISTICAL APPLICATIONS QUALITY ASSURANCE "LESS THAN" VALUE PROBLEM
DOCUMENTATION
Figure 2. Population Dose Assessment Comparability Problem Areas-II
33
ADEQUATE ENVIRONMENTAL PATHWAY ANALYSIS EMPHASIS ON MONITORING OF CRITICAL PATHWAYS STATISTICAL EVALUATION OF DATA, INCLUDING DISTRIBUTIONAL ANALYSIS APPROPRIATE BACKGROUND OR CONTROL MEASUREMENTS (IF ENVIRONMENTAL EVALUATION IS BASED ON ENVIRONMENTAL DATA) ASSIGNMENT OF NUMERICAL VALUES FOR MINIMUM ENVIRONMENTAL DOSE LEVELS OF CONCERN (RELATED TO THOSE DEFINED BY NRC AS ALAP FOR LWRs) DOCUMENTATION Figure 3. Key Emphases-Environmental Manual
PROVIDE TIMELY AND USEFUL ENVIRONMENTAL PROTECTION REQUIREMENTS AND GUIDANCE TO FIELD OPERATIONS ESTABLISH REQUIRED NEW ENVIRONMENTAL COMPLIANCE INFORMATION SYSTEMS PROVIDE AN ADEQUATE AND RELIABLE INFORMATION BASE FOR EVALUATING FIELD PROGRAMS ASSURE THE RELIABILITY AND ACCURACY OF DOE ENVIRONMENTAL DATA Figure 4. Environmental Protection Support and Assistance Objectives
34
CONDUCT SPECIAL STUDIES ASSOCIATED WITH OOS ENVIRONMENTAL PROTECTION RESPONSIBILITIES ANALYZE IMPACTS OF REGULATIONS AND STANDARDS ASSESS ENVIRONMENTAL ASSESSMENT METHODOLOGIES IN USE AT DOE SITES ASSIST OOS WITH INFORMATION MEETINGS AND CONFERENCES ASSIST OOS IN PREPARING GUIDANCE ON EFFLUENT AND ENVIRONMENTAL MONITORING AND REPORTING Figure 5. Environmental Protection Support and Assistance General Task Areas
• • • • •
DEFINE SPECIFIC TASK AREAS COLLECT DOE SITE AND OTHER INFORMATION AS NEEDED PREPARE DRAFT DOCUMENTS FOR OOS STAFF AND PEER GROUP REVIEW ISSUE COMMENT DRAFTS AND CONDUCT WORKSHOPS AS NEEDED ISSUE LETTER REPORT, PNL DOCUMENT OR CAMERA-READY COPY FOR DOE PUBLICATION Environmental Protection Support and Assistance General Approach
35
w
DOE/EP-0023 (ERDA-77-24REV)
"A GUIDE FOR ENVIRONMENTAL RADIOLOGICAL SURVEILLANCE AT U.S. DEPARTMENT OF ENERGY INSTALLATIONS,' JULY 1981
DOE/EP-
"A GUIDE FOR EFFLUENT RADIOLOGICAL MEASUREMENTS AT DOE INSTALLATIONS," SEPTEMBER 1982
DOE/EP-0038
"SUMMARY OF ANNUAL ENVIRONMENTAL REPORTS FOR CY 1980. DEPARTMENT OF ENERGY NUCLEAR SITES." AUGUST 1982
PNL-4410 (D.L STRENGE et al)
"ENVIRONMENTAL DOSE ASSESSMENT METHODS FOR NORMAL OPERATIONS AT DOE NUCLEAR SITES," SEPTEMBER 1982
Figure 1 •
Environmental Protection Support and Assistance-Documents
and meaningful as a public disclosure document. PNL-4410(4) documents one of several studies underway which will provide input for a planned major revision of the Environmental Guide. Figure 8 shows active tasks under this project at present. Task 3 provides assistance in the renewal of the Environmental Measurements Laboratory's quality assurance program. Task 6 has been through a number of changes of scope and objectives - i t is intended to assess the feasib i l i t y of establishing and implementing a program for tracking commitments made in EISs and RODs for the protection and enhancement of the environment. Tasks 5A, 8, 9, and 10 are understandable, I believe, but I would like to mention specifically Task 5B, Implementation of a New Reporting System. This is a continuing task which is intended eventually to address several modules as shown in Figure 9. The objective of this task is to develop specifications for a comprehensive environmental reporting system that will provide 00S with the current compliance and reporting status at DOE f a c i l i t i e s and that will present the information in a form that can be readily used in the decision making process. In FY 1982 a scoping document describing the conceptual design of the EDRS was prepared. An operational prototype for the Noncompliance Reporting module was completed and a data base developed that included data for CY 1981 and half of 1982. An example of a report from the prototype system is shown in Figure 10. Specifications for the module were written that described the data inputs and the reporting to be generated. Specification docunents for the NPDES and Environmental Profile modules were completed in FY 1982, however, they are not in final form since the base data for these modules are not yet available and the specifications may have to be modified when the actual data are compiled. The effort for FY 1983 is focused on developing the i n i t i a l data bases in computer readable form, establishing procedures for routinely updating the data base, and working with EG&G-Idaho on implementing the f i r s t three modules. Some effort will be devoted to starting on the next modules to be included in the system. The f i r s t of these will be the Pollution Abatement Record module. Specifications for the EDRS are being developed by PN! to f i t into the Department of Energy's centralized environmental, safety and health information system. This system, designated the Safety Performance Measurement
37
TASK NO. 1
«
2 3 5 5A 5B 6 8 9 10
TASK PROJECT MANAGEMENT AND OVERVIEW, INCLUDES MISCELLANEOUS REQUESTS AND INFORMATION RESOURCES ENVIRONMENTAL GUIDE REVISION (FY 82 - EFFLUENT GUIDE) QUALITY ASSURANCE IMPLEMENTATION ENVIRONMENTAL DATA REPORTING SYSTEMS ANNUAL DOE ENVIRONMENTAL SUMMARIES IMPLEMENTATION OF NEW REPORTING SYSTEM EIS FOLLOW-UP - IMPLEMENTATION GROUND WATER PROGRAMS EVALUATION AND RECOMMENDATIONS ENVIRONMENTAL PROTECTION INFORMATION MEETINGS PLANNING AND ARRANGEMENTS BACKGROUND DOCUMENTATION FOR DOE ORDER CHANGES Figure 8. Environmental Protection Support and Assistance Active Tasks-FY 1983
MODULE 1
MODULE 2
MODULE 3
NONCOMPLIAIMCE REPORTING
NPDES REPORTING
ENVIRONMENTAL PROFILE
MODULE 5
MODULE 6
r I
I
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I1
UNUSUAL
' OCCURRENCES
( |
I I ENVIRONMENTAL' RADIDATION J
MODULE 7 I EIS | I COMMITMENTS j I
,
i
Figure 9. Office of Operational Safety Environmental Data Reporting System
39
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40
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System (SPMS), i s being developed by the System Safety Development Center, EGSG-Idaho, asd w i l l contain several subsystems which are e i t h e r now operational or under development. The subsystems c u r r e n t l y envisioned a r e : Safety Performance Radiation Exposure Health Information Environmental Data
Data System (SPDS) I n f o r m a t i o n Reporting System (REIRS) System (HIS) Reporting System (EDRS)
The SPDS, REIRS and HIS are c u r r e n t l y being developed or ungraded by EG&G-Idaho. As the s p e c i f i c a t i o n f o r the various modules of the EPRS are completed, they w i l l be turned over to EG&G-Idaho f o r implementation. In c l o s i n g , may I stress t h a t the value o f our assistance t o 00S f r e q u e n t l y has depended most h e a v i l y on the exemplary cooperation of other DOE c o n t r a c t o r s and s t a f f . My sincere thanks t o a l l of y o u .
REFERENCES 1.
J . P. C o r l e y , e t a l . 1977. A Guide f o r Environmental Radiological S u r v e i l l a n c e a t ERDA I n s t a l l a t i o n s . ERDA-77-24. U.S. Energy Research and Development A d m i n i s t r a t i o n , Washington, D.C.
2.
J . P. C o r l e y , e t a l . 1981. A Guide f o r Environmental Radiological S u r v e i l l a n c e a t U.S. Department of Energy I n s t a l l a t i o n s . DOE/EP-0023. U.S. Department of Energy, Washington, D.C.
3.
O f f i c e o f Operational Safety. 1982. Summary of Annual Environmental Reports f o r CY 1980: Department of Energy Nuclear S i t e s . DOE/EP-0038. U.S. Department of Energy, Washington, D.C.
4.
D. L. Strenge, e t a l . 1982. Environmental Dose Assessment Methods f o r Normal Operations a t DOE Nuclear S i t e s . PNL-4410. U.S. Department of Energy, Washington, D.C.
SESSION TWO HAZARDOUS WASTE MANAGEMENT
(
2A
THE ARGONNE HIGH SULFUR DRY SCRUBBER
Paul S. Farber Energy and Environmental Systems Division Argonne National Laboratory Argonne, Illinois 60439 BACKGROUND The Argonne steam plant provides 200 psig steam throughout the entire 1,500-acre site for heating and, via evaporative chillers, cooling. The plant consists of five boilers installed over the period of 1948 through 1964. Boilers 1 through 4 each have a nameplate rating of 85,000 lb of steam/hr, while No. 5 can produce a maximum of 170,000 lb
April 1981 ground was broken for the construction of the pollusystem. By November 1982, construction was 95 percent complete and Niro moved into the startup and compliance testing phases of Compliance testing finally took place on February 22 through 1982.
SYSTEM DESCRIPTION The system installed on Argonne's No. 5 Boiler, to treat the flue gas, is a Niro Atomizer-Joy Manufacturing industrial design. This was field modified, during construction, based on Niro's findings at their Northern States Power-Riverside demonstration unit. The system consists of two parts, a wet end and a dry end. In the wet end, pebble lime is held in a 100-ton storage silo with a "live" cone bottom. From this bin, lime is fed through a Wallace and Tiernan weigh-belt feeder into the lime slaker. The weigh-belt feeder is equipped with both a feed-rate indicator as well as a totalizer which allows ANL to keep track of lime consumption. In the slaker, careful addition of potable water causes the calcium oxide (CaO) to react and form calcium hydroxide (Ca(OH)p), or milk of lime. The milk of lime, at about 15 percent solids, is passed through a rotary screen in order to remove the "grits," or inert particles, from the milk of lime. This milk of lime next flows by gravity to a covered, agitated, 45
storage tank. This storage tank, with a 30-minute hold time has two purposes: 1) to ensure completeness of the slaking reaction as well as to even out any inconsistency's in slaker operation, and 2) to ensure a temporary lime supply in case of slaker system failure. The milk of lime is next pumped to the slurry mix tank. In this agitated vessel, recycled waste powder, lime milk, and some dilution water are combined to form an approximately 40 percent (by weight) slurry. This slurry is transferred to the slurry feed tank via another rotary screen to ensure removal of any lumps which might clog the feed-slurry piping system. Feed slurry is pumped to the top of the spray dryer where it is injected into the rotary atomizer. In the spray dryer the slurry droplets contact the hot flue gases, and two events happen somewhat simultaneously; 1) the sulfur oxides react with the lime to form calcium sulfite and calcium sulfate, and 2) the droplets of water surrounding the lime particles evaporate thereby cooling the flue gas. The temperature of the gas exiting the spray dryer is carefully controlled to 18°F above the dew point. This temperature control is very important for several reasons including: 1) the necessity of protecting the baghouse from condensation; Z) the closer one can operate to the dew point, then the lower the stoichiometry for a required removal, and 3) preventing the wetting of the walls of the spray dryer. Some of the powder formed in the spray dryer drops out, and is picked up by a dra-link conveyor. The remainder of the powder, still entrained within the gas stream, enters the baghouse where they are removed by the fabric. Upon exiting the baghouse, the gas passes through a booster fan and then into the existing stack. Tables 1 and 2 list some of the parameters related to the spray dryer and fabric filter. TABLE 1. Spray Dryer Parameters
Niro Atomizer Incorporated 10-Second Residence Time 25' Diameter, 19' straight side Rotary Atomizer--14,000 RPM Dual Gas Inlet—Roof and Central Gas Dispersors Carbon Steel Construction
46
TABLE 2. Fabric Filter Parameters Joy Manufacturing Company Four-compartment Pulse Jet 3.03:1 Air-to-Cloth Ratio 280 Bags/Compartment—6" Diameter—12' Long Huyck Fiber Glass 5,278 ft 2 i-ilter Area/Compartment SYSTEM STARTUP Argonne and Niro commenced starting up the system in November 1981. While it can be said that the startup proceeded fairly smoothly,like all startups, there were problems encountered. The first problem encountered was with the control damper for the booster fan. Cold gas flow tests repeatedly resulted in the damper closing off more than was indicated on the controller. After some examination, we found that the actuator supplied by the fan vendor was undersized and, the fan was literally sucking the damper closed. The situation was resolved after a larger actuator was installed, and the damper linkage modified to aid in operation. The pumps used for transferring milk of lime and slurry are rubberlined centrifugals. Two pairs of these, the lime milk and slurry mix-pumps are low head and are equipped with packing and seal water. The slurry transfer pumps which send the slurry from the recycle building to the head tank in the spray dryer penthouse, are high head pumps and, were equipped with centrifugal expellor seals. Soon after startup the seals on both slurry transfer pumps failed and had to be replaced. In one sense, this failure was to Argonne's advantage since, the decision was made to remove the expellors and replace them with packing and seal water. This resulted in Argonne's having six pumps (three sets of two) which are identical excepting for the motors and belt drives. The startup problems encountered with the pumps, were also accompanied by the slurry mix tank agitator continually tripping out electrically. It was determined that, as Argonne and Niro built up the solids concentration in the mix tank, the power draw on the agitator increased. Unfortunately, the motor and gear drive were not sufficiently large enough to adequately power the agitator as the density increased. In order to build up the density of the slurry being sprayed it was necessary, during startup, to spray milk of lime directly through the agitator and build up powder in the recycle silo. As this powder built up, part of it
47
was then added to the slurry mix tank, and the percent solids to the agitator increased. At the beginning of the start-up, the fluidizing blower and pads in the recycle silo did not present any operational problems. As the powder level in the recycle silo built up, then the outlet pressure from the blower also increased. This was to be expected since, for a positive displacement blower, the outlet pressure is directly proportional to the system resistance at a constant volumetric flow. After a time, it was discovered that the blower motor was drawing too much power and it started to repeatedly trip out. Initially, it was thought that the motor bearings had gone bad and, the blower supplier replaced the motor. When that did not prove successful, the supplier came back in and rebuilt the blower bearings. Even after all of this work had been accomplished an excessive amperage draw was still evident. It was then reluctantly decided to empty the powder from the recycle silo, and to inspect the fluidizing pads for evidence of plugging. No sign of the pads being plugged with hardened powder could be found, and the sheaves on the motor and blower were changed to reduce blower RPM. To date, this reduction in blower speed (and hence air volume) has not affected the fluidization of powder in the silo, and the blower motor has not overloaded. SYSTEM PERFORMANCE The agreement between Niro Atomizer, Inc. and Argonne calls for guarantees to be satisfied based on: 1. Compliance and Initial Performance Tests 2. Sixty-Day, Reliability Test 3. Final Performance Test. The performance tests, which include both inlet and outlet particulate and sulfur dioxide levels, were to be carried out by an independent third party. Based on a list of approved testing companies, supplied by the State of Illinois, a competitive procurement resulted in the selection of Clean Air Engineering to assist in performance testing. The initial tests took place between February 22 and February 27, 1982 and utilized EPA Methods 1 through 6 for determination of emissions. In order to ensure that emissions tests system performance, and boiler operation were accurate, and could be related to each other, coal samples were taken for analysis. Separate samples were taken over each day of testing from the two coal scales feeding Boiler No. 5. Subjected to an ultimate analysis, the average of all samples (with their standard deviations) are shown on Table 3. As can be seen, the coal burned during the tests was of reasonably constant composition and is representative of the coal that Argonne is burning today. All tests for performance were performed at three boiler loads, 35 percent, 70 percent, and 82 percent of Maximum Continuous Rating (MCR). One point which, at this time, requires some explanation is the justification for the maximum load testing at 82 percent of MCR rather than at 100 percent of MCR. Argonne's No. 5 boiler is our base-loaded unit and is normally operated at 120,000 lbs steam/hr (70 percent MCR). Peaking loads are accommodated through use of Boilers 1 through 4, which are gas fired. Over
TABLE 3. Boiler No. 5—Coal Compositions (Clean Air Engineering 1982) Component
Mean
Standard Deviation
Moisture
11.34%
0.95
C
64.39%
0.76
H
4.72%
0.12
N
1.41%
0.03
S
3.02%
0.35
Ash
7,97%
0.44
Cl
0.07%
0.03
0
7.08%
0.31
Btu/lb
11,799
173
Volatiles
36.28%
1.72
Fixed Carbon
44.35%
1.15
the last five years the Laboratory has engaged in a massive energy conservation program which has reduced our maximum loads (summer and winter) to about 220,000/lbs steam/hr. In fact, on January 10,1982 which was the coldest day recorded in Chicago's history (-27±F with a -81°F windchill), the total steam load at Argonne was 230,000 lbs steam/hr, with Boiler No. 5 producing 136,000 Ibs/hr. Since it was therefore evident that our coal-fired unit would probably never be operated higher than this, ANL decided that the operating permit application could be based on 140,000 lbs steam/hr (82 percent MCR). After some discussion, with Niro, it was agreed that system performance, and utilities consumption, would also be based on 82 percent MCR rather than the 100 percent MCR in the contract. During the tests it became apparant that, while the system was performing fairly well, an operational problem was becoming apparent. Somewhere in the slurry mix tank and/or feed tank loops a water leak was in existence. Although not evident at high boiler rates, as tests were conducted at lower rates, a rise in outlet S0 2 emissions was seen (Table 4 ) . Examination of the operating data recorded during emissions testing has given a clue to what is the source of the problem. Figure 1 shows the variation in flowrates of slurry feed to the atomizer, and milk of lime to the slurry mix-tank. As would be expected the slurry feed rate is linear with respect to the boiler load. 49
TABLE 4. Sulfur Oxide Levels As a Function of Boiler Load (Clean Air Engineering 1982) S02-lb/10° Btu Boiler Load (% of MCR) 35
Inlet 6.69
Outlet 1.36
Removal 79.67
70
5.93
0.60
89.88
82
6.13
0.27
95.60
0
20
40
CO
00
100
120
140
160
Steam Rate 1000 Lbs / Hr FIGURE 1. Flows as a Function of Boiler Load Since we were running approximately the same approach temperature (23UF) to dewpoint at all loads, the amount of water needed to cool the gas is proportional to boiler load. At the constant density of feed slurry, it is therefore reasonable to expect this linearity of slurry flow with steam production rate, it is also reasonable to expect that, for the same percentage removal, the stoichiometry should be relatively constant (except for the effect of changes in residence time) over boiler load rate. Therefore, the lime feed rate should also be linear with boiler load. Unfortunately, as can be seen in Figure 1, this was not the situation during our testing. By the tail-off in milk of lime feed rate we concluded that: 1) there must be a water leak into the slurry mix tank, and b) the magnitude of the leak was about 4 gpm. A short time after the compliance testing, it was found that the dilution water flow meter had jammed in such a manner as to send the controller a zero flow signal even through the valve was 25 percent open. 50
TABLE 5. High Sulfur Dry Scrubber Waste Analysis Moisture
2.4%
A1 2 O 3
1.15%
Fe
2.12%
2°3
SiO 2
2.45%
MgO
0.69%
Na 2 0
0.10%
CaS0 3
58.8%
CaS0 4
12.8%
Ca (OH) 2 [as CaOJ
6.62% 12.9%
L.O.I.
Based on an analysis of the wastes^ ' (Table 5) and the data gathered during the performance a determination of stoichiometries can be made. First, let us define two different stoichiometries, an external and an internal one:
Extern*, sto1cM«try -flSff internal stofchi-rtry - M ° "
s
°f
It is important to know the external stoichiometry to determine the amount of sorbent to be purchased, and the internal to understand the extent of recycle and sorbent utilization. Mgure 2 shows both stoichiometries calculated during the initial performance tests at Argonne. The initial findings seem to indicate (as one would expect) that, as complete removal is approached, the system approaches a "straight-through" operation with no recycle. Results from the particulates tests are, of course, very satisfactory (Table 6 ) . The pulse jet unit supplied by Joy Manufacturing has a design airto-cloth (A/C) ratio of 3.03:1 at 100 percent MCR. Since all tests were performed below the boilers' nameplate rating, and hence at low A/C ratio's, it was expected that particulate emissions would be low. Figure 3 shows the system pressure drop as well as the drop across the filter bags themselves. The decrease in grain loading with increased boiler load is probably due to a (a) Personal communication to P.S. Farber on I. V. Conversion Systems, March 1982. 3 JL
83C304-
"3 0388G8C4028078
0.5
1
US
Z
2.5
Stoichiometry FIGURE Z. Percent Removal as a Function of Stoichiometry thicker cake on the bags themselves, although some further analysis is needed to check this conclusion. Another fact, apparant from Figure 3, is that the pressure drops across the filter bags are low and need to be raised to about 4" H O to ensure that fine particles do not build up and (at a later date) cause inordinately high pressure drops. Argonne is working with Niro and Joy Manufacturing to correct this situation. TABLE 6. Particulate Emissions During Performance Testing (Clean Air Engineering 1982) Filter Outlet
Boiler Load (% of MCR)
SDA Inlet (gr/DSCF)
(gr/DSCF)
35
0.0717
0.0057
0.0161
70
0.1037
0.0039
0.0103
82
0.1671
0.0021
0.0052
52
(lb/10D Btu)
1412U 103-
m
8-
I* S0
SO
40
60
SO
100 ISO 140 100
Steam Rate 1000 Lbs / Hr FIGURE 3. Pressure Drop as a Function of Boiler Load CONCLUSIONS Based on the results of the initial performance tests Argonne is pleased with the operation of the system. A 60-day reliability test and a second series of performance tests are planned in the near future. Argonne will be working with Niro in order to "fine tune" the system, and bring operation, reagent, and utility consumptions more in line with contract guarantees. To date, no serious problems have been encountered and, ANL feels that the dry scrubbing system we have installed is an economical and efficient control technique for high sulfur coal combustion. REFERENCES Clean Air Engineering. 1982. "Performance Test Report to Argonne National Laboratory." Farber, P. S. 1981. "Spray Dryer and Fabric Filter Installation—An Industrial Boiler Conversion," Paper 81-9.6 74th Annual Meeting APCA. Philadelphia, Pennsylvania.
2B
TRAPPING OF GASEOUS FLUORIDE EMISSIONS ON SOLID SORBENTS* N. F. Windt Technical Services Division Paducah Gaseous Diffusion Plant ABSTRACT
Topic Area: Environmental Controls Title: Trapping of Gaseous Fluoride Emissions on Solid Sorbents Author: N. F. Windt, Technical Services Division UCCND/Paducah A variety of gaseous fluorides and solid sorbents have been tested in lab and pilot plant trapping experiments. Most of these gases are toxic and their release to the environment must be controlled by proper disposal methods. While several possibilities for gas removal systems exist, solid chemical traps represent a technology which has been successfully applied to control enrichment industry emissions from the environmental control, health and safety, operational and economic standpoints. Recent advances in this technology have been made in the areas of trap monitoring, new sorbent systems, improved operational schemes, and computer modeling in trap design. Full scale oolitic calcium carbonate (CaC03 ) traps have recently been installed on a new smelting process to remove potential hydrogen fluoride (HF) emissions from the process. Calcium carbonate has been shown to be a selective sorbent for removal of certain fluorides present in gas mixtures. Improved sorbents are also showing favorable "zero" discharge operating times for technetium (Tc-99) and uranium removal from enrichment plant gas streams at concentration levels in the low ppm range. The format for this paper is to briefly review the trapping of gaseous fluorides in enrichment industry vent streams. A fundamental theories section will briefly point out the complexities of gas-solid interactions, desirable sorbent properties, the selection process for new sorbents, and engineering design requirements. The major discussions will focus on several recent improvements and new applications. These are the testing of four different uranium hexafluoride sorbents, sorbents for fluorine (F 2 ), chlorine trifluoride (C1F 3 ), and hydrogen fluoride (HF) and improved sorbents for low-level radioactive gases. The paper will briefly conclude with future studies at the three Enrichment Plants. *Based on work performed at Paducah Gaseous Diffusion Plant operated for the U.S. Department of Energy under contract W-7405-eng-26 with Union Carbide Corporation.
55
INTRODUCTION In the enrichment of gaseous uranium hexafluoride (UF^), the consideration of purity was recognized early as a parameter of utmost concern. In the early years of gaseous diffusion plant operations, the primary concerns were with the purity of feed material, product material, and minimizing the airborne release of uranium/uranium compounds. The purity of feed material was assured through the establishment of specifications which the UF6 feed manufacturers had to meet. Vendors, to a limited extent, utilized solid sorbents to remove impurities and produce feed material which met the specifications. The latter two concerns, product purity and airborne uranium losses, have, since the inception of the diffusion plants, primarily been resolved through the exclusive use of solid sorbents in a variety of trap designs. Through the years, changing environmental regulations, health and safety requirements, and the operation of newly improved cascades have demanded continued improvement and development of solid sorbent trapping systems. A variety of gaseous fluorides, in addition to UF^, may be present in gaseous diffusion plant processes. However, since most are toxic, their release to the environment must be controlled by proper disposal methods. Some potential feed stream impurities are detrimental and even extremely low concentrations can cause serious nonrecoverable operating inefficiencies. Others can migrate to the top cascade and contaminate the diffusion plant product. Some radioactive impurities can also present health hazards to workers during maintenance activities. Some are capable of forming potentially explosive mixtures. Therefore, to meet these operational, environmental, and health and safety requirements, it is often necessary to decrease all fluoride emissions from an atmospheric vent line or to selectively remove a specific fluoride gas from a gas mixture. While several possibilities for gas removal systems exist, solid chemical traps represent a technology which has significant impact on the enrichment industry from the environmental control, health and safety, operational, and economic standpoints. Recent advances in the technology for operation of existing trapping systems have been implemented. A computer model has been developed from extensive lab experiments to aid in trap design. Advances in technology for new sorbent systems have also been successful. As a result of recent lab and pilot plant trapping studies, a new full-size trapping system is near completion in Paducah's C-746 smelter facility. These oolitic calcium carbonate (CaCO3) traps will trap some 100,000 pounds of hydrogen fluoride (HF) that will result from the PEM scrap recovery program, thereby meeting environmental and operational requirements. A patent is pending this year based on recent studies in which oolitic CaC03 was utilized as a selective sorbent for removal of active fluorides such as C1F3 in the presence of UF 6 . Calcium carbonate and soda-lime are showing
56
/
promise at Portsmouth as improved trapping agents for technetium (Tc 99 ) and uranium in vent streams and will be considered for applications such as feed stream and product purification. A continued technical effort is being made in the areas of improved trap operation and design, and identification of new trapping materials. Also, this technology has spin-off applications to the other advanced enrichment technologies. The emphasis on future direction of this technical effort will be influenced by operational, environmental, and health and safety factors, as well as the potential for feeding reactor return UF 6 . This ability to develop, improve, and/or modify trapping technology has provided timely and cost-effective solutions to many potentially serious problems. Three-plant consolidation of these efforts by the technical management teams provides continued effective capability at a level consistent with current needs. FUNDAMENTAL THEORIES The scope of this paper deals with dry, two-phase (gas-solid) chemical traps, which involve the processes of adsorption, absorption, chemisorption, chemical reaction, and molecular screening as shown in Figure 1. UNCLASSIFIED
I
SELECTION OF "BEST" SORBENTS IS COMPLEX! LARGE & SMALL SORBATE MOLECULES .ADSORPTION, BUT IF \ ^ A CHEMISORPTION d | DIFFUSION PENETRATES SQRBENT JflPl J\ 7 THEM ApcriDDTinKl ^
^B
«
»\
^~
AREA AVAILABLF TO -> BOTHSORBATF. MOLECULFS
AREA AVAILABLE TO SMALLER I SORBATE
CHEMICAL PRODUCT
MOI F r i l l AR SCREENING
MCLECULES
NKMSSIHU
Figure 1. Gas-Solid Processes
57
It should be pointed out that in many of the trapping systems presented in this paper more than one of the processes may be occurring at the same time and to varying degrees. It is beyond the scope of this paper to discuss in detail all these processes. However, a somewhat general discussion of fundamental theories is in order to point out the complexity of gaseous interactions with solids and solid surfaces. It is already well established that the molecular forces at the surfaces of a solid are in a state of unbalance or unsaturation. That is, the molecules or ions on the surface may not have all their forces satisfied by union with other particles. As a result of this unsaturation, solid surfaces tend to satisfy their residual forces by attracting onto and retaining on their surfaces gases with which they come in contact. This phenomenon of concentration of a substance on the surface of a solid is called adsorption. The substance thus attracted to a surface is said to be the adsorbed phase or adsorbate while the substance to which it is attached is the adsorbent. Adsorption should be carefully distinguished from absorption, the latter process being characterized by a substance not only being retained on a surface, but also passing through the surface to become distributed throughout the solid phase. Chemisorption, or activated adsorption, on the other hand, is the result of chemical interaction between the solid and the adsorbed substance. The strengths of chemical bonds may vary considerably, and identifiable chemical compounds in the usual sense may not form. Neverthe-less, the adhesive force is generally much greater than that found in physical adsorption. Chemical reactions can also occur with the formation of distinct solid reaction products. Before a given gas-solid reaction can be considered economically feasible, both the thermodynamics and kinetics must be favorable. A concept which becomes especially important in determining sorbent capacity is that of "available" surface, that is, surface area accessible to the sorbate molecule. It is apparent from pore size distribution data that the major contribution to surface area is located in pores of molecular dimensions. In addition to sorbent capacity or loading factors, other items must be considered in the selection of a sorbent system. Pressure drop across the sorbent bed is an important design parameter. Affecting this parameter is sorbent or pellet integrity because, as a result of regeneration cycles, the sorbent can deteriorate and restrict the flow, operability, and efficiency of the trap. Reaction kinetics, the rate at which the sorbent can load the sorbate, is important in sorbent selection and dictates superficial gas velocity limits, length of sorption zone, minimum bed length required, and operating temperature and pressure. Thermodynamics
58
)
can be used to predict the extent of chemical reaction between the sorbent and sorbate and also such things as maximum theoretical temperatures that might be obtained in the trap. A tabulation of sorbent properties that must be considered when selecting the "best" sorbent for a particular application is shown in Figure 2. In the next section on recent advances in trapping technology, these properties of sorbents will be illustrated in more detail on specific systems with actual data from the experiments.
UKCIASSIFIEB
I
SELECTION OF "BEST" SORBENTS IS COMPLEX! SORBENT PROPERTIES PARTICLE SIZE PORE DIAMETER SURFACE AREA DENSITV PORE VOLUME SPECIFIC HEAT
PROCESSES ADSORPTION CHEMICAL REACTION ABSORPTION MOLECULAR SCREENING CHEMISOHPTION DIFFUSION
PELLET STABILITY SYNEHGJS7JC EFFECTS REGENERATION CAPABILITY RECOVERY CAPAIIILITV
OPERATING PARAMETERS PRESSURE TEMPERATURE LOADING PROFILE
CQSI CAPITAL OPERATING DISPOSAL
UKCUSSIflfS
Figure 2. Sorbent Selection Factors Testing of Four Different Uranium Hexafluoride Sorbents The effectiveness of the chemical trap largely depends upon the particular sorbent material being used to fill the trap. The use of the "best" UF6 sorbent is clearly demanded by those applications where the chemical trap is the final barrier between the uranium process and the environment. In other situations, proper sorbent selection is simply a matter of good engineering practice. In the past, decisions were made largely on the basis of prior experience or convenience because adequate
59
laboratory and supportive engineering design data simply were not available. Unfortunately, the choice of the preferred sorbent is not intuitively obvious but has to be made on a side-by-side comparison of several performance factors. Pressure losses, sorbent capacity, reaction kinetics, sorbent regeneration/uranium recovery requirements, and the effects of other system components such as HF and F2 are the major parameters upon which the selection should be made. Not all of the performance factors are equally important in every application and one or two might be weighed more heavily in the overall comparison. Clearly, sorbent selection has to be made on a case-by-case basis. Comprehensive laboratory experiments and theoretical modeling studies were recently conducted on the commonly used chemical sorbents, sodium fluoride, high silica alumina, low silica alumina, and chip type alumina to provide a scientific basis for sorbent selection. Performance information is summarized along with a relative ranking of the particular sorbent in each comparison area as shown in Table 1. TABLE 1. Overall Ranking of Chemical Trap Sorbents
NcF
Chemical Sorbent High Silica Low Silica Alumina Alumina
Chip-Type Alumina
3
4
4
2
Uranium Loading Capacity
4
3
2
1
Reaction Kinetics
4
CVI
3
-
Regeneration/ Recovery Capability
4
1
2
1
HF Effect
2
3
3
3
15
16
Effect of Fluorinating Agents
_4
TOTAL:
21
CVI
Pressure Drop
9
NOTE: A ranking of 4 is the best possible. The most startling difference between NaF and the aluminas is the rate at which the individual sorbents can load UF 6 . As can be seen in Figure 3, which depicts a typical single pellet layer loading comparison between NaF and high silica alumina, the NaF trapping mechanism is substantially faster than the high silica alumina. With comparable UF6
60
pressure <50 torr, the NaF pellets attained a 40 weight percent increase within 1 hour, w..ile the alumina required over 50 hours to reach this same level. Chip-type alumina trapping rates were found to be even less than the high silica alumina. These laboratory data corroborate observations at Portsmouth's vent stream trapping of uranium by NaF and chip-type alumina in comparative testing. UNCLASSIFIED
COMPARISON OF URANIUM LOADING PROFILES FOR NaF & HIGH SILICA ALUMINA" 60
NaF
o50 CO
CONDITIONS
§40 <
<50torrUF6 PRESSURE 88°F !30°C)
!30 ALUMINA
3
4 5 6 TIME, HOURS
7
UNCLASSIFIED
Figure 3. Comparison of Uranium Loading for NaF and High Silica Alumina A dynamic adsorption model was developed assuming radial symmetry, negligible temperature gradient, and homogeneity of physical and chemical pellet properties. Sufficient experimental data are now available and a solid theoretical understanding of the UF6 loading mechanism has been developed for the more widely used sorbent materials to enable the engineer to design chemical traps with considerable confidence, Sorbents for Fluorine, C1F;,, and HF Scoping studies of solid sorbents and reaction conditions have been conducted. In a series of statistically designed experiments, the effects of total gas flow, temperature, and fluoride gas flow were studied for eight dry chemicals in a 1-inch diameter fixed bed designed to remove
61
three separate fluoride gas species (HF, fluorine, and C1F 3 ) from an air stream. The trapping materials selected for the test include two brands of four different material types: soda lime, CaC03 , CaO, and activated alumina. Statistical evaluation of the data has identified total gas flow as the most significant variable while bed temperature was the second most significant variable. Fluoride gas flow was not a significant variable in the experiment. The oolitic CaC03 was identified as the best material selection based on cost and performance. The 48-run experiment developed for the study is a replicate 2 3 factorial design, one for each gas type, in terms of the three process variables: total flow, fluoride gas flow, and initial bed temperature. The eight dry chemicals and the two levels of each of the three process variables are listed in Table 2. TABLE 2. The Levels of the Controlled Experimental Variables Controlled Variables A.
Level
Gas Type 2
2. 3. 1. 2.
B. Gas Flow Rate
C1F3 HF 100 seem 200 seem
C.
Total Flow Velocity
1. 800 seem 2. 8000 seem
D.
Reactor Temperature
1. 200°F 2. 600°F
E.
Dry Chemical
1. 2. 3. 4. 5. 6. 7. 8.
A1 2 O 3 , ( a c t i v a t e d ) sample 1 A1 2 O 3 , ( a c t i v a t e d ) sample 2 Soda Lime, sample 1 Soda Lime, sample 2 CaCO3 , ( o o l i t i c ) sample 1 CaC03 , ( o o l i t i c ) sample 2 CaO, sample 1 CaO, sample 2
A complete factorial design that used all combinations of the dry chemicals and process levels would require 58 experimental runs for each fluoride gas and would require an excessive amount of time to complete.
62
Therefore, an experiment was designed to examine the main effects of the dry chemicals and the main or linear effects of the process variables while sacrificing information on the two-, three-, and four-factor interactions among the dry chemicals and process variables. In the statistical evaluation, several responses were considered; however, the percent fluoride loading factor at breakthrough was chosen as the best representation of performance. From this evaluation, the optimum operating conditions are found to be the low level of total flow (800 seem) and the high level of temperature (600°F) with flow being the most significant variable. Fluoride flow was found not to be a significant variable in the experiment. The fluoride load factor responses for each gas were fitted to a linear equation where: L0AD=MEAN + DRY CHEMICAL EFFECT + TOTAL FLOW + TEMPERATURE + RANDOM ERROR. The magnitude of calculated intercepts was used to rank the dry chemicals. Based on this evaluation, activated A1 2 O 3 appears best for F2 and ClF3 removal and soda-lime is best for HF removal. From an operations standpoint, however, the increased plugging which occurred in the soda-lime and CaO runs with HF at 200°F would discredit these materials as the best selection without further development to circumvent these problems. In addition, the HF runs did not show a statistical difference in the bed materials at the 10% significance level. Activated A1 2 O 3 operated at 600°F and a superficial gas velocity of 0.2 ft/sec gave the highest loading factors of all materials and for all three fluoride gases. However, other considerations, such as material cost, should be taken into account when selecting the optimum trapping material and the statistical evaluation should not be used exclusively for optimum material selection. Table 3 compares trapping costs of the four material types using the maximum loading factor obtained and cost information available at the time of this report. It ^s apparent that CaC03 has the best cost benefit even with a somewhat lower loading factor. As a result of these studies, traps have been fabricated to utilize CaCO3 in a fixed bed removal system to control HF emissions from the C-746 Scrap Pretreatment Facility. A New Selective Sorbent Following the completion of the factorially designed experiments, additional scoping tests were initiated to determine the efficiency of these materials in trapping UF 6 . These studies were instrumental in identifying oolitic CaCO3 as a selective sorber of active fluorides in the presence of UF6 and have led to mixed gas trapping studies involving C1F3 and UF 6 . Laboratory runs were made with a 1-inch OD x 12-inch long
63
bed. Total fluoride loading factors, obtained from bed analyses following 100% C1F3 breakthroughs, ranged from 22% to 32% with only 1.8% to 3.0% (% of total bed weight) uranium retention. TABLE 3. Cost Comparison of Fluoride Trapping Materials Material Cost,
Material Activated A1 2 O 3 CaO Soda Lime CaC0 3
$/lb Material
Maximum Fluoride Loading Factor, %
0.23 0.014 0.497 0.005
55 41 37 32
Material Trapping Cost $/lb F 0.42 0.03 1.34 0.02
The use of the limestone as a selective sorber of active fluorides in the presence of UF 6 is also being considered for potential use at Paducah. A New Application in C-746 The C-746 pretreatment facility at Paducah will make use of diluted high-temperature steam to pretreat approximately 20 million pounds of contaminated nickel scrap from the three gaseous diffusion plant sites. The purpose of pretreatment is to remove the fluoride from the metal scrap. If melted with the fluoride present, the scrap would give off fluoride fumes which would shorten the liner life and pose health and environmental problems. The hydrogen fluoride formed by the reaction of steam and metal scrap will be constantly purged out oF the calciner by the incoming nitrogen which also dilutes the steam to 20% by volume. Fixed bed traps utilizing calcium carbonate will be used to remove HF from the gas stream leaving the calciner in the following manner: CaC0 2 + 2HF * CaF2 + H 2 0 + C0 2 200° to 300°F The only gases going to atmosphere are H 2 0, COg, and nitrogen as shown in Figure 4.
64
C-746 PRETREATMENT FACILITY FOR PEM SCRAP
UNCLASSIFIED
H2O+CO2VENT
2HF+CdC(J3— (200-300° P)
18 MILLION lbs
INDUCTION
Ni F2+H2O-»-Ni O+2HF (I100OF)
niMH*
F U R N A C E — INGOTS UNCLASSIFIED
Figure 4. C-746 Pretreatment Facility for PEM Scrap Improved Sorbents for Radioactive Gases Radioactive discharges from plant vent streams are regulated in accordance with DOE MC 0524 guidelines. Emissions at Portsmouth are in compliance with the present standards specifying radiation limits. Recent regulatory trends, however, have been toward more restrictive limits. This fact, coupled with proposed increases in metabolic uptake factors for technetium (Tc^9),could mean more stringent controls for radioactive emissions in the future. Evaluation of various fixed bed trapping agents for reduction of T c " emissions from the top purge vent stream of the Portsmouth plant has been in progress since 1979. Various potential trapping agents have been tested in parallel with activated A1 2 O 3 , the material that has long been used to reduce radioactive emissions in the vent streams. Tests have identified several alternate trapping agents with the potential for improved radioactive effluent control. Several grades of limestone (CaC03) and soda-lime (Na2O-CaO) were particularly effective in terms of "zero" discharge T c " operating times when compared with 21 other solid sorbents similarly evaluated. Figure 5 shows the relative effectiveness of some sorbents.
65
Tc-99 REMOVAL FROM TOP PURGE VENT STREAMS
UNCLASSIFIED
6-16 MESH K LIMESTONE (C
100 ' 90
«.
|
—-y-^r
J.HARD~]—-—.
(iooo F
° »i SINTERED
< < 80 - SOFT I4OO°F) ^- NaF" ACTIVATED NaF O 70 • 1/8"-1/4 INCH— LU PELLETS
w 40 o < 30
I ~*-^J
t"-^d"
No. 9 K. AND • p y \ ' ' No. 8 WATERLOO N j \ \ 4 | 8 M E ' S H -LIMESTONES 1 "">«T™« • \ — N - \ SODA-LIME (CaCO. \ \ i (Na,O-CaO) !SH - \ i i 1/4 INCH-10MESH "AS RECEIVED" "ACTIVATED ALUMINA (AI2O3)
LU
> 20 10 10
1b 20 25 DAYS OF OPERATION
30
35 UNCLASSIFIED
Figure 5. T c " Removal from Top Purge Vent Streams Three-Site Gaseous Diffusion Plant Consolidation Planning A three-site Technical Program Management Team on Materials and Chemistry regularly meets to review and set priorities on chemical trapping studies. Mairitaining open lines of communication among the three sites to avoid duplication of effort is a valuable function of the Team. Some of the consolidation actions already taken by the Team, as well as site contacts, are shown below: * GAT--Laboratory Studies--A. J. Saraceno * PGOP—Pilot Plant 8 Full-Scale Testing--N. F. Windt * ORGDP--Monitoring & Computer Modelling--C. G. Jones SUMMARY A considerable amount of technology exists in the Enrichment Plants on fluoride trapping. However, necessary improvements continue to be made, and a few examples have been presented in this paper. The technology has spin-off applications to other areas and has already been applied to some of the advanced isotope separations systems. A modest technical effort will be maintained at the Enrichment Plants, and the three-site Management Team will control the effort at a level consistent with needs.
66
REFERENCES Buonicor, A. J., and L. Theordore. 1975. Industrial Control Equipment for Gaseous Pollution, Vol. I and II. CRC Press Cleveland. 141-143. Otey, M. G. 1980. Removal of Gaseous Fluorides (KY/L-1070). Union Carbide Corporation, Nuclear Division, Paducah Gaseous Diffusion Plant, Paducah, Kentucky. Otey, M. G., and C. K. Bayne. 1980 Fixed Bed Trapping for Gaseous Fluoride Effluent Control (KY-705). Union Carbide Corporation, Nuclear Division, Paducah Gaseous Diffusion Plant, Paducah, Kentucky. Otey, M. G., and H. A, Lowery. 1978. Personal Communications. Union Carbide Corporation, Nuclear Division, Paducah Gaseous Diffusion Plant, Paducah, Kentucky. Saraceno, A. J. 1981. Personal Communications. Corporation, Portsomouth, Ohio.
Goodyear Atomic
Schultz, R. M., et al. 1981. Sorbent Selection and Design Considerations for Uranium Trapping (K/ET-5025). Union Carbide Corporation, Nuclear Division, Oak Ridge Gaseous Diffusion Plant, Oak Ridge, Tennessee. Schultz, R. M., and M. J. Stephenson. 1980. Purge Cascade Uranium Monitoring (K/ET-5015). Union Carbide Corporation, Nuclear Division, Oak Ridge Gaseous Diffusion Plant, Oak Ridge, Tennessee.
67
2C FERMILAB'S APPROACH TO PCBS
Samuel I. Baker Fermi National Accelerator Laboratory Batavia, Illinois
ABSTRACT A large number of transformers and capacitors are used in the operation of a high energy accelerator facility such as Fermilab's. Many transformers are contaminated with polychlorinated biphenyls (PCBs) and most of the capacitors are filled with askarel which typically contains 60-70% PCB. Fermilab's approach has been to first measure the PCB concentrations in all (approximately 200) large oil-filled transformers and inventory all the transformers and capacitors containing PCB. Next, a spill plan with installation of appropriate containment was developed. Finally, a program of PCB removal was begun. Approximately 40% of the oil-filled transformers contained PCBs in excess of 50 parts per million. Fourteen transformers have been processed at Fermi lab to reduce PCB concentrations, including six of 11 askarel-filled transformers. One of the 11 askarel-filled transformers was shipped off the site for disposal. PCB capacitors have been removed from normally occupied areas where replacement capacitors were readily available. Storage of PCB oil has been eliminated and storage of PCB capacitors reduced to those spares currently needed for maintaining operations.
INTRODUCTION Fermi National Accelerator Laboratory (Fermilab) utilizes accelerators to produce the highest energy proton beams available for basic research. A large number of transformers and capacitors are required to power the accelerators and balance the inductive loads associated with the more than 1000 magnets. In addition, there are conventional utilities transformers which power the areas where experiments and support functions are carried out. Many of these transformers and most of the capacitors contain polychlorinated biphenyl (PCB), the major component of a dielectric fluid called "askarel" which has very desirable fire protection properties,
69
but relatively recently has been found to be hazardous to humans and the environment.1 This paper will describe Fermilab's approach to reducing the hazards from PCBs. Transformer-grade askarels are usually mixtures of PCB and trichlorobenzene. Askarel is very stable and nonflammable. It is persistent, like DDT, when released to the environment. Dioxin (specifically TCDD) and furan (specifically TCDF) have been found in harmful concentrations in a fire which overheated an askarel-filled transformer inside an office building. PCBs are suspected carcinogens, can cause chloracne and liver damage, and can have adverse effects on unborn children and nursing children of exposed mothers. Fermilab's approach to PCBs has been first to measure the concentrations in oil-filled transformers and inventory all the transformers and capacitors having PCBs. Next, a spill prevention, control, and countermeasures plan (SPCC plan) was developed. Finally, a program of PCB reduction was begun. The results of each of these steps will be described below. PCB INVENTORY AND TEST RESULTS Fermilab has approximately 2000 large PCB capacitors and 200 large transformers. Most of the transformers are oil-filled. However, it was known from efforts to purify the oil in the 65 Main Ring pulsed-power transformers by filtration that most of these contained between 1% and 5% PCB. These transformers were filled with askarel for testing at the factory before the PCB hazards were known. Then they were drained and filled with oil for use at Fermilab. Because of the large changes in load associated with pulsed-power operation, these transformers are much more prone to failure than conventional transformers. They are relatively large 5200 I (1375 gal) transformers which the Laboratory has been getting repaired when they fail. The cost and time required to repair are half that required to obtain a new unit. Thus, reduction of PCBs in these transformers is motivated by the desire to repair them when they fail. Fermilab purchased 11 askarel-filled transformers, primarily for use in locations where fire protection was a major consideration. All these were on outside concrete pads, so airborne effluent problems are not so severe as for inside installations. However, containment of leaks was an important consideration. Most of the 2000 capacitors are located either outside in a 732 capacitor "tree" (elevated platform) or inside one of the accelerators (locked and, when in operation, interlocked). There were PCB capacitors inside the service buildings which housed the Main Ring pulsed-power supplies. These failed occasionally by vaporizing the dielectric. This presented a hazard for the personnel investigating the failures. They have since been removed (see below).
70
Samples taken from Fermilab's 198 large oil-filled transformers gave the following PCB concentrations: PCB Concentration (parts per million)
Number of Transformers
10,000-60,000
53
500-10,000
14
50-500
8
The result of 3. % contaminated above 50 ppm is in agreement with U. S. Environmental Protection Agency( EPA) findings;3 however, the number above 500 ppm is much greater than anticipated. SPILL CONTROL PLANS AND EXPERIENCE Following the identification and inventory of transformers and capacitors having PCBs, a spill control plan (SPCC plan) was written and an inspection program instituted. Fermilab visually inspects all PCB transformers and capacitors quarterly with askarel-filled transformers receiving a monthly inspection. Containment curbs were provided for askarel-filled transformers posing a threat to water course. Spill kits were assembled to handle PCB leaks and spills. A sealed containment area was constructed of concrete with 15 cm (6 in) high curbs to store liquid PCB waste until it can be shipped off the site. Also, transfer of PCB liquids and decontamination was do^e within this area whenever possible. A storage area inside a barn on the site was developed for PCB liquids and capacitors which were being held for future use. It consisted of a poured concrete central containment area with sloping floors adjacent to it. The barn had four disadvantages: 1. There was a lot of wood used in the construction of the barn and grain bins inside it. 2.
It was adjacent to the site boundary. Airborne effluent would quickly leave the site.
3. The ceiling was too low for a fork-lift truck. 4. There was no ventilation. A new building was constructed for PCB storage this summer which has none of these disadvantages. Several minor spills have occurred at Fermilab since PCBs were recognized as a hazard. The experience in dealing with these has led to improvements in procedures. In several cases the fluids were less than 50 ppm PCB. Cases greater than 50 ppm follow: 71
1. A vehicle struck a cooling fin on a Main Ring PCB transformer causing approximately 10 I (3 gal; of coolant to leak into the gravel underneath the fin. The leak was sealed with epoxy and the contaminated gravel disposed of in an EPA-approved burial ground. 2. A capacitor failed in the Capacitor Tree and dripped onto the asphalt below. Approximately 0.3 % (0.1 gal) leaked out and was cleaned up. The asphalt was dug up and replaced before results from analysis of the soil underneath were received. The soil contained 190 ppm PCB. This has been noted in the Laboratory's Decontamination and Decommissioning File for future action. 3. An askarel-filled transformer developed a leak from its drain plug onto the pad inside the containment area. Less than 0.005 £ (0.02 gal) was lost. The leak was fixed and the pad cleaned up. REMOVAL OF PCBS To reduce the hazard from vaporization of PCBs in occupied areas, a program of capacitor replacement has been initiated. All 76 capacitors in the Main Ring service buildings have been replaced. Fourteen capacitors used occasionally in Wilson Hall, a high occupancy laboratory building, have been removed, leaving only three. Replacement capacitors have been selected for the one remaining high occupancy area, the Linac Gallery. Actual replacement of these 53 capacitors is awaiting funding and scheduling of an appropriate shutdown period. One askarel-filled transformer failed without leaking. It was shipped off the site for disposal. Also, 210 PCB capacitors and 37,000 I (9800 gal) of PCB oil, most of which were in storage in the barn mentioned above, were shipped for off-site EPA-approved incineration. Finally, 6,160 I (1,630 gal) of askarel and 23,250 I (6,140 gal) were drained and shipped for off-site incineration this summer in conjunction with the processing of 12 transformers using a hot solvent extraction process described below. Storage of PCB oil has been eliminated and storage of PCB capacitors has been reduced to those spares currently needed for maintaining operations. Two techniques have been used for reduction of PCBs in Main Ring pulsed-power transformers. Filtration was tried with some early success. The PCB concentration was reduced from 3000 ppm to less than 500 ppm in one transformer. However, later filtration units from the same source were not as effective. That filtration technique, based on crushed synthetic rubber, has been replaced by a hot solvent cleaning process. The hot solvent PCB removal system is called Zero/P.C./Forty* and utilizes Freon©TF solvent.** •Developed by Positive Technologies, P.O. Box 3636, City of Industry, Calif. 91744 and applied by Power Transformer Services, Inc., P.O. Box 232, Somers, Wise. 53171. **Manufactured by E.I. DuPont deNemours and Company, Freon ^Products Division, Wilmington, Del. 19898. 72
The main advantage of the Zero/P.C./Forty system over the filtration process is the deeper penetration of the solvent in the gaseous ohase. The system alternates between gaseous and liquid phases. The F r e o n © T F solvent is distilled inside the apparatus which can be carried in a trailer to the transformer site (Figure 1). The distillation process leaves the PCBs behind in the apparatus and allows a small volume high in concentration to be drained off for disposal by incineration. To test the capabilities of the process, the Zero/P.C./Forty removal system was applied for five days to a failed Main Ring transformer initially having 25,000 ppm PCB. The transformer was refilled with oil and transformer operation was simulated by heating the oil externally. The temperature of the coil increased approximately 15°C. for a 30°C. rise in oil temperature. An oil temperature of 50°C. is required to simulate inservice use.** The process reduced the PCB concentration in the oil to less than 25 ppm PCB 90 days after refilling. Another similar transformer was processed subsequently. It had 16,000 ppm PCB initially instead of 25,000 ppm. After processing, refilling, and simulating transformer operation, this transformer also had less than 25 ppm PCB in its oil.
Figure 1. Processing of the ML8 Transformer
73
Based on these results, a program of PCB reduction was initiated. Six askarel-filled and six oil-filled operational transformers were processed in place. The askarel transformers were processed for five days each. The Main Ring pulse-power transformers, which contained approximately 25,000 ppm PCB, were processed for four days each. The two oil-filled utilities transformers containing 5090 and 1520 ppm PCB were each processed three days. Five of the askarel-filled transformers were retrofilled (refilled with different fluid to upgrade them) with silicone fluid.* This fluid was used for its fire protection properties. The sixth askarelfilled transformer was filled with oil because its location did not require special fire protection measures. A carbon filter was installed on one of twin transformers filled with silicone fluid.** At the time of installation, approximately 50 days after processing and retrofill, the PCB concentration in the silicone fluid was 736 ppm for the transformer with the filter and 520 ppm for the other. Ten days later the concentration of PCBs in the transformer with the filter had been reduced to 66 ppm. This transformer had only a very light load. Therefore, leaching of PCBs from the coil package during filtration was probably negligible. Results after 90 days of in-service operation are not yet available. Current results are shown in Table 1 with the number of days of powered operation given. It is clear that further processing will be required in some cases to reduce the PCB concentrations in the fluid below the 500 ppm required for repair should these transformers fail.1* The carbon filter can be used on silicone-filled transformers. No effective filter for oil filled transformers is currently available. Other techniques will have to be used until such a filter is developed. The cost of making a significant reduction in hazard from release of PCBs is relatively high. More than $300,000 have been paid to outside vendors to date and more funding is required to complete the program. Fermilab's management has taken the position that reduction of.PCBs should be placed in its proper perspective as one of the many Laboratory priorities and should be accomplished within the available funding. Experience up to the present time supports this approach.
*Manufactured by Union Carbide Corp., Silicones and Urethane Intermediates Division, Old Ridgebury Road, Danbury, Conn. 06817. **Manufactured by AMF Cuno Division, 400 Research Pkwy., Meriden, Conn. 06450. 74
TABLE 1. Effect of Zerc/P.C./Forty Processing
Before Processing Transformer ' i Volume PCB Cone, j Designation ; Fluid (gal) (ppm) ! Fluid U) ! 327 ; =650,000 Silicone CGI 1 Askarel 1238 i 265 CU2 i Askarel 1003 =650,000 Silicone •
NL1O
; Askarel
After Processing Days Powered After Refill Load
PCB Cone. (ppm)
63
Heavy
456
63
Heavy
745
1238
327
s650,000
! Silicone
59
Light
427
1
TGI
| Askarel
889
235
=650,000
S i1i cone
46
Light
736
TG2
i Askarel
889
235
=650,000
Silicone
51
Light
520
PL7
j Askarel
905
239
=650,000
Oil
21
Light
326
ML8
; Oil
1196
316
5s090
Oil
8
Light
18
PL12
j Oil
1238
327
1,520
Oil
24
Light
134
A22
• Oil
5204
1375
23,700
Oil
0
0
69
A23
; Oil
j 5204
1375
21,700
Oil
0
0
51
E23
:
Oil
5204
i 1375
28,700
Oil
0
0
28
F23
I Oil i
5204
1375
25,700
Oil
0
0
41
i
I
REFERENCES 1. "Occupational Exposure to Polychlorinated Biphenyls (PCBs)," DHEW (NIOSH) Publication No. 77-225 (1977). 2. "PCB Transformer Fire - Binghamton, New York," Morbidity, Mortality Weekly Review 30, No. 16, 187 (1981). 3. 40 CFR Part 761, Federal Register 44, No. 106. 31517 (1979). 4. 40 CFR Part 761, Federal Register £7, No. 165, 37357 (1082).
76
2D
RAFFINATE TREATMENT AT THE PORTSMOUTH GASEOUS DIFFUSION PLANT
T. A. Acox Piketon, Ohio
ABSTRACT Raffinate solutions, which contain uranium* technetium, nitrates, and lesser amounts of heavy metals, are produced in the decontamination and uranium recovery operations at the Portsmouth Gaseous Diffusion Plant. These solutions are presently being placed in temporary storage until three treatment facilities are constructed which will produce an environmentally acceptable effluent from the raffinate. These facilities are: 1) The Heavy Metals Precipitation Facility; 2) The Technetium Ion Exchange Facility; and 3) The Biodenitrification Pilot Plant. When the facilities are completed, the raffinate will be treated in 500 gallon batches. The first treatment is the heavy metals precipitation by caustic addition and filtering. The effluent proceeds to the ion exchange columns where the technetium is removed by adsorption onto a strongly basic, anion exchange resin which has been converted to the hydroxyl form. Following ion exchange, the solution is transported to the biodenitrification pilot plant. The biodenitrification column is a fluidized-bed using bacteria-laden coal particles as the denitrifying media. The resulting effluent should meet the limits established by the US EPA for all metals and nitrate. Technetium will be 98+ percent removed and the uranium concentration will be less than one milligram per liter.
77
DESCRIPTION OF PROBLEM In anticipation of increasingly stringent regulatory requirements, a raffinate treatment system was conceived and designed at the Portsmouth Gaseous Diffusion Plant which is operated by Goodyear Atomic Corporation (GAT). With the advent of the second round National Pollutant Discharge Elimination Permit (NPDES, 1980), the more stringent regulations became reality and, along with DOE's continuing concept of "as low as reasonably achievable" or ALARA (DOE 5480.1), became the catalyst behind the installation of the system. Later, the new EPA dose limit of 25 miliirem/year auded further emphasis to the reduction of radionuclides discharged to the environment. Raffinate generation is the result of uranium recovery operations in which uranium is reclaimed from decontamination operations, laboratory wastes, alumina traps and other process operations. These solutions are dilute nitric acid solutions which are first concentrated; uranium is then extracted from the concentrated solutions by an organic solvent. The aqueous effluent, or raffinate, is then discharged if the uranium concentration is below economic recovery limits. The discharged raffinate has the characteristics presented in Table 1 (Holland 1978). Presently, the raffinate is discharged to the X-701B Holding Pond where, following mixing with a lime slurry, metallic hydroxide compounds are removed from the stream by gravity settling with excess lime acting as a settling aid. This process is very effective for pH adjustment and heavy metals control, however there is no effect on pertechnetate or nitrate ion concentrations. It is primarily for these two chemical species that the raffinate treatment system was developed. The improved efficiency of heavy metals control and ease of metallic hydroxide sludge handling are viewed as added benefits of the system. TABLE 1. Raffinate Characteristics
pH
1.0
NO3
40 wt% 1,450 mg/1 120 mg/1 7,100 mg/1 360 mg/1 15 80 620 mg/1 15 210 mg/1 1.05 1.25
25
2 0.1 700
U Tc Fe Cu Ni Zn Specific Gravity
78
HEAVY METALS PRECIPITATION PROJECT Originally, the heavy metals precipitation project was to be an integral part of the biodenitrification pilot plant. Heavy metals removal is necessary to ensure efficient removal of nitrates at the pilot plant (Kowalchuk 1980). The primary reason for separating the metals precipitation from the pilot plant was a potential safety hazard with the technetium ion-exchange project, which will be discussed later. As stated earlier, the raffinate stream from uranium recovery operations contains significant amounts of heavy metals and has a pH >1.0. A considerable amount of developmental effort went into determining the optimum conditions for precipitation and filtering. The design considerations that were investigated included ease of chemical handling, dewatering efficiencies of the various types of filters and filter cloths, sludge handling and disposal and, ultimately, cost. The major problem in the design was the gelatinous condition and the poor dewatering characteristics of the precipitate (Deacon 1979). Laboratory studies determined the optimum conditions. The raffinate would be diluted approximately 1:1 and neutralized with sodium hydroxide to a pH of 8.0 (Deacon 1979). Lime addition formed a precipitate which had better settling characteristics, however, a higher volume of sludge would have been generated than with sodium hydroxide. Filtering was to be accomplished with a rotary drum vacuum filter or a pressure filter with a filter cloth that had an air permeability of 1-2 CFM (Greiner 1981). The addition of a filtering aid or a coagulant did not increase the filtration rate, but filtration efficiency may be slightly improved. Filtration rates ranged in excess of 5 gallons/minute/square foot in laboratory studies. A pressure filter with an indexing filter cloth was ultimately chosen due to the rapid blinding of a filter cloth. Sludge generation will be approximately 250 ml of dewatered sludge per liter of feed to the filter, or about one-half liter of sludge per liter of undiluted raffinate. The filter cake will contain approximately 20% solids by weight. At maximum operational levels, this will amount to about 76 cu. ft. of sludge per 16-hour day. Disposal of the sludge has posed additional problems. Being both hazardous and radioactive, the sludge needs special disposal requirements. The only option which appears to be available is a low level radioactive waste disposal area. Because the regulations
79
for such a disposal have not yet been finalized (DOE 5820.DRAFT), an enclosed temporary storage area is being utilized until a future line item project is completed. Following treatment (Table 2) (Deacon 1980), all heavy metal concentrations have been substantially reduced. Pertechnetate and nitrate ions have been essentially uneffected by any treatment. Any reduction in the concentrations of these two anions are the result of dilution effects. TABLE 2. Comparison of Average Materials Concentrations (mg/ml)
Before Treatment 'C Loop pH N0 3 -N U Tc
<1.0
After Heavy Metals Precipitation Treatment 8.0-8.5
37,100
*
1,230
<0.1
0.63
*
Fe
96
0.4
Cu
96
0.3
Ni
395
1.4
Zn
67
<0.02
*Any decrease in concentrations due to dilution effects. TECHNETIUM ION-EXCHANGE Technetium was introduced into the cascade as a contaminant of recycled reactor material (Williams 1981). Once in the cascade, portions of the technetium are adsorbed or deposited on the internal surfaces of equipment and the remainder passes through the cascade and concentrates at the top of the cascade. When this equipment is decontaminated, the majority of the deposited technetium is so]ubilized into the decontamination solutions as the pertechnetate ion (TcO^"). The pertechnetate ion is then discharged with the uranium recovery raffinates to the X-701B Holding Pond and ultimately to the environment.
80
Goodyear has had a self-imposed limit which is approximately two orders of magnitude below DOE guidelines for the controlled release of t^chnetiurn-laden recovery raffinates. DOE guidelines also specify tnat radioactive discharges must be kept "as low as reasonably achievable," the "ALARA" concept. In keeping with that philosophy, efforts were initiated to reduce the technetium concentrations in the X-7O1B effluents. Several possible methods for the removal of technetium were investigated in the laboratory. Results of these studies indicated that anion exchange is the most promising, practical, and costeffective of the methods investigated (BeCraft, 1979). The other methods that were investigated were: •
conversion to an insoluble species by chemical means;
•
electrodeposition and;
•
solvent extraction.
Removal of technetium from X-705 raffinates by anionic exchange was demonstrated at Portsmouth, both in laboratory and field testing, in which column and batch methods were used. Laboratory studies indicated that technetium was 98+ percent removed in column tests using four-inch columns while a field study using a batch treatment method demonstrated a 93.6 percent removal (BeCraft 1979). At this time, the treatment method was chosen: •
The nitrated form of the resin was to be used.
•
The raffinate+was to be slightly neutralized in order to reduce the (H ) concentration to approximately 0.5 grams per liter (pH "0.5).
These conditions provided optimum technetium removal at a minimum of raffinate treatment. However, after the initial design had been chosen, it was reevaluated in view of the potentially hazardous condition that existed. Several noncritical incidents have occurred involving nitrated ion-exchange resins, the most serious of which occurred at Hanford, Washington, in 1976 (BeCraft 1979). Although it was pointed out that operating conditions would be much less severe at Portsmouth than at some other locations where incidents occurred, it was decided to change the design slightly. It was at this point
81
that the removal of technetium from neutralized (pH = 8.0) raffinate was investigated (Deacon 1981). The resin may be received in the chloride or the hydroxide form. Forms of the resin were studied to determine the efficiency with which technetium was removed. The studies (Table 3) indicated that technetium loading was in the range of 20-23 milligrams of Tc per gram of resin; nitrated resin had a loading of ~1.5 milligrams of Tc per gram of resin (Deacon 1981). However, the chloride resins reacted with the slightly alkaline raffinate to produce large amounts of off-gassing thus making the hydroxyl resins preferable. By making these changes, the potential safety problems are avoided and technetium loading is substantially increased. In addition, the uranium loading on the nitrated resin (using the highly acidic raffinate) is avoided due to the removal of uranium ii the heavy metals precipitation facility. TABLE 3. Resin Comparison - "C" Loop Raffinate Sample Raffinate (pH 8.0) Dowex 1-X8 (OH")
(CT) Amberlite IRA-401 (OH")
(CD
g Tc/ml
D/min/ml
7.35
278,000
0.01 0.22
233 8,150
0.01 0.01
440 245
BIOLOGICAL DENITRIFICATION The primary constituent used in the dissolution of materials for uranium recovery is nitric acid. Although used in concentrations up to 3jN_, the acid concentration is found as high as 25 weight percent in raffinate material. When the raffinate is discharged (assuming uranium and technetium constraints are met), considerable dilution occurs at the X-701B Holding Pond with the influx of other waste streams, however, the total mass of N0 3 is unchanged. Most treatment technology to date has been concerned with the suspended growth type of systems, such as trickling filters, rotating disks, etc. These systems work well if the N0 3 concentration is relatively low, in the range of 60 to 100 mg/1. Because the
82
N0 3 concentrations are much higher at Portsmouth, a fluidized bed system was developed at Oak Ridge National Lab. Fluidized bed chosen because it represents the most cost effective method for nitrate removal from high nitrate waste streams (Kowalchuk 1982). In fluidized bed biological denitn'fication, the wastewater passes upward through a bed of particles such as sand or granular coal at a sufficient velocity to cause motion or fluidization of the media. When added to the influent, bacteria will attach themselves to the media and grow. This type of attached growth will attain much higher biological surface area (or population) than is possible in conventional technology. The higher bacterial population will allow for substrate conversion efficiencies as much as an order of magnitude greater than conventional technology (Kowalchuk 1981). Biological fluidized bed denitrification pilot plant studies were performed at ORNL (Kowalchuk 1981, 1982). These studies characterized process operating parameters and provided data from which bioreactor performance could be calculated. The best performance resulted from the following operating conditions (Table 4 ) : TABLE 4. Conditions for Bioreactor Operation Influent Parameters NOj
EtOHl PO^
NH3
Bacteria Wt. Fraction Temperature pH Fluidizing Velocity
400-600 ppm C to N ratio > 1.4:1 20 ppm 20 ppm
Effluent Parameters <50 ppm
25% 30°C 7.0 1.2 cm/sec
It should be noted that the study demonstrated that biological denitrification is practical at nitrate concentrations up to 10,000 mg/1. The operating conditions in Table 4 were chosen to assure compliance with the NPDES limit of 10 kg N03/day, maximum (5 kg N0 3 /day, average). CONCLUSIONS Effective August 1983, the effluent from the biodenitrification pilot plant is considered a point source by the USEPA and is designated as Outfall 003B. The NPDES limits (NPDES 1980) and the expected concentrations of pollutants are listed in Table 5.
83
TABLE 5, Average Final Effluent Limits and Expected Concentrations NPDES 003B Limits kg/Day Average Nitrate (N) Total Copper Total Zinc Total Iron Total Nickel
2.3 0.007 0.005 0.23 0.018
Expected Concentrations kg/Day 1.4 <0.001 <0.001 <0.001 0.002
As can be seen, the effluent from the raffinate treatment system is expected to meet the NPDES requirements for metals and nitrate concentrations. Additionally technetium and uranium concentrations are to be substantially reduced in keeping with the "ALARA" concept.
TR-120382-NTB
84
REFERENCES BeCraft, F. B., and M. E. Holland. 1979. Ion-exchange Removal of Technetium-99 from Nitrate-Containing Hastes~at the Portsmouth GDP: Safety Aspects. GAT-T-2962, Goodyear Atomic Corporation, Piketon, Ohio. Deacon, L. E., and M. J. Greiner. 1979. Raffinate Neutralization and Dewatering Studies. GAT-521-79-159, Goodyear Atomic Corporation, Piketon, Ohio. Deacon, L. E., and M. J. Greiner. 1980. Augmentai Solids Removal from X-705 Raffinates. GAT-521-80-44, Goodyear Atomic Corporation, Piketon, Ohio. Deacon, L. E. and M. J. Greiner. 1981. Laboratory Investigation of Amber!ite IRA-401 and Dowex 1-X8 Ion-Exchange Resins for Tc-99 Removal; Technology Transfer Bulletin. GAT-521-81-67, Goodyear Atomic Corporation, Piketon, Ohio. Dikeman, 0. C. 1976. Resin Treatment of 'C Loop Raffinate. GAT-832-76-55, Goodyear Atomic Corporation, Piketon, Ohio. DOE Order 5480.1, Chapters XI, XII. DOE Order 5820 (Draft).
"Radioactive Waste Management."
Greiner, M. J., and L. E. Deacon. 1981. Filtration Characteristics of Sludge Removal from X-705 Decontamination Raffinates. GAT-521-81-227, Goodyear Atomic Corporation, Piketon, Ohio. Holland, M. E. 1978. "Characterization of X-705 Raffinate." GAT-521-78-203, Goodyear Atomic Corporation, Piketon, Ohio. Kowalchuk, M. 1982. Bi odenitri ficat i on of Gaseous Diffusion Plant Aqueous Wastes: riu1
Plant,
NPDES Permit, Final. 1980. Portsmouth Gaseous Diffusion Plant, Permit number 0H0006092. Williams, D. L., R. I. Kaplan and C. R. Walker. 1976. Third Report of the Technical Division Technetium Committee. Section III, Part T, GAT-T-2598 (Secret), Goodyear Atomic Cbrporation, Piketon, Ohio.
2E
ASSESSMENT OF DOE INACTIVE CHEMICAL WASTE SITES: ARE ADEQUATE INVENTORY DATA AVAILABLE?
C. J. English and D. H. Denham Pacific Northwest Laboratory Richland, Washington
ABSTRACT During FY-82, PNL developed a method for collecting data describing past chemical waste generation and management at DOE sites. Included in this development was application of the method to the DOE Hanford Site. The need for such information arose from the apparent lack of data available with which to adequately assess potential problems with inactive chemical waste sites under DOE control. Responses submitted by DOE field offices and contractors in compliance with the Comprehensive Environmental Response, Compensation and Liability Act of 1980 (CERCLA) supplied relatively few data with which to assess the nature and magnitude of inactive chemical waste site problems. It was known, however, that past and present activities at DOE sites involved large amounts of chemicals and potentially generated large quantities of hazardous wastes. A method was, therefore, needed to collect data describing chemical waste inventories at DOE sites. This paper describes the approach to data collection developed by PNL. This approach is purposely general to deal with the tremendous site to site variations that exist, both in the nature of activities conducted and in the availability of records describing past waste management. In addition, the paper describes some of the lessons learned in applying this approach at the DOE Hanford Site. This will hopefully be of benefit to similar applications elsewhere. The results of this effort indicate that adequate data should be available at most sites to place bounds on potential uncontrolled chemical waste site problems. Data to. perform detailed site assessments, however, may be unavailable.
INTRODUCTION Identification, assessment and cleanup of uncontrolled hazardous waste sites are activities that are receiving increasing attention by groups responsible for environmental protection. Recent passage of the Comprehensive Environmental Response, Lia-
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bility and Compensation Act of 1980 (CERCLA) demonstrates the increased importance being given these activities at a national level. In response to this, the Department of Energy has begun steps to assess potential problems related to past management of hazardous wastes at their facilities. This paper describes work performed at PNL from late FY80 through FY82 for the DOE Office of Operational Safety (00S) related to management of inactive hazardous waste sites under DOE control. Activities related to management of inactive waste sites are usually organized into three steps. These are identification of such sites, assessment of the environmental risks posed by them and performance of remedial actions at those sites posing unacceptable risks. Performance of each of these steps often requires related data collection efforts. The work initially undertaken at PNL was organized in a similar fashion. Objectives were: .
Identification of inactive hazardous waste sites under DOE control;
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Development of methods for assessing the risks posed by these sites;
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Development of methods for obtaining data necessary to perform risk assessments; and
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Development of methods for evaluating and implementing remedial actions.
The scope of work was necessarily flexible since the need for each step can only be determined by the results of previous steps. For example, remedial actions are unnecessary if no sites pose an environmental risk. To date, efforts have been limited to identification of sites. Further work to develop risk assessment methods will depend on the results of a Department-wide inventory of inactive sites. SUMMARY OF PAST ACTIVITIES Late in FY80, PNL prepared a questionnaire for collecting data describing past hazardous waste generation and management at DOE sites. Along with the questionnaire, guidelines describing possible data sources were also prepared. The main objective of the questionnaire was to collect information quantifying past hazardous waste generation and identifying inactive disposal sites. If this was not possible, it was hoped that the questionnaire would help determine the current state of knowledge of past waste management and identify data gaps. This information would then be used to develop follow-on efforts.
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•. J
(
Plans to issue the questionnaire during FY81 were stalled by passage of CERCLA. Under Section 103(c) of that Act, potential hazardous waste generators were required to respond to a similar questionnaire issued by EPA. While this questionnaire was not as detailed as the one prepared by PNL, it was felt that it might accomplish the same objective. The PNL questionnaire was not issued to avoid possible duplication of effort. The CERCLA questionnaires received identified few inactive disposal sites. To assess the significance of this, results were compared with published hazardous waste generation lists obtained from RCRA Section 3010 notifications. The RCRA data showed that many DOE sites are presently generating hazardous wastes. This suggested several possibilities: 1) hazardous wastes have been generated only recently at DOE sites; 2) very few DOE hazardous waste sites have been closed; 3) data describing hazardous waste generation are much more available than those describing past disposal sites; or 4) most hazardous waste disposal sites fell under the exemptions granted by CERCLA. Since efforts had also been made to obtain data on exempted sites, it was felt that data availability was the main problem. These results indicated the need for an alternate method of collecting data - one that would rely on data sources other than waste disposal records. Since compilation of the CERCLA and RCRA data indicated a great variation in the nature of activities conducted from site to site, it was assumed that there would be similar variations in the types and availability of data at different sites. Therefore, the method would have to be sufficiently general to have application to different sites. METHOD DEVELOPMENT Previous efforts had focused on identifying specific waste sites and had shown that necessary data were generally unavailable. We therefore decided to focus attention on developing a method for obtaining inventories of wastes which had been generated at different facilities. If specific waste site data were unavailable, the inventory would still allow a preliminary determination of the potential problems associated with past disposal and would help direct efforts at identifying particular disposal sites, if necessary. The first step was to review efforts being conducted elsewhere. Unfortunately, these were somewhat limited. Most of EPA's efforts have been related to characterization and remedial action at waste sites which have already been identified. Site identification efforts are limited because of the large number of sites already known to exist. Several states and one EPA Region had previously made attempts to quantify past waste generation and to
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identify sites. The most directly related work reviewed was that being done by the Department of Army for their Installation Restoration program. This has included searches of hisorical records at Army facilities to identify potential problems associated with past hazardous waste generation and disposal. Discussions were also held with DOE Field Office and contractor staff to identify possible approaches to the problem and possible sources of data. Before developing the method, it was necessary to more accurately define the scope of the inventories. The CERCLA questionnaires had indicated some confusion among the DOE field offices and contractors concerning what constituted hazardous wastes. This was especially true with the exemptions granted to sites and activities authorized under the Atomic Energy Act of 1954 (AEA). Therefore, we decided to define the following as chemical wastes: • • •
all all the non
RCRA defined hazardous wastes wastes which would meet RCRA definitions if not for AEA exemption ("mixed wastes") radioactive components of radioactive waste streams.
Not included are non-RCRA solid wastes contaminated with radioactivity (e.g., low level radioactive wastes). The decision to expand this definition beyond that provided in RCRA was made for two reasons. First, many wastes generated at DOE sites are excluded from RCRA because they are generated from activities resulting from the Atomic Energy Act of 1954. Secondly, it was felt that the RCRA definition might not apply to all DOE wastes and sites. Environmental hazard is not an intrinsic property of a waste, but rather depends on several factors, including the toxicity of waste constituents, the quantities of wastes, the physical state of the waste and the disposition of the wastes. The RCRA definitions make certain assumptions concerning these factors which may not have application to all DOE sites and the wastes generated at these sites. That is, they were developed primarily with typical industrial wastes and typical industrial waste sites in mind DEVELOPMENT OF INVENTORY METHOD Based on the results of the CERCLA questionnaire, we did not expect to find many records available which described past disposal of hazardous wastes. Therefore, while the inventory approach would check for the existence of such records, we expected to concentrate efforts on estimating chemical waste inventories from other data. The inventory approach consists of five major steps:
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I
• • • • •
Identify site activities generating chemical wastes Determine the availability of waste management records for these activities If records are available, determine past waste management; if not, determine methods for estimating chemical waste generation for these activities. Determine the availability of data for estimating waste generation and disposal If available, collect data and make estimates; if not, determine alternate estimating procedures.
If adequate records of past waste disposal exist, the entire process may be limited to the first three steps. If not, most of the inventory effort will involve the final step?. In defining site activities generating chemical wastes, it should be helpful to organize activities similarly to site organizational structures. The advantage is that this is typically how records will also be organized. Also, record searches are simplified by starting at the most general level of activity and working down through more specific levels, as needed. The availability of waste management records can generally be determined through site environmental protection staff or records management staff. If records are unavailable, waste generation will have to be estimated. The key element to this is identifying variables describing activities which can be correlated to waste generation. Data quantifying these variables are then collected and used to estimate waste generation. Variables for several example activities are shown in Table 1. Similarly, possible correlations between these variables and waste generation are shown in Table 2. Correlations can be developed theoretically based on knowledge of the activity generating the waste or they can be developed empirically based on existing data. Sources of data for developing theoretical correlations include any detailed description of the activity such as operating manuals, process technical manuals and process flow sheets. These documents usually describe an activity in sufficient detail for the relationship between process variables and waste generation to be determined. An advantage of theoretical correlations is that they can usually be developed to account for changes in the process that have occurred with time. If these changes are well documented, a unique relationship can be developed for each process variation. A limitation to theoretical correlations is that the process generating the waste must be well defined (i.e., a "black box" approach cannot be employed).
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TABLE 1
Examples of Waste Generating Activities and Variables
Waste Generating Activity
Variable
Operation of Site
Number of Employees Budget Process Variation Material Input Production Level Radioactive Waste Output
Operation of Plant
Uperation of Laboratory
Chemical Usage Research Activities Maintenance Schedule
Maintenance
TABLE 2
Variables
Examples of Waste Generation Correlations and Data Sources Data Sources For Developing Correlation
Correlation
Employment Quantity of Chemical Waste Employment Records Level Generated per Man-Year Waste Disposal Records Material Input
Quantity of Chemical Waste Process Technical Manuals Generated per Quantity of Process Flow Sheets Material Input Purchasing Records Waste Disposal Records
Radioactive Quantity of Chemical Waste Process Technical Manuals Waste Generated per Quantity of Process Flow Sheets Generation Radioactive Waste Generated Radwaste Disposal Records Research Reports Production Quantity of Chemical Waste Process Technical Manuals Generated per Quantity of Process Flow Sheets Product Operating Manuals Production Records Waste Disposal Records
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Empirical correlations describing waste generation are developed from current and available past data describing waste generation. For example, if data describing chemical waste generation are available for an activity for the past five years, feed material input data for the same period can be obtained and used to derive a relationship between feed material input and chemical waste output (e.g., pounds waste produced per pound material input). Feed material input data for the entire activity lifetime could then be used to estimate chemical waste production over the site lifetime. Limitations to this method are that adequate data must be available both to derive the correlation and to estimate waste generation. It is also usually necessary that the activity gmerating the waste has remained constant over the period for which data exist. Once again, in attempting to develop waste generation correlations, it is best to start at a very general level of activity and then become more specific as needed. If the above methods cannot be used, waste generation and disposition must be estimated from the best available data describing waste disposal practices and waste disposal sites. In general, information is collected describing the methods that might have been employed to dispose of wastes and the locations of sites where disposal might have taken place. Best judgment is then used to estimate the disposal of wastes at each site. Possible past disposal methods can be identified by considering the disposal methods prevalent at the time the activity was in operation and the methods used to dispose of similar wastes. For example, wastes currently being incinerated were probably disposed of by open burning in the past. Similarly, if possible disposal sites are identified, judgment is used to estimate those that were used. For example, if chemical wastes were known or thought to be landfillecJ, it is most likely that they were taken to the closest available landfill for disposal. APPLICATION OF METHOD TO HANFORD In applying the method to the DOE Hanford Site, we were able to identify a number of chemical waste disposal sites and quantify the contents of many of these. Several general observations were made which may be helpful in conducting similar efforts elsewhere. In identifying waste generating activities, we found it most helpful to begin with current waste management records and data. Historical data were then used to determine hew present activities had changed with time. Discussions with former or long-time employees were very helpful in identifying past activities. Chemical waste management records were found to go back only to the early 1970s. Radioactive waste management records existed
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through site start-up and were very helpful for estimating chemical characteristics of "mixed" wastes. Since "mixed" wastes form the majority of chemical wastes at many DOE sites, this approach could prove helpful elsewhere. In general, useful records were available only for very specific activities, usually corresponding to operation of an individual facility. Selection of correlations to estimate waste generation appears to be governed by availability of data. Therefore, in some cases, it was more useful to determine all the descriptive data vailable fo an activity and then try to select a correlation which would use the available data. The experience of senior employees familiar with the activities was quite helpful in this regard. CONCLUSIONS e
The method should be applicable to all DOE sites. That is, every site should have some sort of data available from which chemical waste inventories can be estimated.
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The accuracy of the inventory data obtained will depend on the accuracy of the data used for estimation and on the nature of the assumptions involved in estimation.
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In most cases, the inventory developed should be accurate enough to place bounds on potential uncontrolled chemical waste site problems.
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In most cases, the inventory developed will not be adequate to perform waste site risk assessment or to design remedial actions. They should be adequate, however, to design follow-on studies needed to obtain assessment or design data.
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The involvement of senior employees familiar with past activities is essential for efficient performance of the inventory.
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PANEL DISCUSSION: HAZARDOUS WASTE MANAGEMENT
V. De Carlo, Office of Operational Safety, Moderator J. !.. 1Bresson, Albuquerque Operations Office N. v . Stas, Bonneville Power Administration J. F. Winq, Oak Ridge Operations Office S. R. Wright, Savannah River Operations Office Vincent De Carlo, Moderator The primary objective of this panektoday is to provide you with a broad spectrum of hazardous waste management activities at DOE facilities. In addition, we would like to discuss what we currently are doing with DOE Order 5480.2 and the other activities in hazardous waste management supported by Operational Safety. Members of the panel include Jim Bresson from the Albuquerque Operations Office, Nick Stas from the Bonneville Power Administration, Jerry Wing from the Oak Ridge Operations Office, and Steve Wright from Savannah River Operations Office.
James E. Bresson, ALO I am going to spend a couple of minutes discussing Albuquerque's approach to the problems of Albuquerque Operations and the hazardous waste regulations. We, of course, are different from all the other field offices in that we have many offsite locations. There are eight Albuquerque-related facilities in seven different states which means we have had to deal with five regional EPA offices and, of course, all the individual state offices. The first part of our policy has been to let our local people deal with the local offices. All the Albuquerque facilities operate, of course, under the Atomic Energy Act, and we feel that very clearly Section 1006 of the regulation exempts our facilities from literal compliance with the hazardous waste regulations. However, we have had to straddle the fence for about two long years now, simply because we have not really known whether or not that exemption would stand a court test. Consequently, we have instructed our area offices and contractors to work with and to try to cooperate with the local authorities. When it came time to send in the notification of hazardous waste activities in 1980, we did that. We got an EPA number for each of our facilities. It turned out that in some cases this action was prudent. We then submitted our Type A applications for interim status consideration for our waste management activities, and figured that it would be several years before we heard anything after that. About five months later, we began having ALO contractor people calling us and saying - "Hey, we got the local EPA people coming down, and they want to come onsite and pull an inspection." 95
Because we have tried to maintain a posture of cooperation, and we have a responsibility for public health and safety in this matter, we have allowed EPA and state inspectors on our sites when they requested to do so. We have shared information with them. We have exchanged correspondence with them, and in almost every case, at least from what I hear, this policy has worked to our advantage. And my understanding is we have reasonably good working relationships with EPA and state offices in all but one Region, and it happens to be our own Region VI. We may get a chance to test our exemption status before very Region VI inspectors visited Sandia and the Lovelace Inhalation Toxicoloqy Institute, and no matter what we said they were doing, they said it wa< an inspection. And they came up with some possible citations. They citec us for failure to provide adequate ground-water monitoring at a small hazardous waste disposal facility, where the water aquifer is 400-to-600 ft down. And they wanted to cite us for failure to produce bilingual signs. And they wanted to call a sewage lagoon a hazardous waste impoundment pond. This "inspection" was performed in April 1982. We have not hc«r: from EPA yet, but we are still waiting to see if we will get a citation notice. If so, this action could lead to a test case for the exemption. But that is the only case so far, to my knowledge anyway, where ALO contractors have had threats of citations. In summary, for two years, we have been cooperative. We have bent over backwards. We have sent information to EPA, and ALO contractor people groan and moan when I tell them to send EPA or primary states more information. We exchange correspondence. On every piece of correspondence, we tell the EPA, ad nauseam, that we are exempt from the regulations, and are providing data for information only. But the problem right now is after two years its getting awfully hard to try and keep straddling the fence. And I for one really would like to know - are we exempt from the regulations or are we not? Now that the DOE Order has been issued, are we going to get an answer from the EPA lawyers on our proposed Memo Of Understanding (MOU) or not? Once the DOE Order has been issued, then given some time and money, we think we can go ahead and comply with the provisions of DOE 5480.2. In fact we think we are not in too bad a shape now, but frankly the biggest problem we have is this continuing attempt to be cooperative, just in case we may be subject to EPA's regulations someday.
There are some cases where our contractors very definitely have to comply with the hazardous waste regulations. Of our eight sites, only two of them dispose of hazardous waste onsite. Consequently, to get rid of hazardous waste at the other six ALO sites, we have to transfer them into the EPA system. Getting identification numbers from EPA early on has helped us. We have to follow all the manifest rules and we have to follow the rules related to disposal at properly approved disposal sites. In these instances, the exemption does not buy us anything. We have to comply with the EPA system.
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With respect to operations in Albuquerque and whether our contractor people are paying attention to hazardous waste, in the last two years in every waste management appraisal (and we get around on a schedule to appraise all contractors), we have included the topic of hazardous waste management. We have made several recommendations which have resulted in hazardous waste management improvements. We have not truly attempted to implement the proposed DOE Order on hazardous waste management as of this date (12-7-83). We are close, but we have not written the program yet; we want to see the Order in final form before we do that. Our policy though has been, and I hope will continue to be, technical but not literal compliance with 40 CFR Parts 260-265. The ALO currently has several environmental monitoring programs and effluent monitoring programs in effect because we knew it was a good idea to have them, and because related DOE Orders require them. The records may not be exactly the same as those required by the hazardous waste regulations, but the information is there, the intent is there, and the protection of public safety is demonstrated. So we do not intend to lay a program on our contractors, unless we are absolutely directed to do so, that is going to meet the letter of the 'aw; it is going to meet the intent of the law.
Nicholas J. Stas, BPA Bonneville Power Administration (BPA) markets wholesale electric power from 29 Federal hydroelectric dams in the Columbia River Basin and from three Columbia River Treaty dams. The BPA service area includes 300,000 square miles, primarily in the states of Washington, Oregon, Idaho, and Montana, with small service areas in California, Nevada, Utah, and Wyoming. The transmission system serves as the backbone for interconnected utilities in the region and is connected with 17 other transmission systems at over 100 locations. It consists of approximately 13,500 circuit miles of high voltage transmission lines and over 360 substations. I am going to talk primarily today about polychlorinated bipheyls (PCBs) which is BPA's major hazardous/toxic waste problem. Dr. Baker did a good job of providing a background on the chemical properties and the potential hazards of PCBs. So I am going to focus my attention on some of the things that Bonneville has been doing related to the management of hazardous waste including PCBs. PCBs have specific properties that allow capacitors to be made smaller, lowering equipment costs. These capacitors are essential to us for conservation, voltage control, system stability, and serve to maximize the capacity of the associated transmission lines. The BPA has approximately 140,000 PCB capacitor cells in its system, which translates to about 5.5 million pounds of PCBs. In addition, we have estimated that about 40,000 pounds of PCBs are in our transformers and coupling capacitors, primarily concentrated in our Pacific Northwest-Southwest intertie. 97
The BPA has been looking at the various types of substitutes for PCBs for some time, including silicone oil, perchloroethylene, and trichlorofluoroethane; like PCBs, these are also highly nonflammable. The way BPA initially approached the PCB problem was by developing a computerized inventory of all capacitors, providing locations and numbers. The advantages of the computerized inventory are that it enables management to get a quick access to information and an update on where PCB capacitors are located, and also aids in reporting requirements. We are somewhat different than other organizations represented on the panel today in that we have no facilities under the AEC exemption. Our General Counsel has advised us that we must comply with those requirements indicated in Executive Order 12088 and related implementing procedures in all of our States. So early on, we took the approach of getting involved by providing early input to the States in the formulation of their various Implementation Plans for hazardous wastes, and this has paid great dividends. The time that we put into this has let our concerns be known, even though some of the States have gone in a somewhat different direction in implementing hazardous waste regulations. For example, Oregon has included PCBs in their definition of RCRA waste streams, which is different than the EPA's approach by which PCBs are regulated under TOSCA. When we first reported our RCRA hazardous waste streams to EPA, we included PCBs in our initial reporting. Later, with the advice of EPA, it was determined that we did not have to report PCBs in the RCRA waste streams, and compliance with TOSCA regulations for PCBs was sufficient at that time. However, now that the States are coming on the line with RCRA Implementation Plans, we have found that we need hazardous waste permits, particularly in the State of Washington which presently has a 400-pound threshold for their small generator exemption. The rate of PCB capacitor failure at BPA has been low. Historically, only 0.2% of the BPA's capacitors fail each year. Of those that tail, 10% rupture. It is believed that one reason BPA has such a low failure rate is that we have a difficult-to-pass, high-potential test which we require our manufacturers to perform. We expected about 30 ruptures each year, releasing approximately 200 pounds of PCBs in small increments throughout the service area. A rigid BPA technical specification includes one requirement for an application of three times the rated voltage for five seconds. This test has eliminated some manufacturers from BPA's qualified suppliers list. We believe that this has resulted in less incidence of PCB ruptures than are experienced by some of the other utilities. One thing that I should point out is that we do expect an increasing failure rate of capacitors, since about 80% of our capacitors are approximately 20-years old, installed around 1962. I would like to take a little more about BPA's relationship with state and local environmental enforcement bodies. In the area of hazardous waste, we have not hesitated to provide recommendations on proposed rulemakings to state agencies and to EPA. I have a few slides* ' that (a) Not included in these Proceedings. 98
illustrate examples in which we had success in resolving problems related to hazardous waste. These first few slides show the large tanks where we have stored oil from transformers. A problem in the past has been contamination from the mixing of PCB-contaminated oil in the tanks. We resolved this issue very early on, and all our tanks that are PCB-contaminated are now marked. One problem we identified with PCB regulations involved the definition of capacitors. Large coupling capacitors and bushings were addressed in the same manner as the smaller type capacitor cells in the original EPA regulations. We succeeded in getting clarification from EPA Region 10. Now after draining large bushings (of the porcelain type), we are allowed to landfill the bushings in the Arlington, Oregon, site. This is the standard type of PCB capacitor cell with a small bushing on the end. Originally EPA had wanted us to transport these in 55-gal drums. Because we wanted to keep down the costs of disposal, some of our maintenance people were knocking off the top bushing and trying to get as many of these into a 55-gal drum as possible, changing an intact nonleaking article into a leaker and increasing the risk of PCB spills. What we did at Bonneville was to develop a container that more closely fits the shape of the capacitor cells and provides better safety in transporting them. This is the system we are now presently using, and it is also convenient to be handled on forklifts and loaded by personnel. One of our concerns has been the safety of the workers working with the PCBs, and we have contracted for some studies to determine what effects, if any, our workers in the field have had from PCB exposure. We have recently constructed new holding facilities for storing PCBs with appropriate containment basins, and also we have added a full-time position at BPA of Hazardous Waste Program Coordinator, which is located organizationally in our Office of Safety. In the area of spill containment, BPA sites or substations that contain capacitors with PCBs have been identified on a priority basis for SPCC-type attention and additional containment and clean-up and prevention measures. A BPA committee was assigned that identified those sites that are near sole-source aquifers or flood plains or high-potential-risk areas. We also have detailed standards dealing with the handling, storage, and clean-up of PCBs. Based on this committee's recommendation, BPA has been adding containment-type catch basins which have been budgeted and programmed on a yearly basis. Are there any questions? Question: You, too, are apparently hauling these things somewhere to dispose of them by incineration, burial, or what?
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Answer: Yes, well both. Of course, the oil and the capacitors are being hauled (we just recently had a contract with ENSCO in El Dorado, Arkansas) and certain PCB articles are presently being landfilled at the chemical security site in Arlington, Oregon.
Jerry F. Ming, PRO /You people in health physics will remember that at one time resuspension was all the rage; for about two or three years, that is all people seemed to talk about. I was commenting about that to Tom Frangos, and I said I hope it turns out that that is the way it is with hazardous waste, that in about two or three years people will just quit talking about it. And he said, "Well, I don't think so, because nobody ever passed a law about resuspension." Well, our own experience, like yours I think, has been pretty much that you started out with inventories. And we started out with inventories. Now I do not want to play "can you top this?", but our Bonneville friends mentioned they have 5.5 million pounds of PCBs. I hate to say it, but we have 8.7 million pounds. I saw an inventory summary that Los Alamos did on hazardous waste. It was a photograph taken at one particular point in time, and Oak Ridge had about 97% of the Agency's hazardous waste. It all depends on what set of cards you are playing with, because when that inventory was taken, chromium-3 (trivalent) was considered to be a hazardous waste. Because we have some very large cooling-tower systems (the largest in the world) that use chromium +3as an inhibitor, and before we get rid of it we try to make s.ure it is all Cr ,we had a very large number to report. Well, since then Cr 3 has been taken off that list, so that we may not be Number One, I don't know. I would like not to be. But inventories change, and it all depends on how EPA makes their definitions. I think because we have large facilities the matter of storage has not been too serious a problem. We continue to have to store because of the lack of getting this Memorandum Of Understanding. We have the same kinds of problems that Albuquerque has. Kentucky has been making noises, at least lawyer to lawyer, about not accepting our draft MOU, and they have been trading words back and forth. Tennessee has started making some noises just in the last couple of weeks. But if there was ever a race I wanted to lose, it would be the one with Albuquerque. I would let you have your test case first. Please. On the matter of disposal, our contractors have a good practice of pre-contract onsite inspection before a deal is cut with a disposer. That site is visited, inspected, and he is nit-picked, and if he is found clean then we do business with him. All this may sound like an excellent idea, and it is. We have learned the hard way that you cannot do that only once, because one disposer came out like a shining knight the first time around, and lately we have learned that he is very likely in deep trouble. What this
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means is before long your can find yourself in a Superfund situation, where you have got to go out and lay out money if they can identify your containers or your manifest of materials that went into that mess, and then they try to get your proportionate share out of your hide. We did have some experience with that. It is not over yet, and it is expected to be a very small one, but that is exactly the way it happened. There is a second one that very likely will cost somebody some money. I think like Albuquerque in the matters of inspection and registration. We have done pretty much the same thing. We did the registration, and like Albuquerque, we think it was a good idea. Without some of those numbers, we would not be able to move anything. On the other hand, with those numbers, it gave some of the states the idea that we had to pay fees. And so we had to write back to them and send their forms back and say "Well, you see we have this Memorandum Of Understanding," and it kind of got like Virginia and Santa Claus; we are still waiting for that letter that says "Yes, there is a Memorandum Of Understanding." I think it is getting to where we all are saying pretty much the same kinds of things. If Ed or Tom or Vince or anybody can do something to get that Memorandum Of Understanding in place, it would surely be helpful to all of us.
Steven R. Wright, SRO To give you some background on what I am going to talk about I want to tell you a little bit about the SRF. Our site is near Augusta; there are other nuclear sites around us. There are 9,000 people onsite, primarily DuPont. The site is a large natural resource with a large forest and an active timber program. We have forestery research underway to show how you can grow bigger and better pines. There are many natural wetlands on the site which must be reviewed in accordance with the Wet Lands Act. We have wildlife and we have programs to manage them. The areas where we use soil for backfill we have to rebuild. The site was declared a National Environmental Research Park by the Agency, consequently we have research going on in many different areas. Now to get down to business, I am just going to summarize our program. We started a solid waste management program several years ago and we were very much involved with the South Carolina Department of Health Environmental Control. The bottom line is that there are two parts to RCRA one of them is nonhazardous and that is a pretty good program in itself. We keep focusing on hazardous wastes, but we must also consider the nonhazardous waste sites also. You have to manage your nonhazardous waste, and not just a hazardous waste program. When the Agency decided that we were going to be our own regulator, the Savannah River Plant already had quite a few years of working with the State and it is very hard to put the brakes on that. It is one thing to not start regulatory compliance efforts, but it is another to stop them abruptly. So we have had a very sensitive situation on how we interact with South Carolina, and essentially what we have been doing is work the same as we have, very closely, just do it has a matter of comity. 101
Under our effort to meet the requirements and intent of RCRA, we have expanded our monitoring program and have identified some major ground-water contamination problems. We have not been able to develop detailed information yet. We evaluated how we were going to do it, and decided that there were enough areas that we had to develop ^n organized approach on how we are going to handle ground-water protection. In addition, there is a concern about Superfund implications; there are requirements from the Drinking Water Act. The state of South Carolina has informally told us that "If you look at the Safe Drinking Mater Act and the Clean Water Act, they do not need RCRA." We have reviewed these Acts and you can interpret them in such a way that they can get access to the ground water through several different laws. We have separated our program into two categories, solid waste and hazardous waste. Our comment to Headquarters was the present Order only addresses one, what are we going to do with the other one? I understand from Vince, they are working on an Order or modification. Eventually they both will have to be covered, otherwise you really end up jeopordizing your position with the State. We are presently at our site also developing a formal RCRA program and we need the Memorandum Of Understanding. The costs we are looking at are from $20-100 million that we think we are going to need in the next 10 years, or 20 years the way we have been getting our funding. The major problem areas that we have are again chlorinated hydrocarbons from seepage basins. We have to evaluate each seepage basin, its migration, and characterize the contents. That is a big program and we won't have enough data available for probably a couple of years. With mercury, we have our work there cut out several thousand pounds of mercury in seepage basins. wastes and what to do about radioactively contaminated inactive waste disposal sites, Superfund, we think may DOE.
for us. We have The questions of mixed hazardous waste, really be a problem for
Right now, one of the biggest areas we are having trouble with is funding. Maybe the panel can discuss that, but we are having a heck of a time getting sufficient funds for high-risk, high-priority projects. And I think that is going to jeopordize many moves that the Agency may take. As to our ground-water protection program, DuPont and our contractor will have a report in January. It has received national recognition. The media immediately picked it up, and there are a lot of people waiting for just what this report is going to say. We have talked to South Carolina about it.
Vincent De Carlo In reference to DOE Order 5480.2, regarding the current status - I always indicate that is is going to be signed in two weeks, and I will say it again. It is going to be signed in two weeks. Honestly, I do think that it 102
is close to being signed/ ' It is currently with the Office of Management and Administration (DMA) with just about every objection removed from the current version. It has not been signed yet, but OMA indicates that it will go to Heffelfinger for signature very shortly. Therefore, let's keep our fingers crossed. In undergoing the review cycle, the Order has been renamed and is titled "Hazardous and Radioactive Mixed Waste Management." Consequently, the radioactive mixed waste responsibility that was to be picked up in Order 5820 will probably stay with 5480.2. As mentioned earlier, there has been a commitment on our part to expand 5480.2 to include solid waste, and this is currently under review by Bill Metz at the Brookhaven National Laboratory. We anticipate releasing a new Order in about six months after the first Order is approved. There are several activities underway to help implement DOE Order 5480.2. These are being carried out at ORNL, and include the identification of the pertinent technical requirements contained in RCRA, and then to take those technical requirements and develop a DOE procedural program that can meet those objectives. This will then be released as an Implementation Plan Guidance Document that should be available to all Field Offices within three months. Oak Ridge National Laboratory will also help us review the Implementation Plans. Should a field office not be able to meet a requirement and seek an exemption, we want to look at these exemptions and make sure that there is some uniformity across DOE in terms of the exemptions that are being requested and the basis of these exemptions. Another area of activity with DOE Order 5480.2 is to identify a hazardous waste coordinating group at Headquarters. We are requesting names of potential members from the Program Offices, and identifying the types of problems that the group will address. The other major area that we are active in is the collection and analysis of hazardous waste data. This is being carried out at LANL under Ken Williamson. One of the things we are doing is establishing DOE's hazardous waste data base. We are also doing some risk assessment work at LANL. The area we are focusing on initially is the heavy metals, and in particular the mixed radioactive heavy metals. That essentially covers all of our activities in hazardous waste management. Now I would like to open the meeting to questions once again for anyone up here in the panel or anyone that is in the audience. Question: Do we want to finalize the draft hazardous waste order, or wait until we get it back from EPA? (a) This Order was issued on December 13,1982.
103
Answer: We would like to finalize it as rapidly as possible. They have not even seen the version that we have at the present time. But with the Order in final form, DOE's hazardous waste program will be implemented. Question: I have a question on the Implementation Plans. One of the questions that was alluded to earlier is the cost to implement the hazardous waste control program and also for remedial action that might be present. I am wondering to what extent the Headquarters Offices are attempting to identify the costs for these programs and how it is going to be done. What I am afraid of is that there is only so much money, and where we have historically spent the money on radiological protection, now that will have to compete with the nonradiological programs. I am just wondering what assistance or what kind of discussion is going on at the Headquarters level to try and identify the situation and hopefully come up with some resolution. Answer: I do not think we have focused yet on the costs that are going to be required in this activity. The EP very nicely inserted into the Order that the Offices should make the funds available to meet the objectives of RCRA. So that somehow it will have to be resolved - they are not out of the loop, they are in the loop to provide those funds. I think that from the environmental protection effort we can provide support to the overall effort. That is why we want to make sure that people are focusing on what we feel are areas that should be focused on in terms of real requirements under RCRA, and to try and come up with a manageable program. When we look at the Implementation Plans we will have a better idea as to how bad the funding situation is going to be for getting this activity underway. But I think that is another step that will soon come.
104
SESSION THREE RADIOACTIVE WASTE MANAGEMENT AND QUALITY ASSESSMENT
3A
QUALITY ASSURANCE FOR ENVIRONMENTAL ANALYTICAL CHEMISTRY AT LOS ALAMOS Ernest S. Gladney, Daniel R. Perrin, and William E. Goode* Environmental Surveillance Group, MS K490 Los Alamos National Laboratory Los Alamos, New Mexico 87545
ABSTRACT The basic structure and philosophy of our program as it has evolved over the past five years is discussed with particular emphasis on traceability and use of certified reference materials. Our typical summary results of the program and our interactive computerized quality assurance system are presented.
INTRODUCTION Quality assurance (QA) seems to be an elusive quarry. The need for a QA program may range from small projects where a single investigator conducts all sampling, analysis, and data interpretation, to the large-scale facilities monitoring programs that require year-to-year continuity and coherence. The former need QA to maintain professional standing while the latter must meet Congressional and/or Agency mandated legal responsibilities. Quality assurance, and particularly how much is sufficient, can become as much a philosophical as a scientific question. Some analysts maintain that the rigorous calibration of their instrumentation with carefully prepared solution or solid standards constitutes a sufficient QA program and that the limit of their responsibility is whether their analysis on a particular sample is accurate with respect to the instrument response at that time. It is our belief that a true QA program is much broader in focus and responsibility. It should apply to all aspects of sample management, beginning at the conception of a project, following through the collection and storage of samples prior to analysis, including the careful calibration mentioned above (which we label quality control), and extending even into the interpretation and final publication of results. The central element of the instrumental •Present Address: DEC, Los Alamos, NM 87544.
107
ELEMENTS H Li Be
B
N
Na Mg
Al Si
Cl Ar
K Ca Sc Ti
Cr Mn Fe Co Ni Cu Zn Ga
As Se Br
Rb Sr
Mo
Sb Te
Zr
Ru Rh Pd Ag Cd In
Cs Ba La Hf Ta W
Os Ir Pt Au
TI Pb Bi
Sm Eu GdlTb Dy Ho Er Tm Yb Lu
Ce Pr Nd Th
Xe
1
U
ISOTOPES
••••
H a
— •
Be 7
No 33 K 40
Rb
•3 •4 Cs 134 137
Cr SI
Mn F»
Co
In
S4
Sr 90
1
Ru 104
131
Ba 140
Ko 33*
Ac 337
Hg
Pb
Bi
Po
30]
310
310
310
<-• M4
lh
Ho
U
Np fu Am
331 3>1 33313» 341 330 33« 337 U M I
OTHER HCQ,, CO,, NQ,t SO4, PO4, Hp~, Qross A#)ha, Gross Beta, Gross Gamma, IDS, Total Altainity, Hardhess, pH, Conductivity figure 1. Analytical Capabilities of the Los Alamos Environmental Surveillance Group 108
analytical phase of our QA program is traceability to the National Bureau of Standards (NBS) or other appropriate agency through the concurrent analysis of matrix-based certified reference materials. LOS ALAMOS ENVIRONMENTAL ANALYTICAL CHEMISTRY SECTION The dimensions of the QA requirements at Los Alamos are a function of our resources and mission. We are a section of only 10 full or parttime analysts (3 of when are professional chemists) within the Environmental Surveillance Group, but we have both a support and a research mission. We endeavor to support a wide array of customers' programs ranging from large-scale overall facilities monitoring of the entire Laboratory, to special projects designed to focus on unique environmental impacts of special facilities (e.g., the Los Alamos Meson Physics Facility), to single investigator, applied research programs in geochemistry or environmental studies (e.g., the redetermination of the cosmic abundance of boron via meteorite analysis). This breadth of customer needs requires us to maintain capabilities to determine almost every element on the periodic table and 38 radioactive isotopes as well as a number of chemical species and established water quality parameters (see Fig. 1 ) . Approximately 30,000 determinations/year are reported. It is this diversity that provides important professional challenge to all our analysts in a business that can become extremely routine if supportservice is pursued to the exclusion of research. We are constantly searching for improved analytical methods to address these demands as well as participating as equal partners in customers' research efforts and initiating independently funded projects wholly within our section. Peer-reviewed publication is equally important as the day-to-day analytical product. QUALITY ASSURANCE MANAGEMENT AT LOS ALAMOS Our ability to address this diverse mission is strongly based on our overall QA management scheme. The current "central nervous system" is a Digitial Equipment Corporation (DEC) POP 11/34 using the RSX-11M operating system (version 4.0). Peripherals include an RA80 121 Mbyte fixed media disc drive, 3 RL02, 10 Mbyte removable cartridge disc drives, 8 keyboard terminals, two magnetic tape drives and two floppy disc drives. A variety of analytical instruments are also interfaced directly to the computer (Fig. 2 ) . Our overall QA operational scheme is shown in Fig. 3. Our in-house software and data bases have been structured around the DEC Datatrieve11 (version 2.0) query, report, and data maintenance system that permits each user to tailor data bases and interactive software to suit his particular needs. A fundamental tenet of our philosophy is that analytical service work is not a linear transfer of samples from customer to analyst and analytical results from analyst to customer. Continuing interaction between these parties is required. Ideally, the
109
POP 11/34 CPU
MEMORY 256KB
/LA-120/
RA-80 DISC DRIVE 121MB
I
UMBUS RL-02 DISC DRIVE 11 MB
D2 11
/
TBMHAIS/
c
4096-Oi\ PHAs J
RL-02
(
RX-01 \ FLOPPY OBC ) DRIVE /
_L
C
PACKARD " \ TOCARB
LSC
RL-02
Figure 2.
DZ 11
)
J
P E
"
]
603 AAj
«H» J
Block Diagram of PDP-11/34 System
analyst is brought into the project at i t s inception and participates i n the experimental design. This gives the customer the benefit of the analyst's perspective on what can be readily determined in a particular matrix, and on sample handling and preservation problems. I t also alerts the analyst to future research needs for new methodology. The analyst then generates computer compatible sample numbers, prints them on stick-on labels, and returns these to the customer. This avoids sample number duplication, eliminates computer incompatible customergenerated numbering schemes, and minimizes transcription errors between customer and analyst. Once samples are taken (this phase should be interactive i f possible), the customer submits them to the analytical chemistry section via an "H-8 Analytical Chemistry Request" multicarbon form (Fig. 4) f i l l e d out during a discussion between customer and analyst. The analyst then transfers this information to the master SAMPLES data base from a keyboard terminal using our interactive program SUBMIT and distributes a color-coded copy of the form to each responsible analyst. SAMPLES is the software heart of our QA system, and has been structured to contain the information shown in Fig. 5. The power of the Datatrieve system becomes important here, for over 30,000 analyses/year flow between i t and our 10 analysts. Datatrieve permits one to search a data base as a function of any single parameter (e.g. FIND SAMPLES WITH SAMPLE-NUM BT 82.08542 AND 82.08587) or any logical combination of parameters (e.g. FIND SAMPLES WITH SAMPLE-NUM BT 82.01421 AND 82.01430 AND [ANALYSIS = "CS" OR ANALYSIS • "HF"]). These data can be transmitted back to the analyst via video terminal or hardcopy line printer. Using logical combinations of simple Datatrieve commands, the analyst can get a real-time assessment of what has been
110
Customer
Analyst /
\
\
Results to 5 Customer
Discussion i
\
/
3
\ Request Form
Samples to Lab
\
2
( Enter )
( Submit)
( Report
Data Reduction Codes
/
CVS REFS
/
/
/
-
SRMS
Figure 3.
Block Oiagram of Overall Quality Assurance Program
111
H • f ANALYTICAL CHEMISTRY REQUEST Los Alamos National Laboratory
3391
SAMPLE NUMtEMS:
Moutrroft MOJECTNUKWER SAMPLE OWNEHIO OATE TOTAL N».«< SAMPLES MATERIAL TYPE
WMfrll or O n * KMtWlml m» Aliquot** cndreoorml
AnalvK
•PECIAt INFORMATION: h Hmplt dinolution required? {clue* ontl
D v«
DNO
Otlwr mmpkt ctm't'on required? liltcmtKilv) Oth« infwmnion: FomiNo H I WAI I
Figure 4. H-8 Analytical Chemistry Request Form requested, what remains unfinished, where samples are located, what procedures have been suggested for certain determinations, etc. This eliminates the necessity for each analyst maintaining his own hard-copy record system, and offers the analytical manager quick access to the overall status of the section. Many of our analytical instruments store their raw data directly on the PDP 11/34 discs for off-line reduction and in a few cases the entire data generation process is now computer controlled. The analyst has a battery of data reduction programs which in most cases generate Datatrieve executable command files that input final analytical results directly into SAMPLES upon analyst approval. For those instruments where raw data are still taken in the form of chart recordings, the analyst uses the interactive ENTER program to transfer his final calculated values back to SAMPLES. When a set of samples has been completed for a given analysis, the analyst uses the REPORT program to generate an "H-8 Analytical Results" report form which the analyst and customer then go over together. Ideally, the analyst follows the data through interpretation and the writing of the final publication. We have several program tools available for sample management that formalize common sets of logical commands mentioned above. SCHEDULE
112
SAMPLE-INFO SAMPLE-NUM OWNER MATRIX ANALYTICAL-INFO ANALYSIS REQUEST-NUM REQUESTOR REQUEST-DATE PROJECT-NUM ANALYST TECHNIQUE SYMBOL RESULT UNCERTAINTY UNITS COMPLETION-DATE STATUS ANAL-COMMENT Figure 5. Structure of SAMPLES Database
permits one to have an overall look at the current section backlog of unfinished work as a function of analyst, requested determination, sample number, request number, or matrix. WORKCOMP is a utility which facilitates rapid generation of reports for higher management concerning the number and type of determinations requested or completed during a given period of time, including cost code information. This program may also be used without modification to summarize the unfinished backlog for the same parameters. The "QA" utility and associated overview (a powerful Datatrieve tool for linking independent data bases) permits the analysts or section manager to assess our success in meeting the analytical QA goals to be described shortly. This can be done as a function of analysis, matrix, completiondate, etc. ANALYTICAL QUALITY ASSURANCE PHILOSOPHY Our approach is based firmly on the concept of traceability through concurrent analysis of matrix-based certified QA materials. Traceability is a hierarchical system developed largely for regulatory purposes whereby transfer of accuracy among various components or institutions of a measurement system is established. Uriano and Gravatt (1977) have examined this concept in detail. Taylor (1981) has recently pointed out that this process has its limitations, since the achievement of an acceptable result on an SRM does not guarantee a unique foundation for the result. Traceability may more often be based upon inference than firmly established scientific fact. Nevertheless, we strongly prefer to maintain traceability to NBS, the US' national standards laboratory, which in turn maintains traceability to international metrology standards where applicable. We cannot achieve this in all cases since NBS has limited resources for certification of matrix-based environmental Standard Reference Material 113
SRMS: SRM-ENTRY MATERIAL ELEMENT CONC UNCER UNITS COMMENT DIS-METH ANAL-METH REF
REF-CODE REF-NUM REFS: REFERENCE CODE N DESCRIP1 DESCRIP2 DESCRIP3 DESCRIP4 Figure 6. Structure of SRMs and REFS DataBases
(SRM) for as many elements as we need to measure. This has led us in two directions: (1) the collection of a large library of reference materials from other national and international agencies (US Environmental Protection Agency, US Geological Survey, Environmental Measurements Laboratory, Canadian Geological Survey, International Atomic Energy Agency, Centre de Recherces Petrographiques et Geochimiques, and the Japanese Geological Survey) in addition to the available NBS biological and environmental SRMs; and (2) the compilation of all available literature data on these reference materials so that "consensus" values for elemental concentrations not agency certified may be developed. We now have over 270 reference materials, some of which have certified or consensus values for as many as eighty elements. Using Datatrieve-11, we have created two data bases (SRMS and REFS) to permit random collection of literature data on reference materials (see Fig. 6 ) . At any time this collection may be sorted as a function of any of the internal elements and printed for our determination of a consensus value. This process is described in much more detail in the references. Our ultimate goal is to develop a program COMPILATIONS, which will automate this now laborous statistical process (Gladney et al., 1979; Gladney, 1980A-C; Gladney, 1981; Gladney and Goode, 1981; Gladney et al., 1981A-B; Gladney et al. 1982A-
•TJ— Some agencies recommend the splitting and analysis of duplicates of 5% of ones samples to be sufficient QA. We believe that this approach is grossly inadequate, providing only an assessment of precision, not accuracy. Our alternative is concurrent analysis of reference materials of similar matrix with every batch of samples. In this fashion both precision and accuracy may be assessed and traceability to a
114
national standard laboratory maintained. The DOE QAP program, managed by EML, is a necessary but not sufficient component of our QA philosophy, since these samples appear too infrequently to insure that day-to-day QA is maintained. We also participate in an EPA-Las Vegas program that distributes radioactivity samples on a somewhat more frequent basis, but these are at low concentrations. We strive to run a minimum of 10% of all analyses as matrix-based QA reference materials. For some determinations (e.g. isotopic Pu) this has proven to be unattainable due to the lack of c e r t i f i e d materials and/or cost. The a v a i l a b i l i t y problem has now been addressed by new NBS Radioactivity in Sediments SRMs. Originally, we f e l t that we wanted the mean of our experimental determinations on reference materials to be within +/- 10% of the c e r t i fied/consensus value before we would consider an analysis to be under control. This is expressed symbolically in Eq. 1: (])
Xe - 7 C ' < O..XC
where "Xe and T are the experimentally determined and certified/consensus mean elemental concentrations, respectively. For routine environmental monitoring this may be adequate, but it is too inflexible when the needs of our research mission are considered, especially when measurements near our analytical detection limits are involved. We are modifying our approach and are searching for some formula that would calculate acceptable control limits as a function of the nearness to the instrumental detection limit, thereby broadening the limits of acceptability as the detection limit is approached. This is shown in Eq. 2: X e - Xj.1 < f(DL)Xc
(2)
where f(DL) is our "magic" detection limit based formula. Additionally, many of the consensus values have single standard deviation uncertainties that differ significantly from +/- 10%. We now consider our analyses under control whenever the absolute value of the difference between our mean and the certified/consensus mean to be inside the propagated single standard deviation uncertainty of the experimental and certified means. This is formulated in Eq. 3: \je - \ \ < \/ ( S e ) 2 + ( S ^
(3)
where Se and S are the standard deviations associated with Xe and X_, respectively. An additional approach for assessing overall performance on a variety of QA materials over longer time_frames (e.q. annually) involves the calculation of mean ratios of Xg/Xc as a function of analysis and matrix type. These data are summarized as a mean +/- one standard deviation for each determination in each matrix type as follows:
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(4) Figure 7 is a brief extract from our annual QA reports showing the results of this approach.
URANUM QUALITY ASSURANCE 1978-1981 1978 1979 1980 1981 Figure 7.
Sicate
Water
BMogfcal
1.0Q±0.07 0.99*0.06 1.00*0.08 1.00 ±0.02
100 ±0.02 1.01 ±0.03 0.98*0l04 1.01 ±0.04
189 ±0.61 1.19 ±0.40 1.02*0.09 1.02 ±0.05
Sample of 4-Year Quality Assurance Record
The "QA" computer program is being written with the above constraints in mind so that the analytical manager can assess the section's success in meeting these c r i t e r i a in a real-time fashion. Following specification of the matrix, analysis, and completion-date time period, this Datatrieve based program searches SAMPLES for both sample and reference material data, determines the per cent of total analyses which were of QA materials (nominal target = 10%), and calculates the quality parameters shown in Equations 3 and 4. This information is summarized and printed for action by the analysts or supervisor. Complexities of instrumentation and analytical methodology impose additional needs. For instance, within the general technique area of atomic absorption the interferences among flame, graphite furnace, hydride generation, and cold vapor are s u f f i c i e n t l y different that the same QA material may not be adequate to address all sub-categories within an overall analytical technology. For example, the EPA-Cincinnati o f f i c e distributes samples for nonradioactive QA analysis on an asrequested basis. These materials can sometimes impart a false sense of QA for atomic absorption analyses. These water samples for trace inorganic analysis usually do not contain Ca, Mg, etc. which create important interference problems with many analyses on the EPA primary pollutant l i s t . The NBS SRM water does contain reasonable levels of these interference producers and we have found that many of our procedures can be adjusted to produce agreement with either one of these QA materials, but not both simultaneously. Therefore, in the production and use of reference materials, both agencies and analysts must realize the importance of matrix matching of interferences as well as the elements of analytical interest.
116
QUALITY ASSURANCE REPORTS Since QA has both a legal and professional dimension, we have begun preserving the individual data in addition to overall performance summaries in annual Quality Assurance Reports (Gladney et a!.t 1981A-B; Gladney, et al., 1982A). Figure 7 shows an extract from one summary table over the four years of our present program. We feel that data of this nature demonstrate that the vast majority of our work is accurate and within the control criteria described earlier. The "QA" computer program also assists with the preparation of these reports. It produces finished tables of current certified/consensus values from CV-DESC and CVS, including the source for each data point from REFS. Summary tables using Eq. 4 and tables of all experimental measurements made on reference materials during that year are also produced. CONCLUSIONS Some agencies are attempting to legislate QA through the promulgation of analytical methods regulations, through licensing requirements for individual analysts, and through laboratory certification programs. In our opinion, the rapid pace of technological change in chemical instrumentaion, the wide variety of sample matrices, and the broad dispersion of procedures throughout the literature makes it difficult to establish a definitive method for any given determination that is not soon outdated. Licensing and certification address the qualifications of personnel and the adequacy of facilities on an infrequent basis at best, and not the question of day-to-day data quality. The development and implementation of a detailed, yet flexible, QA program is a viable alternative to the legislative approach of QA management. This permits each laboratory to utilize its existing expertise and facilities, to employ a variety of methods in the solution of every changing problems, and to encourage continuing innovation and adoption of the latest technology. ACKNOWLEDGEMENTS We are indebted to the other analysts in our section who strive to constructively criticize and implement our program: Richard Peters, Daryl Knab, Colleen Burns, Wanita Eberhardt, William Schweitzer, Diane Noveroske, Richard Robinson, and Carolyn Macdonell. We appreciate the continuing support and encouragement of Wayne R. Hansen and Alan Stoker, our Group management. James Owens, now with the Eagle-Pitcher Co., Miami, OK, figured prominently in the initial phases of developing our concepts. CREDIT This work is performed under the auspices of the US Department of Energy.
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REFERENCES Gladney, E. S., D. R. Perrin, J. W. Owens, and D. Knab. 1979. "Elemental Concentrations in the United States Geological Survey's Geochemical Exploration Reference Samples - A Review." Anal. Chem. 51:1557-7569. Gladney, E. S. 1980A. Compilation of Elemental Concentration Data for NBS Biological and Environmental Standard Reference Materials, Los Alamos Scientific Laboratory Report LA-8438-MS. Gladney, E.S. 19808. Compilation of Elemental Concentration Data for Fourteen Canadian Certified Reference Materials Project Standards! Los Alamos Scientific Laboratory Report LA-8382-MS. Gladney, E.S. 1981. Comparison of Methods for Calculation of Recommended Elemental Concentrations for Canadian Certified Reference Materials Project Rock Standards SY-2, SY-3, and MRG-1. Los Alamos National Laboratory Report LA-8770-MS. Gladney, E. S. and W. E. Goode. 1981. "Elemental Concentrations in Eight New USGS Rock Standards: A Review." Geostandards Newsleter. 5:31-64. Gladney, E. S., J. W. Owens, T. C. Gunderson, and W. E. Goode. 1981 A. Quality Assurance for Environmental Analytical Chemistry: 1976-1979. Los Alamos National Laboratory Report LA-8730-MS. Gladney, E. S., W. E. Goode, 0. R. Perrin, and C. E. Burns. 1981B. Quality Assurance for Environmental Analytical Chemistry: 1980, Los Alamos National Laboratory Report LA-8966-MS. Gladney, E. S., D. R. Perrin, C. E. Burns, and R. A. Robinson. 1982A. Quality Assurance for Environmental Analytical Chemistry: 1981, Los Alamos National Laboratory Report, in press. Gladney, E. S., C. E. Burns, and I. Roelandts. 1982B. "1982 Compilation of Elemental Concentrations in Eleven USGS Rock Standards." Geostandards Newsletter, in press. Taylor, J. K. 1981. "Quality Assurance of Chemical Measurements," Anal. Chem. 53: 1588A-1596A. Uriano, G. A. and C. C. Gravatt. 1977. "The Role of Reference Materials and Referesnce Methods in Chemical Analysis," Critical Reviews in Anal. Chem. 6: 361-411.
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3B
QUALITY ASSURANCE IN ENVIRONMENTAL MEASUREMENTS
T. W. Oakes Department of Environmental Management Industrial Safety and Applied Health Physics Oak Ridge National Laboratory* Oak Ridge, Tennessee 37830 ABSTRACT The following elements of a quality-assurance (QA) program as applied to environmental surveillance activities are presented: (1) a philosophical and conceptual framework for QA, with a detailed assessment of the sources of uncertainty in a monitoring program; (2) the requirements for the formulation of general and technical procedures of quality control; (3) the environmental QA activities implemented at Oak Ridge National Laboratory (ORNL), including details on record keeping, data reduction and compilation, auditing, analytical procedures, inter-laboratory sample comparisons and data interpretation; and (4) the role management must play to ensure a successful program. The QA principles developed here may be applied to any surveillance program.
INTRODUCTION Quality assurance (QA) is an activity. It is a planned and systematic action that is applied to, and functions along with, any other activity or sequence of activities. But it is more than an activity; it is a philosophy [Oakes et al. (23-26)]. Quality assurance (QA) can be defined as the planned and systematic actions necessary to ensure occurrence of techniques and analyses by determining errors and minimizing them [Linch (13) and Sanderson et al. (28)]. QA is also the summation of all programmed events imposed internally and externally in order to ensure that data being generated by a surveillance program are as meaningful as possible [Sanderson et al. (28)]. A QA program should develop and implement procedures that will facilitate proper identification and evaluation of problem areas in order to give the user confidence that the system or component will perform satisfactorily. Many scientific and technical organizations and governmental agencies are committed to the •Operated by Union Carbide Corporation under contract W-7405-eng-26 with the U. S. Department of Energy. By acceptance of this article, the publisher or recipient acknowledges the U.S. Government's right to retain a nonexclusive, royalty-free license in and to any copyright covering the article.
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continuous development of QA standards and procedures [Oakes et al. (24)]. QA exists at the interface between the performer (person, instrument, or machine) and the performance (action, output, or product). As an activity, QA requires a purpose (policies and objectives), scope (planning and programs), procedures (tasks and methods), and audits (tests and verification); as a philosophy, it requires criteria (standards and values) and commitments (dedication and resources). Thus, QA is broad—it is a complete system. The priority objective of an environmental surveillance program is producing quality data. The user of this data expects precisely that. The demand for perfection in environmental data requires that all uncertainties be identified. Quality awareness and the control of quality must start at the planning stage (initial stage) of an environmental surveillance and ends only once the data has been reported (final stage). Quality control is extended into the sampling design phase during the planning. This will be accomplished through a complete quality assurance system establishing quality goals at each stage by implementation of a quality control plan. The product of an environmental surveillance program is data [Oakes et al. (23, 25)]. Therefore, QA applied to an environmental surveillance program is defined in terms of that data and concerns itself with measurement. Measurement involves uncertainties and many of these uncertainties cannot be eliminated completely. Thus, the task of QA is not as much to eliminate uncertainties as to recognize all the uncertainties and whenever possible, to reduce them to acceptable limits. The theory of errors and methods for detecting and preventing errors in environmental measurements have been discussed in detail [Oakes et al. (25)]. A QA Program should develop and implement procedures that will facilitate proper identification and evaluation of problem areas in order to give the user confidence that the system or component will perform satisfactorily. Development of a workable QA program requires the formulation of general and technical specifications and procedures for quality control (QC) and assurance of conformance to these procedures. Quality control involves those QA actions which provide a means to control and determine the characteristics of measurement equipment and processes to meet established requirements; here QA includes QC [NRC (21)]. Whenever possible, the staff members of the operating program should have a part in the development of the QC procedures so they feel that they are part of the program. Once the QA program is structured, it should be integrated into the operating program, and a timetable should be established for implementation of audits. The role of an environmental surveillance program is to provide qualitative and quantitative data on environmental pollution levels. Surveillance programs are operated by federal, state, local, and private agencies. The data from these programs are used for a wide variety of 120
purposes, such as the establishment of guides and standards for enforcement activities. A QA program to ensure the reliability of the data is essential because of the importance of these values and the importance of the actions which result from decisions that are made using these values. Therefore, it is imperative that the precision and accuracy of the data be assured in order that policy decisions concerning environmental quality are based on valid and comparable data [EPA (19)]. In cases where standards and procedures have been developed, the primary function of QA is determining compliance with existing specifications and acceptance criteria. Since environmental surveillance programs sample dynamic systems which include biological, climatic, and human-made changes, an overall QA program must include details concerning the reliability of the sampling program in addition to the usual analytical and equipment QC procedures. Standard QA procedures for a complete environmental surveillance program are essentially nonexistent. In order to show compliance with many environmental regulations, to conform to the policy of keeping radioactivity releases "as low as reasonably achievable," and, above all, to add credibility to the environmental surveillance efforts, it is necessary to develop and implement a complete QA program for environmental surveillance. Two important objectives of a QA program are to promote comparisons of local results with those of other laboratories and to develop cooperative methods with those laboratories and standards organizations. This paper will review the areas that should be considered in developing such a program and discuss the initiation of a QA program for surveillance activities. GENERAL PHILOSOPHY AND QUALITY ASSURANCE CONCEPTS Quality must be defined in terms of the parameter(s) being measured [Linch (13) and Oakes (23, 25)]. The measurement of a physical entity involves uncertainties which cannot be eliminated completely, but when the possibilities of uncertainties are recognized, they can often be reduced to acceptable limits [Katz (12) and Oakes (24)]. Reduction of these uncertainties can be achieved by proper attention to detail and close control of the significant variables. Control must be related to the source of variation, which can be either random or systematic in nature [APHA (27)]. Examples of unrecognized sources of uncertainties are interferences in chemical reaction systems and the appearance of undesirable physical or chemical effects. Absolute analytical values are not directly attainable; therefore standards from which the desired result can be derived by comparison must be established. Uncertainties cannot be reduced to zero; but methods, such as standards and cross-checking, are available for obtaining reliable estimates of a given value; thus a range of acceptable errors can be estimated [Johnson (If)].
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Since QA is concerned with detecting determinate errors and preventing their recurrence, a systematic QA program must be followed. Such a program for analytical laboratories should include:[EPA (20)] 1. Good physical facilities and equipment. 2. Adequately trained and experienced personnel. 3. Procedure verification. a. Use of standard methods. b. Routine analysis of control samples. 4. Frequent calibration and servicing of instruments and equipment. a. Calibration to maintain accuracy. b. Correlation of quantitative tests to confirm accuracy. 5. A knowledgeable and supportive management. Evaluation of the effectiveness of QA in a laboratory requires a knowledge of [Linen (13)] 1. Equipment, instruments, and sampling techniques. 2. Expected ranges of analytical results. 3. Precision of the analytical methods. 4. Control charts for the determination of trends and gross errors. 5. Data sheets and procedures adopted for controlling sample integrity in the laboratory. 6. Quality-control results on a short-term and long-term basis. The Nuclear Regulatory Commission (NRC) has published a regulatory guide in which a basic structure for QA applied to an environmental monitoring program is presented [NRC (21)]. This guide is an excellent starting point for the development of a QA program. It also lists a number of useful references. Other helpful reports are available from the Environmental Protection Agency [EPA (6)], the National Council on Radiological Protection and Measurements [NCRP (18)], and Oakes et al. (23-26). ENVIRONMENTAL SURVEILLANCE PROGRAMS Environmental surveillance programs are structured to accommodate specific tasks which are associated with the facility operations, site characteristics, environmental conditions, and potential exposure pathways. Such a surveillance program should include the following operational steps: (1) planning, (2) sample collection, (3) sample storage, (4) sample preparation, (5) analytical processing, (6) radiochemical analysis, (7) sample counting, (8) data manipulation, (9) statistical data treatment, (10) data interpretation, and (11) reporting. An overall QA program must apply QC elements to all operational steps. It is impossible to entirely eliminate all uncertainties associated with a surveillance program, but recognizable errors can be reduced to tolerable limits by a well-designed QA program. This type of complete program was recommended first by Oakes et al. (23). 122
Sources of Uncertainty The first objective in the development of an overall QA program is to identify the sources of uncertainty throughout the environmental surveillance program [Oakes et al. (23-26 )]. Most efforts in the past have stressed only the analytical uncertainties. The following expression, which is an expansion of the work of Johnson (11), Oakes et al. (23-26) and Waite et al. (33) lists other sources of uncertainty. The dots indicate that other sources of uncertainty may be important. T
u
= N +C
u u
+A
u^ V W W " *
+
i + Su + Tu + •
where T u = total uncertainty associated with a surveillance program, Nu = uncertainty associated with the natural nonuniformity of the sample population, C u = total uncertainty associated with collection strategies, s u = error associated with a limited sample size, lu = uncertainty associated with the choice of sample location with respect to the source, f u = uncertainty associated with the sample collection frequency, A u = total uncertainty associated with analytical processing, stu = variability associated with sample storage, P u = variability associated with sample preparation, al u = variability associated with aliquoting, m u = uncertainty associated with instrument variability, co u - estimate of error associated with counting statistics, 0 u = total uncertainty associated with data treatment and interpretation, S u = total uncertainty associated with sample separation, T u = total uncertainty associated with time of sample processing.. Natural Nonuniformity (Ntl) Probably the greatest source of variability is in the natural distribution4 0 of radionuclides. For example, in studying the distribution of natural K i n soils, it was noted that the values ranged over two orders of magnitude in central and eastern Tennessee [Oakes (22)]. This large variation was found to be due to the multiple geologic formations in central and eastern Tennessee. There are many other factors that influence the distribution of radionuclides in the environment [Corley et al. (4)], such as weathering and sedimentation of soil and climatic conditions. Radionuclide concentrations will also vary with depth. Further, side-by-side sites may experience different rainfall or wind patterns (perhaps due to a local obstruction). These factors must be analyzed simultaneously along with the collection strategy. 123
Collection Strategy (C N ) The objective of sample collection (Cu) strategy is to obtain a portion of material small enough in volume to be easily transported to and processed through the laboratory while accurately representing the material being sampled. Samples collected by poor procedures produce poor results. It is known that accuracy and precision of the analytical procedures in the laboratory cannot compensate for an inaccurate sample. Often, it is much more difficult to formulate and validate sample collection procedures than those related to analytical processes. Sampling frequency is limited by available time, sampling method, and/or sample process. Proper methods often require long sample periods for collection of a suitable sample. Short sampling time in turn increases frequencies and reduces the accuracy. In some cases, compositing techniques for very short collection times are requested to obtain valid sampling averages. For sampling programs that require continuous collection, the sampling time must be sufficiently long to average out normal discharge variations [Murarka (16), Bernhardt (2), and IPDWR (9)]. One of the important considerations in the determinations of sample size (s u ) is the desired minimum detectable concentration (MDC) of the radionuclide of concern. In addition, in the formulation or planning stage of a surveillance program, sampling location (l u ) and collection procedures are established. Procedural variations are often necessary in order to collect representative samples in those circumstances where the original procedure might yield a nonrepresentative sample. Selection of approximate sampling locations should be based on air and water circulation patterns, land and water usage, population distribution, and availability of land. In selecting sampling locations one must avoid local concentrations of natural radioactivity, building wake effects, dripping and puddling of precipitation, heavy dust-raising activities, backwater areas in streams, riverbank springs, and atypical vegetation patterns. One solution to the problem of selecting representative sampling sites and decreasing the location (l u ) uncertainty is to increase the number of sampling stations. Finally, in order to reduce (fu) to a minimum, it is necessary to tailor the collection frequency to the expected variations in discharges, to the half-life of the released radionuclides, and to the nature of the environmental media. Sources of sampling collection information are HASL Procedures Manual (7), Environmental Radioactivity Surveillance Guide (14), Nuclear Regulatory Commission (NRC) Regulatory Guidelines (21), and Recommended Methods for Water-Data Acquisition - Preliminary Report of Federal Interagency Work Group on Designation of Standards for Water-Data Acquisition, [Oakes et al. (22), Watson et al. (33)]. 124
Analytical Uncertainty (An) Quality assurance procedures in the radiochemistry laboratory are designed to complement the overall objectives of the environmental surveillance QA program. The sampling regime produces a wide variety of materials (soils, vegetation, water, air filters, etc.)» each requiring specialized procedures for analysis and each having an analytical processing (A u ) uncertainty. One of the best ways to reduce A u is to use standardized methods, such as those of the American Society for Testing and Materials (ASTM) (1), American National Standards Institute (ANSI) (30), National Academy of Sciences (NRC Monographs on Radiochemistry) (17), the collection of procedures of the Environmental Measurements Laboratory (7), and Oakes et al. (23-26). One major problem in many environmental monitoring programs is that of inadequate coordination of the sampling and analytical programs. Also nonhomogeneity in sample media contributes possible large uncertainties. For example, replication of 10-g aliquots for plutonium analysis from kilogram soil samples containing discrete particles can easily vary several orders of magnitude (5). The uncertainty associated with sample preparation (p u ), storage (st u ), and aliquoting (al u ) is very difficult to estimate. When a sample is received in the analytical laboratory, it is often not in the proper physical form for analysis. The sample may require concentration (water samples), drying (biological samples), or homogenization before aliquots can be processed by a suitable chemical procedure. The sample must be handled and processed in such a way that no significant changes in composition occur before the analysis is performed. Sample storage prior to analysis requires some thought and care to avoid loss of certain radionuclides or to avoid spoilage or sample decomposition. Generally, the shorter the time that elapses between collection of a sample and its analysis, the more reliable will be the analytical results. It is difficult to state exactly the maximum allowable time between collection of a sample and its analysis; this depends on the character of the sample, the particular analysis to be made, and the conditions of storage. Nondestructive methods of analysis should be used in all applicable situations. Gamma-emitting radionuclides are determined in many surveillance samples by the use of high-resolution gamma-ray spectrometry with a minimum of sample treatment. Calibration sources for the spectrometer can be obtained from the National Bureau of Standards (NBS) to reduce measurement uncertainty (m u ). One of the most useful calibration sources is the NBS Mixed Radionuclide Gamma-Ray Emission-Rate Standard (SRM-4216) that is issued annually. Other extended sources in the series are described by Coursey (5). Many calibrations for natural radioelement determinations are performed with standards from the New Brunswick Laboratory, now located at Argonne National Laboratory. Many environmentally important radionuclides (^^Sr, 99f C> 239pu> etc.) require specific radiochemical separations prior to their final 125
determination by counting or spectrometric measurements. Careful attention to fundamental principles by the chemist performing the separations is another facet of the QA program. Among these principles are complete dissolution of the sought radionuclide from the sample matrix, complete exchange with the isotopic tracers used for yield determinations, and complete removal of interfering radionuclides in the final isolation step. Quality assurance is provided by technician training, use of replicate samples, use of reagent blanks, use of standard samples, use of quality control "blind" standards, and the participation in "round-robin" exercises. Among the important interlaboratory comparison programs useful in this connection are those of the National Bureau of Standards (3), Environmental Protection. Agency (10), and the distribution provided by the Department of Energy's Environmental Measurements Laboratory (8). In addition to specific quality-assurance techniques for general analytical procedures, quality control must be exercised on all instruments used for radioactivity measurements (34). Radioactive decay, by its fundamental nature, is a random process resulting in uncertainty due to counting statistics (co u ). Matuszek (15) presented quality-control charts for the background and standard deviation values for a commercial low-background beta-particle counting system showing significant differences from predicted statistical behavior. An important concept for all environmental programs is that of minimum detectable activity (MDA). Lochamy (14) provided a detailed discussion and mathematical derivation of four different types of MDA implications. Data Interpretation (Du) The data treatment (Du) incorporates all the sources of uncertainty. Matuszek (15) indicated that some laboratory staffs normally report all errors. The treatment of environmental data is complex in that skewed and mixed distributions, along with the presence of "outliers" and "less than" values are generally prevalent. Generally, when these "out-of-fit" values are examined, it is found that the majority are obvious computation, measurement, or processing errors. One of the most common problems facing environmental surveillance programs is the problem of handling "less than detectable (LD)" values. To be conservative, many organizations make the assumption that all LD values actually have concentrations equal to the detection limits, and the group average is computed accordingly; other organizations set the LD values to zero. Both methods, however, bias the averages given. A more correct way is through the use of probability plotting (18). Space (Sn) and Time (T,,) The uncertainties associated with space and time should be treated cautiously, since these components may be divisible into subcomponents for time (both short and long intervals) and for scale of geographic association with the facility being monitored (local versus regional). Investigations of these uncertainties result in a continuing risk that the 126
differences between environmental variance locally and regionally may be wrongly assigned, biasing the apparent impact of local discharges [Waite et al. (33)]. Both of these uncertainties involve the differences between environmental variance locally and regionally. Seasonal changes may cause biasing in data interpretation, RECORD KEEPING FOR ENVIRONMENTAL DATA The importance of record keeping cannot be overemphasized. It is important not only from a QA standpoint, but also from a regulatory point of view, for legal proceedings, for visualizing changing trends in the environment, for optimizing environmental monitoring procedures, for research studies, and for other purposes. Record keeping for environmental monitoring at ORNL has been in existence since the formation of the Laboratory (approximately 37 years). Permanent records are kept of routine monitoring of radioactivity present in the environment, and routine assaying for both specific and nonspecific radionuclides present in a wide variety of samples (e.g., biological, soil, rainfall, etc,). Most routine environmental data are stored on computer printout. Since many samples require rather extensive analyses, specific treatment of each sample for sequential analysis is done by various groups at ORNL. Samples must be labeled properly when being routed through these groups in order to maintain an accurate record for each sample. Special forms and data cards have been developed which greatly facilitate accurate record keeping. The advantages of these computer cards are: 1. The card is routed along with the samples each step of the way. 2. All information related to the sample is on one card. 3. The information can be easily computerized. 4. Short-term turnaround for data is provided for all samples in the system. 5. Traceability of samples as they are routed through the system is provided. 6. Documentation of data related to the sample is provided. COMPUTER AID FOR MONITORING SAMPLE FLOW A computerized scheme for monitoring the sample flow has also been developed at ORNL (32). The monitoring of sample flow involves simply >">eping track of the various samples as they progress through all steps, this is an important aspect of QA, because sample loss and mix-up has been a major concern in the past facing the surveillance activities at ORNL. Cards corresponding to collection, sample storage and preparation, analytical processing and analysis, and data interpretation are used. Those specially designed cards are turned in to the sample coordinator following completion of each step. Special sample coding has been established for this program to expedite sample flow. For example, the collection card has spaces for the sample number step ("C" for collection); type of sample ("AF" for air filter); count rate of the air filter on and off the pump in counts per minute (CPM); routine or special 127
sample designation; techniques used (continuous, grab, etc); and initials of the person performing the particular operation. After this information has been transferred onto computer cards, it is processed and assembled on a computer printout. From this printout, one can determine which steps in the sample flow have been completed. A warning signal has been built into the program: when the time in any one period exceeds a preset time for each sample, that sample is flagged. This is very useful for samples with short-lived isotopes. This program has proven to be a very useful tool for the management of sample flow. QUALITY ASSURANCE IN THE ANALYTICAL CHEMISTRY DIVISION (ACD) The radiochemical methods used at ORNL have been tested and documented. New equipment has been acquired, featuring on-line data reduction, which has decreased turnaround time for gamma spectrometry and improved reliability. The germanium-lithium (GeLi) detectors have been recalibrated. The internal consistency between detection methods has been tested by sample comparison using independent measurement methods. Improvements in the analytical QA program are planned through increased documentation and a more rigorous routine QA schedule.
AUDITS Written procedures and scheduling plans for audits are scheduled annually, and audits are performed in accordance with these procedures. The procedures include such items as checking the air monitors bi-weekly, analytical balances semi-annually, how to take and prepare soil samples to avoid cross-contamination, etc. The Environmental Management staff at ORNL is committed to taking the action necessary to correct the deficiencies identified by these audits. The audit structure is divided into three sections: (1) procedures and documentations, (2) special investigations, and (3) maintenance schedules. The structure and scheduling are reviewed monthly to identify areas where changes are needed. When the audit information is evaluated, analyzed, and tabulated, it is passed on to management so that the appropriate action can be taken. The QA coordinator keeps track of the information received from the audits and makes sure that corrective action is taken. DATA TREATMENT Plotting data on probability paper is a very good technique to ca1cu1','b averages and standard deviations, to quickly show whether the distribution choice was correct (i.e., Gaussian vs. log normal), to determine whether data belong to one of the two statistical populations, and to permit the confident reporting of averages when the values are close to or below the detection level.
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To arrive at a more quantitative decision, however, one must employ standard statistical tests. These tests are usually performed on the central or mean values of the respective distributions; the most common test is the Student's "t" test. THE ROLE OF MANAGEMENT A strong QA program must include a management system to control operations and specialized techniques to facilitate beneficial decisions. Proper organization for the flow of information is a key factor in ensuring that corrective action is taken in the proper time frame. The management elements and corresponding elements of QA are (1) Planning: quality assurance programs, instructions, and procedures; (2) Organizing: organization; (3) Execution: procedural control, document control, control of purchased materials, equipment and services, and identification and control of sample flow; and (4) Monitoring: inspection, test control, corrective action, and audits. Management needs must identify the program objectives and take the steps necessary to attain these objectives. Management should work together with the working staff to plan and organize the program; this makes it possible for the staff to know what is required of them and gives them the opportunity to participate in the feedback loop. The working staff must be encouraged to identify problems affecting quality. CONCLUSIONS Experience has demonstrated the necessity of a complete QA program for improving environmental monitoring (31). A QA program provides a mechanism for ensuring efficiency in everyday activities. Further, the QA approach could and should be used as a basis for establishing regulatory guides for a monitoring protocol requiring the regulated facilities to collect and analyze samples with comparable procedures. Such a QA program would require regulated facilities to review monitoring procedures and methods rather than collect information for the sole purpose of compliance with regulatory guides. In addition, such a program would be beneficial to regulatory agencies by identifying nonessential collection requirements. Although the development of a QA program could initially increase operating expenses for environmental surveillance, we believe that, once fully implemented, a well-designed program will be cost-effective. For example, ORNL sample loss has been reduced by 25% following implementation of a sample tracking system. Sample loss also is reduced because the QA program requires management to keep track of the QA information received from audits and to make sure any needed corrective action is taken. Also, technical procedures are updated and tested for effectiveness. As part of this process, feedback loops should be
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established and used by all staff members and management for an effective QA program. The QA program at ORNL has identified many areas in the environmental surveillance program that needed improvement and has been very successful in ensuring more reliable data from the environmental surveillance program. The confidence of all groups involved in the surveillance activities at ORNL has been improved, and the program has helped identify common problems. The QA program at ORNL has provided the staff members with a vehicle for better communication about program policy input and has also given management a better overview of the sample flow and the impact of changes in program direction on the quality of environmental surveillance.
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REFERENCES 1. American Society for Testing and Materials, 1977 Annual Book of ASTM Standards, Part 3: Steel Plate, Sheet, Strip and Mire: Metallic Coated Products, ASTM, Philadelphia, 1977. 2. Bernhardt, D. E., Evaluation of Sample Collection and Analysis Techniques for Environmental Plutonium, Report ORP/LV-76-5 (PB-253 960), Office of Radiation Programs, NTIS, May 1976. 3. Colle, R., AIR-NBS Radioactivity Measurements Assurance Program for the Radiopharmaceutical Industry, in Measurements for the Safe Use of Radiation, Proceedings of an NBS 75th Anniversary Symposium Held in Gaithersburg, Maryland, March 1-4, 1976, S. P. Fivozinsky (Ed.), pp. 77-82, NBS Special Publication 456, National Bureau of Standards, Washington, D. C , GPO, November 1976. 4. Corley, J. P., D. H. Denham, R. E. Jaquish, D. E. Michels, A. R. Olsen, and D. A. Waite, A Guide for Environmental Radiological Surveillance at DOE Installatfons, DOE/EP-0023, Pacific Northwest Laboratories, July 1981. 5. Coursey, B. M., Use of NBS Mixed-Radionuclide Gamma-Ray Standards for Calibration of Ge(Li) Detectors Used in the Assay of Environmental Radioactivity, in Measurements for the Safe Use of Radiation, Proceedings of an NBS 75th Anniversay Symposium Held in Gaithersburg, Maryland, March 1-4, 1976, S. P. Fivozinsky (Ed.), pp. 77-82, NBS Special Publication 456, National Bureau of Standards, Washington, D. C , GPO, November 1976. 6. Environmental Monitoring and Support Laboratory, Environmental Radioactivity Laboratory Intercomparison Studies Program, FY 1977, Report EPA/600/4-77/001 (PB-263 900), NTIS, January 1977. 7. Harley, J. H. (Ed.), HASL Procedures Manual, USAEC Report HASL-300, Health and Safety Laboratory, NTIS, 1972. 8. Health and Safety Laboratory, Quarterly Report of the Energy Research and Development Administration, Division of Safety, Standards, and Compliance—Quality Assurance Program, ERDA Report HASL-319, NTIS, May 1977. 9. Interim Primary Drinking Water Regulations: Proposed Maximum Contaminant Levels for Radioactivity, Fed. Regist. (Wash., D. C ) , 40(158:34324-34328 (Aug. 14, 1975).
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10. Jarvis, A. N. and D. G. Easterly, Environmental Radioactivity Standards, in Measurements for the Safe Use of Radiation, Proceedings of an NBS 75th Anniversary'Symposium Held in Gaithersburq, Maryland. March 1-4. 1976, S. P. Fivozinsky (Ed.), pp. 77-82, NBS special Publication 45b, National Bureau of Standards, Washington, D. C , GPO, November 1976. 11. Johnson, L. 0., Elements of Quality Assurance in Environmental Surveillance, in Proceedings of the Third Environmental Protection Conference, ChicagcT, Illinois, September 23-26, 1976, ERDA Report ERDA-92, Vol. 1, pp. 77-88, NTIS, December 1975. 12. Katz, M. (Ed.), Methods of Air Sampling and Analysis, 2nd ed., American Public Health Association, Washington, D. C , 1977. 13. Linen, A. L.s Quality Control for Sampling and Laboratory Analysis, in The Industrial Environment: Its Evaluation and Control, M. 0. Amdur (Ed.), pp. 277-29?, National Institute for Occupational Safety and Health, GPO, 1978. 14. Lochamy, J. C , The Minimum-Detectable-Activity Concept, in Measurements for the Safe Use of Radiation, Proceedings of an NBS 75th Anniversary Symposium Held in Gaithersburg, Maryland, March 1-4. 1976, S. P. Fivozinsky (Ed.), pp. 77-82, NBS Special Publication 456, National 3ureau of Standards, Washington, D. C , GPO, November 1976. 15. Matuszek, J. M., Environmental Measurements and Regulatory Responsibilities, in Measurements for the Safe Use of Radiation, Proceedings of an NBS 75th Anniversary Symposium Held in Gaithersburg. Maryland, March 1-4, 1976, S. P. Fivozinsky (Ed.), pp. 77-82, NBS Special Publication 456, National Bureau of Standards, Washington, D. C , GPO, November 1976. 16. Murarka, I. P., Environmental Monitoring and Statistical Considerations, in Proceedings of the Third Environmental Protection Conference, Chicago, Illinois, September~23-26, 1975, ept( ERDAReport ERDA-92, Vol. 1, pp. 92-116, NTIS, December 197b. 17. National Academy of Sciences-National Research Council, Nuclear Science Series: Monographs on Radiochemistry and Radiochemical Techniques (a series of monographs on the radiochemistry of essentially all the elements and on radiochemical techniques), published by the U. S. Atomic Energy Commission, and later by the U. S. Energy Research and Development Administration, as flAS-NS publications. 18. National Council on Radiation Protection and Measurements, Natural Background Radiation in the United States, NCRP Report 45, 1975. 132
19. National Environmental Research Center, Environmental Radioactivity Laboratory Intercomparison Studies Program 1973-1974. Report EPA/680/4-73-001-B (PB-24O 962), NTIS, February 1974. 20. National Environmental Research Center, Handbook of Radiochemical Analytical Methods, Report EPA/680/4-75-001 (PB-240 621), NTIS, February 1975. 21. Nuclear Regulatory Commission, Regulatory Guide 4.15, Quality Assurance for Radiological Monitoring Programs (NormaT Operations)—Eff1uent Streams and the Environment, 1979. 22. Oakes, T. W., K. E. Shank, and C. E. Easterly, Natural and Man-Made Radionuclide Concentrations in Tennessee Soils, ERDA Report CONF-761031-3, Oak Ridge National Laboratory, NTIS, 1976. 23. Oakes, T. W., K. E. Shank, and J. S. Eldridge, "Quality Assurance Applied to an Environmental Surveillance Program," Proceedings of the American Chemical Society 4th Joint Conference on Sensing of Environmental Pollutants, 226-231 (1978). 24. Oakes, T. W., K. E. Shank, and J. S. Eldridge, "Quality Assurance Applied to Environmental Radiological Surveillance," Nuclear Safety 21_(2):217-226 (1980). 25. Oakes, T. W., K. E. Shank, and J. S. Eldridge, Quality Assurance Applied to an Environmental Surveillance Program, USDOE Report, CONF-771113-8, ORNL, NTIS, 1977, 26. Oakes, T. W., K. E. Shank, and J. S. Eldridge, "Quality Assurance in Environmental Measures," Health Physics 35:920 (December 1978). 27. Oakes, T. W., M. A. Montford, K. E. Shank, E. B. Wagner, T. G. Scott, and J. S. Eldridge, Methods and Procedures Utilized in Environmental Management Activities at ORNL, ORNL/TM 7212 (March 1981). 28. Quality Control in Chemical Analysis, in Standard Methods for the Examination of Water and Wastewater, American Public Health Association, American Waterworks Association, and Water Pollution Control Federation, 14th ed., pp. 26-33, American Public Health Association, Washington, D. C., 1976. 29. Sanderson, C. G., "Quality Assurance for Environmental Monitoring Program," Upgrading Environmental Radiation Data, EPA 520/1-80-012 (1980).
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30. Seidel, C. W. and J. M. R. Hutchinson, American National Standards Institute Quality Assurance Program in Radioactivity Measurements, in Measurements for the Safe Use of Radiation, Proceedings of an NBS 75th Anniversay Symposium Held in Gaithersburg, Maryland, March 1-4, 1976, S. P. Fivozinsky (Ed.), pp. 77-82, NBS Special Publication 456, National Bureau of Standards, Washington, D. C , GPO, November 1976. 31. Shank, K. E., T. W. Oakes, and J. S. Eldridge, "Quality Assurance Applied to Surveillance," in Proceedings of the 1980 UCC-ND and GAT Waste Management Seminar, CONF-800416 (April 1980). 32. Stephenson, R. L., T. W. Oakes, K. E. Shank, A Computer Program for Monitoring Sample Flow from Environmental Surveillance Activities at ORNL, ORNL/TM-6599 (December 1978). 33. Waite, D. A., et al., "Statistical Methods for Environmental Radiation Data Interpretation," Upgrading Environmental Radiation Data, EPA 520/1-80-012 (1980). 34. Ziegler, L. H, and H. M. Hunt, Quality Control for Environmental Measurements Using Gamma-Ray Spectrometry, Report EPA/600/7-77/144 (PB-277 377), Environmental Monitoring and Support Laboratory, NTIS, December 1977.
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3C QUALITY CONTROL ACTIVITIES OF THE HANFORD ENVIRONMENTAL SURVEILLANCE PROGRAM
K. R. Price R. E. Jaquish Pacific Northwest Laboratory Richland, Washington
ABSTRACT A comprhensive approach,to quality control (QC) has been developed by the Pacific Northwest Laboratory'8' for the Hanford Environmental Surveillance Program. The framework of quality control for the surveillance program has been documented in a QC implementation guide wherein QC requirements are specified and specific responsibilities and authorities are described. Subjects in the guide include the collection, analysis, and reporting of samples as well as equipment calibration and maintenance, training, audits, and record keeping. A QC file and library have been established to store pertinent documentation, records, and references for ready access.
INTRODUCTION The Department of Energy's (DOE) Hanford Site in southcentral Washington is a large industrial complex occupying about 1500 km 2 and involving eight independent government contractors. At least four of the contractors are engaged in work that might result in a significant environmental impact. While each contractor is responsible for protecting the environment within and around its own operating area, the Hanford Environmental Surveillance Program is responsible for assessing environmental impacts and for calculating potential radiation doses to members of the offsite public in accordance with DOE orders and other applicable regulations and standards. The surveillance program is a function of the DOE's Pacific Northwest Laboratory (PNL) operated by Battelle Memorial Institute. The PNL personnel collect samples, analyze data, and report results with most of the analytical work performed by the U.S. Testing Co., Inc. (UST) under subcontract.
(a) Pacific Northwest Laboratory is operated by Battelle Memorial Institute for the U.S. Department of Energy Under Contract DE-AC06-76RL0 1830
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The surveillance program consists of routine activities to assess potential offsite radiation dose, verify the onsite containment and control of radioactive waste, and assure the public that site activities are monitored and that potential impacts are identified and assessed. The routine surveillance program is not "research" in the sense of hypothesis testing, and as such does not enjoy the luxury of built-in quality control (QC) that is inherent to research projects. The QC needs of the surveillance program are unique and entail much more than assuring the precision and accuracy of numerical values. An holistic approach to QC has been developed for the llanford Environmental Surveillance Program to assure quality at all levels of work. QUALITY ASSURANCE AND QUALITY CONTROL The term quality assurance (QA) identifies the overall program used to assure the quality of a product. Quality control involves those procedures and activities used to implement the QA program. At PNL, QA is organized through the Quality Assurance Office according to DOE Order 570Qt6A and the local directive RL 5700.1. Requirements and specifications within PNL are published in a QA manual by the QA Office, and each project or activity is covered by a specific QA plan also prepared by the QA Office. Quality assurance pl?ns relate the general requirements given in the QA manual to the specific project or activity and indicate the degree of QA that is necessary. Operating or line organizations are responsible for implementing their own QA plans. The QA office audits compliance with the QA manual and QA plans. The PNL subcontract to UST for analytical services requires that UST have a QA program and the contractor has organized and published its own QA and QC manuals. QUALITY CONTROL IMPLEMENTATION A sound QA program must have a convenient method to implement quality control. The following diagram illustrates the method used for the Hanford Environmental Surveillance Program. PNL QA Manual Environmental Surveillance QA Plan (Implementation)
/
V
Technical Criteria and Procedures
QC Implementation Guide
Work Performed The Technical Criteria and Procedures document is a formal statement by the Program Manager describing the criteria and general technical procedures
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(including QC) to be used in the operation of the program, whereas, the QC implementation guide consists of a series of specific procedures on how to implement QC. The guide also outlines specific responsibilites and authorities. Information is presented in a logical manner following the normal sequence of sampling, analyzing, and reporting. Included also are other aspects of the program, such as, work statements (to accompany work orders), equipment calibration and maintenance, training, audits, and record keeping. Each chapter of the guide contains a general statement about the subject, a section listing the applicable QC requirements, and another section describing how to accomplish the requirements. The QC guide is divided into nine subject areas, each of which is described below. Work Orders The work conducted for the surveillance program is handled either in-house by program personnel or by subcontract and work orders to others. The statement of work for the UST subcontract is renegotiated each year and is altered as necessary to reflect new QA-QC requirements. Each work order entailing significant cost (>$2000) and written to PNL personnel outside of the surveillance program also is accompanied by a statement of work. Work statements are prepared by the person requesting the work and typically are written for technical support, analytical services, or craft services. The statements usually are brief (two to four pages) and specifically identify the work to be accomplished, any special instructions (descriptions of special or unusual samples), quality control requirements (if different from normal QC procedures), scheduling (time constraints), reporting (type of report and due date), and cost accounting (no cost overrun without authorization). Results from the use of this technique have been very encouraging especially when requesting special help with radiochemical analyses or microelectronics from research and development groups who may not be accustomed to providing such services. Work statements eliminate misunderstandings and help keep work on schedule. Environmental Monitoring Measurements and Sampling Routine radiation and nonradiation measurements and environmental sampling are conducted according to established procedures under the direction of the Environmental Monitoring Supervisor. Requirements for the supervisor include preparing a specific work schedule to coincide with the annual sampling schedule; reviewing daily work sheets; and retaining equipment purchase and inventory records. The supervisor must advise the program manager in writing of any deviation from established procedures, sampling locations, or prescribed schedules. The specific manuals to be maintained and who is responsible for maintaining them also are discussed.
137
Analytical Laboratories Samples are analyzed by UST or in a variety of PNL laboratories. These analytical laboratories are required to have written procedures and are expected to follow accepted quality control methods and recognized standards of practice. In addition to written procedures, the laboratories are required to participate in interlaboratory comparison programs and use standards traceable to the National Bureau of Standards. These laboratories are also required to identify any deficiencies in procedures or instrumentation through an internal QC program and then document corrective actions. The results of the internal QC program and the corrective actions taken are audited each year by the Program Manager or his designate. Data Management and Calculations Analytical results are automatically entered into a computerized storage and retrieval system, i.e., data base. The system also preprints sample labels according to the work schedule and tracks samples by their identification numbers until results are received from the analytical laboratories. A subprogram automatically identifies anomalous data from individual results that are above or below predetermined action levels. The system is capable of performing statistical analyses and other data manipulation tasks and of printing summary reports. Quality control requirements include a review and double check of all dose calculations prepared for the annual surveillance report. Documentation of computer codes by keeping a log of the version in current use also is required. Quality control implementation includes preparing an annual summary of anomalous data and using data log books to record results of special studies. A users' manual is maintained for the computerized data base. Calibration and Maintenance All measurement and analytical equipment used in the surveillance program is calibrated periodically and receives routine maintenance. Calibrations are carried out using primary or secondary standards either in situ or at a calibration facility. Maintenance is conducted either in situ or at a maintenance shop. Calibration and maintenance records are required, and tags and labels are used to indicate the last date of service. The Environmental Monitoring Supervisor is responsible for field equipment, and laboratory supervisors are responsible for laboratory equipment. The individual worker is responsible for noting the presence of a valid calibration or preventative maintenance tag before using the equipment. The use of source checks is required for radiation detection equipment. Written procedures are used for calibration and preventative maintenance routines. Procurement and Acceptance Testing Performance specifications for equipment and supplies are included in all purchase requisitions and reference national consensus standards whenever appropriate. Receiving inspections and acceptance tests are conducted by the
138
receiver of the equipment. The results of acceptance tests are briefly documented if favorable. If the equipment is unacceptable, a nonconformance report is filed and the equipment returned to the supplier. It is important to document nonconformance with the purchasing department so that suppliers with poor records can be identified. Operating manuals, calibration charts, and other information supplied by the manufacturer along with acceptance test results are retained by the primary user of the equipment. Personnel Training Environmental monitoring personnel participate in training programs consisting of formal and on-the-job training. Staff members are encouraged to attend seminars, society meetings, and short courses related to their special interests and duties and to participate in continuing education programs. Records of formal training and participation in on-the-job training activities are kept as documentation in personnel files. Training effectiveness is evaluated through oral discussions or observations of performance by the trainer, supervisor, or third-party auditor. Quality Control Audits Compliance with documented procedures is audited periodically to assure that procedures are adequate and are being followed. Audit results are discussed with the party audited before a written report is issued. Corrective actions are reviewed at a follow-up audit or during the next regularly scheduled audit, depending on the urgency of the finding. Requirements also include the operation of a "blind audit program" wherein samples of milk, water, vegetation, and soil are spiked with known amounts of radioactive materials and periodically submitted blind for analysis. Laboratory results are recorded on control charts and isotope and nuclide ratios are calculated and evaluated. While these results do not assure analytical accuracy, they do allow precision, continuity of results, and timeliness of reporting to be evaluated, as well as providing a direct check on laboratory performance. Documentation, Records and Reports Procedures, procedural changes, logbooks, and environmental monitoring work sheets are retained for 5 years. Calibration and maintenance records, acceptance tests results, audit reports, and other pertinent materials are retained at least 3 years. A QA-QC central file has been established to contain or reference all current records related to QA-QC activities within the surveillance program. A historical file contains letters and other significant communications as well as a complete set of all formal reports issued. The historical file and the computerized data base are permanent files. A small library of reports and manuals relating to QA-QC is maintained for use by staff members.
139
QA-QC AUDITS Several layers of audits are used to assure that quality is maintained in all phases of the surveillance program. The Quality Assurance Office performs audits to assure that provisions of the QA Manual and QA Plans are met. Occasionally, external third-party audit teams are invited to review surveillance program activities. Internal QC audits are conducted by surveillance program personnel on specific phases of the program including the analytical laboratories. Internal audits are most effective when they are viewed as an aid in doing a job correctly and not as a means of finding fault. Findings permit the Program Manager to focus program activities in the most productive manner.
140
3D
DESTRUCTION OF URANIUM-CONTAMINATED WASTE OIL
L. F. Hary Goodyear Atomic Corporation Piketon, Ohio
ABSTRACT The Portsmouth Gaseous Diffusion Plant routinely generates quantities of uranium-contaminated waste oil, mainly from vacuum pump operation and pump maintenance activities. The diffusion plant currently generates 5,000 gallons of the waste oil per year. This oil is presently disposed of by landfarming on "open field" oil biodegradation plots with a total surface area of 0.44 acres. The new Gas Centrifuge Enrichment Plant will generate an estimated 17,000 gallons of waste oil per year at full operation with limited waste oil generation from this facility to begin in September 1983. Due to the environmental concerns associated with "open-field" biodegradation and the projected increased generation of uranium-contaminated waste oil, studies were conducted to better characterize the waste oil and to determine the optimum biodegradation conditions for disposal of these oils by the currently used landfarming method. The following biodegradation parameters were examined in the optimization studies: microbial strains, soil media, nutrient requirements, and degradation environment (e.g., temperature, moisture). Additional studies were conducted to determine the feasibility and effectiveness of the "open-field" landfarming as compared with a proposed greenhouse-like enclosure for the landfarming plots which would also operate using the optimized biodegradation parameters. The conclusions of these studies are that "open-field" landfarminq biodegradation of waste oil under optimum conditions is an environmentally acceptable and cost effective disposal method. The greenhouse-like enclosure offers a modest efficiency increase for the expenditure, but the assurances of containment offered by this type of structure may justify the cost. Finally, a comparison was made between the biodegradation disposal method and incineration, a more extensively used method for disposal of waste oils. Included in the comparison are discussions on the physical and chemical characteristics of the waste oil and the effects of these characteristics on the use of the alternative methods.
141
DESTRUCTION OF URANIUM-CONTAMINATED WASTE OIL INTRODUCTION The Portsmouth Gaseous Diffusion Plant has been landfarming uranium-contaminated waste oil for approximately seven years. Operations and maintenance of the landfarming plots has been based on recommendations presented in available biodegradation literature (Rumble 1976). Currently, the oil is applied to plots ten feet wide covering an area of 0.44 acres. The plots are only used during warm weather months, fertilizer is added as a mineral nutrient, and the plots are routinely cultivated to promote aerobic conditions. During the latter part of 1981, a study (Hary 1982) was initiated to determine ways to improve the biodegradation process; emphasis was given to enclosing the oil plots with a greenhouselike structure. In addition, other disposal alternatives, such as incineration, have also been evaluated (Hary and Rolph 1982). Based on past experience operating the oil biodegradation plots and on the results of the studies referenced in the previous paragraph, it was determined that further studies were required on the optimization of conditions for the biodegradation disposal method. The Battelle Memorial Institute, Columbus, Ohio, was the consulting contractor selected to perform waste oil biodegradation rate tests under varying conditions and an extensive literature review on biodegradation (Van Voris 1982). This study was initiated during May 1982. The following landfarming technology areas were researched: •
waste oil biodegradation rate determinations by C0 2 evolution;
•
literature review on biodegradability of oil, uranium toxicity, and uranium mobility in soils;
•
design considerations for a biodegradation facility;
•
and nuclear criticality safety considerations.
The final report report was to be issued by November 30, 1982. DISCUSSION Waste Oil Generation and Characteristics Uranium-contaminated waste oils are currently generated at the diffusion plant (5000 gallons/year), and will be generated by the
142
Gas Centrifuge Enrichment Plant (17,000 gallons/year) which is presently under construction. Oils from the diffusion plant have been estimated to have an average of 5000 mg/1 of uranium at 11.5 percent uranium-235. Oils from the Gas Centrifuge Enrichment Plant are predicted to contain 10,000 mg/1 of uranium at 1.0 percent uranium-235. All these waste oils will have a hydrocarbon composition. During June 1982, a total of five oil samples were submitted to Battelle for biodegradation rate studies. The five samples consisted of two types of unused oil and three uranium-contaminated waste oils. The unused oils (a Sohio and a Lapine product) were of the type used to refill vacuum pumps after maintenance. Samples of the waste oil were collected from three distinct vacuum pump maintenance activities. These activities were: maintenance on large yacuum pumps (sample number EZ-741), maintenance on small vacuum pumps (sample number EZ-742), and routine flushing of large vacuum pumps (sample number EZ-743). Sample number EZ-743 was the only one containing solvents; this sample had 35 percent 1, 1, 1,-trichloroethane and 5 percent p-dioxane. A number of the chemical and physical characteristics of the aforementioned oil samples are listed in Table 1. The hydrocarbon fraction breakdown for the unused oils is approximately 95 percent aliphatic, 2 percent polar, and 3 percent aromatics; waste oils have approximately the same composition. TABLE 1. Characteristics of Oil Samples Submitted to Battene Parameter
Sohio
Lapine
EZ-741
EZ-742
EZ-743
Uranium, mg/1
0
0
592
7,870
34
Assay, % U-235
0
0
31.6
2.76
9.88
Heating value, BTU/lb.
19,600
19,600
17,500
18,100
14,000
29.9
29.9
28.2
24.1
7.3
309
353
364
245
49.1
Specific gravity, API/60°F Viscosity, SSU 100°F
Optimized Biodegradation Parameters A diverse group of bacteria and fungi have the ability to degrade petroleum hydrocarbons. Research supporting this conclusion
143
has been carefully detailed in a recent review of the literature (Atlas 1981). Battelle (Van Voris, et al. 1982) has examined research literature applicable to Portsmouth's waste oil situation and has developed optimum biodegradation criteria. More than 100 species representing 30 microbial genera have been discovered as being capable of utilizing hydrocarbons in both aquatic and terrestrial environments (Atlas 1981). The number of hydrocarbon-utilizing microorganisms increases dramatically when oil is applied to the soil. Typically, hydrocarbon utilizers may constitute less than 0.1 percent of the microbial population; however, in oil-polluted ecosystems they can constitute up to 100 percent of the microorganisms. In one study of oil decomposition in soil (Pinholt, et al. 1979), an increase from 60 to 82 percent in oil-utilizing fungi and an increase from 3 to 50 percent in oildegrading bacteria was found after a fuel oil spill. Soil conditions have a significant influence on oil biodegradation (Dibble and Bartha 1979). In soils, pH appears to be a primary influence in microbial degradation of oil. Past studies have indicated that a soil pH close to neutral resulted in the highest C0 2 evolution rate. Soil moisture does not seem to have an appreciable impact on biodegradation rate. Most studies have concluded that the addition of nitrogen and phosphorous-containing fertilizers enhances biodegradation rates. An intermediate initial fertilization rate which brings the C:N ratio to 60:1 and the C:P ratio to 800:1 have resulted in high biodegradation rates. Higher rates of mineral nutrient addition have resulted in depressed biodegradation activity. Petroleum is a complex mixture of hydrocarbons composed of a saturated or aliphatic fraction, an aromatic fraction, and an asphaltic or polar fraction. The degradability of these oil fractions varies (Atlas 1982). The saturated fraction includes n-alkanes, branched alkanes, and cycloalkanes. The n-alkanes are usually the easiest to degrade microbially. Cycloalkanes, however, are particularly resistant to microbial attack. The cycloalkanes are among the most resistant components of petroleum spillages. Light aromatic hydrocarbons are subject to both evaporation and microbial degradation in a dissolved state; however, extensive substitution can inhibit initial oxidation. Condensed ring aromatic hydrocarbons are fairly resistant to enzymatic attack. Metabolic pathways for the degradation of asphaltic components are not well understood. Temperature also can have a significant effect on the rates of hydrocarbon degradation (Atlas 1981). Hydrocarbon degradation has
144
been shown to proceed at a rate an order of magnitude faster at 25°C than at 5°C. Seasonal shifts in the microbial community, however, may permit low-temperature degradation. Temperature effects appear to be interactive with other factors, such as the character of the hydrocarbon mixture and composition of the microbial community. Experimental Design Based on the optimized oil biodegradation parameters discussed in the previous section, an experiment was designed to establish the destruction rate of Portsmouth oil by quantifying the C0 2 evolution rate. The experimental units were specially designed respirometer jars. Each jar was loaded with 50 grams of soil from the Portsmouth site; soil was adjusted to 70 percent waterholding capacity using distilled H 2 0 and to a pH of 7«5 using 0.1M KOH. The respirometer jars consisted of 1-pint jars containing a tightly fitting rubber stopper with an ascarite trap. This design allowed for passage of C02-free air in and out of the jar as well as pressure equilibration with ambient conditions. An alkali trap containing 0.6 N NaOH was used to collect the evolved C0 2 . All tests were performed in triplicate. Tests on the five oils were conducted at three loading rates of 1, 10, and 20 percent (weight of oil-carbon/weight of soil); this is equivalent to loading rates of 3050, 30,500, and 61,000 gallons/acre, respectively. In addition, three glucose ammended controls and three abiotic controls were included. Glucose controls were included to confirm the presence of microbial activity. All oils were air-stripped prior to application to soils. The five oils were also treated at the above three loading rates using fertilized soil. Optimized fertilization was achieved by adjusting carbon:nitrogen to 60:1 with NH[tNO3 and carbon:phosphorous to 800:1 with K 2 HP0 4 . Glucose and abiotic controls were similarly fertilized. Experimental Results Carbon dioxide evolution in the respirometer flasks were determined for a 10-week test period (Van Voris, et al. 1982). Biodegradation rate test results show that landfarming is a viable method for the disposal of Portsmouth uranium-contaminated waste oil. Of all the waste samples, sample number EZ-741 showed the highest C0 2 evolution (biodegradation) rate. Figure 1 presents C0 2 evolution (cumulative) versus time for the three waste loading rates for unfertilized soil. Curves for the no dose and glucose amended cases have been added for comparison. Enhanced biodegradation rates were observed for this oil when fertilizer was added to the soil (see Figure 2). A comparison of the biodegradability of
145
Glucose
cr>
No Dose 20% Dose 1% Dose
SAMPLE PERIOD (WEEKS) Figure 1 . C02 Evolution for EZ-741 in Unfertilized Soil
10% Dose 20% Dose Glucose Control
No Dose 1% Dose
SAMPLE PERIOD (MEEKS) Figure 2.
C02 Evolution for EZ-741 in Fertilized Soil
unused oil versus waste oil is shown in Figure 3. At the end of 10 weeks, the unused oil (Sohio) evolved 192 milligrams of C0 2 while the waste oil (EZ-741) evolved 153 milligrams; consequently, destruction rates are similar. Sample number EZ-742 also showed acceptable biodegradation; however, sample number EZ-743 had substantially lower C0 2 evolution rates. It is believed that this is the result of the toxic effect of chlorinated solvents on hydrocarbon utilizing bacteria. Figure 4 shows a profile of the depressed biodegradation rates for sample number EZ-743. Based on these experimental results, a waste loading rate of 10 percent (30,500 gallons/acre) and fertilized soil (C:N=60:l and C:P=800:l) result in optimum biodegradation. Biodegradation Facility Design and Operations The size of any permanent oil biodegradation facility will depend on the biodegradation rate. Since biodegradation rates are enhanced at higher temperature, two types of facilities were examined: •
open-field plots exposed to the elements; and
•
plots enclosed by a greenhouse-like structure.
The oil destruction rates for these two facilities were determined using optimum biodegradation parameters and the Arrhenius equation for the rate of reaction with temperature. For the openfield case, monthly average soil temperatures for the southern Ohio area were used. The calculated waste oil disposal rate for openfield conditions was approximately 100,000 liters/acre/year. In the second case, which assumed that a greenhouse constructed over the oil application plots would utilize waste plant heat, the disposal rate was calculated to be 190,000 liters/ acre/year. It was assumed that the soil temperature would be maintained above 15°C during the winter months. Battelle also examined types of greenhouse structures for landfarming. In the past, conventional greenhouse structures have been constructed from glass panes supported by wood or steel frames. Recently, a new concept in greenhouse construction technology incorporating a minimum of fixed, rigid structures has been introduced and marketed under the generic term of "air-supported greenhouse." These structures, when erected, resemble an air-inflated bubble, and often consist of air-supported twin layers of vinyl supported by stainless steel aircraft cables. Air forced between vinyl layers by a series of fans keeps the structure rigid and inflated. A cost comparison of biodegradation facilities (open-field and two greenhouses) types is shown in Table 2.
148
(Sohio)
EZ-741
SAMPLE PERIOD (WEEKS) Figure 3. CO2 Evolution f o r EZ-741 and Sohio Oil ( a t 10 Percent Loading and in Fertilized Soil)
Glucose Control
10% Dose en o
0 Dose 1% Dose 20% Dose
SAMPLE PERIOD (WEEKS) Figure 4. C02 Evolution for EZ-743 in Fertilized Soil
An evaluation of uranium toxicity on microorganisms, uranium mobility In soil, and nuclear criticality potential was also performed (Van Voris, et al. 1982). Based on the physical and chemical characteristics of the wastes, it has been determined these radiological concerns would not inhibit landfarming operations. TABLE 2. Summary of Performance and Capital Costs for Landfarming Alternatives System
Estimated Microbial Approximate Capital Degradation Rate, 1/acre/yr* Cost, Dollars**
No greenhouse
100,000
37,000
Glass greenhouse
190,000
714,000
Flexible fabric greenhouse
190,000
190,000
*Based on Battelle data **1982 estimates for facility handling 22,000 gallons/year of waste oil. CONCLUSIONS Based on Battelle's experimental results and the development of biodegradation facility design criteria from these results, biodegradation is a viable alternative for the disposal of Portsmouth's uranium-contaminated waste oils. Open-field biodegradation, and even the use of a flexible fabric greenhouse, are more economically attractive than incineration whose capital cost has been estimated to be approximately $850,000 (Hary and Rolph 1982).
151
REFERENCES Atlas, R. M. 1981. "Microbial Degradation of Petroleum Hydrocarbons: an Environmental Perspective." Microblal. Rev. 45(1):180-209. Dibble, J. T., and Bartha, R. 1979. "Effect of Environmental Parameters on the Biodegradation of Oil Sludge." Appl. Environ. Microbiol. 37(4):729-739. Hary, L. F. 1982. Controlled Waste Oil Biodegradation at Existing Drying Bids. Report No. GAi-S-26, tioodyear Atomic Corp., Piketon, Ohio. ""Hary, L. F., and D. J. Rolph. 1982. Disposal Options for UraniumContaminated Waste Hydrocarbon Oils. Report No. GAT-A-Z32, Goodyear Atomic Corp., Piketon, Ohio. Pinholt, Y., et al. 1979. "Microbial Changes During Oil Decomposition in Soil." Holartic Ecol. 2:195-200. Rumble, B. J. 1976. Review of GAT Oil Biodegradation Facilities. Report No. GAT-R-599, Goodyear Atomic Corp., Piketon, Ohio. Van Voris, P. 1982. Laboratory Studies on the Potential of Enhancing the Biodegradation of Uranium-Contaminated Waste Oil in Soil. Proposal fiol 526-J-7341, Battelle Columbus Laboratories, Columbus, Ohio. Van Voris, P., et al. 1982. The Biodegradation of Uranium-Containing Waste Oils and Degreaser Sludge in Soil. (UKAFIJ, BattelTe Columbus Laboratories, Columbus, Ohio.
TR-120382-WTB
152
3E THE WATER MONITORING PROGRAM AT THE ROCKY FLATS PLANT
R. L. Henry Rockwell International Golden, Colorado
ABSTRACT An overview of the water monitoring program at the Rocky Flats Plant will be presented. The Rocky Flats Plant is a government owned facility that is operated by the Energy System Group within Rockwell International. The plant is a key Department of Energy Facility that produces components for nuclear weapons. Surface water runoff from the plant flows into the Great Western Reservoir which is located about 3.2 kilometers (2 miles) downstream from the plant. This Reservoir is a source of drinking water for the City of Broomfield. The increased demand for water and tb/i concern for the quality of the water resources of the United States has led to an increased interest to improve water sampling and analysis programs. In particular, concern has been shown recently by the public for the amount of radioactive discharges into waters from nuclear facilities. Discussion will include what the Rocky Flats Plant is doing and has done to prevent the pollution of the streams and reservoirs surrounding the PlantAnnual average radioactivity concentration data for the period 1975 through 1981 will be presented that indicates that Rocky Flats Plant has had no adverse environmental impact. Compliance with the National Pollutant Discharge Elimination System (NPDES) permit program will also be discussed.
INTRODUCTION Prior to construction of the Plant, a water sampling study was done in 1951 to develop a baseline for plutonium concentrations in water bodies and vegetation surrounding the Plant site. Water samples were obtained from 27 locations within an area of approximately 144 square miles. Twenty-two of the sampling locations were within a 6-mile radius of the Plant site. 153
The samples were taken from springs, creeks, ditches, canals, lakes and reservoirs during July, August, September and November of 1951. In 1951, an analytical method for specific analysis of plutonium did not exist so a combined plutonium-uranium radiochemical method was used. The uranium concentration was determined by fluorometry and the plutonium concentration was calculated by subtracting the uranium concentration from the total. A comparison of the average plutonium-uranium with uranium only activity levels shows that they are statistically identical so that within the measurement uncertainty no plutonium was present in any samples (Bokcwski, et al, 1981). Water sampling and analyses for radioactive species and some nonradioactive species, such as metals, continued throughout the 1950's and 1960's. FEDERAL REGULATIONS ENACTED The Federal Water national goals for the These goals were to be of the Act until 1985. 1. 2. 3.
Pollution Control Act (FWPCA) of 1972 established protection of water bodies (Federal Act, 1972). achieved on a periodic basis from the beginning For example:
By July 1, 1977, all nonpublic point sources must meet effluent standards based on "best practicable control technology currently available," publicly owned treatment works must attain secondary treatment by July 1, 1977, and attain zero discharge by 1985.
The Clean Water Act (Clean Water Act, 1977), an amendment to the FWPCA, was enacted by Congress to clarify some aspects of the FWPCA and (1) provide a time extension for those industries and municipal dischargers who failed to comply with the July 1, 1977 deadline, and (2) simplify methods for establishing toxic pollutant standards by specifying a less formal rulemaking procedure. REGULATION COMPLIANCE The Rocky Flats Plant has continually progressed as one of the "front runners" in complying with the intent of the Acts. Prior to 1972, discharges from the south side of the Plant including cooling tower blowdown, water treatment plant filter backwash water, and steam condensate were all discharged to Woman Creek. Laundry water and cooling tower blowdown water from the central section of the Plant were treated in the sanitary sewage plant before being discharged to four settling ponds in the South Walnut Creek drainage. Process waters and cooling tower blowdown from the north side of the Plant were discharged to two settling ponds in the North Walnut Creek drainage. Waters that 154
were discharged downstream were retained in the settling ponds for a definite time period to allow settling to take place; then the water was discharged to Walnut Creek which flowed to Great Western Reservoir. A map of the settling ponds and surrounding area is shown in Figure 1. PROJECTS COMPLETED The Rocky Flats Plant was granted a National Pollutant Discharge Elimination System (NPDES) permit in September, 1974 after completing a series of projects required by the FWPCA. These projects included (Internal Communication) the following: 1.
Process water and laundry waste water were rerouted from the sanitary sewage treatment plant to the process water system.
2.
All cooling tower blowdown was repiped to either the sanitary sewage plant or the process water system.
3.
Settling ponds to hold the filter backwash water from the water treatment plant were completed.
4.
The sanitary sewage treatment plant was expanded from secondary treatment to include tertiary treatment.
ZERO DISCHARGE Zero discharge, as defined by the FWPCA, means no discharge of pollutants that may be present in water used in industries or municipalities, the Rocky Flats Plant has gone a step further in defining zero discharge as no release of water *r pollutants used at the Plant. Since December, 1979, there has been no offsite discharge of sanitary sewage treated water. All water has been spray-irrigated or used as makeup water in the cooling towers. To help achieve zero discharge, a Reverse Osmosis (RO) Water Recycling Plant was constructed. A general description of the treatment process in the Reverse Osmosis plant consists of filtration to remove suspended solids and colloids, water conditioning to remove dissolved calcium, and magnesium by sodium substitution, chlorine addition, pH adjustment to 5.5 by the addition of sulfuric acid, temperature increase to approximately 25°C, treatment by reverse osmosis to reduce the dissolved solids, and a final pH adjustment to 7.5 by the addition of sodium hydroxide. Sanitary treated wastewater is used as makeup water for the RO system. After RO treatment, this water is used for cooling tower makeup, with any excess used for spray irrigation on plant land.
155
A-4NEWDAM* RETENTION BASIN
NEW REVERSE OSMOSIS PLANT B-5 NEW DAM * RETENTION BASIN
TERTIARY TREATMENT PROCESS (ADDED TO EXIST. SEWAGE TREATMENT PLANT) INTERCEPTOR
FIGURE 1.
Holding Ponds and Liquid Effluent
C-2 NEW DAM & RETENTION BASIN Watercourses
SURFACE WATER CONTROL SYSTEM The Rocky Flats Plant surface water control system was expanded as a means of improving water retention capability in the event that surface runoff water proved unsuitable for discharge from the Plant. Construction was completed in 1980. Additions to the system consisted of Pond A-4 with 93 acre-feet capacity on North Walnut Creek, Pond B-5 with 80 acre-feet on South Walnut Creek, Pond C-2 with 64 acre-feet capacity on Woman Creek and an interceptor canal to collect surface runoff water from the southern part of the Plant site and route it to Pond C-2. Each of these ponds has a stand pipe connected to the outlet structure, minimizing the amount of sediment discharged downstream. As stated earlier, all Plant process and sanitary wastewater is recycled or used for spray irrigation. Therefore, only surface runoff water due to rain or snow melt is impounded in the new control ponds. Prior to discharge, this water is retained to allow settling. Samples are then collected and analyzed for several control parameters. The data are reviewed by Plant management to make sure that none of the Plant control guides have been exceeded. Approval for discharge is then given by Plant management. WATER MONITORING PROGRAM The primary objectives of the Rocky Flats water monitoring program are: (1) protection of the public against unexpected radioactive contamination, (2) monitoring for contamination of water, (3) prompt detection and control of any accidental environmental releases, (4) maintenance of records and evaluation of data, and (5) compliance with federal and state laws. To meet these objectives at the Rocky Flats Plant, extensive sampling is performed at many locations which are listed in Table 1. The sampling program includes sampling and analyses of incoming water, water in use, storage ponds, surface runoff water and community waters. RESULTS FROM MONITORING Graphically illustrated in Figure 2 are recent plutonium concentration data for Great Western Reservoir and its corresponding drinking water supply, the City of Broomfield. These data are below all regulatory guides - Standards for Radiation Protection, DOE 5480.1 and the Federal Safe Drinking Water Act. Statistical comparison of plutonium concentrations from Great Western Reservoir to other raw water bodies have shown no significant difference in plutonium activity (Bokowski, et al, 1981). Likewise, statistical comparison of the City of Broomfield drinking water to other drinking water supplies have shown no significant differences in plutonium activity.
157
TABLE 1. Sampling Locations for the Rocky Flats Water Monitoring Program 1. Water Treatment Plant Raw Water 2. Water Treatment Plant Treated Water 3. Sewage Treatment Plant Influent 4. Sewage Treatment Plant Effluent 5. Pond A-l Bypass 6. Pond A-3 7. Pond A-4 8. Pond B-3 9. Pond B-4 10. Pond B-5 11. Pond C-l 12. Pond C-2 13. Landfill Pond #1 14. Process Waste Treatment Pond 15. Landfill Bypasses (North and South) 16. Holding Tank Test Holes 17. Solar Evaporation Ponds 18. Hydrologic Test Holes 19. Building Footing Drains 20. Walnut Creek at Indiana 21. Great Western Reservoir 22. Standley Lake
158
TABLE 1. Sampling Locations for the Rocky Flats Water Monitoring Program (continued) 23. Broomfield 24. Boulder 25. Westminster 26. Arvada, Denver, Golden, Lafayette, Louisville, Thornton 27. Regional Lakes and Streams
159
PCi/l Legend
0.04
Great Western Reservoir •V Broomfield Tap Water 0.03
0.02
O
a • £ o-S 3 5 3 « u o S 1979 FIGURE 2.
an. Feb. Mar. Apr. May
0.01
1980
1981
Plutonium Concentration in Broomfield Tap Water and Great Western Reservoir
Listed in Tables 2 and 3 are the average annual plutonium, uranium, americium and tritium concentrations for Great Western Reservoir and Standley Lake, respectively, for the period from 1975 to 1981. The radioactive species for both Great Western Reservoir and Standley Lake are near background. At these low concentrations, the level of activity that can be measured is difficult to distinguish from statistical fluctuations in background samples. CONCLUSIONS The Rocky Flats Plant has made a concentrated effort to minimize its impact on the environment surrounding the Plant by (1) complying with the goals of the Federal Water Pollution Control Act, (2) adopting a policy in regard to radioactive discharges of "as low as reasonably achievable," (3) sampling and analyzing for pollutants in the surrounding area, (4) installing the Reverse Osmosis Plant to eliminate discharges from the Plant, and (5) constructing three new ponds downstream to impound all surface runoff water from the Plant. REFERENCES Bokowski, D. L., R. L. Henry and D. C. Hunt. February 21, 1981. History and Evaluation of Regional Radionuclide Water Monitoring and Analysis at the Rocky Flats Installation. RFP-3019. Hornbacher, D. D. 1975 through 1981. Annual Environmental Monitoring Reports. Rocky Flats Plant. Environmental Master File Reports. Internal Communications.
Rocky Flats Plant.
U.S. Congress, 1972. October 18, 1972. Federal Water Pollution Control Act. PL92-5OO. Federal Register. 1976. Drinking Water Regulations, Radionuclides. Vol 41, No. 133, pp. 28402-09, July 9, 1976. Standards for Radiation Protection. August 13, 1981. DOE Order 5480.1, Chapter 6, Chapter XI, Department of Energy. U.S. Congress, Clean Water Act of 1977 (PL95-217).
161
TABLE 2. Average Annual Plutonium, Uranium, Americium and Tritium Concentrations Observed in Great Western Reservoir Parameters (pCi/1) Year
Plutonium 3
Uraniumb
Americiumc
Tritiumd
1975 1976 1977 1978 1979 1980 1981
<0.1 <0.06 <0.1 <0.1 <0.009 <0.02 0.011 ± 0.004
2.0 0.9 1.9 3.2 2.7 <3.0 3.1 + 0.1
<0.03 <0.02 <0.1 <0.1 <0.017 <0.03 0.000 ± 0.003
2300 <950 <700 <700 <700 <500 <300
TABLE 3. Average Annual Plutonium, Uranium, Anericium and Tritium Concentrations Observed in Standley Lake Parameters (pCi/1) Year
Plutonium a
Uranium'3
Americiumc
Tritium^
1975 1976 1977 1978 1979 1980 1981
<0.04 <0.04 <0.1 <0.01 <0.007 <0.01 0.002 ± 0.002
2.7 1.8 2.3
<0.03 <0.05 <0.1
5.2
<0.01
3.3 <4.0 2.6 ± 0.1
<0.02 <0.03 0.004 ± 0.004
1000 <650 <600 <700 <600 <500 <300
a. The Radioactivity Concentration Guide (RCG^,) for soluble plutonium in water is 1,667 pCi/1• b. The RCQ^. for soluble uranium in water for the periods 1975 through 1980 was 10,000 pCi/1 and for 1981 was 200 pCi/1. c. The RCGy, for soluble americium is 1,330 pCi/1 • d. The RCGjJ for tritium in water is 1,000,000 pCi/1 and the State of Colorado Primary Drinking Water Regulation limit is 20,000 pCi/1
162
DETAILED SITE STUDY PHASE: AN ENVIRONMENTAL APPROACH TO PLANNING AND DESIGN M.L. Brown Office of NWTS Integration Battelle Project Management Division INTRODUCTION This paper describes the planning of detailed studies that w i l l aid in determining the s u i t a b i l i t y of a s i t e for development as a high-level nuclear waste repository. Environmental considerations weighed heavily i n planning and design of t h i s phase of exploratory work. Some background information w i l l be presented f i r s t to acquaint the reader with a purpose of the overall radioactive waste repository program. The plan for s i t i n g repositories w i l l then be summarized as part of the general background information to place the "Detailed Site Study Phase" in proper perspective. The detailed s i t e study phase w i l l be described i n more d e t a i l , including information on the schedule, a c t i v i t i e s , and reports that w i l l document the process. F i n a l l y , the interaction between environmental scientists and design engineers w i l l be described and results of that interaction w i l l be discussed. BACKGROUND The N a t i o n a l Waste Terminal S t o r a g e (NWTS) Program was e s t a b l i s h e d i n 1976 by DOE's p r e d e c e s s o r agency, t h e Energy Research and Development
Administration. The objective of the NWTS program is to develop f a c i l i t i e s to permanently dispose of commercial high-level radioactive waste in a manner that w i l l (1) protect public health and safety and (2) be environmentally acceptable. A decision on a disposal method was needed for the program to reach this objective. After considerable debate and study, an Environmental Impact Statement (EIS) was published which evaluated ten disposal options. Following public review and comment, a f i n a l EIS concluded that the technology for emplacement of radioactive waste in geologic formations can be developed and applied with minimal environmental consequences. U ) This evaluation resulted in the DOE decision that research and development should focus on development of geologic repositories.'2) The repository (Figure 1) w i l l , in some ways, be similar to a conventional mine. Surface structures w i l l be built to create access through shafts to the underground rock. Corridors and rooms w i l l be excavated, but for the purpose of waste emplacement, not for maximum extraction of ore as in the case of a mine.
163
\ FIGURE 1.
CONCEPTUAL GEOLOGIC REPOSITORY
164
NATIONAL OR PROVINCE SURVEYS
SITE SCREENING -< PHASE
REGION SURVEYS
AREA SURVEYS
LOCATION SURVEYS
DETAILED SITE STUDIES PHASE SITE SELECTION PHASE
DETAILED SITE CHARACTERIZATION (INCLUDING EXPLORATORY SHAFT) SITE RECOMMENDATION AND SELECTION
FILE LICENSE APPLICATION WITH NUCLEAR REGULATORY COMMISSION RECEIVE CONSTRUCTION AUTHORIZATION AND BEGIN REPOSITORY CONSTRUCTION
REPOSITORY OPERATION PROCEEDING LICENSING PHASE REPOSITORY OPERATION
REPOSITORY CLOSURE PROCEEDING
REPOSITORY CLOSURE
FIGURE 2 .
REPOSITORY SITING AND LICENSING PHASES
165
The DOE published i t s plan for identifying suitable sites for a repository as a draft i n February, 1982(3) and requested comments from state and local o f f i c i a l s and the interested public. The draft s i t i n g plan # described a three-phase s i t i n g process consisting of s i t e screening, det a i l e d s i t e studies, and s i t e selection (Figure 2 ) . The s i t e screening phase consists of a set of decisions made sequentially to identify sites favorable for waste disposal. Site screening usually begins by considering a limited number of factors over large land areas to identify places that exhibit characteristics favorable for waste i s o l a t i o n . A stepwise method is used to reduce both the number and size of land units under consideration. At a few sites (3 to 5), studies w i l l progress to the detailed s i t e study phase. During the detailed s i t e study phase, information about the physical, chemical, geologic, and human environment necessary to judge s i t e s u i t a b i l i t y w i l l be collected and evaluated. In the s i t e selection phase, alternative sites w i l l be compared, and one or more w i l l be recommended for a repository. Public review of the recommended s i t e w i l l occur before the DOE makes the final selection and prepares i t s license application to the NRC. THE DETAILED SITE STUDY PHASE DOE w i l l prepare a Site Characterization Report (SCR) for NRC review prior to beginning the detailed s i t e study phase. The SCR w i l l summarize how DOE selected the s i t e for detailed study, what is known about the s i t e from the screening phase, what issues remain to be resolved to determine the s i t e ' s s u i t a b i l i t y for development as a repository, and the plans for resolving those issues. During the detailed site study phase, the data needed to s c i e n t i f i c a l l y determine s i t e s u i t a b i l i t y w i l l be gathered. # Geologic study methods to characterize sites w i l l include# borehole d r i l l i n g and laboratory testing of cores, geologic f i e l d mapping, geophysical borehole logging, conceptual modeling, and an exploratory shaft. Field mapping of l i t h o l o g i c units w i l l be performed at the s i t e . Geochemical and isotopic-dating analyses of selected f i e l d and core samples may also be performed. Surface geophysical surveys may include high-resolution, seismic-reflection, and e l e c t r i c a l - r e s i s t i v i t y methods. An environmental sampling and socioeconomic program w i l l also be conducted in t h i s phase. The characterization w i l l include* studies of atmospheric conditions, water quality, background radiation, noise, demographic characteristics, socioeconomic and cultural resources, land and water use patterns, and ecology. Potential socioeconomic effects w i l l be studied and may serve as a'basis for development planning on communities potentially affected.by the repository.
166
The major a c t i v i t y of the detailed s i t e study phase, however, w i l l be the construction of an exploratory shaft which w i l l permit i n s i t u testing. Exploratory shafts w i l l be d r i l l e d at three sties during 1983 and 1984: one at the Hanford s i t e i n Washington, one at the Nevada test s i t e , and one at a yet-to-be-determined salt s i t e (see Figure 3 ) . The shaft at the Hanford s i t e w i l l be d r i l l e d to a depth of approximately 3,300 to 3,600 feet and w i l l be 8 feet i n diameter. The shaft at Nevada w i l l be approximately 1,200 feet deep and 10 feet i n diameter. The exploratory shaft to be constructed i n salt w i l l be between 2,000 and 3,000 feet and w i l l be 12 feet i n diameter. The remainder of t h i s paper focuses on the salt shaft, and especially on the interaction between project engineers and environmental s c i e n t i s t s involved in the planning and design of the exploratory shaft.
DESCRIPTION OF THE SALT EXPLORATORY SHAFT The exploratory shaft to be constructed at a salt s i t e i s , of course, constrained by the features specific to the s i t e . Sites s t i l l under consideration include those at Davis Canyon, Utah; Deaf Smith County, Texas; Ri chton Dome, M i s s i s s i p p i ; and Vacherie Dome, Louisiana. The surface area which w i l l need to be cleared for shaft a c t i v i t i e s ranges from 30 t o 40 acres. Some of the salt removed from a shaft w i l l be stored on s i t e ; some w i l l be hauled away for disposal. Some of the other rock material, as well as the top s o i l , w i l l also be stored on s i t e . The salt storage p i l e w i l l be approximately 3 acres; top s o i l w i l l be stored on approximately 1 acre. The mud p i t for d r i l l i n g f l u i d w i l l occupy approximately 1 acre of the s i t e . The d r i l l r i g i t s e l f , power station and diesel generators, and temporary o f f i c e units w i l l occupy the remainder of the s i t e .
AN ENVIRONMENTAL APPROACH TO PLANNING AND DESIGN Because the NWTS program has established a NEPA Compliance Plan f o r i t s s i t i n g process, i t was recognized at an e a r l y stage t h a t a NEPA document would be required before shaft work began. Environmental s c i e n t i s t s assigned t o the p r o j e c t began discussing d e t a i l s of the design with p r o j e c t engineers very e a r l y i n the p r o j e c t , i n f a c t before the concept u a l design was completed. As a r e s u l t , environmental values were i n corporated i n t o the design instead of becoming "add ons" w i t h the assoc i a t e d schedule delays and economic i n e f f i c i e n c i e s .
167
a1
'85'
'891
'•,»
'..'
'93'
'95'
'97'
'99'
' o , «
' o a ''
«
'
T & E Facility Planning
"T 4 ~
" Construct
4
Ofterete
1
I 1
| '—Design
t Hanlord Basalt
Shtlt A
J,
1 ! 11 1 I
,_,
1 t
NTS Tulf
1
U
Shall j LA.
CA.
ta
. . . .
O.L.
i
1
t
First Repository
1
Bedded Sail
-
Shall | A. a m
a
w T
nU
Design
h
Construct
Salt
Dome Salt
Sites 4 and 5
Shaft
T •
Io *
1 1
U
START COMPLETE STARTUP REPOSITORY SITE SUITABILITY DETERMINED SITE SELECTION OPERATING DATA INPUT TO LICENSING PROCESS
FIGURE 3. Note:
Hara Rock
Sites Available lor Later Selection LA. LICENSE APPLICATION C.A. CONSTRUCTION AUTHORIZATION O.L. OPERATING LICENSE
REPOSITORY SITING SCHEDULE
Three candidate sites, at least one having other than salt as the repository host rock, are currently required by the procedural rule, 10 CFR 6 0 , of the U.S. Nuclear Regulatory Commission.
After reviewing preconceptual designs, possible areas of environmental impact were l i s t e d . Impacts in the areas of solid waste storage and disposal, a i r q u a l i t y , water resources, and noise/aesthetics were thought to be possible. The reasons for concern, and resulting design modification are discussed below for each area of possible impact. Environmental scientists were concerned with the way i n which salt would be stored on s i t e . Salt i s classified as hazardous material under RCRA.(5) Proper storage of salt was especially important considering the environments of the potential s i t e s . Seventy percent of land i n Deaf Smith County, Texas, has been classified as prime or unique farmland. Blowing or d r i f t i n g salt could substantially affect local crop production. Mississippi and Louisiana are characterized by frequent r a i n f a l l and improperly stored salt could be washed into nearby waters or could contaminate ground-water sources. The salt piles were, therefore, designed to be lined with an impervious material. All runoff from a salt p i l e w i l l be contained in a l i n e d , bermed p i t for evaporation or decantation as appropriate. Some discussion of covering or otherwise s t a b i l i z i n g the salt p i l e against erosion has occurred, but a s t a b i l i z a tion method has not yet been selected. In the area of a i r q u a l i t y , i t is assumed, for purposes of t h i s paper, that the exploratory shaft w i l l be d r i l l e d . Actually, a decision of whether t o use the d r i l l i n g method or the d r i l l and blast method has not yet been made. Project engineers planned to power the d r i l l r i g with 11 diesel generators producing a maximum of 5,000 horsepower. Power for support functions would be supplied from e l e c t r i c power lines in Texas, Mississippi, and Louisiana, while additional diesels would be needed for support power at the potential Utah s i t e . Envii onmental scientists compared data on ambient a i r quality at the potential sites with e s t i mates of probable diesel emissions. Serious air quality impacts were considered unlikely for sites located i n areas designated as Class 2 under the Clean Air Act. The potential Utah s i t e , however, i s located close to a Class 1 area and dispersion models, though crude, indicated that restrictions on operating horsepower and d r i l l i n g duration would be necessary. Other alternatives, such as alternative power sources for d r i l l i n g , were also investigated. Possible impacts on water resources were alleviated by l i n i n g and berming the salt runoff retention pond and the d r i l l i n g mud p i t s . The pot e n t i a l for ground-water contamination or mixing in the d r i l l i n g operat i o n was discussed. Methods for sealing aquifers were b u i l t into the d r i l l i n g operating procedures. Additional water requirements for cont r o l l i n g f u g i t i v e dust were added to the estimates of water requirements at the potential Texas and Utah s i t e s . Noise from the d r i l l i n g rigs was not expected to exceed EPA's guidel i n e s , but the low ambient noise levels at the potential Utah site and the p o s s i b i l i t y of affecting residents near the other potential sites
169
resulted in plans for baffles to surround the d r i l l i n g r i g and generat o r s . Other alternatives, such as l i m i t i n g the operating hours of the d r i l l i n g r i g and/or generators, may be possible. This exercise also confirmed the need for obtaining data on local meteorology at the potent i a l Utah s i t e . SUMMARY
In summary, the design of one of the phases in the repository s i t i n g process has been influenced through early consideration of possible enviromental consequences. Early dialogue between project engineers and environmental scientists was possible because a NEPA Compliance Plan exists for the program. This NEPA Compliance Plan requires an environmental review of the Detailed Site Study Phase prior to the start of the phase at a particular s i t e . Modifications to preconceptual designs in the areas of solid waste disposal and storage, a i r q u a l i t y , water resources, and noise/aesthetics were possible through this interaction. REFERENCES 1.
U.S. Department of Energy, 1980. F i n a l Environmental Impact S t a t e ment; Management o f Commercially Generated r a d i o a c t i v e Waste, D0E/EIS-0046F, Washington, DC, October.
2.
U.S. Department o f Energy, 1981. "Program o f Research and Development f o r Management and Disposal of Commercially Generated Wastes; Record of D e c i s i o n ( t o adopt a s t r a t e g y t o develop mined g e o l o g i c r e p o s i t o r i e s . . . ) " , Federal R e g i s t e r , V o l . 4 6 , No. 9 3 , Washington, DC, May.
3.
U.S. Department of Energy, 1982. " N a t i o n a l Plan f o r S i t i n g H i g h Level R a d i o a c t i v e Waste R e p o s i t o r i e s " ; P u b l i c D r a f t , DOE/NWTS-4, February.
170
SESSION FOUR REGULATORY IMPACTS
4A
ENVIRONMENTAL PROTECTION ANALYSIS AND PLANNING FOR PROPOSED SAVANNAH RIVER PLANT ACTIONS T. V. Crawford Savannah River Laboratory E. I. du Pont de Nemours & Co. Aiken, SC J. C. Tseng Savannah River Operations Office Department of Energy Aiken, SC
ABSTRACT The National Environmental Policy Act (NEPA), Council on Environmental Quality Regulations Implementing NEPA (40 CFR 1500-1508); DOE Guidelines for NEPA Implementation (45 FR 20694), as amended; order 00E 5440.IB; and the DOE Environmental Compliance Guide (DOE/EE-0132) require the early consideration of environmental factors for all proposed Federal actions. This paper presents a system which has been implemented at the Savannah River Plant (SRP) for facilitating NEPA considerations for a wide variety of activities.
NEPA HISTORY AT SRP Since the implementation of NEPA, major projects at the Savannah River Plant have received appropriate NEPA consideration on a case-by-case basis. This has been done either in the form of environmental impact statements (EIS) (e.g., Waste Management Operations, Savannah River Plant, ERDA-1537, September 1977, and Defense Waste Processing Facility, Savannah River Plant, Aiken, S.C., DOE/EIS-0082, February 1982), and environmental assessments (EA) (e.g., Naval Reactor Fuel Materials Facility, DOE/EA0170, March 1982, Waste Form Selection for SRP High-Level Waste, DOE/EA0179, July 1982, and L-Reactor Operation, Savannah River Plant, Aiken, S.C., DOE/EA-0195, August 1982). Brief environmental assessments were also prepared as a part of the conceptual design studies for line-item construction projects costing more than $1 million. These environmental assessments were incorporated as part of the conceptual design reports to support the budget requests. There was no system for considering environmental factors or for complying with the procedural requirements of NEPA.
173
ORGANIZATION The Savannah River Plant is operated by E. I. du Pont de Nemours & Co. for the Department of Energy to produce special nuclear materials for national defense. Technical support and process development for SRP production operations are provided by the Savannah River Laboratory, also operated by Du Pont. Within Du Pont, NEPA activities are focused in the Environmental Analysis and Planning Division (EAPD) of the Savannah River Laboratory. All NEPA efforts are closely coordinated with the Environmental Coordinator's Office of the Savannah River Plant organization. Within the DOE-Savannah River Operation's Office (SR), the NEPA Activities Branch (NAB) - Office of Environment (OE) coordinates environmental planning and documentation activities. Close interactions are maintained between SRL-EAPD and SR-NAB in developing this NEPA implementation system. Technical analyses are completed by Du Pont. NEPA strategy considerations are evaluated by both Du Pont and SR. The final determination of the appropriate NEPA documentation is made by SR or by DOE Headquarters (HQ). DU PONT NEPA SYSTEM In the summer of 1982, a plan was Du Pont. This plan requires that Savannah Laboratory Project Custodians complete a necessary submit additional information Evaluation (EE). An EE will be prepared:
developed and implemented by River Plant and Savannah River checklist (Figure 1) and if as part of an Environmental
o
for all orders, except direct work/repair orders, costing $200,000 and above (general plant projects, capital equipment, and cost orders - including indirect work repair orders, purchase requisitions, service orders, etc.);
o
for all test authorizations;
o
for all other activities which have high potential for environmental impact regardless of the dollar limits.
This checklist portion of the EE is designed to be completed as early as possible in the planning stages of the project (e.g., prior to conceptual design). Most projects will only need the completed checklist. For direct replacement projects which will have equal or less environmental impact, only a project description is required and a statement is included in the project description stating this fact. At Du Pont, copies of completed checklists are sent to the Environmental Analysis and Planning Division of SRL and to the Environmental Coordinator's Office of SRP. The checklists become the final NEPA document for proposed actions with clearly no significant impacts or direct replacement projects; this fact is indicated on the project approval papers. For those projects which 174
ACTIVITY NAME: ACTIVITY NUMBER: ACTIVITY CUSTODIAN:
Date: (Organization)
(Name)
ACTIVITY DESCRIPTION: (include purpose, need, location (provide map or figure if applicable), schedule, cost, etc.)
Checklist Instructions: Place an X under the NO or YES column as appropriate for the activity under evaluation. NO YES Is the activity a direct replacement with equal or less environmental or safety impact? ENVIRONMENTAL EVALUATION o Air Emissions (except uncontaminated ventilation air) o Liquid Releases to Streams, Seepage Basins, Storm Drains, Process Sewers, Wetlands, Groundwater, Ponds, Lakes o Solid or Liquid Waste Burial or Storage o Storage of Chemicals or Petroleum Products o Site Clearing, Excavation, Filling or Draining Wetlands/Floodplains, Borrow Pits, etc. o Liquid Withdrawals
Figure 1. Project Custodians Environmental Evaluation Checklist
175
have a "yes" checked indicating that additional data is required, the project custodian then consults with EAPD and provides appropriate supplemental data. Additional data is supplied in as quantitative a form as possible. It includes estimates of releases, rates, locations, areas impacted, etc. Emphasis is placed on using figures and diagrams. From this information, EAPD performs an impact analysis to quantify the potential environmental impacts as related to applicable environmental, health or safety standards; the analyses will be documented as internal Du Pont documents - environmental evaluation and impact assessments (EEIA). Based on this impact analysis, EAPD develops a recommendation on the NEPA strategy. If the environmental impacts are judged not to be significant, EAPD will recommend a Memo-to-File (MTF) as the appropriate NEPA documentation. This recommendation to SR will also include the draft MTF. If appropriate, SR will approve this course of action by preparing and issuing the final MTF. Upon receipt of the DOE-SR final MTF, EAPD will notify project custodians of the completion of the NEPA process. For all test authorizations, which are an internal Du Pont management analysis for departures from current operation, EAPD prepares the final MTF, if required, with a carbon copy to DOE-SR. If the significance of the action's environmental impact is uncertain, EAPD and SR will discuss the NEPA strategy based on the EAPD prepared impact analysis. If the issue cannot be resolved, EAPD will prepare a draft Action Description Memorandum (ADM) for transmittal to SR. SR uses the draft ADM to prepare a final ADM for transmittal to DOE-HQ for determination of the appropriate NEPA documentation. MTF preparation will be by EAPD as specified above. EAPD will coordinate as necessary the Du Pont preparation of a draft EA. EAPD will coordinate as necessary any Du Pont data input to EIS's prepared by an independent SR contractor(s). Final EA's and EIS's are reviewed, approved and published by DOE. The EA or EIS process usually cannot be completed before the Conceptual Design Report is sent to DOE-HQ. Following EAPD input, the project custodian will indicate the NEPA process status in the Conceptual Design Report. This replaces the earlier practice of including an environmental assessment in the Conceptual Design Report. SR NEPA SYSTEM The SR plan was developed in accordance with Order DOE 5440.IB and the NEPA compliance guidelines. Implementing this plan for proposed actions at SR will enable compliance with NEPA requirements and with the CEQ NEPA Implementing Regulations. An SR Supplement to Order DOE 5440.IB will be prepared and issued to all SR contractors for implementation. Within DOE, the determination of appropriate NEPA documentation is based on an evaluation of the significance of potential environmental effects. The CEQ Implementing Regulations for NEPA are used to define significance. In addition to an evaluation of the contractors quantitative technical impact analyses, SR will make a qualitative impact significance determination. The SR qualitative determination is based on the 176
contextual setting of the proposed action within overall SRP operations. For example, qualitative analyses will consider questions of legality, precedence, public controversy (perception), etc. If, following the SR evaluations, a final MTF is prepared, it will be written by OE and approved by the appropriate SR line Assistant Manager. SR intends to provide the Office of Defense Programs (DP) and the Office of Environmental Protection, Safety and Emergency Preparedness (EP) on a periodic basis a listing of MTF's which have been approved. If SR determines that an EA/EIS is needed, a recommendation will be made to HQ for determination. Upon concurrence by HQ of SR recommendations, SR will arrange for the contractor to prepare the NEPA document, identify and manage environmental studies, and provide guidance for the preparation and review of the document. Details of Du Pont, SR, and HQ actions and responsibilities are illustrated on Figure 2. This figure indicates a rather complicated flow of paper. The formalized paper flow is facilitated by regular discussions between Du Pont and SR regarding the status of NEPA projects and strategies. The interactions with DOE-Headquarters by SR are with the program office involved on specific projects or proposed actions or with the Office of Environmental Compliance on clarification concerning DOE requirements/ procedures.
EXPERIENCE This system was implemented by Du Pont in late July 1982 by a memo prepared and signed by EAPD and the Environmental Coordinator's Office and directed to all SRP/SRL department superintendents. Checklists are being received. In addition, several projects were caught in transition and the process applied to them even before the process was implemented. Table 1 is a list of projects to which this system has been applied through October 1982. Only about four out of this list would have been systematically considered (with EAPD input) without this system. If the present trend continues, approximately 80 projects will be considered annually. SR has received documentation from Du Pont for eleven proposed actions, Six MTF's have been prepared by SR and filed. Documentation for the remaining actions have not been determined. CONCLUSIONS Implementation has ensured early consideration of environmental factors for proposed actions at the SRP site and provided a system for documentation of these considerations. Implementation has also provided assistance to the custodians in pointing out when they need to get appropriate construction and operating permits for pollution control facilities from the state. The
177
S?P and SRL Activities Du pont Activity Custodian Prepares Checklist
I
• Inpacts - NO
Environmental Inpacts - Yes
File Checklist
EAPD and Custodian Perform Enviromental Evaluation Inpact Analysis (EEIA) Activity Covered by Test Authorization
,h DOE-SR and Du Pont Independently and/or Jointly Review Actions and Consolidate Caments
Clearly no Significant Enviromientai Inpact
Pont (EAPD) Recoimends to DOE-SR Recormends EA/EIS
Prepares Draft ADM
Prepares Draft MTF
Non-Section D Actions Uncertain of: o Significant Inpact o Type of NEPA Documant Needed
Section D Action DOE-NEPA Guidelines, e.g., Major New Construction or Probable Significant Environmental Inpact (for Non-Section D Actions)
MTF Prepared by EAPD Action Proceeds
Clearly no Significant Environmantal Inpact
Draft ADM Prepared by EAPD DOE-SR Prepares Final ADM and Sends to DOE-KJ 1. Are Inpacts Significant? .2. What Docuirents Needed?v 1.
?
2. EIS Not Certain
\ Draft MTF Prepared by Du Pont (SRL-EAPD)
1. Yes 2. EIS Needed Draft EA Prepared by Du Pont (EAPD) EAPD Coordinates NEPA Action with DOE-SR
Figure 2.
\ EAPD Coordinates NEPA Action with DOE-SR
NEPA Documentation Flowchart
178
V MTF Approved and Issued by DOE-SR Action Proceeds
TABLE 1. NEPA Activities, August through October 1982
Activity
Total
No Additional Documentation Direct Al] Nos Replacement
Project
22
4
Service Order
1
1
16
7
Test Authorization
3
4
MTF Issued SR
Du Pont
6
--
4
170
In Progress SR Du Pont 5
4
1
Environmental Coordinator's Office within SRP is the focal point for such permit interaction, although SR is responsible for obtaining the actual permit. The system is working well with a minimum of workload impact.
ACKNOWLEDGMENT The information contained in this article was developed during the course of work under Contract No. DE-AC09-76SR00001 with the U.S. Department of Energy.
180
4B NUCLEAR WASTE DISPOSAL AND NEPA COMPLIANCE
Barry H. Smith, M. L. Brown* O f f i c e of NWTS I n t e g r a t i o n B a t t e l l e Project Management D i v i s i o n
The purpose of t h i s paper i s t o present the Nuclear Waste Terminal Storage (NWTS) NEPA Program f o r s i t i n g h i g h - l e v e l r a d i o a c t i v e waste r e p o s i t o r i e s . This program was developed i n response t o the r e q u i r e ments of the National Environmental Policy Act (NEPA)' 1 ), Council of Environmental Q u a l i t y Regulations (CEQ) f o r Implementing t h e Procedural Provisions of the National Environmental Policy A c t ( 2 ) , and DOE Guidel i n e s f o r Compliance with t h e National Environmental Policy Act (DOE G u i d e l i n e s ) ( 2 ) . This paper w i l l address the f o l l o w i n g : t h e r e q u i r e ments of NEPA and implementing r e g u l a t i o n s , the NWTS Program plan f o r meeting these requirements, t h e implementation of the NWTS NEPA Program, and t h e possible changes t o t h e NWTS NEPA Program i f nuclear waste management i s enacted. The paper assumes t h a t the reader i s generally f a m i l i a r w i t h the s p e c i f i c requirements of NEPA, and the implementing CEQ r e g u l a t i o n s . Therefore, these requirements are b r i e f l y discussed while the DOE guidelines are explained i n greater d e t a i l . NEPA Requi rements NEPA imposes several duties upon federal agencies. One of t h e more prominent i s found i n Section 102, which requires each agency t o : Include i n every recommendation or report on proposals f o r l e g i s l a t i o n and other major Federal actions s i g n i f i c a n t l y a f f e c t i n g the q u a l i t y of the human environment, a d e t a i l e d s t a t e m e n t . . . o n (i ) (ii) (iii) (iv) (v)
The environmental impact of the proposed a c t i o n , Any adverse environmental e f f e c t s which cannot be avoided should the proposal be implemented, Alternatives t o the proposed action, The relationship between local short-term uses of man's environment and the maintenance and enhancement of long-term p r o d u c t i v i t y , and Any i r r e v e r s i b l e and i r r e t r i e v a b l e commitments of resources which would be involved in the proposed action should i t be implemented.
*The opinions expressed i n t h i s paper are the authors' and do not represent those of the Department of Energy.
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The "detailed statement" has come to be known as the Environmental Impact Statement (EIS). For over a decade, federal agencies have been dealing with the question of what actions require the preparation of an EIS. The response to t h i s controversial point has resulted in numerous court decisions and the promulgation of the CEQ regulations. There are two salient features of the CEQ regulations which are particularly relevant to the NWTS NEPA Program for deciding the question of when to prepare an EIS. The f i r s t is the CEQ regulation requiring a federal agency to identify Classes of Actions t h a t ( ^ ) : (1) normally do not require either EAs or EISs, (2) normally require EAs but not necessarily EISs, and (3) normally require EISs. For the Waste Management Program, DOE has i d e n t i f i e d the following Classes of Action. The second important feature is the CEQ requirement that agencies integrate the NEPA process into the agency decision-making process. DOE Classes of Actions Generally Applicable( 5 ) to Nuclear Waste Management Program
Normally doesn't requi re EAs or EISs
Normally requires EIss
Normally requires EAs but not
necessarily EIss
Exploratory and site characterization a c t i v i ties which by virtue of resource commitment or elapsed time for completion may foreclose reasonable site alternatives. Land acquisition activities solely for the purposes of reserving possible candidate sites and which do not prejudice future programmatic site selection decisions.
DOE actions resulting in the site selection, construction, or operation of major treatment, storage, and/or disposal f a c i l i t i e s for transjranic and high-level nuclear waste and/or spent nuclear fuel such as spent fuel storage facilities and geologic repositories (NoteClarifies an existing class of action).
The demonstration or implementation of intermediatedepth burial of low-level waste at DOE sites. (Note - Proposed on July 16, 1981, 46 FR 36884). 182
has implemented t h i s obligation by establishing three cateaor decision making. They are: policy level decision ma"•• ievel decision making, and project level decision making. Policy level decisions encompass broad strategies such as proposals for legislation or statements of national energy policy. Program level decisions involve DOE deciding on a variety of approaches to implement specific policies or statutory authorities, e . g . , adoption of a program plan. Project level decisions include deciding on specific action to executive a program or to perform a regulatory r e s p o n s i b i l i t y , e . g . , approval of projects. To assure integration of NEPA into the DOE decision process, DOE has also established general and specific procedures which are set forth in the DOE Guidelines. For the purpose of this paper, a discussion of the general procedures is sufficient to provide an overview of how t h i s i s accomplished. The specific procedures provide direction to o f f i c i a l s within the agency relative to each category of decision making. For additional information on specific procedures, one should examine the DOE Guidelines and the DOE Environmental Compliance Guide.( 6 ) The general procedures require DOE o f f i c i a l s to f i r s t i d e n t i f y and evaluate environmental factors and alternative courses of action. This is followed by a determination of the appropriate level of environmental r e v i i - (using as guidelines the Classes of Action discussed above). The next i ep is to commence the document as soon as possible and complete i t in . 'vance of final decision making. As the proposal is developed, DOE must ""ssure i t s e l f that a l l the alternatives being considered are encompass*. : in the relevant environmental document (EA or EIS)*. F i n a l l y , DC1-! must circulate the environmental document along with other decision-making documents during internal review. I f an EIS is prepared, a record of decision must be published. A requirement of the CEQ regulations which is found in the general procedure is that DOE u t i l i z e " t i e r i n g " in the preparation of NEPA documents. Tiering i s defined in the CEQ regulations as: "the coverage of general matters in broader environmental impact statements (such as national program or policy statements) with subsequent narrower statements or environmental analyses (such as regional or basinwide program statements or ultimately site-specific statements) incorporating by reference the general discussions and concentrating solely on the issues specific to the statement subsequently prepared. Tiering is appropriate when the sequence of statements or analyses i s
*CEQ regulations refer to EA's and EIS's as "environmental documents". 40 CFR 1508.10. This paper w i l l not use this nomenclature. Instead i t w i l l refer to them as "NEPA documents."
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(a)
From a program, plan, or policy environmental impact statement to a program, plan, or policy statement or analysis of lesser scope or to a site-specific statement or analysis.
(b)
From an environmental impact statement on a specific action at an early stage (such as need and s i t e selection) to a supplement (which i s preferred) or a subsequent statement or analysis at a later stage (such as environmental mitigation). Tiering i n such cases is appropriate when i t helps the lead agency to focus on the issues which are ripe for decision and exclude from consideration issues already decided or not yet r i p e " . ( 7 )
In other words, t i e r i n g i s using a broad EIS to evaluate general and non-site-specific impacts. Subsequent EISs or EAs w i l l incorporate by reference the general matters and focus on the issues which are r e l e vant to the proposed action presently contemplated by an agency. One can i l l u s t r a t e t i e r i n g as a series of building blocks. The f i r s t document has very general information and forms the foundation. The next series of documents have specific information which rely on the foundat i o n for support. The documents continue t o be placed upon one another u n t i l the f i n a l one i s set i n i t s place. While t h i s is a simple concept, i t s application can be complex. DOE's application of the above requirements to the NWTS Program and the examples of t i e r i n g are discussed in the next section.
NWTS NEPA Program The NWTS NEPA Program was published in the Final Environmental Impact Statement; Management of Commercially Generated Radioactive Waste( 8 ), and i n the DOE Statement of Position f i l e d in the NRC Waste Confidence Rulemakingw). After the Statement of Position was f i l e d , NRC promulgated 10 CFR Part 60 Procedural Rules and Amendments to 10 CFR Part 51 relating to high-level waste disposal. These rules and amendments require in s i t u testing during s i t e characterization at a potent i a l candidate s i t e , as well as alternative s i t e s . This regulatory approach caused DOE t o modify i t s s i t i n g schedule to conform t o these general requirements. Consequently, the nature of the decision being made at specific stages of s i t i n g changed as did the NEPA requirements. The current NWTS NEPA Program i s found in the Draft National Plan for High-Level Radioactive Waste Repositories. The DOE's NWTS Program has b u i l t i t s compliance with NEPA on the foundation of a broad programmatic EIS which evaluated the various alternatives for disposal of high-level nuclear waste generated by commercial nuclear power plants. The Final Environmental Impact Statement: Management of Commercially Generated Radioactive Wasted) examined the environmental effects of selecting a program strategy,
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including the selection of a preferred concept for waste disposal. The Record of Decision for selecting the mined geologic-disposal program alternative was published on May 14, 1981 (46 Fed. Reg. 26677). Building upon t h i s foundation, DOE has determined the major decision points in the s i t i n g process and the appropriate level of environmental review at each point. The anticipated NEPA documents for the major decisions are as follows: o
EA for adoption of a National Plan for Siting High-Level Radioactive Waste Repositories.
o
EA (or EIS) for selecting candidate sites for detailed s i t e studies, including exploratory shaft, and action to protect potential sites from other uses.
o
EIS for selection of a s i t e for a repository.
The purpose of the NEPA documents and how they are intended to f u l f i l l the requirements of NEPA are set forth below. EA for Draft National Plan for Siting High-Level Radioactive Waste Repositories The draft National Plan for Siting High-Level Radioactive Waste Repositories (National Siting Plan) describes the methodology, c r i t e r i a , and steps for screening regions, areas, and locations for potential sites to be studied i n detail prior to s i t e selection. The s i t i n g process was divided in three phases: the f i r s t step i s s i t e screening; the second is detailed s i t e studies; the final step is site selection. Note th?t the planned NEPA documents correspond with these phases. The EA evaluates the environmental impacts of the s i t i n g strategy and reasonable alternatives. I t also evalutes the anticipated range of environmental impacts of f i e l d a c t i v i t i e s which could be carried out during the s i t e screening phase. Environmental Evaluations (EE) or Environmental Assessments (EA) were prepared on a site-specific basis for f i e l d a c t i v i t i e s which were conducted prior to the publication of t h i s EA. In addition, checklists have been and w i l l continue to be prepared to evaulate the potential impacts of d r i l l i n g (the screening phase a c t i v i t y with the greatest potential for impact) on a s i t e specific basis. Both the EA and the draft National Siting Plan were published for public comment. These comments are now being reviewed by DOE. A decision w i l l soon be made as to whether a Finding of No Significant Impact (FONSI) can be made. The EA for the National Siting Plan is an example of u t i l i z i n g the t i e r i n g concept. Although the s i t i n g process consists of three phases,
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the EA only evaluates in detail the impacts of the s i t e screening phase. This i s not to say that the EA does not acknowledge future a c t i v i t i e s — i t does. However, because a future decision point has been i d e n t i f i e d , the detailed environmental evaluation of future a c t i v i t i e s w i l l be accomplished in a later NEPA document with a narrower focus. For example, the impact of constructing an exploratory shaft w i l l be examined in the NEPA document accompanying the decision to conduct detailed s i t e studies. EA or EIS for Detailed Site Characterization The next anticipated NEPA document is an EA (or EIS) t o accompany the decision to conduct detailed site studies, including construction of an exploratory shaft. The NEPA document w i l l review the process leading to selection of sites for detailed studies and quantify the potential impacts associated with detailed s i t e characterization, including construction of an exploratory shaft. Because t h i s phase i s very s i t e - s p e c i f i c , i t is possible that an EA is sufficient at one s i t e and that an EIS would be required at another. EIS for Site Selection Under the current NWTS NEPA Program, an EIS w i l l be prepared to accompany the decision to select a s i t e for a repository. This EIS would evaluate the site-specific environmental impacts of constructing and operating a repository, transporting and emplacing wastes, decommissioning, and reasonable alternatives to the proposed repository. This EIS w i l l more l i k e l y than not be integrated into the environmental report which w i l l accompany DOE's license application to the NRC. Land Protection During the detailed site study phase, DOE may take steps to protect the land from conflicting land use in order to protect the i n t e g r i t y of the s i t e and the investment of public monies in exploration work. The EA (or EIS) prepared for the s i t e characterization phase of the s i t i n g process w i l l most l i k e l y to be used as the NEPA document required for land acquisition or withdrawal. As a part of the s i t e selection process, DOE w i l l take action to acquire permanent ownership of a repository s i t e . The site selection EIS would be the logical NEPA document for land acquisition or permanent withdrawal.
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Implementation of the Program The program as outlined i n the preceding section is being implemented and appears to be working. As can be expected in such a long and controversial project, however, challenges to the planned program have been voiced. Examples of how the program is working, and on what basis i t has been challenged are discussed below. A successful program usually requires cooperation of various parties. In the NWTS Program, the Bureau of Land Management (BLM) has been involved with DOE's site screening efforts in the State of Utah. Because DOE requires access to lands managed by BLM, i t has been necessary to obtain authoriziation to conduct work. At the Area and Location phase of site screening, BLM prepared Environmental Assessments prior to t h e i r decision to allow DOE work to commence. These NEPA documents provide an example of cooperation in the NEPA process. This also demonstrates that even though federal agencies may have different responsibilities under NEPA, an agency objective can s t i l l be accompli shed. Although several issues have been raised which challenge the planned NWTS NEPA Program, only two of them are discussed here. One issue is that DOE i s "segmenting" the NEPA process by having early NEPA documents that address only the impacts of a proposed action and that do not look at the impacts of future actions. Some have urged that DOE examine site-specific repository impacts during the screening phase. Another issue is whether a programmatic EIS be prepared t o accompany the National Siting Plan. These issues are now being reviewed by DOE. Pending Legislation As t h i s paper i s w r i t t e n , the Congress is debating several b i l l s concerning the management of nuclear waste. Most b i l l s contain a "road map" for the NEPA Process. While i t not known what the legislation w i l l eventually require or i f any w i l l be enacted, this section provides examples of what may be required. A major element in a l l the b i l l s is that Congress is ready to declare that detailed site studies, i . e . , construction of exploratory shafts, is not an a c t i v i t y that requires an EIS. An EA would be prepared, but this would not be an EA as i t is contemplated i n the CEQ regulations. I t would evaluate environmental impacts and an e f f o r t would be made to mitigate significant adverse impacts. However a Finding of No Signficiant Impact (FONSI) would not be required. A number of b i l l s also require that the environmental s u i t a b i l i t y of potential sites as repositories be examined before a s i t e i s recommended for detailed s i t e characterization, u t i l i z i n g an exploratory shaft. The results would be documented in either a site recommendation report or the EA di scussed above.
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All b i l l s require that an EIS be prepared for the selection of a repository. The various b i l l s carefully define the content and scope of this EIS. For example, some b i l l s restrict the siting alternatives to be considered and preclude consideration of the need for a repository. An important element in almost all the pending b i l l s is the role of the NRC. Tha various bills caution that NRC should not duplicate DOE's effort. It should adopt, to the extent feasible, the EIS prepared by DOE. The proposed legislation also defines the scope of any EIS prepared by NRC. In summary, i f legislation is enacted, i t will certainly aid in defining the NEPA process for repository development. Conclusion In summary, DOE is complying with the requirements of NEPA by: (1) integrating the process into the overall NWTS decision-making process and (2) tiering i t s NEPA documents. The program for the disposal of nuclear waste has many components; the NWTS NEPA program is only one, albeit a very important one.
REFERENCES 1.
The National Environmental Policy Act of 1969, as amended, 42 U.S.C. Section 4321-4347.
2.
40 CFR Parts 1500-1508.
3. 45 Federal Register 20694 (March, 1980), as amended by 47 Federal Register 7976 (February, 1982). 4.
40 CFR Section 1501.4.
5.
47 Federal Register 7978.
6.
OOE/EV-0132 Vols. 1 and 2.
7.
40 CFR Section 1508.28.
8.
U.S. Department of Energy, 1980. Final Environmental Impact Statement: Management of Commercially Generated Radioactive Waste, D0E/EIS-0046F, Washington, DC, October.
9.
Statement of Position of the United States Department of Energy f i l e d in the Matter of Proposed Rulemaking on the Storage and Disposal of Nuclear Waste (Waste Confidence Rulemaking) Parts 50-51 (44 Federal Register 61372) DOE/NE-0007.
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4C
ROCKY FLATS PLANT INTERACTION WITH THE COLORADO DEPARTMENT OF HEALTH D. D. Hornbacher Rockwell International Golden, Colorado ABSTRACT The Rocky Flats Plant maintains a continuous interface with the Colorado Department of Health. The basis for this interface is contained in a memorandum of understanding. Monthly, a public meeting is held to share information. Typical topics discussed are Rocky Flats monitoring data, Colorado Department of Health environmental surveillance data, and water surveillance data from Broomfield, a city located downstream from the plant. Agenda items also include changes in plant operations and materials, and the plant monitoring program. Special topics of common interest such as new construction activities and health effect studies are periodically presented. The meeting is commonly attended by city, county, state and federal officials as well as the public. At times, the news media also attends. This interface with the State of Colorado and others provides a means to demonstrate cooperative actions and generally is beneficial to the plant. However, the potential for negative benefits is always present. This presentation will summarize the plant experiences gained through public meetings with the State of Colorado.
INTRODUCTION The Rocky Flats Plant is a government-owned nuclear weapons component fabrication facility that is operated by Rockwell International, Energy Systems Group, for the United States Department of Energy. Operations at the plant involve the use of radioactive materials such as plutonium, uranium and americium. Beryllium, a known toxic material, is also handled at the plant. The plant is located on a high plateau about 26 kilometers northwest of the Denver, Colorado metropolitan area at a location equidistant from the cities of Boulder, Broomfield, Westminster, Arvada and Golden. Windflow patterns from the plant are generally toward the Denver metropolitan area and surface water flows into two downstream reservoirs that are used for drinking water. 189
Because of the plant activities and its location, there has been a long-standing public interest in plant operations and assessment of any environmental impact that may be caused by these operations. In response to the public interest, plant officials and officials from the Colorado Department of Health have been routinely meeting for about 13 years to exchange environmental monitoring and surveillance information. This paper describes the nature of this meeting and the experiences gained through continued interface with the public. EXCHANGE MEETING HISTORY In 1970, officials from the Rocky Flats Plant and the Colorado Department of Health agreed to routinely held a technical monthly meeting to review and discuss environmental monitoring data. At that time, attendance was limited to personnel from the Rocky Flats Plant, local Atomic Energy Commission office (now Department of Energy), State Health Department and the U.S. Public Health Service. Typically, a total of about 10 people was in attendance and the meeting was rather informal. In the mid 1970's, public interest in plant activities and environmental protection increased and attendance at the exchange meeting was extended to the public. Attendance then increased to about 30 people including people from two counties, from neighboring city governments, from the U.S. Environmental Protection Agency, from the U.S. Department of Housing and Urban Development, from a Rocky Flats Monitoring Committee appointed by the Governor of the State of Colorado, and the general public. Typically, representatives from the various news media also attended. During this period, it became apparent that there was a need to formalize the exchange meeting and to provide better direction for response to the variety of questions and concerns that were expressed by the expanded audience. Often, the questions raised could not be adequately answered by those present without review of data and records. For these reasons, the exchange meeting was formalized to include an agenda. Following is the meeting agenda and a discussion of each agenda item. EXCHANGE MEETING AGENDA A.
List of Materials This is a listing Rockwell provides that shows the types of radioactive materials currently handled at the plant. It is provided tc "isure that the Ro ky Flats Environmental Monitoring Program and the Colorado Department of Health (CDH) surveillance programs are adequately designed.
B.
Monitoring Data 1. 2.
City of Broomfield Colorado Department of Health 190
3.
Rocky Flats Plant
During this part of the meeting, formal reports are presented and summarized that show the analytical results of samples collected during the previous month. C.
Engineering Update The engineering update is presented by Rockwell listing the status of plant construction projects that might have some type of positive or negative environmental impact. Some examples of projects that have been discussed in the past are: modifications of the plant sanitary waste treatment facility, construction of surface water control dams, construction of a reverse osmosis facility, solar pond decontamination and decommissioning, and soil removal activities.
D.
Changes to the Monitoring Program This agenda item is used to announce any changes that have been made to the plant environmental monitoring and control program during the previous month. Annually, a new catalogue of the environmental monitoring program is published and presented at the exchange meeting.
E.
New Agenda Items Frequently, the meeting attendees request information on specialized topics not routinely covered during the exchange meeting. In the past, impromptu responses often did not adequately give all of the necessary information. For this reason, this exchange of information was formalized and now consists of a prepared presentation by an appropriate expert from the plant, CDH, county or city to whom the request was directed. Following are some typical examples of topics that have been presented as special agenda items: Wind Flow Patterns at the Rocky Flats Plant, Evaluation of the Rocky Flats Plant-Designed Ambient Air Sampler, Air Handling and Filtration Technology, Historical Environmental Data Trend Evaluations, the Rocky Flats National Pollutant Discharge Permit, Environmental Data Calculation Methodology, Environmental Impact from Medical Facilities, Uranium Mining Operations and Private Industry, and Health Effect Studies. As can be seen from this listing, this exchange meeting agenda item provides an excellent opportunity to share knowledge and information on environmental monitoring, control and protection topics that are of common interest to the plant, public health officials and the public.
PUBLIC PERSPECTIVE AND UNDERSTANDING One of the challenges of the exchange meeting is to provide technical data and information in a manner that is meaningful and understandable to the public and the news media. To some extent, this is accomplished by
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data comparison with the radioactivity present in household items such as smoke detectors and camping lantern mantles, cosmic radiation, medical X-rays and by comparison of data to data from background locations. Also, comparisons are made to historical data. However, public review and acceptance of the data becomes a feedback process of acceptability for current plant operating practices. As the values decrease, the acceptable range of values are also reduced and even slight deviations above the lowered norm for expectation may give rise to public inquiry. From this process, the public essentially has an influence in the "As Low As Reasonably Achievable" practice. Experience has shown that public perspective of the information presented depends to a large extent on the manner in which the information is presented and the choice of words used. The audience will quickly become aware of a person who is not comfortable with the data or does not display confidence. To use an old cliche, bedside manner is extremely important! Public perspective also is highly dependent on the choice of words used for the exchange meeting presentation. Words must be very carefully selected that accurately describe the data. For example, a person responsible for an environmental control program has a working vocabulary of words that are used to command the appropriate attention to work situations that can affect the environmental control of effluent discharges to the public. If there are activities that could and should be done to control effluents, environmental data might be described with words such as significantly higher, marginally acceptable, disturbing or not acceptable. As can be imagined, the list is infinite. Assuming that the environmental monitoring data are well within publicly accepted guides and standards, the use of these adjectives in a public setting can convey a sense of urgency not appropriate to the reported data. POSITIVE AND NEGATIVE ASPECTS There are both positive and negative aspects to the exchange meeting. Some examples of each follow: A.
Time Approximately 80 man hours of time are required to prepare and print a formal report and to travel and participate in the. actual meeting.
B.
Data Validation Occasionally, the validity of some data may be in question at the time for the meeting. Time constraints limit the amount of additional work, such as reanalysis and evaluation, that can be conducted to verify the results.
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C.
Audience Background The audience consists of people with both highly technical and nontechnical backgrounds. This makes it difficult to present the information in an understandable manner for all present.
D.
News Media Releases Negative news media releases may occur, which is a risk associated with any public meeting.
Positive Aspects A.
Routine Data and Information Exchange The meeting provides for continued routine exchange and comparison of environmental monitoring data between the Rocky Flats Plant and the Colorado Department of Health. This interface prevents the possibility of surprises and serves to maintain an open working relationship.
B.
Public Interface The exchange meeting provides a means for continued public interface. It gives the public the opportunity to be better informed and allows the opportunity to raise questions and resolve concerns. This is a mutual benefit to all meeting attendees.
C.
Environmental Protection Through the use of special agenda items, the exchange meeting gives the plant operating staff a means to describe ongoing projects that are being conducted to enhance environmental control and public protection. This leads to public confidence and an understanding that the plant is a good neighbor.
D.
News Media Releases The news media has the responsibility to disseminate information to the public. Due to the exchange meeting, information is obtained that does result in positive news media releases.
E.
Credibility The monthly exchange meeting is an excellent means to maintain continued credibility with public health officials, the public and the news media. This is probably the most important positive aspect of the meeting.
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SUMMARY AND CONCLUSIONS Rockwell International has a strong commitment to continually keep the public informed regarding possible environmental impacts from plant operations. The monthly environmental exchange meeting is one of several methods used to meet this objective. Experience gained through many years of participation in this meeting indicates that it is an excellent communication tool and that it is mutually beneficial to those who attend.
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40 THE INFLUENCE OF PCB AND HAZARDOUS WASTE REGULATIONS ON THE OPERATIONS OF THE WESTERN AREA POWER ADMINISTRATION
Ken E. Mathias J. John Kirby Western Area Power Administration Golden, Colorado
ABSTRACT The Western Area Power Administration (Western) has made operational, economic, and administrative adjustments to comply with recent regulations. Federal regulations dealing with PCB's (Toxic Substances Control Act, PCB Final Rule, 40 CFR 761) and hazardous waste (Resource Conservation and Recovery Act, Hazardous Waste Regulations 40 CFR 261-265) influence operational, economic, and administrative procedures. Additional Federal and State regulations add to the complexity. Dissemination of information by training and environmental protection audits to Western's f i e l d personnel provide awareness of requirements so that they may be incorporated into routine operating procedures. Inspections, handling, transportation, and disposal are operational requirements that are regulated and for which Western Headquarters Division of Environmental Affairs provide support. Economic impacts are primarily from the disposal of PCB wastes and the replacement of equipment. A Western-wide 10-year PCB capacitor replacement program was developed at an estimated cost of 6.75 m i l l i o n dollars. A Western-wide PCB o i l treatment contract i s being considered to reclassify equipment i n which approximately 300,000 gallons of 10 to 4,200 ppm PCB o i l is contained. A Western-wide contract, rather than by current individual procurements, i s planned for a more economic bulk transportation and disposal of PCB Items which cannot be handled by other treatment methods. Administrative adjustments to direct compliance such as directives, guidelines, and direct problem assistance, are carried out at the Headquarters Division of Environmental A f f a i r s . Audit inspections under DOE and Western Order 5482.1 are directed
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from the office. A Western Hazardous Waste Management Committee, consisting of Headquarters and area personnel has been formed to provide a coordination, Information, and problem solving forum for local and Western-wide PCB and hazardous waste problems. Application of regulatory requirements, keeping in mind realistic operational, economic, and administrative constraints, has resulted In regulatory compliance which does not cause system disruption or extra personnel inconvenience.
INTRODUCTION The Western Area Power Administration, one of the five Department of Energy power marketing administrations, 1s responsible for the transmission and marketing functions in 15 states; an area of 1.3 million square miles. Over 225 substations, and numerous maintenance facilities are required to assure operation of the transmission system. Some materials utilized in the equipment, as well as some materials used to service and maintain the equipment, have been classified for regulation by Federal and State bodies. Federal regulations dealing with polychlorinated biphenyls (PCB's) (Toxic Substances Control Act, PCB Final Rule, 40 CFR 761) and hazardous wastes (Resource Conservation and Recovery Act, Hazardous Waste Regulations, 40 CFR 261-265) have mandated operational, economic, and administrative adjustments to assure compliance. Other Federal and State requirements provide further complication. This paper summarizes the impact of the regulations and describes the methods utilized to effectively deal with those impacts. OPERATIONAL Western's field personnel have had to incorporate environmental regulatory requirements into their operation and maintenance regimes. Providing awareness of these requirements has been achieved by offering training courses that deal with the PCB and hazardous waste regulations and implementing an environmental protection audit program. An inspection program is a common requisite for both the PCB and hazardous waste regulations. Inspections of PCB equipment, PCB storage areas, and hazardous waste storage facilities are conducted in conjunction with routine daily, monthly, and quarterly operations and maintenance procedures. The recently released (August 25, 1982), PCB Final Rule has had a major Impact on the PCB Transformer Inspection program. The former PCB rule requirement of quarterly inspections for all PCB Transformers
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(greater than 500 ppm PCB) has been reduced to annual inspections. Only PCB Transformers greater than 60,000 ppm PCB or posing an exposure risk to food or feed must continue to be inspected quarterly. Approximately 98 percent of Western's PCB Transformers are less than 60,000 ppm, which decreases the level of e f f o r t in the inspection program by almost 75 percent. Transporting PCB's and hazardous wastes for disposal requires preparat i o n , execution, and monitoring until the wastes are properly disposed of. Western's f a c i l i t i e s are provided with functional support from Headquarters Division of Environmental Affairs on selecting experienced and economical vendors to transport their wastes for proper disposal. Correct l a b e l i n g , manifesting, and recordkeeping are required for every shipment. The Area (or D i s t r i c t ) Offices are responsible for providing necessary personnel to supervise a l l transport and disposal operations from preparing s p e c i f i cations to assuring that the wastes have reached their proper disposal destination. ECONOMIC The economic impacts of the PCB and hazardous waste regulations are primarily from the disposal of wastes and replacement of equipment. The majority of the costs incurred thus far have been from PCB capacitor replacement and disposal. In 1978 Western began t h e i r own 10-year capacitor replacement program prior to the April 22, .1982, PCB Proposed Final Rule calling for the phasing out of Large High-Voltage Capacitors over a 10-year period. (The August 25, 1982, PCB Final Rule allows the usage of PCB Large-, High-, and Low-Voltage Capacitors that do not pose an exposure risk to food or feed, that are within a restricted access and/or indoor f a c i l i t y , for the remainder of t h e i r useful l i v e s . ) The Western replacement program was i n i t i a t e d for two reasons: one-third of the substations that contained capacitor banks were nearing or exceeding their 20-year service l i f e ; and the discarding of PCB's within Western's system would eliminate future operational requirements, such as the need for PCB storage f a c i l i t i e s and individual transportation and disposal procurements. The cost estimate for the PCB capacitor replacement program is approximately 6.75 m i l l i o n in 1982 d o l l a r s . Eleven to twelve thousand PCB capacitors w i l l be replaced, e f f e c t i v e l y removing a half m i l l i o n pounds of PCB's from potential environmental exposure. Disposal and transportation cost for the capacitors w i l l be one-half to three-quarters of a m i l l i o n d o l l a r s . Replacing PCB capacitors with non-PCB capacitors w i l l cost approximately 4.8 m i l l i o n d o l l a r s . Another 1.2 m i l l i o n dollars w i l l be spent on labor, materials, t e s t i n g , and engineering. Western has approximately 300,000 gallons of PCB o i l (concentrations of PCB's ranging from 1G ppm to 4,200 ppm) that can u t i l i z e one of the
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commercially available PCB oil treatment methods for lowering the PCB concentration. This PCB oil is located in bulk oil storage tanks, oil circuit breakers, reactors, regulators, and transformers. Some of the oil-filled PCB electrical equipment (up to 69 kV energized) could be treated while inservice. A Western-wide oil-treatment contract is being contemplated. It is expected that a large volume Western-wide contract would result in considerable cost savings over treatment by individual area locations. Cost estimates for the oil treatment is 1.3 million dollars. A Western-wide contract is also planned for the transportation/disposal of PCB Items that cannot use the oil treatment method. These items comprise 75 percent of Western's total PCB inventory. A single contract to remove PCB Items would enable Western's Area and District Offices to budget their disposal costs based upon stable prices, with an expected cost savings because of a large volume contract. Other PCB equipment not covered under the capacitor replacement or oil treatment program is currently being inventoried and tested to determine the best environmental/economic method(s) for managing them (i.e., pure PCB Transformers and miscellaneous PCB Contaminated Electrical Equipment). ADMINISTRATIVE Most of the administrative adjustments to assure compliance with PCB and Hazardous Waste Regulations are carried out at Western Headquarters Division of Environmental Affairs. Headquarters has an environmental compliance staff which maintains contact with the fifteen state and six Federal environmental agencies responsible for various PCB and hazardous waste regulations. All environmental regulations that affect Western's operations are continually reviewed for their impact. Headquarters analyzes, interprets, and issues functional directives to field offices to help assure compliance. For example, the recent PCB Final Rule (August 25, 1982), had changes that affected Western's field operations. Headquarters reviewed the regulation and issued summaries to the Area and District Offices. The summary consisted of interpretations of specific parts of the regulation that would directly affect Western's operations. Other day-to-day assistance by Headquarters includes preparation of a toxic and hazardous materials handbook, preparation of reporting forms, and direct interaction with the various regulatory agencies on behalf of the field offices. An environmental protection audit program as directed by DOE and Western Orders 5482.1 has been instituted to assure that field offices comply with applicable environmental regulations. Audit inspections are 198
conducted regularly and are closely coordinated with the field offices and Headquarters. The audit inspection program has been effective in assisting field facilities to solve and/or prevent environmental problems. In addition to the audit program, a Western Hazardous Waste Management Committee consisting of Headquarters and Area Office personnel has been formed. The Committee meets regularly to discuss and solve local and Western-wide PCB and hazardous waste problems. SUMMARY The numerous influences of PCB and hazardous waste regulations, namely to operational, economic, and administrative phases of Western's activities, have demanded development and implementation, usually at relatively short notice, of compliance methods and physical resources. Through application of regulations with realistic operational, economic, and administrative constraints kept in mind, it has been discovered that adherence to compliance requirements is possible, without system disruption or excess personnel inconvenience.
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4E COAL CONVERSION:
THE CH EXPERIENCE
Vicki L. Alspaugh U. S. Department of Energy Argonne, Illinois
ABSTRACT
In FY 1980, reflecting concerns of economy and national energy policy, the Chicago Operations Office reconverted Argonne Boiler No. 5 from gas/ oil to coal firing. Efforts to promote this conversion required extensive legal and technical coordination, and consisted of (1) designing and performing modifications to the physical plant; and (2) securing state construction and operating permits. To satisfy the requirements of the National Environmental Policy Act of 1969, an Environmental Assessment was conducted, which resulted in a Finding of No Significant Impact (FONSI). Environmental and cultural evaluations also were performed under the National Historic Preservation Act of 1955; Executive Order 11593, Protection and Enhancement of the Cultural Environment; and Executive Order 11988, Floodplains Management. However, the major obstacle to operation of the converted plant was provisions of the Clear Air Act.
BACKGROUND Argonne National Laboratory (AND is a multidisciplinary research and development laboratory managed by the Chicago Operations Office of DOE and operated by the University of Chicago as a Government-owned, Contractor-operated facility. It is located on a 1700 acre site in Downers Grove Township, DuPage County, Illinois, 22 miles southwest of Chicago. The boiler plant at ANL is located at the northeast corner of the site. It consists of five boilers that provide the steam requirement for heating and refrigeration for the entire laboratory. Boiler No. 5 is the largest and newest of the five boilers, providing some 58% of the
201
steam requirement. It was installed in 1965, and was originally designed for coal firing. In 1973, it was converted to natural gas and oil firing for environmental reasons. The investigation into feasibility of reconverting Boiler No. 5 to its original coal firing capability was initiated in FY 1978. The major concern was that, although conversion of the boiler from gas/oil to coal firing and upgrading of the coal handling equipment could be completed before the FY81 winter heating season, the necessary additional pollution control equipment could not be installed and properly adjusted for operation before April of 1982. Meanwhile, emissions of particulate matter and sulfur dioxide would exceed applicable emission standards, even utilizing low sulfur, low ash Illinois or eastern coal. Relief was sought so that DOE could coal fire Boiler No. 5 during the interim period between conversion of the boiler to coal firing and installation and operation of the additional pollution abatement equipment without being subject to civil and criminal enforcement proceedings under the Clean Air Act. REGULATORY FRAMEWORK Under the mandates of Executive Order 12088 of October 13, 1978, Federal Agencies must comply with State permit requirements adopted under the authority of a Federal environmental statute. The 1977 Clean Air Act Amendments (P.L. 95-95) explicitly require Federal facilities to comply with all State requirements "in the same manner, and to the same extent as any nongovernmental entity" (42 USC §7481). The definition of federal and state roles will, of course, differ from state to state. In Illinois the basic framework for pollution control is set forth in the Illinois Environmental Protection Act (the Act), which creates three functionally distinct state agencies with environmental protection/pollution control responsibilities. The Illinois Environmental Protection Agency (Illinois EPA or IEPA) is the largest of the three agencies, and administers the state permit system. It also gathers and analyzes data, and on the basis of these data, proposes substantive regulations to the Illinois Pollution Control Board (IPCB or Board), which functions in a legislative role (adopting the state's administrative pollution regulations) in addition to sitting as an "environmental court". IEPA also makes recommendations to the Board on whether to grant or deny a variance from a substantive standard. The third state agency, the Illinois Department of Natural Resources, is an environmental research institute which also performs the economic impact studies on proposed Board regulations.
202
The Illinois air pollution regulations are found in Chapter 2 of the Illinois Pollution Control Board Rules and Regulations. They are enforceable as state law, but beyond that, they form the heart of the Illinois State Implementation Plan (SIP) under section 110 of the Clean Air Act (CAA), 42 USC 7410, and are enforceable as federal law. The Illinois Environmental Protection Act also explicitly adopts the provisions of: 42 USC 7411 (New Source Performance Standards) and 42 USC 7412 (National Emission Standards for Hazardous Air Pollutants), and has adopted rules which are in substance identical with federal regulations to implement those sections (Section 9.1 of the Act.) The Board also has adopted regulations establishing a permit program meeting the requirements of Section 173 of the Clean Air Act, 42 USC 750. In an effort to identify our obligations and options relative to emissions from the modified powerplant, we met with representatives of USEPA and Illinois EPA in November of 1978; additional meetings were held subsequently. IEPA advised that three state permits would be needed for the facility: (1) a construction permit for the coal storage area; (2) a construction permit for the dry scrubber and baghouse; and (3) an operating permit for the modified facility during the "interim period" (between reconversion to coal-firing capability and construction of the additional pollution abatement equipment). The Chicago Operations Office thereupon contracted with PaDCe Environmental, of Cincinnati, Ohio, to conduct worst case computer modelling analyses required by Illinois EPA. Baseline levels had to be determined, and air quality modelling studies conducted to predict emissions during the interim period and final operations. Worst case emission rates and air quality impacts were calculated assuming no sulfur dioxide (SO 2 ) controls and an approximate removal efficiency of 85% for particulate (TSP) emissions from the existing multicyclone system. The analysis predicted, for each type of coal, the concentration of both TSP and SO2 to occur at or beyond ANL's property line. TABLE 1.
Coal Eastern (Low S)
Sulfur Content .72%
Illinois
1.47
High S
3.16
POTENTIAL TO EMIT
Ash Content
Heat Content
TSP Emission Rate
so 2
Emission Rate
6.0%
13,000 Btu
12 .05 g/s
28. 17 g/s
7.6
11,750
16 .90
63.64
11,595
22 .31
10
(PEDCo 1980(b))
203
137. 7
On tht- basis of the PEDCo report on the "Air Quality Impact of Coal Conversion for Boiler No. 5", Illinois EPA determined that during the interim period, Boiler No. 5 would violate certain regulations of the Pollution Control Board, and stated that on that basis, they were prohibited from issuing a permit for the operation of Boiler No. 5 on coal. AIR QUALITY IMPACTS The ANL steam plant is located in the Metropolitan Chicago Interstate Air Quality Control Region (AQCR). This portion of DuPage County has been designated by the Illinois Environmental Protection Agency (45FR6786) as an attainment area with respect to the primary national ambient air quality standards (NAAQS) for sulfur dioxide and total suspended particulates, but as nonattainment with respect to the secondary NAAQS for TSP. Portions of adjacent counties of Will and Cook have been designated as nonattainment areas for both pollutants. National Ambient Air Quality Standards (NAAQS) The modelling analysis indicated that the contribution of Boiler No. 5 emissions to the existing air quality (worst case values) in the vicinity would not cause a violation of the NAAQS for SO2. However, the projected increases to ambient air concentration for TSP for Boiler No. 5 would contribute to a "significant" violation of the secondary NAAQS for that pollutant. The Illinois Environmental Protection Agency has established Rules for Issuance of Permits to New or Modified Air Pollution Sources Affecting Nonattainment Areas which provide deminimis ("insignificant") increments of 1 ug/nH (annual average) and 5 ug/m3 (24-hr, average).) However, we could not be certain of meeting that limitation (Table 2). TABLE 2. MAXIMUM INCREASES TO AMBIENT AIR CONCENTRATIONS Eastern Coal
TSP
Ann. Georo. Mean 24-hr. Average
so 2
Ann. Geom. Mean
1.5 ug/m^
12. 3.5 ug/m3
Illinois Coal 2.1 ug/m3
17. 7.9 ug/m3
24-hr. Average
28
64.
3-hr. Average
87
196.
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Air Programs The primary programs under the Clean Air Act for new source regulation are: Prevention of Significant Deterioration Rules, New Source Performance Standards, National Emission Standards for Hazardous Air Pollutants, and the Offset Policy for Nonattainment Areas. Each of these programs was considered during the course of this project, consistent with our environmental protection mandate. A number of the programs did not apply to the modified boiler. New Source Performance Standards for fossil fuel fired steam generators were not applicable because Boiler No. 5 has a heat input rate of less than 250 million Btu/h (40 CFR 60.40(a)(l)). The National Emission Standards for Hazardous Air Pollutants are emission limitations on asbestos, beryllium, mercury, and vinyl chloride, which were not pertinent to this case (40 CFR 61). Therefore, the only air quality regulations to be met for Boiler No. 5 during operation with, and without controls, were the SIP emission limits, the PSD consumption increments, and the state emission limits. STATUTORY RELIEF The Powerplant and Industrial Fuel Use Act of 1978 (FUA), 42 USC 8301 et. seq., (which superseded the Energy Supply and Environmental Coordination Act of 1974 (ESECA), 15 USC 791 et seq.) provides the Secretary of Energy with the authority to prohibit the use of petroleum or natural gas as a primary energy source in any existing major fuel burning installation which previously had the technical capability to use coal and which could be reconverted to such capability without substantial physical modification of the unit. This authority has been redelegated to the Economic Regulatory Administration of the Department of Energy. The plants subject to the FUA are defined as "major fuelburning installations" (MFBI's), and include any stationary unit consisting of: a boiler, gas turbine unit, combined cycle unit or internal combustion engine which has a designed fuel heat input rate of 100 million Btu's per hour or greater. 42 USC 8202(10(A) The major focus in on boilers. The Fuel Use Act is very closely tied to the Clean Air Act, and provides temporary relief from Clean Air Act mandates. For example, Section 111(a) of the Clean Air Act, 42 USC 7411(a) Standards of Performance for New Stationary Sources, provides:
205
(8) A conversion to coal (A) by reason of an order under section 2(a) of the Energy Supply and Environmental Coordination Act of 1974 or any amendment thereto, or any subsequent enactment which supersedes such act, ...shall not be deemed to be a modification... (emphasis added) Section 112 of the Clean Air Act, 42 USC 7412, National Emission Standards for Hazardous Air Pollutants (NESHAPS) provides that the terms "stationary source", and "modification" have the same meaning as such terms have under section lll(a), therefore incorporating the exclusion for mandatory conversions of boilers under FUA. Regulatory Exclusions Consistent with their statutory mandate, USEPA developed regulations implementing these sections of the Act, which expressly exclude coal conversions under FUA. New Source Performance Standards (40 CFR 60.14) expressly exclude "Conversion to coal required for energy considerations, as specified in section lll(a)(8) of the Act" from consideration as a modification under that part. Likewise, under USEPA regulations for Prevention of Significant Deterioration, a physical change (necessary for a "major modification" and applicability of the program) shall not include "Use of an alternate fuel or raw material by reason of an order...(under FUA)..." (40 CFR 51.24, 52.18). The same exclusion is present in the definition of major modification for application of new source review (40 CFR 42.24, 51.18). Furthermore, in determining compliance with a maximum allowable increase under PSD, the Administrator may exclude "concentrations attributable to the increase in emissions from stationary sources which have converted from the use of petroleum products, natural gas, or both by reason of an order in effect under sections 2(a) and (b) of the Energy Supply and Environmental Coordination Act of 1974 (or any superseding legislation) over the emissions from such sources before the effective date of such order..." 40 CFR 52.21(f). Incidently, the same regulations also exclude from the definition of major modification, the use of an alternative fuel or raw material by a stationary source which it was capable of accommodating before a certain date (for PSD purposes, January 6, 1975 under 40 CFR 51.24, 52.18; for new source review purposes, December 21, 1976 under 40 CFR 51.18, and July 1, 1979 under 40 CFR 52.24).
206
Delayed Compliance Order (DCO) However, the most important statutory relief is found in section 113(d)(5) of the Clean Air Act, 42 USC Section 7413(d)(5), which provides for an Order postponing the date for final compliance "of any requirement under an applicable implementation plan," for a stationary source which is prohibited from burning petroleum products and natural gas under an order pursuant to the provisions of FUA. Therefore once the Economic Regulatory Administration issued a proposed Prohibition Order for Boiler No. 5, the Manager of the Chicago Operations Office requested USEPA, Region V, to prepare a Delayed Compliance Order pursuant to 42 USC 7413(d)(b). The request sought permission for DOE to delay compliance with certain requirements found in Part II of Chapter 2 of the Illinois Pollution Control Board Rules and Regulations insofar as such requirements had been incorporated into the Illinois SIP: (1) Rule 202(b), Visual Emission Standards and Limitations for All Other Emission Sources; (2) Rule 203(g)(l)(A) Particulate Emission Standards and Limitations for Fuel Combustion Emission Sources; and (3) Rule 204(c)(l)(A) Sulfur Dioxide Emission for Existing Fuel Combustion Sources. USEPA drew up a proposed Delayed Compliance Order during the summer of 1980, which granted the requested postponement, and established a compliance schedule and a number of interim requirements, including use of the best practicable system of emission reduction for the period during which such order was to be in effect. The proposed DCO was sent to EPA headquarters, and to the Governor of Illinois for concurrence; the latter action was taken pursuant to their interpretation of §15 USC 792, which was passed as part of P.L. 95-95, and states: ...any certification or notification required to be given by the Administrator shall be given only when the Governor of the State in which is located the source to which the proposed order under section 7413 (d)(5) of the Clean Air Act is to be issued gives his prior written concurrence. State Variance By letter dated July 2, 1980, Illinois EPA submitted comments to the Manager of the Chicago Operations Office regarding the proposed
207
USEPA delayed compliance order. Briefly, the Agency stated that an air quality violation (in this case, Rule 307(a)(2)(B), secondary ambient air quality standard for particulate matter) prevented it from issuing a permit without a variance from the Illinois Pollution Control Board. Further, the letter continued, "Until a variance is obtained, it is this Agengy's recommendation that the Governor of the State of Illinois not concur in the delayed compliance order proposed to be issued by the Administrator in this matter". As a result of the Agency's position, we filed a Petition for Variance with the Illinois Pollution Control Board on August 28, 1980, requesting that the Board grant relief from the sulfur dioxide and particulate emission rules, and the secondary standard for particulate matter. This Petition was dismissed on September 4, 1980, for lack of Board jurisdiction. The Board did not immediately recognize its authority to grant a variance from an Ambient Air Quality Standard, and rules 203(g)(l) and 204(c) (1)(A) had been ruled invalid by the First District Appellate Court in Commonwealth Edieon v. PCB, 25 Ill.App.3d 271, 323 NE2d 84 (1974), and in Ashland Chemical Corp. v. Pollution Control Board, 57 Ill.App.3d 1052, 373 NE2d 826 (1978). (The rules were remanded to the Board to correct procedural deficiencies in their adoption, and are being reconsidered in a pending rulemaking (R80-22). Meanwhile the rules are unenforceable as state law, although valid federal law as part of the SIP.) However, IEPA agreed that ambient air quality standards are enforceable limitations, and therefore variances from them may be granted upon presentation of adequate proof that compliance with them would impose an arbitrary or unreasonable hardship. Therefore, the Petition was reopened on October 2, 1980, and granted (after public hearing and over an IEPA recommendation to the contrary) on December 4, 1980. Conclusion The Final Delayed Compliance Order was published in the Federal Register on January 27, 1981. Under the terms of the DCO, Boiler No. 5 had to achieve final compliance with the requirements of the Illinois SIP by March 15, 1982. A permit for the operation of Boiler No. 5 during the interim period was issued February 4, 1981, and Boiler No. 5 was put on the line firing low sulfur, low ash Illinois coal on February 10, 1981. The new dry scrubber was dedicated on March 22, 1982, after initial operation began in January of 1982. A final Prohibition Order was never issued. POSTSCRIPT The Omnibus Budget Reconciliation Act of 1981 (0BRA), p.L. 97-35 Subtitle B, amended section 301 of the Powerplant and Industrial Fuel
208
Use Act of 1978 (42 USC 8341) by revoking ERA authority to take hostile action against existing electric powerplants. As a result, on Wednesday, November 10, 1982, the Economic Regulatory Administration issued revisions to its final rules under 10 CFR 500 et seq., substantially amending the prohibition order process regarding existing facilities (47FR50846). This is consistent with OBRA and DOE's announced policy of not issuing involuntary prohibition orders to existing MFBI's under section 302 of FUA. (The amendatory language for section 301 relating to electric powerplants is notably similar to the language in section 302 relating to major fuel burning installations (MFBI's) as it was implemented in 1978.) OBRA did not, however, affect the Administration's Clean Air Act jurisdiction. Prohibition Orders, despite their limitation to voluntary conversions, continue to provide relief from PSD, BACT, and NSPS requirements. This is a significant inducement. Therefore, to the extent that the surviving legislation remains tied to the Clean Air Act, the prohibition order is still a viable mechanism, and conversions will continue to be economically feasible. Regarding the Clean Air Act, which is before Congress for renewal this year, at this time it appears unlikely that there will be any attempt to eliminate those "breaks", due to the stalemate between factions representing the conflicting concerns of energy independence and acid rain legislation.
209
4F
ENVIRONMENTAL REGULATORY ASPECTS OF ALTERNATE LIQUID FUEL USE AT RROOKHAVEN NATIONAL LABORATORY
M. J . Bebon Brookhaven Area O f f i c e - IISDOE Upton, New York
J. R. Naidu L. C. Emma Brookhaven National Laboratory - All I Upton, New York
ABSTRACT Toxic chemical wastes in the environment represent one of the most serious problems facing society today. The need, therefore, for proper disposal and control of hazardous wastes is well recognized. Effective control techniques are available, however, the best means of disposal for toxic organic chemical wastes is hiqh efficiency combustion. This is so because many of them have significant heating values and are valuable substitute or supplemental fuels. Regulating the combustion of hazardous wastes is a complex matter. This arises from the unknowns resulting from complex chemical reactions and/or molecular recombinations, which are exotic potential contaminants. Locally the New York State Department of Environmental Conservation (NYSDEC) has regulated this disposal, through combustion, of such chemical and hazardous waste. BNL which depends heavily on #6 oil for steam production, has been a forerunner in developing a patented system of integrating chemical wastes and other surplus or reprocessed fuels with virgin #6 oil and successfully used a blend of these materials as a fuel in the Central Steam Facility. The question of staying within the boundary limits of State Regulations while maintaining economic viability has been an educational experience. This paper explores the approach taken by DOE and BNL to meet the stringent reguirements of regulation, environment and costs.
211
INTRODUCTION The Alternate Liquid Fuels (ALF) program at Brookhaven National Laborat o r y (BNL) was developed during the period 1976-1979 in response to shortaqes of v i r g i n fuel o i l , rapidly escalating prices, and the dictates of the Fuel Use Act of 1978. Since the inception of the ALF program i t has been possible to avoid the use of over 18 m i l l i o n gallons of c r i t i c a l v i r g i n fuels at a savings of approximately 4 m i l l i o n d o l l a r s . Alternate Liquid Fuel as defined by BNL is a blend of non-virgin low viscosity liquids called l i g h t feedstocks (LFS) with a v i r g i n No. 6 or nonv i r g i n residual o i l . Due to viscosity limitations of the fuel handling equipment and burners the LFS portion of the blend has been limited to 50%. Modifications currently nearing completion w i l l eliminate this r e s t r i c t i o n and allow f i r i n g of 100% LFS. Typical characteristics of LFS are shown in Table 1 . 1 . Table 1.1.
LFS Characteristics BTU/qal Ash Lead Sulfur Flash Point
90,000 - 140,000 0.04 - 2.0% 10 -
130ppm
0.02 - 0.5* 70 - 16OF
Cadmium
0
-
4ppm
Chromium
0
-
lOOppm
Total Haloqens PCB
0.02 - o.«s% <10ppm
I t should be noted that the ALF blend contains significantly lower ash and sulfur than No. 6 fuel oil resulting in lower sulfur and particulate emissions. Also, there are good reasons to believe that ALF constituents contribute to improved combustion, functioning as catalysts to promote a more complete burn of the as-fi~ed fuel material. Perhaps the most unique feature of the ALF blending process is the long term homogeneity of the fuel product. Samples prepared in 1976 (and stored since then) have not exhibited any stratification to date.
212
The primary consumer of ALF at RNL is the Central Steam Facility (CSF) which provides steam for heatinq, humidification and air conditioning to facilities in the central core area of the site. Four boilers provide a total plant capacity of 345,000 lbs/hr, 125 psiq saturated steam. Plant output peaks at about 130,000 lbs/hr during the winter season. Fach boiler has its own stack. No air pollution control equipment has been required or installed on the boilers. Data on the CSF boilers is presented in Table 1.2 below. Table 1.2
Central Steam Facility Roiler Data
Roiler No.
Installed
Manufacturer
Nameplate Ratinq (lh/hr)
1A
1963
Rabcock X Wilcox
45,non
3
1955
Combustion Enqineerinq
fin.noo
4
1961
Combustion Engineering
6n,non
5
1965
Combustion Enqineerinq
lRn.oon
In addition to the economic and environmental advantaqes mentioned above, the ALF process provides an efficient method for the beneficial recovery and re-use of spent resource materials such as solvents and surplus fuels which miqht otherwise pose severe problems if released to the environment in a uncontrolled manner. This aspect is particularly meaninqful in areas such as Lonq Island where the indiscriminant disposal of such material could easily cause significant contamination of the qround water which is our sole source of potable water supply. Recognition of its benefits underscores the need to place appropriate controls on the alternate liquid fuel program. Indeed the chemical waste disposal industry has been a favorite target for those individuals who would subvert the go*.ls of environmental protection for their own personal economic gains. In liqht of this, DOE and BNL continue to implement those measures contained in applicable federal, state, and local regulations which mandate analytical testing and environmental monitorinq in connection with the use of these materials. In this reqard we have pursued continuing dialogue with both the federal and state authorities to maintain the application of sensible and yet not overly restrictive regulations on the use of alternate liquid fuel materials.
213
REGULATORY ASPECTS The environmental regulatory agency for New York State is the Department of Environmental Conservation (NYSDEC). Within this organization, the Division of Air Resources is responsible for regulating air emissions points under their 200 series regulations. These regulations are promulgated under the authority of the Clean Air Act and thus, under Executive Order 1208a, are applicable to U.S. Department of Energy Facilities. Another aspect of ALF program regulation is the solid waste area under the Resource Conservation and Recovery Act (RCRA), and NYSDEC's 360 series regulations. Since the applicability and jurisdictional aspects of the latter are currently being discussed with NYSDEC, this paper will concentrate on the air emissions aspects. One last regulatory aspect of importance is the Toxic Substances Control Act (TSCA) PCB regulations. NYSDEC issued the most recent certificates to operate the CSF boilers in December, 1978 with an expiration date of November 30, 1981, It was the pending expiration of these certificates, coupled with the expansion and maturity of the ALF program that prompted tr,e NYSDEC to develop special conditions to supplement the renewal certificates. To their credit, the NYSDEC Air Division representatives approached this task with an open mind and a willingness to develop conditions unique to the ALF program and the physical facilities at the BNL-CSF. NYSDEC began discussions with us in December 1980. At that time their policy for regulating non-virgin fuel was in its early stages of development and their familiarity with the ALF program, minimal. As a result, the initial proposed conditions were restrictive and in some cases, inapplicable to the supply aspects of the non-virgin fuel marketplace. ALF was incorrectly characterized as "waste oil and solvents". During the last two years we have participated, with NYSOEC, in a process of modifying and refining those originally proposed conditions to achieve our goal of "sensible yet not overly restrictive" regulations on the use of ALF. In addition to on-site orientation tours for NYSDEC regional and headquarters representatives and frequent meetinqs and discussions, we arranged for NYSDEC to visit our ma.ior non-virgin fuel supplier and to discuss various ALF program aspects with EPA Office of Federal Activity representatives. Our success in meeting our goal can be seen in Appendix A which compares the original and final conditions. The development of the special conditions for BNL's CSF certificates parallelled NYSDEC's development of their policy on combustion of non-virgin materials as a fuel or fuel supplement. Initially a draft policy memorandum on Combustion of Waste Crankcase Oil was prepared April 6, 1981. A draft
214
memorandum on Use of Burnable Chemical Waste as Fuel followed on December 8, 1981. The latter document classified burnable chemical waste (BCW) into four categories including "low chlorine content RCW" which proved to be an appropriate cateqorization for LFS as well as the blended ALF. The current NYSDEC regulatory emphasis on the ALF program is on application of their 360 series "Solid Waste Management" regulations. We have questioned their jurisdiction in this area in light of DOE's position visa-vis RCRA as well as the fact that the 360 proqram has not received FPA approval. As an act of comity we have submitted a permit application however it was qualified as having been submitted "under protest", Our application is currently under review. While NYSDEC headquarters has approved the latest proposed conditions for the CSF boiler Certificates to Operate, the NYSDEC reqional administrator has directed his staff to delay issuance of the certificates until the solid waste issues are resolved. During this interim period, BNL is voluntarily complying with the proposed conditions of the Certificates to Operate. ENVIRONMENTAL ASPECTS Underlying our lengthy discussions with NYSDEC and EPA was the general perception on their part that BNL had been exercising responsible measures to assure that the ALF program would not have a detrimental effect on the environment. Early in the implementation of the program at the CSF, BNL's Environmental Monitoring Group assessed the effect of ALF usage on ambient air guality at the site boundary. Table 3.1 lists the caTculated concentrations of SO2, N0 x and' particulates released at the CSF stacks and the resultant site boundary concentrations for the year 1979. Table 3.2 indicates the projected site boundary concentrations of lead and cadmium under various ALF blending and usage assumptions.
215
Table 3.1.
1979 BNL Emission of SO2, N0 x and Participates Central Steam Facility Average
EPA Primary
stack
boundary
Air finality
concentration8
concentration
Standard
Calculated
Effluent
S02
3.25
x 105
270 ppm
1.06
x 10-3 ppm
0.03 ppm
N°x
1.44 x lf)5
156 ppm
6.33 x 10"* ppm
0.05 ppm
Particulates a
Total kq
3.! x 10"
0.09 q m"3
0.35 yq m"
from
75 pq nr 3
3
liSEPA AP-42 Compilation of Air Pollutant Emission Factors
Table 3.2.
1979 BNL Projected Emissions of Pb and Cd Central Steam Facility under Varying ALF Usage Modes Lead (Pb) Minimum
Maximum
<1
273
<3.8xlO-6
<1.9xlO"6
from
Cadmium (Cd) Minimum
Maximum
1.04xl0-3
<0.38xl0"6
1.52xlO"5
0.52x10-3
<0.19xl0" 6
O.76xlO"5
Concentration in ALF Sample (ppm) Concentration at Site Boundary (yq m"3) a) if only ALF was burned b) if burned at the maximum possible mixture 50% ALF/50% #6 Standard ( g m"3)
a
USEPA
b
NYSDEC
216
In both tables above, the site boundary concentrations are compared with ambient air standards. I t was clearly demonstrated that the usage of ALF did not pose a health or environmental r i s k . These studies were based on s i t e specific meteorological studies and CSF stack heights. The results of these early analyses were subsequently confirmed when NYSDEC diffusion models indicated allowable as-fired heavy metal levels far beyond what is or has been experienced in LFS deliveries. These higher levels were incorporated into the proposed conditions as indicated in Table 3.3. Table 3.3.
NYSDEC Allowable Heavy Metal Levels vs. Typical Light Feedstock Values
NYSDEC Max. Allowable (ppm)
LFS Typical (ppm)
Lead (Pb)
2,500
130
Cadmium (Cd)
5,500
4
500
100
Chromium (Cr)
In addition to the conventional air contaminants, combustion of nonvirgin fuels or burnable chemical wastes creates the potential for formation of more exotic contaminants throuqh complex chemical reactions and/or molecular recombinations. Factors influencing the probability of these contaminants include, total halogens and PCB content as well as fuel combustion efficiency. BNL had been testing for total halogens and this was carried over into the conditions. NYSDEC focused on the application of the TSCA PCB regulations to the ALF program durinq the winter of 1980-1981 when the problem of PCB-contaminated natural gas line condensate became a regional concern. NYSDEC's i n i t i a l requirement was for BNL to sample each incoming LSF tank truck for PCB's prior to allowing i t to off load. When we advised them this was not feasible due to PCB testinq turn-around times of 24-36 hours and associated tank truck demurrage costs, they requested hold-up tanks. The cost of twelve, ten-thousand gallon holding tanks with associated pumps, piping, dikes and l i n e r s , the increased risk for a s p i l l , and the long construction lead time made this alternative similarly unattractive as a short-term solution. At this point, we made a counter proposal that NYSDEC allow a l l incoming material to be offloaded in the 210,000 gallon "blending" tank and a composite sample tested. We cited 40 CFR 761.10 (g)(2) as a basis which
217
states " — s a m p l e s of waste oil may be taken from a common container—". NYSDEC conditioned their approval of this procedure on EPA's concurrence. EPA subsequently approved the approach. The procedure has been in use since early 1981 and has been refined to testing a proportional sample composited from aliquots taken from each incoming LFS tank truck. This composite sample is analyzed for PCB's and the LFS are not blended with residual oil until the results are known. A further requirement added by NYSDEC is that a composite sample represent not more than 100,000 gallons of LFS. One interesting aspect of the NYSDEC requirement to analyze LFS samples for PCB's has been the difficulty we experienced in obtaining accurate analyses. Local laboratories were accustomed to analyzing dielectric fluids and when confronted with LFS samples containing many organics, took an ultra conservative approach and labelled unidentifiable GC peaks as PCB's. Subsequent analyses by laboratories in the metropolitan NY-NJ area more familiar with LFS-type materials proved these results erroneous. As a result BNL has developed in-house capability for PCB analysis. BNL has also constructed a small holding tank facility using six, tenthousand gallon surplus tanks, which provides additional flexibility for storaqe and analysis of incoming LFS. An additional 720 thousand gallons of storage capacity is expected to be on line by sprinq 1983, as we move toward 100% LFS usage. NYSDEC included requirements for a minimum combustion efficiency of 99.9% and continuous CO and 0£ monitoring of each stack in their initial proposed conditions. As their policy development advanced, and their familiarity with the ALF program grew, they relaxed these requirements to 99.0% combustion efficiency with an option of in situ or portable monitorinq equipment. Two recent testing programs were conducted at the CSF to measure fuel combustion efficiency. The first was conducted in conjunction with the CSF modifications to allow firing of 100% LFS. This testing proqram was performed by Fuel and Energy Consultants, New York, New York in December 1980 and consisted of steady state Orsat readings at various load plateaus. Boilers 1A, 3 and 4 were tested using viroin #6 oil; boiler 5 was tested using both virgin #6 oil and ALF. The results are shown in Table 3.4 below. Calculated temperature and residence time data are also included.
218
Table 3.4.
Central Steam F a c i l i t y - Fuel Combustion E f f i c i e n c y vs. Output, Retention Time, and Temperature Calculated
Boiler
% of
First Pass
First Pass Bulk
Fuel
Output
Rated
Retention
Gas Temperature
Combustion
(Lb/Hr)
Capacity
Time (Seconds)
Range (°F)
Efficiency(%)
10,000
22
4.45
#6 Fuel Oil 19,000
42
2.33
99.2805
28,500
63
1.56
97.3509
33,700
75
1.31
98.0392
45,000
100
0.98
2600 °F
98.6666
14,500
24
8.17
2200 °F
100.0000
#6 Fuel Oil 32,000
53
3.70
97.2222
40,000
67
2.92
98.6301
54,500
91
2.15
97.3333
60,000
100
1.96
2600 °F
98., 6666
18,000
30
6.53
2200 °F
100.0000
#6 Fuel Oil 32,000
53
3.70
100.0000
40,000
67
2.92
98.6111
54,000
90
2.17
97.3684
61,000
102
1.92
2600 °F
95.8904
33,500
19
8.94
2200 °F
95.0413
#6 Fuel Oil 68,000
38
4.47
97.9729
90,000
50
3.40
99.3197
133,000
77
2.20
96.1538
175,000
97
1.75
2600 °F
97.2973
5
35,000
19
8.94
2200 °F
99.2000
ALF
64,000
36
4.72
100.0000
(50/50
108,000
60
2.83
98.6301
165,000
92
1.84
97.2222
200,000
111
1.53
1A
3
4
5
blend)
219
2200 °F
2600 °F
100.0000
98.6666
Subsequent testing of Boiler No. 5 on ALF was accomplished in April 1982 specifically for the permitting process. The testing was accomplished by Reliance Energy Services Inc. and consisted of continuous monitoring of CO and 02 over the boiler load range using a continuous infrared analyzer (MSA Lira Model 303). Results of from 99.19% to 99.97% fuel combustion efficiency indicate there will be l i t t l e difficulty in complying with the revised 99.0% criterion. Similar testing of boilers 1A, 3, and 4 is scheduled for early 1983. CONCLUSIONS The dialogue that has gone into certifying the CSF to burn ALF has been a learning experience. Clearly DOE and BNL have shown their desire to respect environmental concerns and have made every effort to live within the limitations of a cost-effective methodology. In doinq so, major achievements have been made in analysis of ALF samples for PCB, Pb, Cd and in developing a tank farm for the purpose of storage and blending of LFS. Modifications are under way to utilize LFS at its full strength and thereby gain further independence from imported oil. NYSDEC has been able to use this experience as a testing area for their regulatory development and presenting BNL as a lead agency in disposing of material in an environmentally sound manner.
220
APPENDIX A Comparison of December 1980 and Auqust 1982 Certificate Conditions for BNL Central Steam Facility December 1980 Auqust 198? The waste oil/solvents mus t The spent oil/solvent must contain less than 50 be PCB free. ppm of PCBs. No solvents or waste oils The concentration of total haloqens in the fuel containinq total haloqens in as fired must, not exceed 0.5% by weioht. excess of o.5% may be burned. The concentration of heavy metals in the fuel as fired must not exceed the followinq limits: Lead - 2,500 PPM Cadmium - 5,500 PPM Chromium - 500 PPM A fuel sample and analysis Fuel samplinq/analysis conditions: will be performed weekly by A spent oil/solvent sample shall be taken the supplier. The results from each tank truck delivery to BNL. of the analysis will be submitted monthly to the NYSDEC From the individual tank truck samples menRegion 1 office by the suptioned above, a composite sample representinq plier. In addition, on a not more than 100,000 qallons shall be made quarterly basis an indepenon which the fuel analysis will be conducted. dent certified laboratory will sample and analyze the The individual tank truck samples will not be fuel as sold to Rrookhaven entirely composited or disposed of until the National Laboratory and for- analytical results of the composite sample ward the results directly to confirm its acceptability. the NYSDEC Reqion I office. The above samplinq to take The analysis of all composite samples will be place at supplier's depot. performed by a laboratory acceptable to the The analysis shall as a min- NYSDEC. The results of the analysis must be imum report concentrations kept on record at BNL and made available for of PCB's, total haloqens and inspection by NYSDEC or its aqent. heavy metals. The fuel samplinq/analysis for PCBs must be In addition to fuel samplinq completed before the ALF blending process at the supplier's depot, beinqs. samplinq may be conducted periodically ab Brookhaven The fuel sampling/analysis shall as a minimum report concentrations of PCBs, total haloNational Laboratory at the qens, lead, cadmium and chromium. discretion of the NYSDEC or its aqent. Samplinq may also be conducted periodically at BNL at the discretion of the NYSDEC or its agent; accordinqly, a sampling valve must be provided as fired basis is required.
221
Appendix A, Continued A stack test or diffusion analysis is to be completed within sixty (60) days from the date of issuance of the permit, The stack test or diffusion analysis will be performed for particulates and for heavy metals. Brookhaven National Laboratory will provide continuous exhaust gas monitoring instruments to monitor their process exhaust gas emissions for carbon monoxide and excess oxygen. Periodic manual tests for carbon dioxide will be performed and correlated with the CO and O2 measurements. Combustion efficiency will be computed from these measurements. Combustion efficiency must be no less than 99.9%.
Brookhaven National Laboratory will provide exhaust gas monitoring instruments to monitor their exhaust gas emissions for carbon monoxide and excess oxygen; combustion efficiencies will be computed from these measurements. If continuous monitors are installed carbon monoxide and excess oxygen measurements will be taken at least once for. every three months of boiler operation. If portable monitors are used carbon monoxide and excess oxygen measurements will be taken at least once annually. Combustion efficiency: Combustion efficiency of 99% must, be demonstrated by 12/31/82. If a combustion efficiency of 99% must be demonstrated by 12/31/82 If a combustion efficiency of 99% cannot be demonstrated and BNL does not wish to pursue the trial burn/testing option, this certificate may be subject to revocation.
A sampling valve must be provided at Brookhaven National Laboratory as well as at the fuel supplier's storage depot.
The name and address of all spent oil/solvent suppliers must be kept on record at BNL to be available for inspection by NYSDEC or its agent.
222
Appendix A, Continued The fuel supplier for this facility is [name fuel supplier^)]. In the event of a chanqe to a different supplier, the NYSOEC Region I office must be notified thirty (30) days prior to use of any of the new suppliers product. Permit period not to exceed one year.
The facility must be in compliance with 6NYCRR Part 360, 364 and 365. The quantity and composition of fuel burned will be restricted so as not to contravene any of the followinq standards: Federal and/or primary ambient, air quality standards. Federal and/or state emission standards. Incremental increases in air pollution emissions in excess of the de minimus values established by "Federal Prevention of Significant Deterioration" requlation. Acceptable ambient levels as may be established by the Commissioner (listed below) specifyinq maximum concentrations of air contaminants he deems allowable to protect the public health, welfare, or confort of residents of the state. Contaminant Cadmium Chromium Hydrogen Chloride
223
Acceptable (Annual) Ambient Level 2 uq/m 3 3 0.17 uq/m 2.33 uq/m 3
SESSION FIVE EFFLUENT MONITORING
5A
NUCLEAR AIR CLEANING: THE NEED FOR A CHANGE IN EMPHASIS Eugene H. Carbaugh .. Pacific Northwest Laboratory^' Richland, Washington 99352
ABSTRACT The nuclear industry now has over 35 years of experience in nuclear air cleaning. This experience covers technology development, system design, operations, and maintenance. Much of the past experience has been directed towards technology development with particular emphasis on high efficiency particuiate air (HEPA) filters. Implementation of this technology has lagged its development by a number of years. A recent study examines the causes and frequencies of HEPA filter changeouts and failures. These data lead to a conclusion that a shift in emphasis from technology development to the training of personnel and the designing and maintaining of such systems is needed. Some highlights of the data and a discussion of topics which should be addressed in training will be presented.
INTRODUCTION Nuclear air cleaning has long been recognized as an essential process for containment and control of radioactive material. V i r t u a l l y all f a c i l i t i e s handling radioactive materials include some engineered form of air cleaning to control radioactivity and minimize releases to the environment. The most common method of nuclear air cleaning is the use of high e f f i ciency particuiate a i r (HEPA) f i l t e r s . Application of these f i l t e r s ranges from single f i l t e r s on glove-box vents to multiple banks of hundreds of f i l t e r s on building exhaust systems. The industry now has over 35 years of experience in a l l phases of a i r cleaning including research, development, applications to f a c i l i t y design, system operation, and maintenance. Much of this experience has been
(a) Operated by Battelle Memorial Institute for the U.S. Department of Energy under Contract DE-AC06-76RL0 1830. 227
documented in the thousands of pages of proceedings from the 17 nuclear air cleaning conferences which have been sponsored by the Department of Energy (DOE) and i t s predecessors, the Energy Research and Development Administration (ERDA), and the Atomic Energy Commission (AEC). Long overdue topical access to these proceedings has been provided with the recent publication of an index to the f i r s t 16 conferences (Burchsted 1981). In addition, nuclear air cleaning has been the subject of American National Standards I n s t i t u t e (ANSI) standards (ANSI 1980a,b), U.S. Nuclear Regulatory Commission (NRC) Regulatory Guides, (U.S. NRC 1978) and a handbook (Burchsted, Kahn and Fuller 1976). Widespread application of HEPA f i l t e r s led to questions concerning the magnitude and reasons for f i l t e r f a i l u r e s . As an aid to direct future technology development e f f o r t s , the Pacific Northwest Laboratory assisted the DOE's Airborne Waste Management Program Office by surveying HEPA f i l t e r experience at DOE s i t e s . Data was requested on the numbers of and reasons for HEPA f i l t e r changeouts and failures for the years 1977-1979. The survey data was tabulated and published in a 1981 report (Carbaugh 1981) and presented at the 17th DOE Nuclear Air Cleaning Conference (Carbaugh 1982). Some implications of the survey data are the subject of t h i s paper.
HEPA FILTER SURVEY RESULTS Survey questionnaires were distributed via DOE Operations Offices to all DOE site contractors reporting 1979 releases to the DOE Effluent Information System data base. Data were requested for systems operated for the years 1977-1979. Usable data was sent from 24 DOE site contractors and covered 342 filter banks ranging from 1 to 790 filters each, for a total of 9,154 filter applications. Several sites indicated that compiling data to complete the questionnaires was a difficult process because records were not readily retrievable or were not available. One contractor indicated no data was available regarding changeouts or failures. Other sites did maintain adequate records to complete the questionnaire. The total number for HEPA filter applications, changeouts, and failures are listed in Table 1. As used in this paper, an application is considered to be a single slot for a HEPA filter. A system consisting of one filter is considered a single application as is a single slot within a bank or array of filters. The relatively low filter failure incidence {\?.%) over three years would appear to indicate th .. filters are generally performing their intended task. Furthermore, the ratio of total filters changed to filters failed (approximately 6 to 1) indicat d that most filters are changed out prior to failure. Further evidence that HEPA filters are performing well was provided by the data from the questionnaire on the reasons for filter changeouts. It is presented in Table 2. From this data it is apparent that the largest
228
TABLE 1.
HEPA F i l t e r Experience Summary
Description
Number
% of Total
Filter applications
9,154
100
Applications witlj no changeouts
3,870
42
Applications in which changeouts occurred
5,284
58
Applications having more than one changeout
1,610
18
Filter failures
1,105
12
TABLE 2.
Filter Changeout Summary
Reason for Changeout
Number
% o f Total
High differential pressure (AP) across f i l t e r
4,333
63
Leak-test failure
1,020
15
Preventive maintenance service 1ife>
871
13
Suspected damage
376
5
Radiation buildup
256
4
38
_<1
6,894
100
Other reasons (unspecified) TOTAL
majority (63%) of f i l t e r changeouts were attributed to high differential pressure (AP) across the f i l t e r , indicative of f i l t e r plugging. This reason was four to five times higher than the next most prevalent reasons which were leak-test failure and preventive maintenance service-life l i m i t s . Other reasons for changeouts were relatively infrequent compared to the preceeding ones. I t can be concluded from Table 2 that 15% of a l l changeouts were performed based on actual evidence of failure ( i . e . , leakage). This number compares favorably with the incidence of reported failures identified in Table 1.
229
The remaining 85% of the changeouts were performed for a variety of preventive maintenance reasons. Of particular interest, are those changeouts identified as preventive maintenance service life, because they assume that changeout is made only on the basis of filter age and not because of particular evidence of failure or plugging. The close approximation of service life changeouts to reported failures would seem to indicate that reasonably good judgment has been exercised in establishing preventive maintenance service lives. Exposure of filters to a single significant environmental factor is tabulated with changeout incidence in Table 3. The vast majority of filters were reported as exposed to no distinguishing environmental characteristics (i.e., they filtered essentially clean, dry, air environments similar to those that might be found in typical building ventilation systems or in systems with good pre-HEPA treatment features). The highest frequency of changeouts appeared to occur in hydrofluoric acid (HF) or high moisture environments. These environments also reflected the highest frequency of changeouts for leak-test failure and high &P. Improvement in the acid/ moisture resistance of filters was the most frequently stated development need, specifically cited by four sites. Filter failure modes and their reported incidence are listed in Table 4. Data indicate that the majority of failures occurred for unknown or unreported reasons, and that most filter failures are either not investigated as to cause, or, if such an investigation is performed, it is not documented in a sufficiently retrievable manner. When the unknown failure modes are eliminated from Table 4, it is noted that failure due to handling or installation damage is approximately equal to all other modes of failure combined, and at least three times more prevalent than any one of the component failure modes (frame, gasket, media, construction, or sealant). The apparent predominance of handling and installation damage was supported by three sites which offered subjective comments on that failure mode. The incidences of frame failures, gasket or seal failures, and filter media ruptures were approximately equal; each constituted 5-6% of all filter failures. Where frame failure was identified, essentially all failed frames (58 out of 65) were wood. No observation of steel frame failure was reported, and the remaining 7 failures were of unspecified frame type. The predominance of wood frame failures could be due to frame warping or cracking caused by overtightening hold-down clamps. Data concerning gasket or seal failure was somewhat inconclusive. Gasket seals were reported in much wider use (3,920) than fluid seals (151), and proportionally more filters with gasket seals failed (422) than those with fluid seals (14). However, only 40 gasket seals and three (3) fluid seals were specifically identified as failing. Calculating the ratios of seal-type failures to seal-type applications can lead one to
230
TABLE 3. Single Environment Versus Changeouts
Environment A l l single environments
Filter Applications
Tota Changes
Leak Test
Other Penet.
High AP
Radiation Buildup
623 (0.15)
14 (<0.01)
2,059 (0.49)
4,190
4,042 (0.96)
5
4 (0.80)
High moisture
82
327 (3.99)
60 (0.73)
High dust
85
199 (2.34)
6 (0.07)
Solvent
Grease/Oil
Suspect Damage
Service Life
Visual
192 (0.05)
376 (0.09)
766 (0.18)
12 (<0.01
134 (1.63)
107 (1.30)
1 (0.01)
25 (0.30)
129 (1.52)
18 (0.21)
4 (0.80)
—
46 (0.54)
6
6 (1.00)
High temperature
12
12 (1.00)
1 (0.08)
.__
Hydrofluoric (HF) acid
12
44 (3.67)
26 (2.17)
18 (1.50)
.
108
114 (1.06)
6 (0.06)
108 (1.00)
7 (0.06)
__.
—
3,880
3,336 (0.86)
524 (0.14)
1,660 (0.43)
67 (0.02)
364 (0.09)
695 (0.18)
j
Co
Other acid
No Distinguishing
6 (1.00)
14 (<0.01)
11 (0.92)
_--
12
(a) Entries are reported occurrences with the calculated frequency ratio of changeout to filter applications given below each occurrence.
TABLE 4.
Filter Failure Modes
Failure Mode
Number
% of Total
Failure mode unknown
702
64
Handling or installation damage
213
19
Frame failure
65
6
Gasket or seal failure
62
6
Media rupture
54
5
Filter construction
6
<1
Media to frame sealant failure
3_
<1
1,105
TOTAL
100
conclude that little difference may exist between gasket and fluid seal failure rates. Media rupture could be due to a variety of causes including media breakdown, separator sagging, frame damage, or sealant failure. Possible differences in filter media were not addressed in this survey since it was assumed that essentially all filters used fiberglass paper. However, out of 54 reported occurrences of media rupture, several correlations with other filter component characteristics were noted. Specifically, all media ruptures occurred in wood-frame filters, and practically all (48) had aluminum separators and polyurethane foam sealants. Sealant failure was not identified as a significant filter failure mode (3 occurrences reported out of 1,105). A general lack of reported cases of filter failure attributed to faulty construction (6 in 1,105) would appear to indicate that vendor and contractor quality control programs are collectively, functioning well. Failure modes as correlated with single characteristic environment exposure are presented in Table 5. Noteworthy in this table are the high frequencies of failure in HF acid and high moisture environments. Ratios of filters failed to filter applications in which these failures were experienced are several times higher for HF acid and high moisture environments than for environments having no distinguishing characteristics, or the average of all single environment exposures. Also, the incidence of media rupture appears to occur most frequently in HF acid applications.
232
TABLE 5.
Environment All single environments
ro
2,017
Filters Failed
fedia Rupture
Sealant Failure
Frame Failure
Gasket or Seal Failure
Filter Construction
Hbndlinq or Installation Damage
620 (0.31)
48 (0.02)
2 (0.001)
7 (0.003)
6 (0.003)
2 (0.001)
19 (0.009)
536 (0.27)
1 (0.029)
47 (1.34)
35
48 (1.37)
High dust
1
1 (1.00)
High temperature
6
1 (0.17)
High moisture
CO CO
Filter Applications
Hydrofluoric (H 7 ) acid
12
26 (2.17)
Other acid
63
12 (0.19)
1,900
532 (0.28)
No d i s t i n g u i s h ing characteristics
Single Environment Versus F a i l u r e Mode(a)
Unknown
1 (1.00) 1 (0.17) 26 (2.17)
22 (0.012)
2 (0.001)
7 (0.004)
5 (0.003)
—-
10 (0 .159)
2 (0.032)
2 (0 .001)
8 (0 .004)
486 (0.25)
(a) Entries are reported occurrences with calculated frequency ratio of failure mode to f i l t e r applications qiven below each occurrence.
IMPLICATIONS OF THE DATA The foregoing discussion indicates that filters properly installed in systems with pretreatment sufficient to give relatively clean, dry, air prior to HEPA filtration can be expected to have minimum frequencies of changeouts and failure. Improvements in HEPA filter acid or moisture resistance can be expected to reduce changeout and failure frequencies to some extent. However, more significant reductions in changeout and failure frequencies, appear possible by improving system designs and worker training programs. HEPA filters are not to be mere additions to process effluent streams like garbage disposals are to kitchen sinks. Rather, they are the heart of an air treatment system which may be required to function under a wide range of normal and abnormal facility and process conditions. An air treatment system, like any engineered system, requires careful evaluation of the operating conditions, especially the nature of the effluent being treated. The data show that HEPA filters perform particularly well when exposed to relatively clean, dry air. This implies that pretreatment features should be incorporated into any air cleaning system design to remove the harshest air stream contaminants. For the most part these pretreatment systems are straightforward; they include scrubbers for chemical and particulate removal, moisture traps, de-entrainers, heaters to lower the relative humidity, and good quality prefilters to remove bulk quantities of particulates. Particular attention to the adequacy of air cleaning systems is needed when facilities are modified or new processes added. Occasional design engineering tradeoffs can result in a relatively short HEPA filter service life. One example of such a tradeoff is the sizing of fans or blowers for minimum filter resistance. This results in capital cost savings for initial designs by using smaller fans. However, such cost savings can rapidly vanish if HEPA filters must be replaced more frequently due to partial dust loadings which limit fan capacity. Human factors of system design can also contribute substantially to filter life. Systems must be designed for periodic filter testing and replacment using a minimum of worker effort. Filter banks must be easily accessible and stacked at a reasonable height; typically not more than three high. Adequate clearance from obstructions must also be included in system designs so that filters are not bumped and damaged during installation. The above considerations are not new. The Nuclear Air Cleaning Handbook has provided excellent detailed design guidance for some years. The challenge seems to be to make better use of the guidance presently available. Finally, there is the issue of worker training. With the majority of identified filter failures attributed to improper handling and installation, significant reductions in filter failure frequency appear possible
234
through improved worker training programs. While it is assumed that most sites provide some minimal training to those workers involved with filter handling and installation, the scope, content, and quality of this training may be highly variable. Subjective responses to the HEPA filter survey questionnaire indicated that quality training programs had reduced HEPA filter damage by workers. Another site indicated a continuing need for trained, experienced HEPA filter workers. Worker training is a subject which has not received much attention at air cleaning conferences. This was evident from a search through the index of the first 16 conferences which revealed that there were no entries under the keyworkd "training." Training programs are needed to develop an awareness by those handling filters, of the delicacy of the filter media and the major impact small leaks can have on overall system operation. Filters must be stored and installed in the proper orientation, with separators vertical and not horizontal or flat which can cause sag and rupture the media or separate it from the frame. The tendency to overtighten filter hold-down mechanisms must also be addressed; it can result in crushed, broken, or warped frames that may not be readily visible. Training programs must also be flexible enough to meet the needs of small sites where one person might be responsible for the entire air cleaning system and also for large sites where several craft trades might be involved. CONCLUSIONS High efficiency particulate air filter technology appears to have developed to the point where filters can be expected to have relatively slight chance of failure. This is contingent upon three factors: (1) adequate pretreatment of air prior to the use of HEPA filters, (?.) proper installations and (3) periodic surveillance and preventive maintenance. While some reduction in the incidence of filter failure can be accomplished with improved technology, the most significant reductions appear possible through improved training. This training should address two distinct groups: 1) those responsible for the engineering and design of nuclear air cleaning systems, and 2) those involved with handling, installing, and maintaining HEPA filters. The review, definition and implementation of this training is the challenge now presented to air cleaning. REFERENCES American National Sandards Institute/American Society of Mechanical Engineers. 1980a. Nuclear Power Plant Air Cleaning Units and Components. N509, New York.
235
American National Sandards Institute/American Society of Mechanical Engineers. 1980a. Testing of Nuclear Air Cleaning Systems. N510, New York. Burchsted, C. A. 1981. An Index to the 1 st Through 16th AEC/ERDA/DOE Nuclear Air Cleaning ColvFerence. DOE/TIC-11405, Oak Ridge National Laboratory, Oak Ridge, Tennessee. Burchsted, C. A., J . E. Kahn and A. B. Fuller. 1976. Nuclear Air Cleaning Handbook. ERDA 76-21, Oak Ridge National Laboratory, Oak Ridge, Tennessee. Carbaugh, E. H. 1981. Survey of HEPA F i l t e r Applications and Experience at Department of Energy Sites. PNL-4020, Pacific Northwest Laboratory, Richland, Washington. Carbaugh, E. H. 1982. A Survey of HEPA F i l t e r Experience. PNL-SA-10213, Pacific Northwest Laboratory, Richland, Washington, po" be published in proceedings of the 17th DOE Nuclear Air Cleaning Conference). U.S. Nuclear Regulatory Commission. 1978. Design, Testing and Maintenance C r i t e r i a for Post Accident Engineered-Safe'ty-Feature Atmosphere Cleanup System Air F i l t r a t i o n and Absorption Units of Light-Water Cooled Nuclear Power Plants. Regulatory Guide 1.52, Office of Standards Development, Washington, D.C.
236
5B
OPERATIONAL EXPERIENCE WITH TWO TRITIUM EFFLUENT MONITORING SYSTEMS John S. Haynie and Jose A. Gutierrez Los Alamos National Laboratory Los Alamos, NM
ABSTRACT The Los Alamos National Laboratory has designed and built two new tritium stack monitoring systems. The operational experience ofga wide^ range detector with a useful range of a few pCi/m to 10 yCi/m , and a second monitoring system using an improved Kanne chamber and a new electrometer, called a Model 39 ElectrometerChargemeter are discussed. Both tritium chambers have been designed to have a reduced sensitivity to tritium contamir-ation, a fast response, and an integrating chargemeter with digital readout for easy conversion to microcuries. In addition, this paper will discuss the calibration of these monitors and will point out advantages of using these chambers over conventional systems.
INTRODUCTION Los Alamos National Laboratory handles megacuries of tritium gas annually. The limitations of conventional tritium effluent monitoring systems prompted the development of improved designs. This paper discusses two of these systems. WIDE RANGE TRITIUM EFFLUENT MONITOR The wide range t r i t i u m monitor has been in routine use at the Laboratory's tritium processing f a c i l i t y since January 1980. At this f a c i l i t y there exists a potential for large unexpected gaseous releases of t r i t i u m , therefore suiting the application for which this instrument is designed.
Wide Range Detector A schematic of the ionization chamber is shown in Figure 1. A series of parallel grids form two active volumes, labeled low and high range chambers. The low range section has a geometric volume of %1.0 i; the high range section -».0.1 *. The high range section has reduced grid spacing as compared to that of the low range section to increase the electric field
237
WIDE RANGE TRITIUM DETECTOR
LOW RANGE CHAMBER ftWa CERAMIC
-DEIONIZER-t
-M ^-HffiH RANGE CHAMBER
FIGURE 1. Schematic of Wide Range Tritium Monitor strength and to reduce recombination effects at high concentrations. A deionizer is built into the intake end of the chamber to remove ions from the air before entering the active volume. A particulate filter at the intake also serves to remove dust and heavy ion pairs. Reduced Chamber Effect In tritium effluent monitoring systems, the problem of internal chamber contamination from tritiated oil or condensation of tritium water vapor is always of concern. The increase in background current significantly increases the lower limit of detectability and makes effluent release calculations difficult and erroneous at low concentrations. The wide range detector is designed to minimize the contamination effect through internal electrode spacing. At Los Alamos, because of the reduced barometric pressure (t75% of sea level), the maximum range of the 18 keV betas is about 10 mm in air. Based on this, the inner wall of the chamber was designed to be 10 mm from the active volume. The chamber walls are at the same potential as the collection grids, therefore tritium-contaminated walls will not contribute to the signal (Anderson 1981). Further reduction of the contamination effect is accomplished by making the electrodes out of fine grids thus effectively reducing their surface area. Los Alamos Model 600 Electrometer-Chargemeter The Model 600 Electrometer-Chargemeter measures current and integrates charge. There are two electrometers, one that accepts a current signal from the low range chamber and the second that accepts a signal from the high range chamber. Each electrometer feeds its own 4-decade logrithmic amplifier. The final 8-decade analog current output is achieved by mixing the outputs of the two logrithmic amplifiers. The low range electrometer spans 1 fA to 10 pA and through a discriminator circuit the high range electrometer takes over at 10 pA to 100 nA. A normalized chamber current
238
is displayed in analog form on the face of the Model 600. Figure 2 is a photograph of the chassis, low range electrometer amplifier and the wide range chamber. Charge measuring is achieved by taking the linear outputs from the electrometer amplifiers and feeding these signals into a voltage to frequency converter and counting the pulse train. This makes it possible for digital signal integration, which is a measure of charge. Operational Experience Since its installation, the wide range tritium effluent jnonitor has measured monthly releases ranging from 1.5x10 yCi to 5.3x10 yCi. The largest accidental release experienced to date has been, a few hundred curies, resulting in a maximum concentration of %350 mCi/m . The electrometer autoranged into the second decade of the high range electronics. For a backup, this system has a conventional Kanne chamber and six decade pi coammeter monitoring the effluent, thus providing an opportunity to compare the results and evaluate the response time of the two systems. In general, the wide range system responded faster, thus, tracking the total release more accurately. For the 24 hour period that included this release, the total integrated effluent was 12.6% higher than that of the conventional monitoring system. The advantages of using the wide range system when an acute release occurs are a) results are available almost immediately because of the integrating capability, b) there is more confidence in the data-jdue to fast response time, and c) the upper limit of response (100 Ci/m ) will assure adequate documentation of unexpected large releases.
FIGURE 2. Model 600 Chassis, Electrometer Oven, and Chamber
239
The sampling point to chamber distance was minimized to take advantage of the one second current/meter time constant for tracking of sudden high concentrations of tritium. It is advantageous to keep that distance as short as possible. The chamber was found to be microphonic and shock mounting of the low range electrometer and the chamber was required to eliminate mechanical noise. The tritium effluent monitoring system has an inherent current that is composed of a) current produced from gamma interactions with the ion chamber, b) current resulting from variations in the environmental radon levels, c) current produced from chamber contamination (if any), and d) an offset current from the electrometer, i.e, a slight positive bias current from the electrometer is necessary to keep the analog meter from being driven negative. This current is usually on the order of several femtoamperes (fA). Typical inherent background currents, over a six month period in 1982, ranged from 8-30 fA» which correspond to a tritium equivalent concentration 3of 9-32 yCi/m with a mean minimum detectable concentration of 12 pCi/m . Because of the background fluctuation, it was found advantageous to take daily fresh air backgrounds to establish the currents due to radon level fluctuations or any contamination effects. Calibration The wide range tritium monitor was first calibrated at the Los Alamos Gamma Calibration Range. Radiation fields of 1.9x10 to 1.2x10 R/h were used. The output current, automatically normalized to the 1 * chamber, was recorded. The results of the gamma calibration are shown in Figure 3. At the range change, marked on the figure, there was a slight deviation from a linear line. This is believed to be caused by a submillivolt offset in the zero of the high range electronics (Anderson 1981). The wide range monitor was then calibrated to tritium gas using the diagram showruin Figure 4. Since the range of detectability is from a few pCi/m to 10 yCi/m , it was necessary to use two tritium sources and a calibrated instrument for the tritiumxalibration/linearity check. For the low end of the range (up to 400 pCi/m ) , a Johnston calibrator was used as the source. For the higher concentrations, (2.5 mCi/m and higher), a high concentration (>500 yCi/m*) gas sample was introduced into the closed loop system via a hypodermic needle and the current reading plotted as a function of the calibrated Johnston 111 meter reading. The calibration/ linearity check of the wide range monitor is performed to 10 uCi/m and the complete response curve mentioned above is shown in Figure 3. IMPROVED KANNE TRITIUM EFFLUENT MONITOR The way in which Kanne chamber currents are measured has changed very little over the past two decades. The,current is fed to an electrometer and its logrithmic output, usually 10" to 10" A, drives a six decade strip chart recorder. The quantity of tritium passing through the Kanne chamber is calculated by hand integrating the area under the logrithmic
240
TRITIUM CONCENTRATION -710
10*
I03
I0 4
I05
IP6
I07
10
• TRITIUM • RADIATION
I0" 4
10"3
I0"2
10''
I
10
I02
103
RADIATION FIELD(RZh) FIGURE 3.
Wide Range Chamber Response Curve PUMP
HIGH RANGE CHAMBER
MOD-600 ELECTROMETER
FIGURE 4.
Typical Calibration Diagram 241
trace. This is very time consuming and can introduce inaccuracies from individual interpretations. The improved Kanne chamber and Los Alamos built Model 39 ElectrometerChargemeter have become the Laboratory's principal tritium effluent monitoring systems. This tritium monitoring system is described. Improved Kanne Chamber In the late seventies, Los Alamos redesigned the Kanne chamber to 1) reduce the problem of internal contamination from tritiated oil and tritiated water vapor, and 2) to make it a viable chamber for easy decontamination (Anderson 1980). The conventional Kanne chamber as described by Hoy (Hoy 1961) consists of three concentric cylinders, with the inner and outer cylinders at, or near ground potential while the intermediate cylinder is operated at about 200 V. The region between the outer and intermediate cylinder acts as an ion trap. The inner region is the ion chamber and the inner cylinder acts as the collecting electrode. The purpose for redesigning'the chamber was to reduce the effective surface area of these cylinders whose contamination would affect the signal from the chamber. Figure 5 shows a detailed drawing of the improved Kanne chamber with the high voltage cylinder of the conventional Kanne chamber replaced by a wire cylinder. The conventional 76 mm diameter cylinder electrode was replaced with a 6.4 mm diameter aluminum rod. Because of theoretical recombination effects at high concentrations, we are considering a larger (>6.4 mm) diameter collecting electrode with minimal surface ar&a.
DETAM. OP m m o V I O KANNC CHAMBER
KtuwmriM LHJCTMi Mffffi HW
ttlUKKCU
Fig. 5. Detailed drawing of the improved Kanne chamber
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Reduced Contamination Effect As in the wide range chamber, the redesign of the conventional Kanne chamber has taken into account the range of the tritium betas for reduction of the contamination effect. This was accomplished by keeping all surfaces outside of the high voltage cylinder (i.e., connecting rods and chamber wall) greater than 1 cm from the wires. Since the outer support rings are grounded, the ionization due to their contamination will terminate in the ring and will not contribute to the signal. The supported end of the collector electrode is shielded as shown in Figure 5 to prevent increased signal from contamination on the support rings or chamber ^end. The sensitive area of the improved Kanne chamber is less than 266 cm while the sensitive area of a conventional Kanne chamber is about 11,000 c m . This is a reduction in sensitive area by a factor of 40 (Anderson 1980). The improved Kanne chamber design enhances decontamination if needed. The high voltage wires can be decontaminated by passing a heating current through them. If necessary, the central collecting electrode can easily be removed for decontamination or replacement. The improved Kanne chamber does not use an internal deionizer and, therefore, one must be provided. An external HEPA filter is used for removal of dust and oil. Los Alamos Model 39 Electrometer-Charqemeter The Los Alamos Model 39 Electrometer-Chargemeter and temperature controlled oven containing the preamplifier are shown in Figure 6. A selector switch dictates any one of three choices of full scale current range. The range selected then spans four decades of electrometer output- The electrometer is designed to measure currents as low as 1 fA (10 A ) , and to integrate these currents. However, as normally used, the electrometer sensitivity is switched one decade lower because of the background currents experienced. Hence, for normal operation, the most sensitive range becomes 100 pA full scale. Features Common to Both Systems Both the Model 600 and Model 39 chargemeter ..systems have a readout that covers a 10 decade digital display. From 10 to 10 C/digit full scale. Readout, on the chassis, is with three decades of digital indicators and exponent multipliers. Charge readout selection is either manual or autoranging whereby the three most significant digits with nonzero information are displayed along with the correct exponent multiplier (Anderson 1980). Both electrometers have the capability to suppress unwanted signal currrents (e.g., from chamber contamination), which reduces the accumulation of charge over long periods of time.
243
• ' « » "
Fig. 6. Model 39 chassis and preamplifier oven The systems described allow for short distances between the ionization chambers and their associated electrometer amplifiers for optimal low current measurements. The main chassis for both systems may be remoted from their points of measurement. Operational Experience Improved Kanne chamber background currents at Los Alamos vary considerably with location and environmental factors. One system, over an eight month period in-1982 displayed a mean fresh air background current of 100 fA (l.OxlO"1 A) with a range of 40-180. fA. The mean tritium equivalent background concentration was 1.8 uCi/m . This variation is the result of a mild radon contribution verified by corresponding barometric pressure readings. On the other extreme, a Kanne chamber system in an underground storage j*auH has shown background,chamber currents ranging from 150 fA (1.5x10"" A) to 900 fA (9.0 x 1O" 10 A ) . This corresponds to a tritium equivalent background concentration ranging from 2.7 uCi/m to 16.4 pCi/m respectively, as the result of large and variable radon ^contributions. The mean fresh air background current was 210 fA (2.1x10 A}. The environmental factors influencing background current were found to be inversely proportional to the barometric pressure and were attributed to radon. Calibration The initial calibration of, the improved Kanne chamber system defined the response as 1.8x10 uCi/cm per ampere compared to 2x10 jiCi/cm per ampere for the conventional chamber (Hoy 1961 and Anderson 1980). The implied effective volume for the improved chamber is 56.6 liters.
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REFERENCES Anderson, D. F . , and R. D. Hiebert, 1981. A Wide Range Tritium Monitor. LA-UR-81-2987, Los Alamos National Laboratory, Los Alamos, New Mexico. Anderson, D. F . , and R. D. H i e b e r t , 1980. An Improved Kanne T r i t i u m Monitoring System. IAEA-SM-245/19, Management of Gaseous Wastes From Nuclear F a c i l i t i e s , International Atomic Energy Agency, Vienna, Austria. Hoy, J . E., Health Phys. 6:203 (1961).
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5C ON-LINE LIQUID EFFLUENT MONITORING OF SEWAGE AT LAWRENCE LIVERMORE NATIONAL LABORATORY
M. Dreicer, J. L. Cate, D. W. Rueppel, C. J. Huntzinger, and M. A. Gonzalez Lawrence Livermore National Laboratory Livermore, California
ABSTRACT On-line effluent monitoring is important to prevent releases of potentially toxic material to the environment. At Lawrence Livermore National Laboratory (LLNL) the sewage leaving the laboratory is treated by the Livermore Water Reclamation Plant (LWRP). The laboratory has the responsibility to protect the reclamation plant and local environment from the accidental discharge of toxic materials into the sewers. An automatic on-line, sewage effluent monitoring system has been developed at LLNL. A representative fraction of the total waste stream leaving the site is monitored for pH, radiation, and metals as it passes through a detection assembly. This assembly consists of an industrial pH probe, Nal radiation detectors, and an x-ray fluorescence metal detector. A microprocessor collects, reduces and analyzes the data to determine if the levels are acceptable by established environmental limits. Currently, if preset levels are exceeded, a sample of the suspect sewage is automatically collected for further analysis, and an alarm is sent to a station where personnel can be alerted to respond on a 24-hour basis. Since at least four hours pass before LLNL effluent reaches the treatment plant, sufficient time is available to alert emergency personnel, evaluate the situation, and if necessary arrange for diversion of the material to emergency holding basins at the treatment plant. We will present information on the current system, and report our progress in developing an on-line tritium monitor as an addition to the assembly. *Work performed under the auspices of the U.S. Department of Energy by the Lawrence Livermore National Laboratory under contract No. W-7405-ENG-48. 247
INTRODUCTION Since its beginning in 1952, the Lawrence Livermore National Laboratory (LLNL), has supported an environmental surveillance program to determine its impact, if any, on the local environment. The Laboratory, located in a suburban valley 65 km east of San Francisco, is involved in widely varied research and development programs for the U.S. Department of Energy. The main efforts are in nuclear and nonnuclear weapons, magnetic and laser fusion energy, biomedical research, and nonnuclear energy technologies, such as geothermal power and fossil fuel utilization. These programs and the support efforts involved in carrying them out generate waste products that could present a negative impact on the environment if not properly managed. The Laboratory's sanitary sewer system is a possible route for the escape of toxic materials. Liquid effluents are released to the city of Livermore sanitary sewer system at an outfall located at the northwest corner of the site. The effluent is treated at the Livermore Water Reclamation Plant (LWRP). The plant is a secondary treatment operation that returns most of the water to the San Francisco Bay via a transport pipeline. The remaining portion is used for irrigating vegetation along the roadways and the local golf course. Although the Laboratory has an extensive program to control and dispose of all potentially toxic materials at the source, the environmental monitoring program serves as a check on control v procedures. Liquid effluent samples are taken daily at LLNL a n d \ v LWRP. These are analyzed for radioactivity. Monthly composites of these samples are analyzed for heavy metals. In addition, quarterly samples are taken and analyzed for the parameters specified in the LWRP's permit (National Permit Discharge Elimination System). This program gives us documentation of the content of our effluent and provides evidence that the effluent concentrations are below established regulations. However, with this system harmful material could be released for a significant time before we became aware of the situtation and acted to correct it. The sewer monitoring system was designed and constructed to detect toxic material releases and to facilitate immediate response action when they occur. This is necessary in preventing damage to the city treatment plant and regions receiving discharged effluent. If toxic materials are detected at levels that exceed predetermined LLNL alarm limits, signals are sent to a central alarm station that is manned 24 hours a day and a sample of the suspect toxic effluent is automatically collected. Since at least four hours pass before LLNL effluent reaches the treatment plant, sufficient time is 248
available to alert emergency personnel, evaluate the situtation and, if necessary, arrange for diversion of the material to emergency holding basins at the treatment plant. The toxic waste can be treated in ^ e basins or removed without destroying or reducing the efficiency of the treatment plant. An automatic on-line sewage effluent monitoring system has been developed that diverts a representative fraction of the total waste stream leaving the site. This portion is monitored for pH, radiation, and heavy metals as it passes through a detection assembly. The assembly consists of an industrial pH probe, two sodium iodide radiation detectors, and an x-ray fluorescence metal detector. A microprocessor collects, reduces and analyzes the data to determine if the levels are acceptable by established environmental limits. MONITORING SYSTEM COMPONENTS The on-line monitoring system consists of several components. The flow route to these components starts at a point in g manhole where all Laboratory sewage discharge lines converge. As the sewage flows through a Parshall flow-measuring flume, approximately 40 liters per minute is pumped to an aboveground building where the detection instrumentation is located. Inside this building, the sample enters a tank housing the high- and low-energy radiation detectors, pH probe, and a sample line leading to the metal analyzer unit. After it is scanned, the sample is returned to the sewer via an outlet pipe. Radiation Detectors Radioisotopes being used at LLNL that could be released in quantities that would exceed concentration guide levels are 9 0 Sr, 235 Uf 238u, 237 N p > 238pu, 239 Pu , 241 Am , 244 Cm , and 3H. All of these, with the exception of 90Sr and 3H, emit heavy element x-rays and low energy gamma rays during decay. For 90Sr, low-energy bremsstrahlung photons give an indication of specific activity. Tritium detection will be discussed in another section. With this in mind, the radiation detection system was designed (Cate and Hoeger, 1972). The primary detector, a 3 x 127 mm sodium iodide crystal separated from the sample by a thin polycarbonate plastic sheet, allows recognition of the low-energy x-rays, gamma rays, and bremsstrahlung photons in the energy range of 10 to 100 keV. In addition, a 50 x 50 mm sodium iodide crystal detector is mounted immediately adjacent to the sample tank to provide detection of high-energy (100 - 1000 keV) events. This system gives us a minimum detectable activity in a 10-minute count time (95% confidence level) of 1.6 x 10- 6 uCi/ml
249
or 0.016 of the concentration guide for 23 9pu, and 3 x 10" 6 uCi/ml or 0.3 of the concentration guide for Sr^O (17-minute count at 95% confidence level) (DOE Order 5480.1). The electronics associated with the detectors includes single-channel analyzers and a Digital Equipment Corporation LSI-11 microprocessor which compares the data to preset alarm levels. The count rates are checked every minute. If the current rate is greater than the alarm levels, the count rate over the past 60 minutes is averaged. If this average exceeds the preset levels an alarm is sent. This is a way of monitoring for low-level continuous releases. pH Monitor There are many locations at LLNL where acid and basic solutions are routinely used. An accidental release of extreme pH levels would cause damage at the LWRP. Therefore, the pH is monitored continuously by a commercial industrial pH probe housed in the tank holding the radiation detectors. Output from the probe is recorded on a seven-day circular chart and sent to the DEC LSI-11 microprocessor to be evaluated. The average pH level is checked every minute. If the high or low level is exceeded the pH from the previous 30 minutes is averaged. An expotential function is used to take into account the fact that pH is a log function. The low and high limits are set at the pH values of 4 and 10. Metal Detection The deleterious effect of excess concentrations of heavy metals on sewage treatment plant operations has been demonstrated in a number of studies. The Laboratory has had incidents in the past when inadvertent releases of toxic metals have created operating problems at the sewage treatment plant. These events prompted a search for a continuous metal-detection system that would prevent the recurrence of plant down-time caused by a metal release to the sewer. Commercial units available appeared to be unsuitable for sewage analysis. A monitoring unit capable of detecting hazardous concentrations of ions found to be most harmful to the bacteria in the treatment plant process, specifically copper, nickel, chrome and zinc, was then designed (Gate, Matthews and Rueppel, 1979). The system meets the requirements placed on a continuous system: it is fairly reasonable in cost, especially when contrasted to the cost of reseeding a treatment plant; works on-line and in real time; does not require extensive pretreatment of the sample, thereby preserving representability; and requires minimal maintenance.
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The unit providing metals detection is an x-ray fluorescence analyzer (XRFA). Its design is based on the principle that elements emit characteristic x-ray lines when excited by a radiation source. These x-ray lines can be measured by energy dispersion techniques to determine their energy, which permits species identification, while the intensity of the lines is proportional to concentration. A portion of the sample stream that flows through the tank described previously is routed through a macerator to reduce any solids to particles 100 microns or less in diameter. The flow is then introduced through a nozzle into a flume inclined at 45 degrees. The flow spreads to a sheet 1 mm in depth before it reaches the source/detector region of the unit. A thin plastic window backed by air is positioned under the stream in this region to prevent detection of the chrome and nickel in the stainless steel flume. A lO^Cd source is used to excite the elements of interest if they are present in the stream, and a commercial xenon-C02 mixture x-ray proportional counter is the detector used. The output from the detector goes to amplifiers, then through an analog-to-digital converter interfaced to a DEC LSI-11 microprocessor. A dual register system for the data output allows both a rapid response to a high metal concentration (500-3600 sec count) and a more sensitive response to low concentrations released over a longer time frame (5000-36,000 sec accumulated count). The longer count length is used as the activity of the !09cd source decays. The data counting system also provides for advisories at concentrations that are elevated, but not at alarm levels. Maximum permissible discharge concentrations for copper, chrome, nickel, and zinc have been established at LLNL based on studies of the effects of toxic metals on sewage treatment and the dilution factors resulting from the intermixing of LLNL's effluent with the domestic sewage from the Livermore city population. These concentrations, as well as the short and long count time alarm limits for the elements of concern, are shown in Table 1. Tritium Monitor The quantities of tritium used at LLNL could result in a liquid effluent release that would be greater than the established concentration guides. A release of tritium could contaminate the LWRP and be released into the local environment when the treated water is used for irrigation. To continue to improve our capabilities we are developing an on-line tritium waste watermonitor to detect the low-energy beta radiation. The specifications for the system include: a detection sensitivity of 1-20 pCi HTO/ml; the capability of providing a representative and particulate-free
251
TABLE 1. LLNL Maximum Permissible Metal Discharge Limits and XRFA Alarm Limits
Single Metal
LLNL Limit (ppm)
Cr Cu
100 10 50
In Hi Hg
As
Pb Total of above
XRFA alarm limit (ppm)(a) Short Count Long Count 107 17 50 23 16 10 12
10 ----
50 10 50 10 10 10 4
100
(a) Count times vary from 500 - 3600 second (short) and 5,000 36,000 second (long), depending on the source strength.
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liquid sample; economical to operate and maintain on a continuous basis; and the detector can be connected to the existing DEC LSI-11 microprocessor system. The system can be divided into two subunits: sample collection and radiation detecton. To collect small portions of the waste stream that flows down through the x-ray fluorescence analyzer, a 49-mm-diameter prototype cross-flow filter was installed in the chute. Cross flow filters differ from conventional "dead-end" filters in that the sample stream continuously washes the filter face, greatly extending its useful life. Various pore size filters are being tested for optimal performance in filtering the effluent and not clogging too quickly. A one micron pore-size teflon filter disk, wetted with ethanol, has lasted the longest, approximately 10 days, at a filtrate flow rate of 0.23 ml/minute. Liquid scintillation counting with a continuous-flow cell will be used for the low-energy beta detection. The sample stream will be mixed with a precise amount of liquid scintillation cocktail before being pumped through a flow cell detector. The detector element is a thin-walled, teflon tube coiled flat and sandwiched between two matched photomultiplier tubes. The usable cell volume is 2 ml. The system will consume approximately one gallon of scintillator a week. At this point in the development the maximum counting efficiency for tritium is 22.6%. This corresponds to a minimum detectable activity (MDA) of approximately 50 pCi/ml. Further measurements such as a graded metal shield and improved light transmission from the flow cell to the photomultiplier tubes should lower the MDA further. To cut down on biofouling in the system, an ultra dynamics Model 500 ultraviolet flow sterilizer has been installed in the effluent line before it reaches the XRFA assembly. At this time the cross-flow filter and the liquid scintillation flow cell have been tested separately. The next step will be testing the operation of the combined system. Grab Sampler The identification and concentration of the toxic material is needed to evaluate an alarm situation. An automatic grab sampler extracts a sample from the waste stream as soon as the alarm is tripped. This mechanism can also be activated manually. System Monitor Since the operation of the whole monitoring system depends on tht. pumping of a continuous sample flow through the instrument tank, we installed a sensitive differential pressure switch on the tank.
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Changes in pressure drop across the system indicate changes in flow rate and will trigger an alarm. Other problems such as count rate lower than background and power failure will also send an alarm signal to the central alarm station. CONCLUSIONS The on-line sewer monitoring system cannot in itself prevent the accidental discharge of toxic materials into sewers. However, by rapidly detecting releases and sounding alarms to alert emergency personnel to respond, actions can be taken to prevent damage to the environment. Because it is a real-time system. Laboratory personnel can more easily locate the source of a toxic discharge and take corrective action to prevent its recurrence.
DISCI.AIMKK This documeni was prepared as an account of work sponsored h> an agency of the I nited Stales (ioiernmi.nl. Neither the ( nited States (ioternmenl nor the I niversilj ,pf ( 'alifornia nor an> of their employees, makes any warranty. express or implied, or assumes any legal liability or responsibility for ihe accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately o»ned rights. Reference herein to any specific commercial products, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by Ihe I nilvd Stales Government or the I niversily of ( alifornia. The views and opinions of authors expressed herein do not necessarily slate or reflect those of the I niled Stales Government thereof, and shall not be used for advertising or product endorsement purposes.
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REFERENCES Gate, dr., J. L., M. A. Matthews and D. R. Rueppel "A Prototype On-line X-Ray Fluoresence Analyzer for Detection of Metals in Sewage", Proceedings of the 34th Industrial Waste Conf.t Purdue Univ. 1979. Ann Arbor Science, Cate, Jr., J. L., and T. 0. Hoeger. 1972. "A Radioisotope Monitoring System for Sewage Effluent." Amer. Indust. Hygiene October 1972: 693-699. DOE Order 5480.1 Standards for Radiation Protection.
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5D GENERIC PARTIOJLATE MONITORING SYSTEM FOR RETROFIT TO HANFORD EXHAUST STACKS
Jerry W. Cammann RockwelI Hanford Operations Richland, WA 99352 Eugene H. Carbaugh Pacific Northwest Laboratory Rlchland, WA 99352 ABSTRACT Most Hanford exhaust stack sampling and monitoring systems were designed and built prior to the availability of currently recognized design standards and guidelines. As environmental concerns became increasingly important, Improved airborne effluent sampling and monitoring for radioactive partlculates was recognized as being essential In determining the adequacy of engineered effluent control systems and providing a record of environmental releases. Systems once adequate, contain areas where modifications can be made to Improve sampling system efficiency and reliability. Evaluations of 72 sampling and monitoring systems were performed as the Initial phase of a program to upgrade such systems. Each evaluation included determination of theoretical sampling efficiencies for particle sizes ranging from 0.5 to 10 micrometers aerodynamic equivalent diameter, addressing anlsoktnetic bias, sample transport line losses, and collector device efficiency. Upgrades needed to meet current Department of Energy guidance for effluent sampling and monitoring were then identified and a cost for each upgrade was estimated. A relative priority for each system's upgrade was then established based on evaluation results, current operational status, and future plans for the facility being exhausted. Common system upgrade requirements lead to the development of a generic design for common components of an exhaust stack sampling and monitoring system for airborne radioactive partlculates. The generic design consists of commercially available off-the-shelf components to the extent practical and will simplify future stack sampling and monitoring system design, fabrication, and installation efforts. 257
This paper will emphasize evaluation results and their significance to system upgrades. A brief discussion of the analytical models used and experience to date with the upgrade program wlil be Included. Development of the generic stack sampling and monltorfng system design Is outlined. Generic system design features and limitations are presented. Requirements for generic system retrofitting to existing exhaust stacks are defined and benefits derived from generic system application are discussed. 1.0
INTRODUCTION
Rockwell Hanford Operations (Rockwell) has a commitment to maintain complete and accurate records of radioactive discharges to the environment and to operate facilities In compliance with Department of Energy (DOE) established criteria. Accurate and representative sampling of effluents Is required for comparison with established release limits to determine compliance with "As Low As Reasonably Achievable" (ALARA) radioactive discharge philosophy and provide confident reportIng of releases. An integral part of obtaining accurate release documentation fs defining the technical requirements and capabilities of sampling and monitoring systems, accompanied by routine evaluations to verify that system performance meets the Intent of the requirements. Most Rockwell exhaust stack sampling and monitoring systems were designed and built primarily for process monitoring and control prior to the availability of currently recognized design standards and guidelines. As environmental concerns became increasingly important, Improved effluent sampling and monitoring were recognized as essential to determine the adequacy of engineered effluent control systems and provide a reliable basis for reporting environmental discharges. An engineering study was performed to evaluate Rockwell's exhaust stack effluent sampling and monitoring systems with respect to standards and guidelines established by the DOE and Rockwell. The DOE directives i,2,3present general guidelines which outline the type of effluent sampling and monitoring to be performed but do not provide design specifics. American National Standards Institute (ANSI) guidelines, 10.11 which are referenced and endorsed In DOE directives, present specifics on sampling and monitoring system designs. Rockwell's Environmental Protection Manual kcomb I nes Rockwell's Interpretation of DOE directives and ANSI guidelines Into detailed requirements for sampling and monitoring system designs and Is the measure for compliance at Rockwell. The Engineering Study defined system capabilities. Identified deficiencies, and addressed corrective actions. The measurement of radioactive materials collected Is not addressed in this paper. Exclusion of analytical laboratory measurement capabilities must not be construed to imply that measurement of sam258
pies Is of lesser Importance than sampling. Accurate measurement and analysis are vitally Important, but can be valid only to the extent that samples collected depict existing conditions accurately. 2.0
PARTICLE PENETRATION ANALYSIS
The most Important goal of a sampling program Is to accurately detect and Indicate radioactivity In effluent streams so that Individuals may be protected from excessive exposure to airborne radioactive materials. Therefore, the need for representative sampling systems cannot be over emphasized. Consistent with this goal are other desirable objectives which can be realized in a well developed sampling program. °
Detection of deteriorating equipment, faulty processes, or other conditions resulting In loss of effective control of airborne materials in an operation, and subsequently to determine the effectiveness of corrective actions.
°
Determination of quantitative releases of radioactive materials to the environment. Sampling of airborne effluents will assist in evaluating and controlling radioactive discharges to provide assurance that people In the surrounding environment are not exposed to levels of airborne radioactive materials exceeding established limits.
A truly representative and accurate effluent sample will not be extracted from the exhaust stack unless sample air Is withdrawn parallel to the stream flow from a well mixed portion of the alrstream, the air velocity through the sample probe nozzles Is Identical to the airstream velocity at the sample extraction point, and probe nozzle edge effects do not disturb the aerodynamic streamlines of the approaching effluent air detrimentally. Sampling under these conditions Is commonly referred to as "Isoklnetlc Sampling. 11 In actual practice, however, these requirements are seldom. If ever, absolutely achieved necessitating some degree of compromise. Sample concentration biases can evolve from anlsokinetic sampling conditions, resulting from the collection of nonrepresentatlve effluent samples with distorted particle size distributions. When sampled air Is extracted at a "SuperIsoklnetlc 11 rate, small particles which would have otherwise not been sampled will be preferentially collected from the surrounding alrstream. Large particles will be subject to Inertial effects and are siot able to follow streamlines into the nozzle. The net effect can be determination of release concentrations which are lower than actual discharges. Samples collected at a "SubIsoklnetlc" r=vte will result In preferential collection of large particles. Siral I particles have a tendency to follow predominant air flow patterns around the nozzle and are diverted. The net 259
effect can be determination of release concentrations which are higher than actual discharges. Subsequent to entering the sample probe, the sample Is transported to the collection medium. Ideally, the collector should be positioned at or very near the point of sample extraction to minimize particle deposition losses In tran.^ort tubing. However, excessive background radiation levels Inside the duct and other operational difficulties preclude direct In-stack sample collection In many cases. Consequently, a compromise must be reached by Incorporating sample transport lines In system designs. Care must be exercised In the design and operation of systems utilizing sample transport lines so that excessive deposition losses can be avoided. The degree of particle loss In the sampling system Is a function of particle size, particle density, particle distribution, sample extraction velocities, effluent conditions, and sampling system configuration. Except In unusual situations, particles smaller than an aerodynamic equivalent diameter (AED) of about 155 micrometers are able to follow the streamlines of the air closely. Work performed by L. C. Schwendlman and J. A. Gllssmeyer of Pacific Northwest Laboratory - Battelle led to the Issuance of BNWL - 1921, "Appraisal of Airborne Radioactive Effluent Sampling Systems - Comprehensive Analysis Format and I n - P l a c e S a m p l e r Validation." 8,9,12 T n [ s document presents an accumulation of theoretically and empirically derived equations describing anisokinetlc sample extraction bias and gravity settling and turbulent deposition of partlculates In gaseous effluent sample probes and transport tubing. The techniques and models detailed In BNWL - 1921 were applied by Rockwell, within the guidelines provided by J. A. Gllssmeyer, to evaluate existing Rockwell stack sampling and monitoring systems. The mathematical relationships used to estimate particle deposition are valid for any single particle size. Normally aerosols to be sampled will be polydlsperse containing a wfde spectrum of particle sizes. Total system deposition, therefore, is obtained by calculating loss factors for each particle size In the spectrum and summing to get the total. The Rockwell engineering study addressed six different particle sizes independently, which are of concern from a health physics standpoint, and estimated particle deposition associated with each. A computer program utilizing Basic computer language was developed to Implement the repetitive calculations for particle deposition within the sampling system. This program incorporated deposition models for estimating particle retention In straight and bent tubes. The program continued by summing the effects of each sampling system component Into total system deposition and presented the results In
260
tabular form opposite respective particle sizes. Calculations determining the anisokinetlc sample extraction bias were not Included In the program and were performed on a programmable hand calculator. 3.0
STACK SAMPLINB AND MONITORING SYSTEM ASSESSMENTS
A total of 12 active exhaust stack sampling and monitoring systems were evaluated at Rockwell to determine their conformance to prescribed system design features delineated In ANSI guidelines, DOE directives, and Rockwel I standards. This assessment Indicated that upgrades could be Implemented to Improve existing sampling and monitoring systems performance. 3.1
Assessment Methodology
The evaluation of Rockwell exhaust stack sampling and monitoring systems consisted of both quantitative and qualitatlve assessments of each system. Two assessment sheets were developed to document system features and results of the particle deposition computer model. The first sheet evaluated the overall final effluent sampling and monitoring capability, subjectively evaluated the adequacy of each system, made recommendations for system upgrades, provided an order of magnitude estimate for cost, and established a priority for system upgrade. The second sheet contained objective evaluations of the characteristics of each sample associated with an exhaust stack discharge and presented results of the particle penetration analysis through the sample train. A quantitative rating system was employed to establish general sampling system upgrade priorities. Numerical values were assigned to the following sampling system particulars with the sum totals defining the relative priority for system upgrade. STACK FLOWRATE - (1-5 Rating) Ventilation and air balance records of exhaust stack flowrates were Investigated to determine current air flowrates for systems being evaluated. The following flowrate ranges and associated rating values were assigned to evenly distribute exhaust stacks Into five classifications: (1) (2) (3) (4) (5)
0 501 2 ,001 10 ,001 20 ,001
-
261
500 2,000 10,000 20,000 UP
ft?/mfn ftf/mln ft^/mln f-r/min
HIGH RELEASE SAFETY HAZARD - (1-3 Rating) The potential for the exhaust stack to release levels of airborne radioactivity to the environment In excess of criteria established In DOE Directive 5480.1 was investigated. Operations being ventilated, extent of filtration being employed prior to discharge, and historical release levels were taken Into consideration. DEGREE OF NONCOMPLIANCE - (1-5 Rating) The degree of variance of stack sampling and monitoring system designs from criteria established In DOE directives, Rockwell environmental protection standards, and ANSI guidelines was investigated. Items of particular Interest included sample extraction system design, sample transport line configuration, and sample collection methods. PROJECTED LONG-TERM STACK OPERATIONS - (1-5 Rating) The scheduled future of exhaust stack operations and/or facility operations was Investigated. Based on available funding and resources, some technically feasible upgrades were not considered economically practical due to limited remaining service life or use of a particular exhauster. PROJECTED INCREASED FACILITY OPERATIONS - (1-4 Rating) This category evaluated near term operations planned for Rockwell facilities. Consideration was given to facility startups, Increased operations of support facilities, and increased activity within facilities scheduled for decontamination and decommissioning. TOTAL PRIORITY INDEX - (5-22 Rating) The summation of the five quantitative rating categories formed the basis for the overall upgrade priority assigned each exhaust stack sampling and monitoring system as follows: Low Priority
-
5-10 Rating
Medium Priority
-
11-15 Rating
High Priority
-
16-22 Rating
262
3.2
Assessment Results
Stack sampling and monitoring systems currently operating at Rockwell represent several generations of philosophy and technology. .This fs attributed to past efforts to keep current with preferred sampllng methodologies of the time. Increasing environmental concerns, however, have resulted In Issuance of more definitive system design guidelines. Consequently, systems once adequate are now found not to conform totally to present-day practices. In general, sample extraction probes were not designed for Isokfnetlc sampfe extraction, sample transport lines contained excessive small radius 90° bends and obstructions upstream of the sample collection point, airborne radiation monitors In use were not reliable, and no provisions were made for Independent record sampling. Sample extraction probes In use at Hanford exemplify the diversity of sampling concepts used since the Hanford site was commissioned for operation. Probe designs vary from pipes welded to the side of exhaust stacks, to pipes with holes drilled in them, to ANSI N13.1 designed multiple nozzle fsoklnetfc probes. Sample transport lines were Installed with unnecessary runs and bends In almost all cases. This is attributed to the lack of specific Information on Installation drawings detailing sample transport line Installation. In the past, transport lines were "field run" as necessary leaving transport line Installation to the discretion of Installation crews. Many of the Rockwell stack sampling and monitoring systems use a Hanford designed airborne radiation monitor device. This device has remote high airborne radiation alarm c a p a b i l i t y o n l y . State-of-the-art airborne radiation monitoring systems, however, have Improved reliability and provide local and remote alarm capabilities signaling high airborne radiation, monitor high voltage failure, and monitor detector failure. In many cases, airborne radiation monitor samples were being used as record samples. This Is not recommended because design considerations which allow detector positioning near the collection medium may obstruct the transport of sampled partlculates to the filter paper. Investigative studies performed by r e s e a r c h e r s at P a c i f i c Northwest Laboratories - Battelle Indicate that particle deposition losses In airborne radiation monitoring devices used at Rockwel I can be significant. A preliminary reportllfwas issued showing particle deposition results for the Hanford designed beta-gamma monitor and the Radeco beta-gamma monitor for 0.8, 2.5, and 6.0 micrometer particle diameters. The following Information was presented:
263
PERCENT OF INLET SAMPLE PARTICLES DEPOSITED ON COMPONENTS OF THE HANFORD DESIGNED COLLECTOR DETECTOR UNIT Particle Diameter Micrometer
Percent Deposited Stainless Steel Inlet Distributor Ring And Cavity Walls Hose Connector 0.47
Total 0.94
0.8
0.47
2.5
6.4
27.0
33.4
6.0
4.7
54.7
59.4
PERCENT OF INLET SAMPLE PARTICLES DEPOSITED ON COMPONENTS OF THE COLLECTOR DETECTOR UNIT, 2 F^/MIN RADECO UNIT Particle Diameter Micrometer
Percent Deposited Inlet Tubes
Cavity Walls, Etc.
Total 0.41
0.8
0.34
0.07
2.5
20.9
2.9
23.8
6.0
33.5
6.1
39.6
In order to obtain collection efficiency values for additional particle sizes, the existing data was curve fit to exponential, logarithmic, and power curves using a programmable hand calculator. Given the three experimentally obtained data points. Tables 1 and 2 were generated using curve fitting techniques to provide col lection efficiencies corresponding to different size particles for the Radeco and Hanford designed radiation monitors. Values determined for the Radeco unit were used as being characteristic of commercially manufactured airborne radiation monitor units of similar design. In both cases, comparing the coefficient of determination and the experimental values to the Inferred values, It appears that the experimental data is most nearly represented by the logarithmic curve.
264
Table 1.
Hanford Design Beta-Gamma Monitor Drawing No. H-2-33904
Particle Size Micrometer
0.5 0.8 1.0 2.0 2.5
5.0 6.0 8.0 10.0
99.52 94.66 91.55 77.46 71.26 46.93 39.71 28.43 20.36 Y = aebx a = 108.2 b = -.17 r 2 = .98
Table 2.
Power Curve***
Exponential* Logarithmic**
125.53 102.17 92.66 68.39 62.02 45.78 42.27 37.26 33.79
112.86 99.23 92.76 72.67 66.20 46.11 40.82 32.49 26.02
0.5 0.8 1.0 2.0 2.5 5.0 6.0 8.0 10.0
inferred Col lection Efficiency Values 100.
66.6
99. 92. 70. 66.
40.6
46. 40.
99.06
34. 28.
*** Y = ax c
** Y = a + blnx a = 92.76 b = 28.99
a = 92.66 b = -.44
r 2 = 1.00
r 2 = .98
Radeco unit Beta-Gamma Monitor Model GH-221
Particle Size Micrometer
Experimental Data Points
Exponential* 97.87 95.21 93.48 85.29 81.46 64.77 59.10 49.19 40.95
Logarithmic** 108.40 99.24 94.88 81.37 77.01 63.50 59.94 54.33 49,98
Power Curve*** 112.36 100.02 94.65 79.73 75.44 63.55 60.74 56.57 53.53
* Y = ae b x a = 102.46 b = -.09
** Y = a + blnx a = 94.88 b = -19.50
r 2 = .94
r 2 = 1.00
265
Experimental Data Points
Inferred Collection Efficiency Values 100.
99.59
76.2 60.4
*** Y = ax D a = 94.65 b a -.?5 r 2 = 1.00
99. 94. 80. 76. 63. 60. 55. 50.
Application of the theoretical model Indicates that sampling system efficiencies can be greatly Improved by eliminating unnecessary sample line runs and bends. This Is especially true for participate matter greater than about 5 micrometers AED In slze^ Particles smaller than 5 micrometers AED behave essentially as a gas and are able to follow predominant air flow patterns closely. Larger particles are subject to Inertlal effects and are more likely to be deposited. Stack exhaust systems which are normally or potentially contaminated with radioactive effluents are required to have High Efficiency Partlculate Air (HEPA) filters as a final means for controlling radioactive releases to the environment, Large particles, therefore, would not be expected In effluent alrstreams downstream of HEPA filters under normal operating conditions. Limited Rockwell experimental data, however, suggests that large particles can become entrained In exhaust stack effluents due to scaling of contaminated ductwork downstream of the HEPA filters. Furthermore, there Is a potential for filter leakage and resuspenslon of agglomerated submlcrometer particles which could result in larger partleu I ate matter in exhaust stack effluents. Sampling and monitoring system upgrade requirements vary from stack to stack but generally consist of replacing sample probes, sample transport lines, airborne radiation monitors, and providing Independent record sampling capability. Although design of Isoklnetlc sample extraction probes Is dependent upon existing flow conditions In a particular exhaust stack, the Instrumentation portion of the system iends Itself readily to generic design. Most Rockwell stack sampling and monitoring system upgrades require Installation of fixed filter head record samplers and either alpha and/or beta airborne radiation monitoring devices with appropriate support equipment. This need provided the motivation for development of a generic design for common components of a stack sampler and airborne radiation monitor. 4.0
GENERIC SYSTEM DEVELOPMENT
The development of the generic design for a stack sampling and monitoring system was accomplished fn three phases. Phase one Identified the functional design criteria specifying desired system capabilities. Phase two consisted of producing an approved detailed generic design suitable for specifying required components of the stack sampling and monitoring system. Individual system components were tested during this phase to verify their suitability fn the final design. Phase two resulted In the completion and approval of 40 drawings which detail the Rockwell generic stack sampling and monitoring system. Phase three Involved fabrication of a prototype assembly to Identify and eliminate problem areas which couid Inhibit generic system fabrication efforts.
266
4.1
Design Features
The term "Generic Stack Sampling System" is actually a misnomer and does not mean that one universal system is suitable for all installations, Exhaust stack sampling and monitoring system requirements differ from facility to facility depending on which radlonuclides are of concern. Upgrade requirements vary by the degree to which existing stack sampling and monitoring systems conform to current standards. The standard design, therefore, was made capable of retrofit to existing systems having adequate sample probes and transport IInes. The generic design consists of 13 baste assembly drawings which when brought together comprise six sampling and monitoring system options. Each generic design consists of a series of modular components housed In either a weather tight enclosure or an open rack assembly. Individual components are commercially available off-the-shelf Items to the extent practical. The six sampling and monitoring systems available include the following: 0
Alpha/Beta/Record Cabinet Assembly
°
Alpha/Beta/Record Open Rack Assembly
°
Beta/Record Cabinet Assembly
°
Beta/Record Open Rack Assembly
°
Alpha/Record Cabinet Assembly
°
Alpha/Record Open Rack Assembly
Open rack assemblies are designed for Indoor use while the cabinet assemblies are designed for outdoor use Incorporating heating, cooling, and ventilation capabilities to provide assurance that continuous air monitoring devices are kept within prescribed operating temperature limits. The Instrument cabinet is a double-door, single-access metal cabinet measuring 6 feet high, 4 feet wide, and 2 feet deep. When fully equipped, the cabinet assembly will weigh approximately 1000 pounds. Lifting hooks ore provided to facilitate moving and positioning of the cabinet assembly. The sample vacuum system is shock mounted and resides on the floor of the instrument enclosure. Each sample collected (alpha, beta, record) has an independent vacuum source. Each vacuum source
267
utilizes a "Tee" Inlet connection with two quick disconnect connectors. This provides Interim backup vacuum capability In the event of a single pump failure. Provisions have also been made for the use of house vacuum supply systems If preferred. Sample flowrate monitoring and control Is provided with a rotameter, gas flow totalizer, flowrate regulator, and a flow switch. The gas flow totalizer Is temperature compensated and system pressure Indicators are provided for additional f lowrate correction to standard conditions. The sample flowrate regulator !s provided to maintain a constant sample flowrate to accommodate filter loading effects. The pressure regulator holds a constant pressure drop across an In-line orifice by varying a bypass valve Into the vacuum pump. This system allows the pump to work at minimum head drop extending the pumps useful life. The orifice Is adjustable, allowing flow rate adjustment from near zero up to the pump maximum capacity. Flow alarms are provided on each vacuum pump leg. Normal flow results In the movement of a magnet which, when flow Is lost, falls Into a region containing a magnetically activated flow switch. The magnet causes the switch to open, breaking the current supplied to the relay, thus causing a contact closure which activates the alarm. A 47-mm fixed-filter head record sampling device Is used to collect samples for discharge Inventory measurement and airborne radiation monitoring units are provided for process monitoring and control. The generic system utilizes the commercially available EBERLINE ALPHA 5 and the EBERLINE AMS-3 for detection of above standard airborne levels of alpha and beta radiation, respectively. A slide out shelf Is provided for the beta monitor to facilitate maintenance work and Installation of this device which weighs approximately 160 pounds. The record sample extraction loop Is designed to operate at 2.2 ft 3 /m!n. This number Is based on a laboratory requirement of 370 (ft 3 /mln)«(hrs) to obtain 10* of Table If detection capability. Continuous air monitor sample extraction loops are designed to operate at 2.0 ftVmln per manufacturers recommendation. A sample single sample
machined sample flow splitter Is provided to evenly divide the stream between the record sample and the monitor sample If a extraction probe Is used. Independent sample probes for each are preferred, however.
A centralized alarm annunciator panel with latching alarms is located within the cabinet or rack assembly and provides Indications for high airborne radiation (alpha and/or beta), radiation monitor unit failure (alpha and/or beta), loss of sample flow (alpha, beta, record), and high or low cabinet temperature. The latching alarm feature enables detection of Intermittent system problems. Once activated they must be manually reset to clear the condition. Audio and visual Indicators (bell and beacon) are provided to alert personnel of
268
Record Sample Rotameter
System Failure Beacon Flow Splitter Record Sampler \ \
Local Alarm Panel
oooooooo
O OO
High Radiation — I Bell
Record Sample Flow Totalizer
Metal Cabinet or Open Rack
3
EBERLINE ALPHA-5 MONITOR
©-
EBERLINE AMS-3 BETA MONITOR
O O O —
rn rn rn
Slide-out Shelf
Equipment Connection Box — & Elapsed Timer
-Pressure Gage •Flow Switch — Flow Regulator
Filtered Inlet Air Fan (Cabinet Only)
Power Panel — Connection Box I1
'I
I
1
-
[\
—Cabinet Heater i
Quick Disconnect Vacuum Hose Connectors
i
Vacuum Pump
GENERIC STACK SAMPLING AND MONITORING SYSTEM TYPICAL EQUIPMENT LAYOUT 269
high airborne radiation or system failure at the local Contacts are provided for remote alarm signal transmission.
level.
Two adjustable thermal switches are Installed within the cabinet and will activate an alarm If the operating temperature limits for the airborne radiation monitors Is exceeded. The lower limit Is set at 50°F to Inhibit condensation of moisture and associated problems, and the upper limit Is set at 125°F, near the upper limit for reliable monitor unit operations. The generic system requires 220-volt, single-phase, 2-pole, 3-wIre, 30-amp power for operation. The power for this system Is arranged as a split bus (240/120 VAC) with one side of the system feedIng the Instrumentation and utilities, and the other feeding the vacuum pumps. This allows the record sample pump to be activated by an outside contactor relay which Is coordinated with stack blower operation. Thus, when the exhaust stack Is activated, the record sample pump draws a sample of the effluent alrstream. Conversely, when the exhaust stack blower fan stops the record sample pump Is deactivated to eliminate dilution of the sample collected. An elapse timer Is paralleled with the record sample vacuum pump to Indicate elapse time of operation. The airborne radiation monitors and associated vacuum pumps are designed to operate continuously. The system has power supplied for heat trace, cabinet heater, and cabinet cooling fan operations which are all activated by thermostatlc controllers. Power for Instrumentation and the alarm system Is Intended to be supplied continuously by connection to normal and emergency power buses. 4.2
Installation Requirements
5
The generic design does not eliminate the need for site specific design work. The following design considerations must be addressed for the design review prior to system Installation: 0
Identification of sample extraction probe requirements and sample transport line Interfaces with the generic design.
°
Generic system alarm Interfaces with existing facilities for focal and remote operations.
0
Tie-Ins to existing facility power supplies and vacuum suppi !es If applI cable.
°
Issuance and performance of the acceptance test procedures to verify system operability.
270
The acceptance test procedures (ATP) have been standard fzed for each generic system option and cover all possible Installation configurations. Applicable sections of the generic ATP must be Identified based on site specific Installation requirements. Sample transport line runs are not Included In the generic design due to their site specific nature. Installation drawings, therefore, must show details required for Installation Including type of tubing, welding requirements, fittings, exact path of transport line run Including bend radii, tubing supports, heat tracing, and polnt(s) of entry Into the generic system cabinet. The generic systems provide sampling and monitoring Instrumentation only and are capable of retrofit on existing sample extraction apparatus. Existing sample probe and transport line design adequacy must be determined. Sample probes not conforming to established standards will require redesign and replacement. Audio and visual remote and local alarm annunciation capability Is provided In each generic system option. Details for remote alarm wiring to continuously or frequently occupied areas, however, Is site specific and must be addressed by the end user. The availability of electrical power and Its Interfacing with the generic system will also differ from site to site and must be Identified. Installation power wiring diagrams must show the power system Interface to the generic system and show contactors and wiring back to existing drawings that depict the power source used for the given lnstaliatI on. 4.3
Special ApplI cat Ions
7
The generic systems provide constant flow partlculate sampling and monitoring capabilities for exhaust stacks with adequate sample probe designs. The systems do not apply directly to the following situations without additional modifications and/or administrative controls: 0
highly corrosive exhaust stack effluents
0
high humidity exhaust alrstreams
c
variable flow exhaust alrstreams
°
alrstreams routinely containing large partlculate matter (>5 micrometers)
271
4.3.1
High Corrosion
If the exhaust alrstream contains reactive vapors such as radfolodlne, care must be taken to prevent absorption or reaction with other materials. Rubber vacuum hose Is used downstream of the record sample filter holder to facilitate filter paper changeout and downstream of all sample collection points to simplify generic system fabrication and minimize vibration effects associated with the vacuum pumps. Rubber vacuum hose Is susceptible to radiolodlne attack. The rubber degradation, however, will not affect the sample accuracy In any direct way. The rubber vacuum hoses will have to be Inspected periodically for excessive deterioration and replaced when necessary to prevent sampling system failure caused by hose breakage. 4.3.2
High Humidity
When the air to be sampled Is nearly saturated with water vapor, the temperature In the sample line must be maintained above the dewpotnt or water vapor will condense out in the sample line providing an effective barrier for particle transport to the collection medium. Additionally, excessive moisture may degrade the filter media either by blocking the pores or weakening It to the extent that it tears easily. When heavy moisture loading is anticipated, sample lines must be heat traced sufficiently to Inhibit moisture condensation. !f heat tracing Is not effective, alternate means of moisture removal will be required. Caution must be exercised that moisture removal does not also remove sample partlculates, or that designs account for such sample loss. 4.3.3
Variable Flow
The generic stack sampling system is designed to sample at a fixed rate and, therefore, Is not suitable for stacks with flow variations so great that the anlsokinetic sampling error would be significant In the overall sampling accuracy. A determination of this sort requires knowledge of particle size and distribution characteristics for a particular effluent alrstream. Limited Information available concerning Rockwell exhaust stack effluents suggest that the generic system Is adequate for stack flow variations not exceeding ±20?. Calculations show that anlsokinetic biases Introduced for such flow variations result In an Insignificant error for particle sizes expected in Rockwell exhaust stacks. Development work is currently underway to design Into the generic system the capability for automatic Isokinetic flow proportional record sampling for variable flow exhaust afrstreams. Existing hot-wire anemometer technology is being Investigated for this application.
272
4.3.4
Large PartFries
The Rockwell generfc stack sampling system Is not designed to effectively sample large part I cu I ate matter (>5 micrometers). Thfs is characteristic of any sampling system in which delivery lines are used to carry the sample to the collection medium. Large particles will be preferentially removed either through gravitational settling, when the flow Is too low, or through turbulent Impactlon, when the flow Is too high. Typically, particles smaller than an AED of 5 micrometers can follow the streamlines of. the sample air and particle loss Is not substantial . Stack exhaust systems which are normally or potentially contaminated with radioactive effluents are required to have HEPA filters as a ffna! means for controlling radioactive releases to the atmosphere. In the case of filter failure the overall measurement accuracy would be good only If the bulk of the particulate matter was less than 5 micrometers. As particle size Increases the accuracy would decrease as larger particles sampled would have a tendency to plate out in the sample line. 5.0
GENERIC SYSTEM IMPLEMENTATION AT HANFORD
6
Stack sampling and monitoring system upgrades at Rockwell are being accomplished in five phases. The Initial phase involves preparation of a site specific design criteria document which provides information necessary for generic system retrofitting and Identifies the number and type of generic systems required to complete the upgrade effort each year. The second phase, purchasing and procurement. Involves preparation of purchase requisitions required for specifying generic system components. Once the equipment is received on site it is cal ibrated to assure proper Instrument operabiiity. The third phase, fabrication, Is approached with an assembly line concept In mind. Initial efforts were directed towards completing one system entirely in the early stages of the upgrade program to Identify and eliminate problem areas. This Initial effort has now been completed. The fourth phase, design and drafting, utilizes criteria Identified in phase one to generate a drawing package detailing Installation of each system at selected sites. This phase terminates with a formal design review of system Installation drawings and supporting documentation. The final phase, installation, Involves site preparation and installation of the generic sampling system per design phase specifica273
tlons. Once Installed, an acceptance test Is performed to verify system operabllIty. 6.0
CONCLUSIONS
The purpose of this paper was to provide an overview of the analytical methods and results of an effort which evaluated the adequacy of airborne radioactive partlculate sampling and monitoring systems at Rockwell. System designs and operations were compared to DOE and Rockwell established standards and guidelines for conformance. Common upgrade requirements lead to the development of a generic stack sampling and monitoring system for radioactive particulates. This paper presents highlights of generic system evolution and application at Rockwell. These highlights lead to the following general conclusions: 1. The computer model was a valuable tool for comparing systems to determine relative Inadequacies, aiding system upgrade prlorltlzatlon efforts. Experimental and theoretical uncertainties, however, require caution to be exercised In considering or using the derived sampling system efficiencies as absolute values. 2. The generic stack sampling and monitoring systems being installed at Rockwell are significant Improvements over their predecessors, incorporating state-of-the-art technology to the extent practical. These systems will provide Improved system reliability, more accurate sample coliection, and ^111 foster a higher degree of confidence that release data reported Is Indicative of actual exhaust stack radioactive discharges. 3. The generic design allows a major portion of the design and review effort to be streamlined by eliminating duplication of efforts. Additional cost savings can be Incurred by standardizing fabrication and Installation requirements, having similar operating procedures, maintenance and calibration requirements, and Interchangeable spare parts. 4. Purchase and fabrication phases are scheduled Independent of the design phase due to the availability of the approved generic design. Therefore, necessary equipment can be ordered against these approved drawings prior to design phase completion and sampllng systems can be fabricated. This enables a smooth transition into the Installation phase minimizing procurement time constraints.
274
During the design phase of the stack sampling system upgrade process a site specific evaluation should be perfomed with respect to alrstream corroslvfty, humidity, flow variability, and expected particle sizes to determine If the generic system Is suitable for application. Adverse conditions may require additional system modifications or enforcement of administrative controls over system operations to assure representative sampling. REFERENCES 1. United States Department of Energy Directive 5480.1, Environmental Protection,. Safety, and Health Protection Program for DOE^Operations, August 13, 1982. 2.
United States Department of Energy Directive 5484.1, Environmental Protection,, Safety, and Health Protection Information Reporting Requirements,. February 24, 1981
3.
United States Department of Energy - Rich I and Operations Office Order 5820.2, Radioactive Waste Management, July 26, 1979.
4. Rockwell Hanford Operations Health, Safety, and Environment Function, RHO-MA-139, Environmental Protection Manual, June 1981. 5. Evenson, R. J., September 1980, RHO-MA-241, Installatlon Criteria for the Generic Airborne Radioactive Contamination Sampling and Monitoring System, Rockwell Hanford Operations, Rfchland, WA. 6.
Cammann, J. W., April 24, 1980, RHO-CD-971, 200 Area Stack Sampler-Monitor Systems Upgrade Approach, RockwelI Hanford Operations, Rich!and, WA.
7. Cammann, J. W., Geler, C. J., November 1980, RHO-CD-1092, 200 Area, Stack Sampler-Monitor Systems Upgrade; Generic System ApplIcat Ions, Rockwell Hanford Operations, Richland, WA. 8.
Schwendiman, L. C., G. A. Sehmel, and A. K. Postma, November 1963. "Radioactive Particle Retention In Aerosol Transport Systems." Proceedings of the International Conference on Radioactive Pollution of Gaseous Media. Vol. II, pp. 373.
275
9.
Schwendfman, L. C , September 1976, BNWL-1921/UC-70, Appraisal of Airborne Radioactive Effluent Sampling Systems - Compreshenstve flr;qlysis Format and In-Placa Sampler Validation, Pacific Northwest Laboratorles-BattelIe, RichI and, WA.
10. American National Standards Institute, 1969, ANSI N13.1, Guide to Sampling Airborne Radioactive Materials In Nuclear Facilities,. American National Standards Institute, Inc., New York, NY. 11. American National Standards Institute, 1974, ANSI N13.10, Spec! float Ion, and Performance of On-Slte Instrumentation for Continuously Monitoring Radioactivity lr> Effluents, Institute of Electrical and Electronics Engineers, Inc., New York, NY. \2. Scfwendlman, L. C. and J. A. Gllssmeyer, August 1976, "An Analysts Format and Evaluation Methods for Effluent Particle Sampling Systems In Nuclear Facilities,." Proceedings of the Fourteenth ERDA Air Cleaning Conference. Vol. 1, pp. 507-527, CONF-780819, available from NT IS, Department of Commerce, Springfield, VA. 13. Kabat, M. J., "Deposition of Airborne Radioactive Species on Surfaces of Metals and Plastics," 17th DOE Air Cleaning Conference. 14.
Schwendlman, L. C., February 5, 1975, January Progress Report on ARHCO Gaseous Effluent Sampling Studyr Pacific Northwest Laboratory, Rich I and, WA.
15. Bergson, B. P., December 1980, "Use Isoklnetlc Methods for Sampling and Monitoring Nuclear-Plant Stack Effluent," Power MagazIne, pp. 41-44, Bechtel Power Corporation.
276
SESSION SIX ENVIRONMENTAL MONITORING I
6A
SOME PROBLEMS IN EVALUATING RADIATION DOSE FOR EFFECTIVE ENVIRONMENTAL ASSESSMENT E. C. Watson, B. E. Vaughan, J. K. Soldat, R. G. Schreckhise and D. H. McKenzie Pacific Northwest Laboratory Richland, Washington
ABSTRACT This review is concerned with radiation dose assessment methods and their accuracy. Although the models used in radiation dose assessment are inherently capable of better absolute accuracy, in practice, a calculated dose is usually overestimated by from two to six orders of magnitude. The principal reason for so large an error lies in using "generic" concentration ratios in situations where site specific data are needed. Major opinion of the model makers suggests a number midway between these extremes, with only a small likelihood of ever underestimating the radiation dose. Factors influencing dose include: source considerations (i.e., physical and chemical status of released material); dispersal mechanisms (atmospheric, hydrologic and biotic vector transport); mobilization and uptake mechanisms (i.e., chemical and other characteristics affecting the biological availability of radioelements); and critical pathways. Examples are shown of confounding in food-chain pathways, due to indiscriminate application of concentration ratios. The pathways models may also require improved parametrization, as they are not at present structured adequately to lend themselves to validation. The extremely wide errors associated with predicting exposure stand in striking contrast to the error range associated with the extrapolation of animal effects data to the human being.
This afternoon, I want to comment on radiation dose assessment methods (somewhat generically) and the accuracy with which such methods determine the calculated risks of incurring health effects from exposure to radiation. An extensive review by Vaughan, et al. of environmental radiation dose assessment methods has been published in Environmental Health Perspectives (1). I will not discuss the problems associated with estimating human health effects by extrapolation of animal data to humans. Such an extrapolation has been treated elsewhere (2).
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Radiation dose estimation, as used for assessment purposes, depends critically on how exposure factors are determined. Several prevalent misunderstandings about the significance of radiation dose estimates have suggested a need for comprehensive review of dose estimate methods. Those responsible for performing dose assessments have usually been well aware of constraints limiting their accuracy (3). However, many other persons concerned with dose data, in the biomedical, engineering and legislative communities, tend to believe that the mathematical models used to generate dose estimates are capable of a degree of accuracy that is impossible to achieve in practice. This belief is especially enhanced if the models are exercised by a sophisticated computer system. Only a few people have noted that the model developers, themselves, believe the absolute errors in the use of their models led to doses overestimated by anywhere from two to six orders of magnitude (3, 4). As contrasted with uncertainties from two-fold to ten-fold in the prediction of human health risks from laboratory data (2), it is therefore highly desirable to refocus attention on the models used to estimate dose and the factors used to integrate these estimates into projections of human health effects. At a time when calculated annual radiation doses were often only a small fraction of the then allowable limits, inaccuracy of dose estimates although large was not a big problem. Subsequent developments have changed this situation drastically; i.e., issuance of FRC guidelines in 1960, and particularly the gradual development of the philosophy of maintaining doses at levels aslow-as-reasonably-achievable (the so-called ALARA philosophy). At the same time, environmental concentrations resulting from human endeavors decreased for the most part, submerging into natural background levels. With the application of ALARA, a convergence of radiation standards and environmental concentrations has occurred. The response to this has been to attempt to improve accuracy in calculating doses by increasing model sophistication; i.e., inclusion of more exposure pathways of increased complexity. Yet, the real need is to improve the quality, specificity, and realism of both model parameters and data. A more judicious selection of existing data, some additional research, and a reconsideration of present model structure are all involved. Time does not permit me to describe the mathematical models and the computer codes that implement them. Most of these have been described in detail in the literature (5-8). Many of their elements are provisional, but the models provide the only practicable way of accounting for an extremely large number of variables (see Figure 1). Such models have utility where the substance under study 1) is comparatively stable, 2) is noxious at low levels, 3) disseminates through a multiplicity of environmental pathways, and 4) requires comparatively long time intervals for tissue accumulation and induction of health effects. The most serious constraints in using the models are the needed development of a sometimes impracticable data base required for implementation, and the comparative lack of data specific to real locations. When measurable concentrations of radionuclides in air, water and foods existed at measurable levels near large AEC installations, it was a relatively
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HYOROLOGIC MODEL
ATMOSPHERIC MODEL
A SOURCE
WATER (SURFACE. SUBSURFACE. INTERSTITIAL)
ATMOSPHERE (GASES. PARTICIPATES!
J
D SOILS AND SEDIMENTS
I FOOD CHAIN MODEL 7
I
MICROBIOTA lOECOMPOSfRSI
f PLANTS
TERRESTRIAL ANIMALS 12F
C AruATIC ANIMALS
12H
J 14C
I HUMANS I
HUMAN DOSE MODEL
'
I
FIGURE 1. Environmental pathways model for dose assessments showing major routes affecting man: (1) aquatic discharge; (2) atmospheric discharges; (3) irrigation, water/sediment interphase exchanges; (4) surface deposition; (5) leaching, erosion, mineralization, sediment/water interphase exchanges; (6) resuspension/evaporation; (7) microbial incorporation; (8) microbial releases, decomposition; (9) skin absorption; (10) irrigation plant surface exchanges; (11) root uptake; (12) ingestion; (13) mortality; (14) inhalation. The representation above delineates relationship between 4 major submodels, on which dose assessment depends. The detailed computer codes are too complicated to permit diagrammatic representation and no one submodel is adequate by itself to determine critical pathways (see reference 1). 281
simple matter to calculate radiation dose by combining the measured concentrations with living and dietary habits. The calculated annual doses were generally a small fraction of the then existing limits, with only a few exceptions. In addition, radioactive fallout from nuclear weapons testing provided measurable concentrations of several specific radionuclides in the environment, 89 90 » S r , 1 3 1 I , and 13lt>llt7Cs to name a few. These two sources of environmental radioactivity provided some of the first (and in certain instances the only) field data on the behavior of radionuclides in the environment. They also, quite naturally, gave rise to empirical equations designed to predict the concentration of selected radionuclides in the human diet (9, 10). To consider a cattle grazing ecosystem, for example, the relevant compartments would include air, water, soil, primary producers (plants), primary consumers (grazers), secondary consumers (predators) and decomposers (bacteria, fungi), any of which can be further divided, e.g., grasses, forbs, shrubs and trees for the primary producers. For food chain evaluation, the requisite level of detail will depend on who is eating and what is being eaten specifically (11). Additional model subcomponents are needed to deal with variabilities in dietary consumption patterns, living habits, and the specific physiology of given radioelements. The necessary computer subroutines are well established in current codes for computing radiological dose (8, 12, 13) and they are under more or less continuous revision. An early ICRP publication, ICRP Publication 2 (14), its numerous subsequent publications, and the deliberations of model builders should be consulted for current details (4). My comments are limited to some of the environmental pathway subcomponents now in need of better focused data. (The reader is referred to reference 1 for a complete discussion.) The modeling problem is one of determining the distribution of pollutant concentrations in a compartment. Functional connections between compartments are thus represented by steady state interchanges. Typical linkage processes include adsorption, absorption, inhalation, ingestion, excretion, decomposition, and dissolution. Few of these processes follow the reaction kinetics generally assumed by model builders. It is clear that seasonal production (vegetation) and reproductive cycles affecting population size (fish, animals and birds) lead to modeling difficulties because of decidedly non-steady state conditions (15). Our current inability to appropriately describe these time dependent processes sometimes leads to excessive variability in projecting dose estimates, and it certainly prevents application of the present modeling approach to predicting ecosystems responses. Attempts at rigorous compartment modeling conducted at several DOE laboratories have been only partially successful (15-21). More work is needed also on the parameterization of transport rates. I would like now to make a few comments about some of these compartments. First, the source term. Particle size, physical state, molecular form, the presence of codisposed organic complexes, and release rates are basic factors controlling biological availability of a radioelement to different biotic
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receptors. Since data of these kinds are costly to obtain, process engineers seldom provide them. If they are provided, some selectivity is required on the part of the analysist. The analysist should prioritize the needed information, based upon an understanding cf the metabolic potential of released materials. Among the factors just mentioned, the biological importance of aerosol particle size is often overlooked. In terrestrial ecosystems, both particle form and size determine subsequent soil/root behavior and interaction on the plant leaf. This is true whether the particle originates in the upper atmosphere or from wind -resuspended soils. Suitable particle size data are rarely available to interpret observed plant uptake of radioelements. As a general rule, the smaller the size of particle deposited on a plant, the greater its likelihood of biological interaction, regardless of its chemical solubility. A leaf can sorb and translocate components of initially anhydrous compounds and compounds of extremely limited solubility (22). Particles of 1 jam or less in size are unlikely to be washed off the leaf (23), a point not generally appreciated. Adherence is probably related to epidermal structures, the surface area/volume ratio, and charge densities on the surface macromolecules of the leaves. Some particles in air, like sodium-plutonium oxide and certain refractories, are unusual in disintegrating on contact with water to form exceedingly small particles (24). It is not known how readily assimilable these smaller particles may be, although in animal body fluids, translocation is reportedly enhanced 45-fold (25). There are very few nuclear industrial data on size of the particulates released and no data on how particulates are transformed in the atmosphere. The EPA requires measurement of the number of suspended particles, but it does not currently require the determination of the particle size distribution (26). Tc is an excellent example of the need for source characterization to talk -bout. The environmental and metabolic behavior of technetium, a potentially large contributor to radiation dose, has been widely investigated in recent years. In 1959 the ICRP (14) defined the metabolic behavior of Tc and indicated that it was taken up less readily by humans than iodine. Nevertheless, environmental analysist considered it to be an iodine-like element. The use of the pertechnetate form of Tc in thyroid function (uptake) tests as a substitute for radioiodine began a few years ago. Since then, the possibility that technetium-99 exists in the environment in the pertechnetate form has stimulated several studies of its form and uptake by plant roots (27). Several researchers also reported surprisingly high Tc uptake by plants from soil, based on laboratory experiments (28, 29). All of this led to heightened concern that the potential doses from Tc-99 in the environment were being grossly underestimated. However, more recent research has indicated that the appropriate concentration ratio (CR) for use in the food-chain models are lower than first implied by the laboratory studies and that Tc ingested with food does not behave like iodine in the body (30). This situation illustrates the need for additional research on several other elements where chemical similarity to another element was assumed for modeling purposes.
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A word about dispersal mechanisms. In the terrestrial environment, atmospheric dispersal/deposition, direct discharge to waterways, and leachates from soil-stored wastes are governing pathways for soil and water concentrations. The first principles of these processes are fairly well understood, and it is mainly the specific interactions of particular elements that present problems. Airborne materials that deposit directly onto surface waters usually make only a small contribution to water concentrations compared to material deposited onto land in the watershed and subsequently washed into the river or lake (31). Regarding the atmospheric dispersal compartment. Because people and green plants can act as integrators over time for the low level, long-period release of radioelements, those concerned with food-chain assessment are usually interested in annual mean deposition rates about the point of release. Several atmospheric models have been constructed for the prediction of annual deposition of trace metals from combustion facilities and for radioelements applicable to nuclear fuel cycle facilities, principally for open terrain (31, 32). These submodels (see Figure 1) have not been validated for deposition in forest or field canopies, and deposition on these canopies is therefore probably underestimated by their use. Recycling of radioelements from soil deposits to the green leaf (via wind resuspension or gaseous diffusion from soil) also has not been measured. However, resuspended soil would appear to provide a 5- to 10-times greater source to the plant than the root/soil interface (33), and these data should be established. Mobilization and uptake mechanisms depend on biological availability which in turn depends on chemical composition/interaction of a contaminant with its immediate environment, its uptake route, metabolism and resuspension (1). Most of these mechanisms are circumvented by using concentration ratios that, more often than not, have been determined for circumstances which differ significantly from those under investigation. Concentration ratios are significantly influenced by biological availability processes and uptake routes. The ratios themselves represent operational definitions that (in the case of plant/soil ratios) lump together soil desorption, root uptake, foliar uptake, microbial and soil solution chemical equilibria. Where the published concentration ratios vary over 4 or 5 orders of magnitude (4, 34, 35), each of these factors may need to be evaluated. The concept of critical pathways has been used to permit needed simplication of an already too complicated environmental model (36, 37). If calculations using generic concentration ratios leads to minor dose contributions from a radioelement being transported in certain pathways, these pathways might be feasibly ignored. Attention then might be focused on getting specific data to determine transport in the dominant pathways. Today, there is discussion of abandoning this approach for regulatory purposes, and instead attempting a total mathematical model computation. Considering the many possible sources of error and the provisional nature of much of the present radiological data base, this choice, I believe, is ill-advised.
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Regarding the accuracy of dose estimates. I believe the authors of the environmental support document to the Generic EIS for Commercial Waste Management (3) have expressed it well when they said, The consensus of those individuals contributing to dose assessments is, for any given dose estimate, the actual dose that would be received by the regional population in the reference environment would not be more than 10 times the stated value, nor would it be less than 1/100 of the stated value. Thus, the likelihood of actual values exceeding estimates is low, whereas the likelihood of values actually being substantially less than estimated is rather high.
In another comment (3) Because of additional uncertainties in modeling for a worldwide dose, the consensus is that doses received by the worldwide population would not be more than 100 times the estimate given, nov less than 1/1000 of the value given. Also (3), ...doses presented in this report are best estimates of the doses; it would be improper to multiply all doses by 10 when it is just as likely that the true dose is 1/100 of the stated dose. Some investigators have estimated that the absolute error band in calculated radiation exposure to a population exposed through aquatic food chains might be as much as a million-fold (4). The solution was recognized in the RIME report (38) where Foster stated Perhaps the greatest uncertainties in predictive calculation is the selection of an appropriate concentration factor..., and ... wide variations that have been observed among different environments, and even among closely related species in the same environment emphasize the need for careful consideration of the specific characteristics of each site. Current practice is to use generic concentration ratios. The doses so calculated, in the past at least, have been acceptable for routine determination of compliance (dose to maximum individual) when the computed doses represented a small fraction of the design limit. These ratios may not be applicable for other purposes. Health physicists are forced to make decisions on which parameter values are most applicable to their particular situation when sometimes there is only one value available. The value may have been determined under conditions far different from those involved in the dose calculation.
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Pathway analysis and radiological assessment must be a continuing reiterative processes until there is sufficient confidence in both the mathematical models and the parameter values for unequivocal decision making. In fact, this second need has been recently recognized by Committee 64 of the NCRP, which ,is considering the problem of site-specific data. Considering the above problems, particular caution also needs to be exercised in applying a set of concentration ratios derived from one set of exposure circumstances to a new set of circumstances. When comparing different situations, such as surface stored sources, fugitive emissions from industrial facilities, and accidents, one finds that the relative importance of the several pathways shown in Figure 1 is different for each situation. As for goals and future direction of environmental assessment through model research. Current research on environmental pathways should be aimed at two goals, not readily achievable at the present time: 1) Improving the absolute accuracy of dose estimation, and 2) Validating the predictive capability of dose estimation models, by field measurements under controlled conditions. To accomplish these goals, particular attention needs to be paid to the following considerations: .
Accelerating the collection of site-specific data bases. This must be done for specific configurations of release, for specific molecular forms of the released material, and for specific physical and ecological features of I'^e site. Developing generic quant irative approaches for those descriptive processes that have not .•/•t been quantitated. This includes a determination of the range c.;>r which such variables may change.
.
Improving the structure and particularly the parameterization, of the present dose assessment models. The models are not adequately structured to lend themselves readily to validation. More rigorous evaluations are needed to investigate model sensitivity to changes in sets of interrelated parameters and alternative formulations for individual components.
. Maintaining the critical pathway focus in data collection. Models should be an integral part of the decision and allocation process. They provide a framework for identifying needed information, evaluating its contribution to improving our understanding of the system and thereby developing criteria for appropriate allocation of research efforts. Such criteria will differ depending on specific circumstances of a radiological release.
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REFERENCES 1. Vaughan, B. E., J. K. Soldat, R. G. Schreckhise, E. C. Watson, and D. H. McKenzie. 1981. "Problems in Evaluating Radiation Dose Via Terrestrial and Aquatic Pathways." Environmental Health Perspectives, 42. 2. Bates, R. 1981. "Critical Appraisal of Current Toxicological Approaches Utilized in Context of Health Risk Evaluation. In: Symposium on Health Risk Analysis. P. J. Walsh (ed.) Gatiinburg, TN October 27-30, 1980 Oak Ridge National Laboratory (Publisher). 3. U.S. Department of Energy. Environmental Aspects of Commercial Radiological Waste Management. Vol. 1, page 3.8.1. Office of Nuclear Waste Management, Washington D.C. 4. Hoffman, F. 0. (Ed.). 1978. Proceedings of Workshop on the Evaluation of Models Used for the Environmental Assessment of Radionuclide Releases. CONF-770901, TIC, Oak Ridge, Tennessee. 5. Fletcher, J. F., and W. L. Dotson (compilers). 1971. "HERMES - A Digital Computer Code for Estimating Regional Radiological Effects from the Nuclear Power Industry." USAEC Report HEDL-TME-71-168, National Technical Information Service, Springfield, Virginia. 6. Napier, B. A., W. E. Kennedy and J. K. Soldat. 1980. PABLM - A Computer Code for Calculating Accumulated Radiation Dose to Man From Radionuclides in the Environment? PNL-3209, Pacific Northwest Laboratory, Richland, Washington. 7. Soldat, J. K., N. M. Robinson and D. A. Baker. 1974. "Models and Computer Codes for Evaluating Environmental Radiation Doses." BNWL-1754, Pacific Northwest Laboratory, Richland, Washington. 8. Brenchley, D. L., J. K. Soldat, J. A. McNeese and E. C. Watson. 1977. "Environmental Assessment Methodology for the Nuclear Fuel Cycle." BNWL-2219, Pacific Northwest Laboratory, Richland, Washington. 9. Garner, R. J. 1972. Transfer of Radioactive Materials from the Terrestrial Environment to Animals and Man. pp. 15-20. CRC Press, Cleveland, Ohio. 10. Russell, R. S. 1966. Radioactivity in Human Diets, pp 194, 195, 302. Pergamon Press, Mew York. 11. Rogers, L. E., and W. H. Richard (ed.). 1977. "Ecology of the 200-Area Plateau: A Status Report." PNL-2253, Pacific Northwest Laboratory, Richland, Washington.
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12. Napier, B. A., R. L. Rosweil, W. E. Kennedy and D. L. Strenge. "ARRRG and FOOD - Computer Programs for Calculating Radiation Dose to Man From Radionuclides in the Environment." PNL-3180, Pacific Northwest Laboratory, Richland, Washington. 13. Baker, D. A., G. R. Hoenes and J. K. Soldat. 1976. "Food - An Interactive Code to Calculate Internal Radiation Doses From Contaminated Food Products." Proc: Conference on Environmental Modeling and Simulation. Cincinnati, Ohio, April 20-22, 1976. pp. 204-208. EPA 600/9-76-016. 14.
International Commission on Radiological Protection. 1959. Report of Committee II on Permissible Dose for Internal Radiation. ICRP Publication 2. Pergamon Press, New York.
15. Cadwell, L. L., R. G. Schreckhise and R. E. Fitzner. 1979. "Cesium-137 in Coots (Fulica americana) on Hanford Waste Ponds: Contribution to Population Dose and Offsite Transport Estimates." Low-Level Radioactive Waste Management. Proc. Health Physics Society 12th Midyear Topical Symposium. Williamsburg, Virginia. 16. Mo, T., and F. G. Lowman. 1975. "Laboratory Experiments on the Transfer of Plutonium From Marine Sediments to Seawater and to Marine Organisms." Radioecology and Energy Resources, pp. 86-95. C. E. Cushing (ed.), Proc. Fourth National Symposium on Radioecology, Corvallis, Oregon, May 12-14, 1975. 17. Adams, L. W., G. C. White and T. J. Peterle. 1975. "Tritium Kinetics in a Freshwater Marsh." Radioecology and Energy Resources, pp. 96-102. C. E. Cushing (ed.), Proc. Fourth National Symposium on Radioecology, Corvallis, Oregon, May 12-14, 1975. 18. Vanderploeg, H. A., R. S. Booth and F. H. Clark. 1975. "A Specific Activity and Concentration Model Applied to Cesium-137 Movement in a Eutrophic Lake." Radioecology and Energy Resources, pp. 164-177. C. E. Cushing (ed.), Proc. Fourth National Sumposium on Radioecology, Corvallis, Oregon, May 12-14, 1975. 19. Cummings, S. L., J. H. Jenkins, T. T. Fendley, L. Bankert, P. H. Bodrosian and C. R. Porter. 1971. "Cesium-137 in White-Tailed Deer as Related to Vegetation and Soils of the Southeastern United States." Radionuclides in Ecosystems, pp. 123-128. D. J. Nelson (ed.). Proc. Third National Symposium on Radioecology, Oak Ridge, Tennessee, May 10-12, 1971. 20. Olson, J. S. 1965. "Equations for Cesium Transfer in a Liriodendron Forest." Health Phys. JA:138521.
Eberheadt, L. L., and W. C. Hanson. 1969. "A Simulation Model for an Arctic Food Chain." Health Phys. 17:793.
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22. Cataldo, D. A., T. R. Garland, R. E. Wildung and J. M. Thomas. 1980. "Foliar Absorption of Transuranic Elements: Influence of Physiochemical Form and Environmental Factors." J. Environ. Qua!. J3. 23. Vaughan, B. E., R. E. Wildung, and J. J. Fuquay. 1976. "Transport of Airborne Effluents to Man Via the Food Chain." Controlling Airborne. Effluents from Fuel Cycle Plants. CONF-76-0806, TIC, Oak Ridge, Tennessee. 24. Chatfield, E. J. 1969. "Some Studies of the Aerosols Produced by the Combustion or Vaporization of Plutonium Alkali Metal Mixtures, II." J. Nucl. Mater. 32:247. 25. Strather, J. W., S. Dowden and R. F. Carter. 1975. "Method for Investigating the Metabolism of the Transportable Fraction of Plutonium Aerosols". Phys. Med. Biol. 20:106. 26. U.S. Environmental Protection Agency. National Primary and Secondary Ambient Air Quality Standards. Title 40 Code of Federal Regulations, Part 50, Appendix B: Reference Method for the Determination of Suspended Particulates in the Atmosphere. U.S. Government Printing Office, Washington, D.C, 1979. 27. Garland, T. R., R. G. Schreckhise, R. E. Wildung, L. L. Cadwell, K. M. McFadden and D. A. Cataldo. 1980. Environmental Behavior and Effects of Technetium-99 and Iodine-129. PNL-3300, Part 2, Pacific Northwest Laboratory, Richland, Washington. 28. Wildung, R. E., T. R. Garland, D. C. Cataldo. Technetium by Plants." Health Phys. _32:314.
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"Accumulation of
29. Landan, E. R., L. J. Hart, and R. G. Gast. 1975. "Uptake and Distribution of Technetium-99 in Plants." Bioiogical Implications of Metals in the Environment. H. Drucker and R . E . Wildung (ed.).CONF-750929, National Technical Information Service, Springfield, Virginia. 30. Wildung, R. E., K. M. McFadden and T. R. Garland. 1979. "Technetium Sources and Behavior in the Environment." J. Environ. Quality 8:156. 31. Vaughan, B. E., K. H. Abel, D. A. Cataldo, J. M. Hales, C. E. Hane, L. A. Rancitelli, R. C. Routson, R. E. Wildung, and E. G. Wolf. 1975. "Review of Potential Impact on Health and Environmental Quality From Metals Entering the Environment as a Result of Coal Utilization." Report to Battelle Energy Program Directors, Battelle Memorial Institute, Columbus, Ohio. 32. VanHook, R. I., and W. D. Shuts (ed.). 1977. "Effects of Trace Contaminants From Coal Combustion." ERDA 77-64, p 23, National Technical Information Service, Springfield, Virginia.
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33.
Romney, E. M., W. A. Rhoads, A. Wallace, and R. A. Wood. "Persistence of Radionuciides in Soil, Plants and Small Mammals in Areas Contaminated With Radioactive Fallout.: Radionuciides in Ecosystems, pp. 170-176. 0. J. Nelson (ed.). CONF-71O5O1-P1. National Technical Information Service, Springfield, Virginia.
34.
Cataldo, D. A., and R. E. Wildung. 1978. "Soil and Plant Factors Influencing the Accumulation of Heavy Metals by Plants." Environmental Health Perspectives 2Jj 149.
35. Hoffman, F. 0., and C. F. Baes, III. 1979. "A Statistical Analysis of Selected Parameters for Predicting Food Chain Transport and Internal Dose of Radionuciides." NUREG/CR-1004, National Technical Information Service, Springfield, Virginia. 36.
Soldat, J. K. 1071. "Modelling of Environmental Pathways and Radiation Doses From Nuclear Facilities." BNWL-SA-3939, Pacific Northwest Laboratory, Richland, Washington.
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Hoffman, F. 0., C. W. Miller, D. L. Shaeffer and C. T. Garten, Jr. 1977. "Computer Codes for the Assessment of Radionuciides Released to the Environment." Nucl. Safety 18:343.
38.
Panel on Radioactivity in the Marine Environment of the Committee on Oceanography. 1971. National Research Council. pp. 154-159, 247. Radioactivity in the Marine Environment. National Academy of Sciences, Washington, D.C.
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6B STANDARDS FOR CONCENTRATIONS OF RADIONUCLIDES IN SOIL: BASIS AND IMPLEMENTATION Roger C. Brown Gary S. Kephart Rockwell Hanford Operations Rich land, WA 99352
ABSTRACT Levels of acceptable concentrations for specified radionuclides in surface soil for release from a surface contamination status (soil standards) are in use by Rockwell Hanford Operations. They were developed for use in the two areas where the fuel and waste processing and waste management activities are located; these areas are referred to as the 200 Areas, (The soil standards, expressed as average activity per unit mass, are based on criteria to limit the total dose equivalent to acceptable levels.) The soil standards provide assurance that no one will receive more than 500 mR per year (continuous occupancy) from external gamma radiation and that uptake of radionuclides by ingestion of soil particulates, and drinking of surface contaminated water will be less than acceptable standards. These soil standards assume (1) U.S. Department of Energy control of the areas and (2) that no food or forage crop production or grazing will be permitted. Implementation of the appropriate standards suggests a two tiered investigation approach. Three general categories of spatial contamination distribution are possible. An implementation guide has been proposed which would first identify which spatial distribution best characterizes the contamination on a given site. The spatial distribution of the first tier would then form the basis for statistical sampling/survey design and compliance testing (the second tier). INTRODUCTION Acceptable concentrations of radionuclides in surface soil for release from a radiologically controlled status are addressed at Rockwell Hanford Operations (Rockwell) by soil standards contained in an environmental protection manual. The need for such standards was originally fulfilled through the development of the 200 Areas soil standards (Boothe, 1979). At that time, it was anticipated that 600 Area soil standards would be developed in the future as a second level of contamination standards for inclusion in the environmental protection manual.
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The existing soil standards are in use in the two areas where the chemical processing and waste management activities are located. These areas are referred to as the 200 Areas. The remaining unoccupied areas of the Hanford Reservation for which Rockwell has assigned responsibilities are known as the 600 Area. It is important to recognize the intent of the standards discussed here. Both the 200 and 600 Areas contain inactive waste sites such as cribs and burial grounds on which we anticipate remedial action may be performed. The standards discussed in this paper were developed for use as decision levels such that decontamination efforts could be planned with a clearly defined goal. The following sections will discuss (1) the bases for setting the 200 Areas soil standards, (2) some representative radionuclide concentrations which result from these standards, and (3) an implementation plan for compliance with the above standards. Although there is a great deal of information available relating to soil contamination standards for both facility operations and decommissioning activities (Mueller et al. 1981), direct comparison of these standards is difficult since most criteria are directed toward specific uses thereby employing different bases of development. Bases The 200 Areas of the Hanford Reservation encompass two areas where the chemical processing and waste management activities are located. Waste management facilities contiguous with the 200 Areas are also encompassed under the 200 Areas soil standards. Control of personnel access to the 200 Areas is restricted on the basis of classification of these areas as controlled areas (DOE 5480.1). The 600 Area is not routinely occupied by personnel on a permanent basis. Access is not as restricted as for the 200 Area although administrative controls are in effect to limit access. The 600 Area occupies the vast majority of the 570 square miles encompassing the Hanford Reservation (ERDA 1538). There are sources of radioactive materials in the 600 Area from inactive burial sites. These differences in usage and personnel occupancy are factors in setting acceptable standards for concentration of radionuclides in the surface soil. The criteria for determining the average radionuclide concentrations in soil are presented in Table 1. It should be noted that no provision is made for uptake of radionuclides from food. U.S. Department of Energy control is assumed with prohibitions on the production of food or forage crops.
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TABLE 1. Criteria for Determination of Average Concentration of Radionuclides in Soil. Criteria
Pathway External Exposure Internal Exposure (a) Direct Ingestion of Soil (b) Inhalation (c) Drinking (d) Food
500 tnR/yr 0.1 body burden/yr Table II MPC Table II MPC* Not Considered
NOTE: The most restrictive of the constituents of internal exposure is used to determine the average concentration exclusive of external exposure considerations (Boothe, 1979). The rationale for adopting these soil standards should be expanded. One alternative approach would be to provide one set of soil standards for unrestricted release of previously contaminated sites. However, it would be impractical to implement such standards where the areas adjoining the site are not known to be below these same soil standards. That is, unconditional release of a site would require recognition that the possibility of recontamination via transport from adjacent areas be negligible. We were unwilling to make such a generic assumption in the Hanford case; therefore, prior evidence of localized compliance of soils with unconditional release standards would be of limited value. Food and forage crops are not and probably will not be grown at Kanford. In view of these factors coupled with uncertainties in modeling (Little and Miller, 1979), "interim" standards which explicitly avoid unconditional release have been proposed. Development of such interim nil standards has resulted from the specifics of the Hanford situation described. Derivation of Radionuclide Concentrations in Soil Application of the criteria presented in Table 1 allows the average concentrations of radionuclides in soil to be calculated on a dry weight basis. The approach to the inhalation pathway for the 2Q1 Areas assumes a resuspension factor of 10~° cm"' to estimate the concentration of soil particulates in the air (Boothe, 1979). For the soil standards the maximum exposure rate of 500 mR per year must be ascertained by measurement. Only if this condition is met can the average soil concentration be compared with the calculated values. The compliance to the values of average soil concentrations will not in and of itself assure compliance with the maximum exposure rate limitation. These soil standards use the lowest soil concentration from the three pathways considered for internal exposure. Soil standards for selected radionuclides are presented in Table 2. 293
TABLE 2. Average Concentrations of Selected Radionuclides in Soil. Average Concentration (pCi/g)
Radionuclide Cobalt Strontium-90/Yttrium-90 Cesium-137 Plutonium-239
300 400 400 60
IMPLEMENTATION Having arrived at standards for acceptable radionuclide concentrations in surface soils, the next obvious step is to develop a methodology for their implementation. We have done some preliminary work in developing a technically defensible application of standards in testing site soils for compliance. Measurement capable of detecting the limiting concentrations for all radionuclides of concern over all the soil mass on the site would be the ideal compliance test. The challenge in implementing soil standards is to approach this ideal sufficiently to make a safe determination of compliance within the constraints of available funds and technology. Many complications can arise in design of a suitable survey/sampling plan. Factors to be considered in design of a site specific survey/sampling plan include the following: 1) Radionuclides known or suspected as being present. 2) Detection capabilities of instruments or analyses relative to item 1 above, the concentration limits of interest, etc. 3) The expected spatial distribution of the contaminants over the sites; i.e., both the areal and depth distributions are of interest. 4) Cost of otherwise equivalent approaches. 5) Level ofa confidence required of the site evaluation as a decision basis.' ) (a)For example, an investigator might require 95% confidence that any accumulation of contaminated soils (greater than the specified concentration limits) of size larger than some specified level (of mass, volume, activity, etc.) have not escaped detection by the survey/sampling plan. Likewise some confidence (say 75%) may be desired in the determination that controls will not be maintained erroneously on a site for which soils are actually within the limiting concentrations. 294
We will discuss items 1, 2, and 3 in greater detail here, assuming the significance of cost is obvious and that the acceptable level of required confidence is basically a policy decision (Barnes, 1981). Radionuclide(s) of Interest Knowledge of what one is looking for is basic to any search. In this case, identifying the radionuclide(s) of concern is basic to selection of appropriate instruments or analyses of soil samples. Established chemical, physical, and biological properties of the radionuclide(s) may lead to assumptions regarding their distribution in the site soils. Where radionuclide mixture ratios are well defined it may be possible to simplify survey requirements. Thus there are radionuclide specific aspects of the instrument and distribution factors discussed in later sections. Since multiple radionuclides could present a cumulative hazard, it is important that the site be evaluated for all species which may be present in any significant fraction of their respective concentration limits. Instrument Limitations In situ evaluation of site soils would generally be preferred over laboratory analyses in terms of economy. However, particularly at the concentrations of interest in our application, instrument limitations must be recognized. Measurements of alpha or beta emitting radionuclides in soil are obviously limited by their high attenuation. Characterization of gamma emitting radionuclide in surface soils is considerably more achievable in situ and some efforts have been reported in the literature. Even so, field-of-view and count time considerations may impact economical survey of large areas, A ratio technique measuring gamma constituents in the field and estimating associated alpha and/or "pure" bets emitting radionuclide concentrations may be feasible where such ratios can be established in intial laboratory analyses or estimated using conservative assumptions. Any in situ technique is subject to a magnitude of environmental parameters such as geometry variations, soil moisture content fluctuations, temperature effects on detectors, etc. Reported efforts to correlate field (in situ) measurements with laboratory analyses have shown large variations (Gilbert, 1979).* A mobile laboratory approach coupled with sequential analysis statistical techniques (Howe, 1982; Wald, 1947) may be the most economical and technically suitable compromise in view of these problems. In sequential analysis the null hypothesis is tested continuously (i.e., as each sample analysis or field measurement is added to the data base). All data to date provide the bases of accepting or rejecting the hypothesis (soil compliance) or finding that the evidence is insufficient for either decision (continue sampling). Mobile laboratory capabilities *(Gilbert and Barnes, 1979) 295
would allow practical application of this statistical technique for minimization of sample numbers. In addition, sample locations could be biased based on in situ measurements and samples prepared as necessary to eliminate some environmental variables. Spatial Distribution Spatial distribution concerns present the most complex obstacle to adequate «sign of any site soil "release" plan which utilizes a statistical approach. This complication reduces to the basic question: what is a reasonable mass (volume) of soil over which activity measured should be averaged to arrive at a concentration determination? For example, depth distribution of contamination must be recognized to avoid dilution error. If an airborne release results in a thin surface deposition of high concfMtration, sampling of a 30 cm core followed by homogenizing, aliquotiny anJ analyses would seriously dilute or complete1v mask the nature of the problem. The converse situation would arise over cribs or burial grounds where contaminant concentration may increase with depth from the surface (Figure 1). We have also found that for a given depth distribution or depth of interest, the investigator must recognize three general cases of areal distribution and differentiate one from another. For our purposes we have defined the three cases es: Continuous, in which the contamination is present in a relatively uniform distribution across the site; Discontinuous, in which contamination is present in localized targets or hot spots whose actual dimensions, activities, and orientations are related to the grid size'required to locate them with some established level of statistical confidence (Gilbert, 1982); and Spotty, in which contamination is present as individual particles or very small hot spots of a dimension which is some fraction of the typical sample area or survey instrument field-of-view (Figure 2 ) . Handling of every site as if spotty contamination were present would be conservative, but very costly in that virtually the entire mass of site soil would have to be surveyed/sampled in order to achieve a reasonable confidence of locating the contaminant specks. Similarly, treating a discontinuous distribution as if it were continuous could result in a survey or sampling design which fails to recognize localized, highly contaminated problem areas. A Two-Step Approach In urder to achieve a reasonable balance between costs and confidence in characterizing contamination a two-step approach is suggested. In this approach a review of site historical data forms the basis for design of the initial survey/sampling effort. Information such as site dimensions, topography, usage, mechanism of contamination, decontamination efforts, etc., supplies a source for assumptions of radionuclide(s) of concern, their expected distributions, and probable concentrations in soil. The initial survey/sampling plan is then directed at confirming these assumptions and especially toward establishing the areal and depth distribution models applicable to the site. 296
SURFACE u5 CONTAMINATION BURIED CONTAMINATION
a: OJ
o o
OLD FALLOUT or J CONTAMINATED DRAINAGE
VERTICAL DISTANCE FROM SURFACE
Figure 1.
Illustrations of Typical Depth Distributions of Soil Contamination
297
DISCONTINUOUS
SPOTTY
CONTINUOUS .
/
i
;
/
/
/
'
•
.•.
I
y
,y.-. •
•
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•
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,
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'
'
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RADIONUCLIDE CONCENTRATIONS > DETECTION LIMITS RADIONUCLIDE CONCENTRATIONS >. SURFACE SOIL CONTAMINATION STANDARDS
Figure 2.
Examples of Typical Area! Distributions of Soil Contamination
298
The second tier of the survey/sampling plan would then supplement the initial data as appropriate to the areal distribution model selected and the statistical confidence level required. The overall benefit of the two-tiered approach is to reduce the number of samples required to achieve a certain confidence. While much of this effort is necessarily site specific, we have drafted some general guidelines for site release plans and compliance testing. CONCLUSION The development of soil standards for the ?00 Areas of the Hanford Reservation has been discussed and concentrations for some typical radionuclides in soil have been presented. The implementation procedures necessary to assure compliance with the applicable soil standards have also been discussed. Guidelines for application of the soil standards have been developed. However, experience is needed to provide feedback on the adequacy of both the standards and the implementation techniques. The practicality of implementation methods must also be established. A final comment is in order on the relative importance of technical and policy decisions required to implement soil standards. The technical problems, while complex and laden with assumptions, are amenable to solutions, Policy considerations are equally as challenging and may be as time consuming as are the technical aspects.
299
REFERENCES Boothe, Gary F., 1979. Surface Soil Contamination Standards, RHO-CD-782, Rockwell Hanford Operations, Richland, Washington. Barnes, M. G., 1981. Statistics and the Statistician in Nuclear Site Decontamination and Decommissioning-Lecture Notes for a 4-Day Short Course. PNL-SA-9486, Pacific Northwest Laboratory, Richland, Washington. Department of Energy, 1981. Environmental Protection, Safety, and Health Protection Program for DOE Operations, DOE Order 5480.1A, Washington, D.C. Energy Research and Development Administration, 1975. Final Environmental Statement, Waste Management Operations, Hanford Reservation, Richland, Washington; ERDA-1538, Springfield, Virginia. ~ Gilbert, R. 0. and M. G. Barnes, 1979. TRANS-STAT Statistics for Environmental Transuranic Studies. Number 8, PNL-SA-7655, Pacific Northwest Laboratory, Richland, Washington. Gilbert, R. 0., 1982. TRAN-STAT Statistics for Environmental Studies, Number 19, PNL-SA-10274, Pacific Northwest Laboratory, Richland, Washington. Howe, H. L., 1982. "Increasing Efficiency in Evaluation Research: The Use of Sequential Analysis." Amer. J. Pub. Health. 72(7): 690-697. Little, C. A. and C. W. Miller, 1979. The Uncertainty Associated with Selected Environmental Transport Mocfels, ORNL-5528, Oak Ridge National Laboratory, Oak Ridge, Tennessee. Mueller, M. A., W. E. Kennedy, Or., and J. K. Soldat, 1981. Review of Soil Contamination Guidance, PNL-3866, Pacific Northwest Laboratory, Richland, Washington. Wald, A., 1947. Sequential Analysis. John Wiley and Sons, Inc., New York, New York.
300
6C
SCREENING LEVELS FOR RADIONUCLIDES IN SOIL: APPLICATION TO DECONTAMINATION AND DECOMMISSIONING (D&D) CRITERIA Susan K. Rope and Steven R. Adams EG&G Idaho, Inc., INEL Idaho Falls, Idaho
ABSTRACT Decommissioning of nuclear facilities can require reducing residual radioactivity in uncontrolled areas to acceptable levels which are based on dose limits. In developing D&D criteria, it is useful to apply predictive methods for dose calculations in an inverse manner, e.g., to back-calculate radionuclide concentration levels in soil that result in a unit radiation dose to an individual. This approach is being used at the INEL (1) as input to a D&D long-range plan, and (2) to provide perspective to measured radionuclide concentrations in soil from fallout, background, and site operations. In this paper, radionuclide concentrations in soil are presented which correspond to a 1.0 mrem/year dose equivalent to a homesteading individual. Those concentrations are called "screening levels." The theoretical individual resides on the contaminated area and derives all his food from locally-raised sources. Pathways considered are ingestion of meat, milk, and vegetables (no water ingestion pathway), inhalation of resuspended soil, and external radiation dose from soil. The relative significance of each pathway is discussed. Screening levels for each radionuclide are based on limiting the total radiation dose from all pathways to 1.0 mrem/year (specifically 1.0 mrem weighted committed dose equivalent using ICRP 26 methodology). Using sitespecific data for Idaho, screening levels for 13 radionuclides associated with D&D activities are derived. Calculated screening levels are compared with those proposed by the EPA and other authors. The results presented in this paper are preliminary. The screening level derivation is only one factor used in the D&D decision analysis, and the derived values are guidelines, not rigid criteria. Because of the conservative scenario used, the screening level concentrations may be several orders of magnitude low compared to those from a realistic analysis. Screening levels derived in this paper generally cannot be measured with widelyavailable field survey instruments.
301
INTRODUCTION During decontamination and decommissioning (D&D) of INEL facilities, contaminated soils may need to be removed so that remaining levels are low enough to permit unrestricted use of the area. The INEL D&D philosophy requires that for unrestricted release, soil will be decontaminated so that the dose equivalent to an individual of the public during fifty years of continuous exposure does not exceed 1 mrem/year above background (Hine and Chapin, 1982). Unrestricted release is not always selected as the D&D alternative. A cost trade-off is made to select the alternative. "Screening levels" for radionuclide contamination in soils have been calculated assuming the land is released immediately to the public for homesteading, farming, and grazing. Screening levels are also calculated assuming 100 years of institutional control prior to release. METHODOLOGY General Approach The soil concentration for each radionuclide is calculated which results in <1 mrem/year over a 50-year period, due to the sum of ingestion, inhalation, and external radiation pathways. That concentration is called the screening level. Screening levels are calculated using a pathways analysis approach similar to that in NRC Regulatory Guide 1.109. The pathways considered are shown in Figure 1. The hypothetical person is assumed to live on the contaminated area, which is large enough for him to raise all his produce, meat, and milk. As well as obtaining all his food from the farm, the person breathes resuspended contaminated soil, ingests a small quantity of soil, and receives an external radiation dose. Based on other studies of INEL surface soil contamination, we assumed that ingestion of water contaminated from surface soil would be insignificant, compared to the pathways shown in Figure 1. Calculation of Dose "Dose" as used in this paper refers specifically to the effective (or weighted) committed dose equivalent (CDE), as defined in ICRP Publication 26. Internal dose conversion factors are obtained from ICRP Publication 30, which gives the 50-year CDE from a single year's intake. This dose commitment (in mrem) is numerically equal to the dose rate (in mrem/year) at the end of 50 years if the same quantity is taken in every year. The assimilation, distribution, and retention of elements in the body is considered, as well as irradiation of one organ by nuclides deposited in another organ. Organ doses are weighted such that the risk is comparable to the total risk from uniform whole body radiation. These weighted organ doses can then be summed to give the effective whole body dose. This effective whole body dose is intended to represent the same risk as if the entire body were irradiated uniformly. Dose conversion factors for external radiation dose from the contaminated ground surface were obtained from Kocher (1980). The ICRP 26 approach overestimates the annual dose for nuclides with 302
long biological half-lives (e.g., Pu), jrf the exposure period is less than 50 years, or if annual intakes decrease with time due to decreased availability of the radionuclides (e.g., due to radioactive decay or weathering processes). However, dose from those nuclides with relatively short biological half-lives (e.g., Cs, Co) will not be greatly overestimated. The weighted whole body dose is useful for the screening level calculations, since organ doses are normalized according to risk. Therefore, weighted organ doses are additive, and a single dose criterion can be used. However, ICRP 26 methodology has not been fully accepted or implemented by the nuclear community. Previous work on priorities for D&D cleanup at the INEL did use ICRP 26. A strict adherence to ICRP 26 methodology ultimately may not be used in the INEL D&D screening levels. Ingest ion Pathway The ingestion pathway methodology is consistent with U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide 1.109. One pathway was added: ingestion of soil by cattle at a rate of 500 g/day and by man at a rate of 0.1 g/day (Healy, 1974). Generic parameter values for the transfer of radionuclides from plants to soil were obtained from a literature review by Miller et al. (1980). Transfer coefficients from forage to meat and milk were obtained from Ng (1982) except for Th and U, which were obtained from Miller at al. (1980). Growing season, crop yield, storage times, and
Figure 1. Pathways Considered for Determination of Screening Levels.
303
interception fraction of participates on vegetation were all estimated from site-specific data. Further analysis of INEL-specific data, as well as recommendations for field studies to support the site modeling efforts, is underway. Inhalation Pathway A mass loading approach is used for assessment of inhalation of contaminated dust. This approach has been recommended for use in generic assessments (Healy, 1980; EPA, 1977). It is very conservative for small areas, as it assumes that all airborne dust comes from soil in the contaminated area. An "enrichment factor" of 1.5 is used to correct for the relative concentration of radioactivity in the resuspendable soil fraction (EPA, 1977). A mass loading of 100 yg/m air is used for our analysis (EPA, 1977), This value may be revised downward based on site-specific data on particulate loading. The assumed inhalation rate of 0.96 m /hour is for a "standard man." External Radiation Pathway The relationship between soil contamination level and external radiation dose from photons was defined for each nuclide using conversion factors from Kocher (1980). Those factors are based on an infinite plane source geometry and do not account for the shielding effect of soil. Short-lived daughter products are included. External radiation dose was estimated based on only 8 hours/day indoors, where the exposure rate is reduced to 70% of the unshielded value. Screening levels were also corrected for radioactive decay so that the average external dose during the first year is 1 mrem. Screening levels for the external pathway are in pCi/cm . Areal concentrations (pCi/cm ) must be converted to mass concentrations (pCi/g) to compare to the ingestion and inhalation pathways. To make that conversion, the radionuclides are assumed to be uniformly distributed in a 10 cm depth of soil of density 1.5 g/cm « Combinations of Pathways and Radionuclides To limit the total dcse from all pathways, the screening level (SL) must be less than the lowest single-pathway concentration. That is,
where C- 9 , C ., C + are the partial screening levels for ingestion, inhalation, and external pathways, respectively. Screening levels are derived for each radionuclide alone. When more than one radionuclide is present, the residual levels would have to be further reduced for the total dose commitment to be less than 1 mrem. Appropriate cleanup would be decided after concentrations of radionuclides in a particular area are determined. 304
RESULTS AND DISCUSSION Screening Levels Assuming Immediate Release of Area Table 1 shows screening levels for 13 radionuclides associated with D&D activities. The screening levels for each nuclide and pathway are presented which correspond to 1 mrem/year. The values are based on immediate homesteading of the site. Significant exposure pathways (those that affect the screening level) are listed in the last column of Table 1. Several of the assumptions used in the calculation of screening levels are unrealistically conservative in order to ensure that even implausible conditions remain bounded by the analysis. The more significant of those assumptions are (1) the radionuclide distribution in the soil is an infinite plane at the ground surface; (2) all food is obtained from the contaminated site; (3) the chemical form if the radionuclides is the most physiologically hazardous of those described by the ICRP; and (4) the radionuclide concentration in surface soil is not reduced by radioactive decay, leaching, or resuspension during the 50-year exposure period. TABLE 1. Screening Levels Assuming Immediate Release of Area Screening Level (pCi/g) Immediate Release
100 Year Control Period Corrected for Decay
Nuclide Co-60
0.03
Sr-90
0.1
Corrected for . Decay and Weathering
20000
20000
1. 13
1.
MO
13
Cs-134
0.05
Cs-137
0.1
1
1
Ce-144
2
a
a
Pb-210
0.2
6
6
Ra-226
0.05
0.05
0.05
Th-230
2
2
U-235
0.3
0.3
0.3
U-238
2
2
3
Pu-238
2
4
10
Pu-239
2
2
20
Am-241
1
1
1
^10
10
Screening level exceeds specific activity of Ce-144 Weathering is assumed to affect the inhalation pathway only
305
The assumptions used in the external dose rate calculation are important since this pathway is significant for over half of the nuclides considered. An infinite plane source distribution is unrealistic, but it is reliably conservative. In realty, concentrations of radionuclides in aged surface deposits decrease exponentially with depth. That depth distribution can be described by the ""^laxation length," which is the depth required for the radionuclide concentration to decrease by 1/e. Depth distributions of fallout nuclides at the INEL show relaxation lengths of about 2 cm for Cs-137 and Pu-239 and 3-4 cm for Sr-90 (DOE 1979). For Cs-137, the calculated exposure rate at 1 meter above the ground is about 2 times greater when a plane source is assumed than when the source is distributed with a relaxation length of 2 cm (Beck, 1972). The external pathway is not significant for Sr-90 and Pu-239. Some contaminated soil areas at the INEL are subsurface, having resulted from leaking underground pipes. The soil layers above that contamination may significantly reduce the external exposure rate. Sitespecific calculations must be done once the depth profile of a particular contaminated area is characterized. The area of contamination affects the screening levels in that the ability of individuals to raise all of their food decreases with a decrease in area. The external dose rate and total resuspendable activity also decrease with decreasing area. The chemical form of the radionuclide may make an order of magnitude or more difference in the screening level, depending on solubility and retention in the body. The effect of radioactive decay and weathering processes on the screening level is discussed below. Screening Levels Assuming a Period of Institutional Control If we assume that institutional control of the site is maintained for 100 years after D&D, screening levels are higher for the shorter-lived nuclides, due to radioactive decay during the control period. Those screening levels, corrected for radioactive decay during 100 years, are shown in Table 2. During the control period, weathering processes such as resuspension, dispersion, and leaching out of the upper soil layers could also reduce the availability of radionuclides, thereby increasing the allowable screening level. The resuspension rate constant A is defined as the fraction of surface radioactivity removed per unit time. In a recent review of resuspension models, Healy (1980) presented data from a number^of experiments in which the resuspension rate constants ranged from 10" to 10" s" . Using meteorological data from the INEL and an empirical relationship from Healy {1980), the resuspension rate constant weighted for wind speed and frequency is 1 x 10" 9 s"1 Migration of radionuclides into deeper soil layers can also reduce the availability of radionuclides in surface soil. An estimate of the downward migration of surface deposits of radioactivity at the INEL was made from observations on fallout plutonium (DOE 1979). Plutonium should be one of the least mobile nuclides and is therefore a conservative example. The downward migration rate out of the top 1 cm has been somewhat more than 1% 306
per year. One percent/year, or 3 x 1O~ s" , was used in this paper. That value is relatively insignificant compared to the assumed resuspension rate constant. Potential for migration in or out of the entire root zone would be analyzed for a particular site once the depth profile of contamination has been established. However, in this paper, removal from the root zone is assumed to be negligible during the 100 year control period. Screening levels for those nuclides which are limited by inhalation dose can be corrected for removal of resuspendable radioactivity with time. Although resuspension only operates on the top few mm of soil at one time, this is the layer of concern for the inhalation pathway. Correction for weathering (resuspension and downward migration) effectively reduces the significance of the inhalation pathway when release of the site is delayed. The screening level does not change if inhalation is not an important pathway. Table 2 shows that correction for weathering can increase the screening level by up to an order of magnitude. It is assumed that resuspension and migration from the top 1 cm of soil during a 100 year period do not affect the dose for the external and ingestion pathways. TABLE 2. Screening Levels Assuming a Period of Institutional Control Screening Level (pCi/g) Immediate Release Nuclide
100 Year Control Period Corrected for Corrected for . Decay Decay and Weathering 20000
20000
Co-60
0.03
Sr-90
0.1
Cs-134
0.05
Cs-137
0,1
1
1
Ce-144
2
a
a
Pb-210
0.2
6
6
Ra-226
0.05
0.05
0.05
Th-230
2
2
U-235
0.3
0.3
0.3
U-238
2
2
3
Pu-238
2
4
10
Pu-239
2
2
20
Am-241
1
1
1
1. .10 1 3
1. MO 1 3
10
Screening level exceeds specific activity of Ce-144 Weathering is assumed to affect the inhalation pathway only
307
Effect of Daughter Product Ingrowth on Screening Levels Daughter product ingrowth is an important consideration in calculating the screening levels for members of decay chains. This issue has only been partially covered in our analysis. Our external exposure pathway considers short-lived decay products of Ra-226, U-235, and U-238 (Table 1 ) , which are in secular equilibrium with the parent nuclides in less than a year. The daughter products can contribute the majority of the dose. For example, Pb-214 and Bi-214 contribute 99% of the external dose from Ra-226. Over the long term, daughter products which we have not included may dominate. For example, Th-230 decays to Ra-226, which grows into secular equilibrium with Th-230 according to the relationship: A(Ra)=A(Th)(l-expXt), where A is the decay constant for Ra-226. In one hundred years, Ra-226 and daughters are present at only 4% of the Th-230 activity. Because of the lower screening level of Ra-226, its presence at that percentage still reduces the Th screening level by a factor of two below that shown in Table 2. Ultimately, the screening level for Th-230 would be limited (by Ra daughters) to 40 times less than that based on the hazard of Th-230 alone. This example illustrates the importance of defining the time period of interest. We have not yet considered inhalation dose from Rn-222. That dose may reduce the U-238 series nuclides screening levels even further. Comparison of Screening Levels with Other Soil Contamination Guidelines Standards for environmental contamination with radioactivity are normally based on limiting dose to humans. Those dose standards have been translated into maximum permissible concentrations for air and water. No analagous concentration limits exist for soil, but in certain circumstances, soil concentration guidelines have been derived. Our screening levels are compared below with those guidelines reviewed by Mueller et al. (1981) which are appropriate for uncontrolled areas. 40 CFR 190 includes uranium processing sites terion is not given, but detection capability and Ra-226 (^1 pCi/g). The 0.05 PCi/g.
a screening level for residual Ra-226 from inactive which is 5 pCi/g to a depth of 5 cm. The dose crithe level is partially based on field instrument ability to separate contamination from natural Ra-226 screening level derived in this paper is
The other soil guidelines reviewed here are for environmental contamination with transuranic nuclides. The State of Colorado adopted a Pu-239 limit of 2 dpm/g (0.9 pCi/g) in the top 6 mm of soil for protection of construction workers in home building. The basis for that limit, including the dose criterion, is not available for review. Healy (1974) recommended interim standards for plutonium in soils which were believed to be safe at any location, based on unrestricted use of the land and a "reasonable worst case." For any 1 cm layer of soil those recommended levels are: (1) 200 pCi/g in the less than 100 ym diameter soil fraction; and (2) 400 pCi/g in the total soil sample, as long as the first criterion is not exceeded. The U.S. Environmental Protection Agency has proposed an interim screening level
308
for transuranium elements in-the environment (EPA, 1977). The proposed screening level is 0.2 yCi/m , or 12 pCi/g for a depth of 1 cm and a soil density of 1.5 g/cm . The dose criteria for those transuranic screening levels are different than the one used in this paper. When normalized to 1 mrem weighted CDE, the soil concentration proposed by Healy is 0.4-0.8 pCi/g and by the EPA is 1.6 pCi/g. The screening level for transuranics derived in this paper is 1-2 pCi/g, based on immediate release of the site for unrestricted use, and 1-20 pCi/g based on a 100 year control period. Measurements An important practical consideration is whether or not screening levels can be measured. Available field survey instruments used to measure gross alpha, beta, or gamma radiation are impractical for use in differentiating the screening level concentrations above background. Furthermore, they cannot distinguish individual radionuclides, which is how the screening levels are expressed. In-situ gamma spectrometry can probably be used to measure contamination equivalent to 1 mrem/year for some nuclides (Bruns, 1982). However, interpretation of the spectral data will require a thorough understanding of the source distribution in the soil profile. Soil sampling and laboratory analysis may be the only way to measure screening levels of some radionuclides. The Radiological and Environmental Sciences Laboratory at the INEL estimates detection limits for soil samples of 0.04 pCi/g for specific gamma, 0.004 pCi/g for Pu and Am, and 0.09 pCi/g for Sr-90 (DOE 1981). These detection limits appear adequate to measure screening levels, especially if a period of institutional control is assumed (Table 2 ) . However, those types of analyses can be very costly, and a large number of samples may be required to adequately characterize an area. Screening levels which are comparable to ambient concentrations will be difficult to detect by any method. Those nuclides include Cs-137 and Sr-90, which have ambient fallout concentrations 8 times and 4 times higher than the screening levels, respectively. Screening levels for nuclides in the U-238 series are in the range of background concentrations. Ambient fallout concentrations of transuranics at the INEL are far below the screening levels derived in this paper. Summary Screening levels are derived which correspond to 1 mrem/year to an individual using a conservative pathways analysis. Sites having screening levels less than those presented in this paper should be able to be released for unrestricted use. The screening levels presented here are conservative lower-limit concentrations which are intended to be revised upwards with knowledge of the particular site. Contaminated areas of the INEL are presently being characterized as to the radionuclide composition, concentrations, size, and contamination depth profile. Those details will be used to do a more realistic pathways analysis, although based on a similar method309
ology as presented in this paper. The appropriateness of a 1 mrem/year dose criterion for all nuclides may have to be re-evaluated due to the feasibility of measuring the very low concentrations in the presence of background. REFERENCES Beck, H. L. et al. 1972. In Situ (Ge(Li) and Nal(Tl) Gamma-Ray Spectrometry, Report HASL-258, U.S. Atomic Energy Commission, New York. Bruns, L. E. 1982. "Capability of Field Instrumentation to Measure Radionuclide Limits," Nuclear Technology, Vol. 58, pp. 154-169. DOE. 1979. Environmental and Other Evaluations of Alternatives for LongTerm Management of Buried INEL Transuranic Wasted IDO-1008b, Department of Energy, Idaho Operations Office, Idaho Falls, Idaho. DOE.
1981. 1980 Environmental Monitoring Program Report for the Idaho National" Engineering Laboratory Site, IDO-IZQBZ, Department of Lnergy, Idaho Operations Office, Idaho Falls, Idaho.
EPA (U.S. Environmental Protection Agency). 1977. Proposed Guidance on Dose Limits for Persons Exposed to Transuranium Elements in the Genera I Environment, EPA-b2U/4-77/OI6. Healy, J. W. 1974. A Proposed Interim Standard for Plutonium in Soils. Informal Report LA-54 83-Mb, Los Alamos National Laboratory, Los Alamos,
m. Healy, J. W. 1980. "Review of Resuspension Models," In: Transuranic Elements in the Environment (Wayne C. Hanson, ed.), National Technical Information Service, Springfield, Va., pp. 209-235. Hine, R. E. and J. A. Chapin. January 1982. Decontamination and Decommission ing, Long Range Plan, Idaho National Engineering Laboratory. rfrF.erna! Technical Report PR-W-79-OO5, Rev. 2, EG&G Idaho, Inc., Idaho Falls, Idaho, Kocher, D. C. 1980. "Dose-rate Conversion Factors for External Exposure to Photon and Electron Radiation from Radionuclides Occurring in Routine Releases from Nuclear Fuel Cycle Facilities," Health Physics Vol. 38, pp. 543-621. Miller, C W. et al. 1980. Recommendations Concerning Models and Parameters Best Suited to Breeder Reactor Environmental Radiological Assessments, ORNL-5529, Oak Ridge National Laboratory, Oak Ridge, Tennessee. Mueller, M. A. et al. 1981. Review of Soil Contamination Guidance, PNL3866, Pacific Northwest Laboratory, Richland, Washington. Ng, Y. C. 1982. "A Review of Transfer Factors for Assessing the Dose from Radionuclides in Agricultural Products for Environmental Effects," Nuclear Safety, Vol. 23, No. 1, pp. 57-71. 310
6D
THE TEXAS PANHANDLE SOIL-CROP-BEEF FOOD CHAIN FOR URANIUM: A DYNAMIC MODEL VALIDATED BY EXPERIMENTAL DATA
W. J. Wenzel, K. M. Wall work-Barber, J. C. Rodgers Los Alamos National Laboratory Los Alamos, New Mexico and A. F. Gallegos New Mexico Highlands University Las Vegas, New Mexico
ABSTRACT Long-term simulations of uranium transport in the soil crop-beef food chain were performed using the BIOTRAN model. Experimental data means from an extensive Pantex beef cattle study are presented. Experimental data were used to validate the computer model. Measurements of uranium in air, soil, water, range grasses, feed, and cattle tissues are compared to simulated uranium output values in these matrices when the BIOTRAN model was set at the measured soil and air values. The simulations agreed well with experimental data even though metabolic details for ruminants and uraniuii chemical form in the environment remain to be studied. INTRODUCTION The BIOTRAN model was originally developed at the Los Alamos National Laboratory to simulate cesium, uranium, and plutonium biological transport in the Southwest (Gallegos 1980). At present, three large subroutines can be used to simultaneously integrate daily and year]y environmental transport simulations. Figure 1 shows the existing BIOTRAN model components and those under development. BIOTRAN simulations have been used to assess radioactive waste burial alternatives (Walter et al. 1981, Hansen and Rodgers 1982) and for long-term assessments of radionuclides in the environment following, postulated accidents (Wenzel and Gallegos 1982). A major purpose of this study is to compare the BIOTRAN simulation of uranium transport in the human food chain for the Texas Panhandle region to actual measurements of uranium in soil, water, air, forage, feed, and cattle tissues taken in that region. BIOTRAN simulation details, analysis methods, and results can be found i.i the following reports: Wenzel and Gallegos 1982, Wenzel et al. 1982, and Buhl et al.
311
RUMINANT SUBROUTINE RUMINANT SIMULATION USES FORAGE AND FEED SIMULATED IN MAIN PROGRAM
AQUATIC SUBROUTINE (UNDER DEVELOPMENT)
I
MAIN PROGRAM
FORMAN SUBROUTINE
PLANT SIMULATION TWENTY-TWO PLANT TYPE; GRASSES. FORBS, SHRUBS. AND TREES
FOREST CHARACTERIZATION BIOMASS. AGE. COVER
ABIOTIC PROCESSES RAINFALL. TEMPERATURE SOIL, HYDROLOGY
SEDIMENT. PLANKTON. PISH. AND AQUATIC TRANSPORT IN LAKES
FORCUT SUBROUTINE FOREST (TIMBER) MANGEMENT
ATMOSPHERE - ACUTE AND CHRONIC RELEASES RADIOACTIVITY INPUT AND LOSSES HUMAN POSE SUBROUTINE (UNDER DEVELOPMENT) FISH. MEAT. MILK. AGRICULTURAL PRODUCTS. AND INHALATION AS DYNAMIC INTAKE
Figure 1. 8IOTRAN Model Components 1982. Another purpose of this study is to identify the type of experimental data needed to clarify the transport and uptake of uranium. Figure 2 depicts the agricultural food chain simulated. Two measured uranium concentration values, water and soil, were assigned as input to the BIOTRAN model in addition to site specific conditions pertaining to meteorology, hydrology, farming, and animal management. Air, forage, feed, and cattle tissue uranium concentrations were generated as output. These simulated concentrations are compared to uranium measurements taken in the same compartments on the Pantex Plant site near Amariilo, Texas. Each major environmental compartment, water, soil, air, forage, feed, and beef cattle will be dicussed separately and simulation values compared to experimental measurements.
MATER Three water samples were taken; these were the Pantex Plant pasture water trough, the Texas Tech Feedlot tap, and the Bushland Feedlot tap. The values for whole water (dissolved and suspended) ranged from 6.1 to 6.9 fCi/m£ for natural uranium. The 6.1 fCi/mfc value was chosen as input to the BIOTRAN model for the variable determining contaminant
312
IAIR
!
I
[RANGE [ GRASS \
\ \
t
FEED CROPS
-BEEF CATTLE —
FOOD CROPS
MAN
/
WATKR
Figure 2. BIOTRAN Agricultural Food Chain Pathways
I intake via water (WATCON). During the simulations, drinking water for cattle and irrigation water applied to the simulated crops contained a constant 6.1 fCi/m£. This is depicted in Table 1 where the measured values are given for the three samples and assigned as input for two different BIOTRAN simulation years. TABLE 1. Experimental Data Means and BIOTRAN Simulation Output for Water, Soil, Air, Forage, and Feed SAMPLING DATA
SAMPLE TYPE
WATER (ICl/ml) SOIL (fCi/g Dry) AIR U C l / V )
PANTEX PASTURE
BIOTRAN SIMULATION OUTPUT
TEXAS TECH BUSHLAND FEEDLOT FEEDLOT
6.9
6.1
6.5
1 YR. OLD2 YR. OLD CATTLE CATTLE
-
2000 .02
.02
RANGE GRASS « C l / g Dry) WINTER WHEAT
-
FEED (fCl/g Dry)
260
313
150
80
79
14
22
" j
II
SOIL Thirty resuspendable (top 2 cm layer) and thirty 25-cm core soil samples were taken from the Pantex Plant pasture while it was grazed by beef cattle. No statistically significant uranium concentration difference was found between the resuspendable and 25-cm core soil samples. Therefore, for the simulation, the soil concentration in the resuspendable soil layer and the next 25-cm layer wdre set at the same value, 2000 fCi/g of natural uranium. To further assure that the initial simulation conditions were set properly, the movement of uranium from the resuspendable layer down to the next layer (25-cm layer), variable SLOSS, was set at zero. This assured that the concentration of uranium in the resuspension layer would be the same as that in the next (or rooting) layer during the simulations. The variable determining the rate of loss of water to soils below the rooting depth (FLODN) was set at 0.005 to simulate the water holding character of the Panhandle Pullman soil (Unger 1981, Wenzel and Gallegos 1982). AIR Onsite NE perimeter Pantex Plant air samplers measured 33 ± 32 (mean ±1 standard deviation) pg/m 3 total uranium (Buhl 1982). Usirg a conversion 6'f 0.677 nCi natural uranium/g total uranium, this converts to 0.022 fCi/m 3 annual average air concentration of uranium in air near the Pantex Plant pasture. This is within an order of magnitude of another and less precise estimate based on dust mass loading and measured soil concentration. The annual average dust mass loading at the Amarillo, Texas, Airport was 57 yg/m3, which gives 0.11 fCi/m"3 based on 2000 fCi/g dry soil if the dust is assumed derived entirely from soil. Simulation with 2000 fCi/g natural soil uranium gave soil resuspension values in air about an order of magnitude lower than the 0.02 fCi/m3 value. The rate constant in the resuspension equation in BIOTRAN was increased by a factor of e to stabilize air simulation values near 0.02 fCi/m3. This is shown in Figure 3 and Table 1. Saltation and resuspension are shown in Figure 3 to illustrate the mass difference between the two transport mechanisms. Since particle size data were not available, no particle enrichment factor was used for the uranium in air simulation. FORAGE AND FEED The four plant types simulated using BIOTRAN included grain sorghum, alfalfa, winter wheat, and warm season (WS) grasses. The simulated crops were irrigated with water with uranium concentration set at 6.1 fCi/mJl. Figure 4 illustrates 25 years of plant growth. The variation year by year in maximum biomass in g/m3 is caused by stochastic weather changes from the site specific input conditions in the main BIOTRAN program (primarily temperature, rainfall, and insolation). Biomass for alfalfa should be multiplied by three because three cuttings per year are obtained in the Panhandle region. Simulated
314
AIR URANIUM
ltf
RESUSPENSIOM - SALTATION
ie?J\
S ltf
10" 10
15
20
25
TIME.YEARS
Figure 3. Uranium in Air from Resuspension and Saltation Processes Simulated by 8I0TRAN
BIOMASS 3000 SORGHUM WS GRASSES
2500-
WINTER WHBAT
2000-
1500
1000-
500-
r: 0 1 2 3 4 5 0 7 S 910111213141510171819202122232425
TIME.YEARS
Figure 4. Biomass for Grain Sorghum, Warm Season (WS) Grasses, A l f a l f a , and Winter Wheat Simulated by 3I0T3AN
315
productivity for these and other crops was found to agree well with literature values (Wenzel and Gallegos 1982). Figure 5 illustrates the maximum concentration in each crop each year in the live above ground biomass (LAGB) in fCi/g (dry).
CROP AND RANGE URANIUM 80
OQ O
60
40-
20-
0-
SORGHUM WS GRASSES ALFALFA. WINTER WHEAT
-ao0 1 2 3 4 5 6 7 8 0
10111213141516171819202122233425
TIME. YEARS
Figure 5. Uranium Concentration in Crops and Range Grasses Simulated by BIOTRAN Measured uranium values for the feedlot feed mixes were 260 and 150 fCi/g (dry) at the Texas Tech and Bushland Feedlots, respectively. This is about an order of magnitude greater than the concentration in the milo and alfalfa feed fractions simulated by BIOTRAN from sources entirely in the soil and irrigation water. The difference is due to the fact that there were measurable amounts of uranium carried over in the phosphate derived from phosphate fertilizer in feed supplements added to the feed (Reid 1977, Wenzel and Gallegos 1982). The measured and simulated concentration in range grass agreed closely. BEEF CATTLE Figure 6 shows the forage and feed schedule for the BIOTRAN cattle simulation. Cattle are generally pastured on winter wheat in early
316
1
0.8o
0.6-
<
0.4-
OS
w o
GRAZE WINTER WHEAT
GRAZE NATIVE CRASSES
>
0.2-
0
J I F I M I A I M I J ' J I A I S I 0 IN I D
O
1 -
Q W W
0.80.60.40.2-
GRAZE NATIVE CRASSES GRAZE WINTER WHEAT
LAST YEARS ALFALFA
ROLLED M1L0
>
)
1 F 1 M
I A 1M 1 J 1 J 1 A1 S 1 0 1 N 1 D
MONTH Figure 6, Forage and Feed Regime for BIOTRAN Beef Cattle Simulations spring and then placed on native grasses through the remainder of the year. Alfalfa is typically a supplemental feed while on pasture. Feedlot feed at the Texas Tech Feediot contained 78% milo and 12% alfalfa. The feed mix at the Bushlund feedlot contained 69% rolled corn, 16% cotton seed meal and hulls and 7.5% alfalfa. The Texas Tech feedlot milo came from the 1981 grain sorghum harvest near the NE perimeter of the Pantex Plant. Table 2 gives the experimental (see Wenzel et a!. 1982 for methods and analyses) and BIOTRAN simulation results for the Forage and Feed schedule in Figure 6. Figure 7 summarizes the BIOTRAN simulation. The ruminant metabolic model was developed from human models from ICRP 23, ICRP 30, and Wrenn 1977. The model developed by Wrenn appears to fit the data best. The uraniuTi retention functions used are:
317
TABLE 2. Experimental Data Means and BIOTRAN Simulation Output for Cattle Tissues AIR (.017 fCi/m 3 )
2 YR. OLD BEEF CATTLE (531Kg) LUNG (.062 fCi/g)
.06
- • KIDNEY , (.49 fCi/g)
.2
— BONE (.69 fCi/g)
02 001 024
GUT
r F,=.OO2 - • BLOOD
. LIVER (.028 fCi/g) MUSCLE (.017 fCi/g)
.63
URINE
SOIL (2000 fCi/g) INPUT
FECES
MILO (7.4 fCi/g) ALFALFA(42 fCi/g)
SAMPLING DATA SAMPLE TYPE
CATTLE TREATMENT
fCl/g «•!
LUNG LIVER
BIOTRAN SIMULATION OUTPUT
PANTEX TEXAS TECH •USHtAND FEEDLOT PASTURE FEEDLOT (630 Ib.) (WOO Ib.) (1000 th.)
PT
PC
.11 .IS
BONE
2.4
KIDNEY MUSCLE
.012 .OM
AT
. 14 .23
i rt. ou> CATTLE (SIP Ib.)
2 Y l . OLD CATTLE (1200 Ib.)
At
.36 .11 7.4
S.3
19.
1.4
.90
I.S
J»
.10
.071
.090
.096
.037
.062
.019
.028
.90
.69
1.2 .0069
.49 .017
Figure 7. Input and Output Values from BIOTRAN Ruminant Model
318
Bone = 0.2 e " 0 ' 6 9 3
t / 3 0 0
+ 0.02 e " 0 " 6 9 3
Kidney - 0.08 e " 0 ' 6 9 3
t / 1 5
Liver = 0.001 e " 0 ' 6 9 3
t/370
0 693
Muscle - 0.024 e " '
t/3700
°
t/370
°
Lung = Task Group Lung Model
An f1 = 0.002 scaled the tissue concentration the best. values for f1 f o r humans range from 0.2 to 0.002 (ICRP 30).
Literature
Cattle t i s s u e and organ weight to whole body weight (WB) r a t i o s were derived from other ruminant species (Meadows and Hakonson 1981) and from tissue weights (kidney) taken for t h i s study (Wenzel et a l . 1982): Bone = 0.055 (WB) Kidney = 0.0016 (WB) Liver = 0.011 (WB) Muscle = 0.45 (WB) Lung = 0.0014 (WB) CONCLUSIONS Comparison of the experimental data with BIOTRAN simulations has indicated a number of areas where additional information is needed. The metabolic model is p a r t i c u l a r l y s e n s i t i v e to the fi value used. Since a value from 0.2 to 0.002 is c u r r e n t l y in use, i t appears that metabolic experimentation delineating the absorption of p a r t i c u l a r chemical forms of uranium is needed before any f u r t h e r c l a r i f i c a t i o n of uranium k i n e t i c s can be done. Healy, Rodgers, and Wienke (1979) also found from t h e i r l i t e r a t u r e survey a need f o r i d e n t i f i c a t i o n of chemical form. Scripsick et a l . (1983) have found enhanced d i s s o l u t i o n rates in simulated lung f l u i d for "weathered" uranium in s o i l p a r t i c l e s . A major difference was also found between the Wrenn and ICRP 30 model f o r p a r t i t i o n i n g between bone and kidney. The experimental c a t t l e data c o n s i s t e n t l y shows bone concentrations to be greater than kidney concentrations (Table I I ) . Anatomical data for beef c a t t l e are another area f o r improvement. 8one, kidney, and muscle r a t i o s are considerably d i f f e r e n t from man. By far the most important area f o r c l a r i f i c a t i o n is the uranium chemical form under consideration. For t h i s study, intakes by water, airborne p a r t i c u l a t e s , range grasses, and feeds were combined. I t is probable that each intake pathway represents a combination of several d i f f e r e n t chemical forms each absorbed and d i s t r i b u t e d w i t h i n the ruminant in a c h a r a c t e r i s t i c fashion. This could account for the wide experimental v a r i a b i l i t y seen for f r
319
In general, the BIOTRAN model simulated the cattle tissue uranium concentrations quite well. Sealer errors (Kirchner, Whicker, and Otis 1982) caused by poorly known metabolic kinetics remain to be verified when additional metabolic data is available for f : and the retention functions. At present, a value of 0.002 for environmental uranium gut uptake in ruminants appears to give the best fit to experimental data. ACKNOWLEDGMENTS The authors appreciate the efforts of Kathy Oerouin for preparing the manuscript and George Trujillo for MAPPER graphics of tables and figures. REFERENCES Buhl, T., J. Dewart, T. Gunderson, D. Talley, J. Wenzel, R. Romero, J. Salazar, and D. Van Etten. 1982. Supplementary Documentation for an Environmental Impact Statement Regarding the Pantex Plant: Radiation Monitoring and Radiological Assessment of Routine Releases. LA-9445PNTX-C, Los Alamos National Laboratory, Los Alamos, New Mexico. Dunning, D. E., Jr., S. R. Bernard, P. J. Walsh, E. G. Killough, and J. C. Pleasant. 1977. Estimates of Internal Dose Equivalent to 22 Target Organs for Radionuclides Occurring in Routine Releases from Nuclear Fuel Cycle" Facilities, Vol. II. QRNL/NUREG/TM-190/U2, Oak Ridge National Laboratory. Gallegos, A. F., 8. J. Garcia, and C. M. Sutton. 1980. Documentation of TRU Biological Transport Models (BIOTRAN). LA-8213-MS, Los Alamos National Laboratory, Los Alamos, New Mexico. Hansen, W. R. and J. C. Rodgers. 1982. Risk Analyses for Shallow Land Burial and Greater Confinement of Alpha Contaminated Wastes. Los Alamos National Laboratory Report (in preparation), Los Alamos, New Mexico. Healy, J. W. , J. C. Rodgers, and C. L. Wienke. 1979. Interim Soil Limits for D&D Projects. LA-UR-79-1865-Rev., Los Alamos National Laboratory, Los Alamos, New Mexico. ICRP 23. 1975. Report of the Task Group on Reference Man. ICRP Report 23, Pergamon Press, Oxford, England. ICRP 30. 1979. Limits for Intakes of RadjonucTides by Workers. ICRP Publication 30, Part 1, Pergamon Press, Oxford, England.
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Kirchner, T. B., F. W. Whicker, and M. D. Otis. 1982. Pathway: A Simulation Model of Radionuclide Transport Through Agricultural i-ood Chains. Presented at Third International Conference on State-of-theArt in Ecological Modeling, Colorado State University, May 24-28, 1982. Fort Collins, Colorado. Meadows, S. D. and T. E. Hakonson. 1982. Contribution of Tissues to Body Mass in Elk. J. Wildl. Manage. 46(3): 838-841. Reid, D. F., W. M. Sackett, and R. F. Spalding. 1977. Uranium and Radium in Livestock Feed Supplements. Health Physics J. _32_: 535-540. Scripsick, R. C , K. C. Crist, M. I. Tiliery, and S. J. Rotherburg. 1983. Dissolution of Uranium Oxide Materials. Los Alamos National Laboratory Report (in preparation), Los Alamos, New Mexico. Unger, P. W. and F. B. Pringle. 1981. Pullman Soils: Distribution, Importance, Variability, and Management^ B-1372, Texas Agricultural Experiment Station, College Station, Texas and USDA, Agricultural Research Service and Soil Conservation Service. Walker, L. J., W. R. Hansen, 0. C. Nelson, G. Maestas, W. J. Wenzel, F. A Guevara, J. L. Warren, J. C. Rodgers, and J. M. Graf. 1981. Alternative Transuranic Waste Management Strategies at Los Alamos National Laboratory. LA-8982-MS, Los Alamos National Laboratory, Los Alamos, New Mexico. Wenzel, W. J. and A. F. Gal legos. 1982. Supplementary Documentation for an Environmental Impact Statement Regarding the Pantex Plant; Long-Term Radiological Risk Assessment for Postulated Accidents. LA9445-PNTX-O, Los Alamos National Laboratory, Los Ala-nos, New Mexico. Wenzel, W. J., K. M. Wall work-Barber, J. M. Horton, L. C. Hollis, E. S. Gladney, D. L. Mayfield, A. F. Gallegos, J. C. Rodgers, R. G. Thomas, and G. Trujillo. 1982. Supplementary Documentation for an Environmental Impact Statement Regarding the Pantex Plant: Agricultural Food Chain Radiological Assessment. LA-9445-PNTX-M, Los Alamos National Laboratory, Los Alamos, New Mexico. Wrenn, M. E., J. LoSasso, and P. W. Durbiri. 1977. A Metabolic Model for Uranium Metabolism in Man. In: Radioactivity Studies Progress Report. C00-3382-16, Institute of Environmental Medicine, New York University Medical Center, New York, NY.
6E RISK ASSESSMENT AS A MEANS OF EVALUATING ENVIRONMENTAL CONTROLS
Susan A. Dover Rockwell International Rocky Flats Plant, Co
ABSTRACT
A probabilistic risk assessment is a systematic method for estimating the potential for unwanted or negative consequences resulting from an activity. It can focus on any component, subsystem, or system, from the risk of a single part failure to risk of an unwanted hazardous material release for an entire plant. The end result of a risk assessment is a risk value which can be used to judge the adequacy of an environmental control system, or the cost-effectiveness of a proposed new addition to a system. The risk assessment process will also highlight the most effective environmental controls or new controls that have the greatest impact. The information necessary for a risk assessment is developed by performing Safety Analyses as defined in DOE Order 5481.1A. These safety analyses estimate the probability of human and equipment failures, and the corresponding consequences to the structure, workers and environment. As this paper illustrates, this information could easily be used to evaluate the effectiveness and future needs of environmental control systems. The Safety Analyses and Risk Assessments from the Rocky Flats Plant in Golden, Colorado will be used for this illustration.
INTRODUCTION A risk assessment can provide valuable insights and additional decision making c r i t e r i a in c e r t a i n types of environmental c o n t r o l evaluations. Risk concepts are most useful when comparing systems, evaluating systems under extreme conditions, and highlighting areas where system improvement is most effective. They can be based on either actuarial data or hypothetical conditions, however, assessments based on postulated scenarios are less exact and should not be used in applications that require absolute risk values. Risk concepts are extremely valuable in any e v a l u a t i o n t h a t involves comparisons. Additionally, the information required for a risk assessment is already generated and documented in the facility Safety Analysis Reports prepared in compliance with DOE Order 5481.1A. 323
RISK AND RISK ASSESSMENT
The concept of risk involves both the consequence of an undesirable event and the probability of that event occurring. In the dictionary (Stein, 1981), i t is defined as "exposure to the chance of injury or loss". Mathematically, risk is calculated by multiplying the consequences of an event by the probability that the event will occur (Henley, Kumamotto 1981). Any event can be analyzed for risk. Examples range from a decision, to a single part f a i l u r e , to an unwanted hazardous material release from a major processing f a c i l i t y . The consequences of a risky event can be measured in terms of dollar loss, grams released,deaths, or any combination appropriate for the goals of the assessment. A risk assessment is a systematic evaluation of the risk(e) of an event. Evaluation of a complex system such as a major processing f a c i l i t y , involves three basic steps: (1) definition of the types of accidents and releases that could occur, (2) quantification of the probability of occurrence, and (3) quantification of the consequences. (McCormick, 1981). The Department of Energy (DOE) requires t h a t , " . . . those DOE operations which involve hazards which are not routinely encountered and accepted in the course of everyday living by the vast majority of the public" be analyzed for risk and safety. These safety analyses document that "(1) p o t e n t i a l hazards are s y s t e m a t i c a l l y identified, and (2) potential consequences are analyzed ..." and are to be i n i t i a t e d in the e a r l i e s t phases of the l i f e cycle of the DOE operation (DOE 5481.1A, 1981). The information required by t h i s DOE order is contained in Safety Analysis Reports. Additional guidance for the preparation of these reports is provided in AL Order 5481.1 and the DOE Guidelines for Performing Safety Analysis A c t i v i t i e s and for Reviewing Safety Analyses (DOE/EP-0044, 1982). Environmental control systems are essential to the mitigation of hazards to the public and environment and are considered extensively in the Safety Analyses Reports (SAR's). Since the safety analyses are usually initiated in the design stage of a project, the information can be used to evaluate control systems before they are installed. The examples of the use of risk information in environmental decisions in t h i s paper focus on using information from the SAR's. However, these concepts could also be applied to environmental systems that control "accepted" hazards, such as emissions from coal fired power plants, which are not generally covered in the SAR's.
EXAMPLES FOR ENVIRONMENTAL CONTROL There a r e four b a s i c uses f o r r i s k i n f o r m a t i o n from Safety A n a l y s i s Reports i n s u p p o r t i n g environmental c o n t r o l e v a l u a t i o n s : (1) t o p o i n t out t h e s t r e n g t h s and w e a k n e s s e s of t h e c u r r e n t s y s t e m , ( 2 ) t o show s y s t e m adequacy, p a r t i c u l a r l y under extreme c o n d i t i o n s , (3) to measure compliance w i t h t h e "as. low a s r e a s o n a b l y a c h i e v a b l e " (ALARA) p o l i c y p a r t i c u l a r l y in 324
light of new technologies, and (4) as an additional c r i t e r i o n when comparing two or more proposed new systems. Examples from the Rocky Flats Plant ^a' Safety Analysis Reports and Final Environmental Impact Statement will be used to illustrate each of these applications. Highlight System Strengths and Weaknesses One of the f i r s t steps in a safety analysis is to analyze systems c r i t i c a l to the f a c i l i t y safety for hazards and failure modes. This a n a l y s i s must be documented in the Safety Analysis Report (AL Order 5481.1). At Rocky F l a t s , this documentation takes the form of failure modes and effects analyses (FMEA's) as i l l u s t r a t e d in Table 1. Every c r i t i c a l system is broken down into applicable subsystems. This example shows the inert gas system for glove boxes which is a subsystem of the heating, v e n t i l a t i n g , and air conditioning (HVAC) system. Each failure mode is separated and the effects of that failure are examined. In this example, there are no adverse effects resulting from any single failure. However, the analysis shows that if the fan shuts down, and the standby fan f a i l s , the inert system would no longer function. If there were no standby fan, the failure of a single fan would cause the subsystem to fail. The first stage of a failure effects of one failure at a time failure analysis. However, once would result in a system failure,
modes and effects analysis examines the and is technically only a single point it is determined that no single failure multliple failures may be examined.
For all practical purposes, the failure modes and effects analysis is a listing of the strengths and weaknesses of a system. It shows where redundancies exist and where they are needed. An FMEA will also point out the areas where redundancy exists but is unnecessary. When this analysis is performed on an environmental control system before it is installed, i t saves time and effort wasted in i n s t a l l i n g unnecessary equipment or r e t r o f i t t i n g a system with redundancies shown to be required after installation. The next step in the risk analysis focuses on the subsystem failures that could cause system failure and estimates the probability that these failures would occur. The end result is a l i s t of failures in a particular system that are ranked by probability of occurrence. This l i s t could be used to p r i o r i t i z e r e t r o f i t s , or highlight parts of the system that should be watched. If this ranked list is combined with a knowledge of the state-of-the-art equipment for that system, insight into the most effective areas for future research may be gained.
The Rocky Flats Plant, located in Golden, Colorado, is a Governmentowned and contractor-operated facility, which is part of a nationwide nuclear weapons production complex. 325
Table 1 Excerpt From The Rocky Flats Plant Heating, Ventilating, and Air Conditioning System Failure Modes and Effects Analysis Sub-System and Function
Failure Mode and Mechanism
Inert system; provides recirculating nitrogen atmosphere
Fan shuts down; Redundant fan will control failure automatically pick up load. Fan shutdown alarms occur in utilities control room.
Failure Detection and Compensation
FailureEffects and Effect Category
No effect on environment, personnel or building operations; IX
Loss of source nitrogen; high flow cutoff instrument failure
Nitrogen make-up No effect on environment, personnel system will lose or building operapressure. Low tions; IX pressure alarms sound. Oxygen level in inert system rises and alarm sounds when C>2 level reaches 4.5%. It will continue rising until problem is corrected.
Loss of negative pressure; instrument failure
Pressure interlocks Minor loss of operwith make-up nitro- ational capability; VIII gen preventing system from pressurizing. Low differential pressure alarms will sound. Control valve can be manually operated to restore negative pressure.
Loss of cooling; chiller failure
Temperature will slowly rise to equlibrium. Chiller alarms and hightemperature alarms sound.
326
No effect on environment, personnel or building operat ions; IX
Demonstrate System Adequacy In Extreme Conditions After the basic accidents or failure scenarios for a risk assessment are determined, the next step is to determine the probability that the sequence will occur. Usually, the scenario probability will vary with the consequence severity and the relationship between the two can be of great value to the decision maker. For example, the frequency-severity relationship for an aircraft crashing into a Rocky Flats Plant plutonium processing building was developed for the Rocky Flats Final Environmental Impact Statement (DOE-EIS 0064, 1980). The magnitude of the release varies with the postulated size of the airplane, the estimated damage it does to the building and the HVAC system, the probability of f i r e , and several other factors. These same factors also influence the probability. If only the consequences were considered, this analysis might imply that the control system were inadequate because it allows for a 100 gram release. However, if the facility remained as it is now for ten million years, a release resulting in 100 grams would only occur once. From a risk point of view, such a release is not considered credible. Generally a 10 per year , or 1/1,000,000 probability is used as a cutoff for credible accidents. Designing a building for a release that is postulated once in 10 million years may seem extreme yet, i t is comforting to know that the system is adequate for the worst a i r c r a f t crash accident postulated to occur once in 25,000 years. Other frequency severity analyses on other accident scenarios may indicate system weaknesses. The value of a frequency-severity analysis is that it shows the relationship between the estimated frequency of a particular type or class of accident and the magnitude of the consequences. The major drawback of this type of analysis is that it relies heavily on estimated p r o b a b i l i t i e s and postulated consequences for events that have never occurred or occur very rarely. Therefore, there is a great deal of uncertainty associated with the absolute value of the numbers. Another uncertainty is that some accident mechanism with a high probability was not considered in the analysis. These uncertainties may preclude the use of risk guidelines that set some definite inflexible limit such that a l l events will present a lower r i s k . However, the analysis also shows the types of situations, including accidents that have been postulated to occur, that should be considered in a design. 7t can also show areas where retrofit might be considered in the existing system.
Measuring Compliance With ALARA Frequency-severity analyses can be performed on stack effluents as well as extreme accident conditions. Figure 1 is a graph i l l u s t r a t i n g this type of analysis on one of the plutonium process buildings at Rocky F l a t s . The v e r t i c a l scale represents the magnitude of the release (in 327
T
I
i
i
I I i n
M
I r
FIGURE 1 | | LOG-NORMAL GRAPH OF 30 MONTHS OF BUILDING 776/777 TOTAL PLUTONIUM EFFLUENT RELEASE '
1.0 11/100 yr
0.5
o X 0.1 LU Q.
Q HI
«3 0.05 < ui -j
UJ
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s
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0.005
0.001 0.01
0.1
2
5
10
20
40
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PROBABILITY (% less than value)
328
99
99.9
this case measured in microcuries) and the horizontal scale represents the probability of the release occurring. The expected return periods (simply another method for expressing probability) have also been included! This particular analysis expresses the release consequences in uCi released over a 30-day period. However, the consequences could be converted to expected health effects to the public or to stack concentrations if those consequence measures were more useful. As with the extreme accidents, this frequency-severity analysis shows the relationship between the magnitude of the release and i t s expected frequency. However, this analysis is based on historical data rather than postulated scenarios and that l i m i t s the length of the return period to some extent. Compliance with the ALARA policy can be judged by looking at the worst expected release in some time period, for example every year or every ten years, and deciding if that release is acceptable.
Additional Criteria For Comparing New Systems If the worst release in the designated time period were not acceptable, or if a new technology were available, this same type of analysis could be used to compare the two systems. Figure 2 shows how this analysis might look. System 1 is equivalent to the system analyzed in Figure 1 and System 2 represents a hypothetical new system. System 2 is not as effective as System 1 for frequent r e l e a s e s , the systems are approximately equal for the worst annual release, and System 2 will reduce the worst release in ten years by 0.1 pCi. This translates into a risk increase of .004 uCi, 11 times per year, or .044 pCi per year. The same system will decrease the risk of release by 0.1 pCi once every 10 years corresponding to a risk decrease of 0.01 uCi/year (0.1 pCi X 1/10 years = .01 uCi/year). Overall, t h i s analysis shows that System 2 would not lead to a decrease in releases. If the systems were reversed, i.e. System 2 was the existing system and System 1 was the new system, the estimated risk reduction could be used as benefit in a cost-benefit analysis. Another advantage of using risk as a yardstick for comparing systems is that generally the same uncertainties are inherent in the analyses of both systems. So, while the uncertainties in the absolute risk values may be large, the uncertainties in the difference between the two systems will be much smaller. CURRENT USE OF RISK ASSESSMENT WITHIN DOE As a f i e l d of s t u d y , r i s k a s s e s s m e n t i s r e l a t i v e l y new, developed primarily by advanced a i r c r a f t and spacecraft testing programs in the early 1960's. Rockwell International, at the Rocky Flats Plant, recently
329
I I I I I T T » I I M FIGURE 2 1.0
FREQUENCY-SEVERITY GRAPH OF TWO DIFFERENT SYSTEMS
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O
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I
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I 0.01
0.005*-
0.001 0.01
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20
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PROBABILITY {% less than value) 330
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99.9
introduced the position of risk manager to help prioritize some of the planned engineering changes with the degree of risk reduction in mind. To my knowledge, the concept of risk has not been formally used to evaluate environmental control systems.
SUMMARY Risk concepts can provide some useful insights into system evaluation and decisions. Risk can be used as an additional criterion or parameter when comparing, describing, or evaluating a system. The preliminary analyses that provide the basis for a risk assessment show the strengths and weaknesses of a system and provide insights into the types of system failures that could reasonably be expected. It is not a panacea for d i f f i c u l t problems and must be used carefully. Risk guidelines are currently not incorporated into DOE regulations and probably will not be; primarily because of the inherent uncertainty in this type of analysis. However, information required to incorporate r i s k concepts into environmental control decisions is already available in the f a c i l i t y Safety Analysis Reports and Environmental Impact Statement and should be used where applicable to improve other evaluations and provide additonal information for decision making.
REFERENCES Henley, Ernest J., and Hiromitsu Kumamoto. 1981. R e l i a b i l i t y Engineering and Risk Assessment. P r e n t i c e - H a l l , I n c . , Englewood C l i f f s , New J e r s e y . McCormick, Norman J.. 1 9 8 1 . R e l i a b i l i t y and Risk A n a l y s i s . P r e s s , I n c . , New York, New York.
Academic
S t e i n , J e s s . 1980. The Random House College Dictionary Revised E d i t i o n . Random House, I n c . , New York, New York. U.S. Department of Energy. 1982. Guidelines f o r Performing Safety Analysis A c t i v i t i e s and for Reviewing Safety Analyses. DOE/EP-0044 UC-2 & 1 3 , U.S. D.O.E., Washington, D.C..
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6F
FEASIBILITY OF AN EIS FOLLOW-UP PROGRAM*
I.C. Nelson, R.E. Jaquish, D.G. Watson Pacific Northwest Laboratory Richland, Washington
ABSTRACT In the course of preparing environmental impact statements (EISs) and records of decisions (RODs) commitments to protect the environment or to mitigate environmental impacts are often made. A study was initiated to assess the feasibility of establishing a program for tracking commitments made in EISs and RODs. This paper presents progress and conclusions of the study including criteria for selecting commitments warranting tracking and a tracking system.
INTRODUCTION DOE and its predecessor agencies have made commitments in Environmental Impact Statements (EISs) and Records of Decision (RODs) to take certain actions for the protection and enhancement of the environment. Review of existing EISs and RODs reveals that the significance of the impacts for which commitments for mitigation have been made vary widely; from impacts that are potentially serious to those that are trivial. Most commitments for mitigating action appear warranted; however, some are vague or merely implied by such modifiers as "should" or "if possible". Because EISs and RODs are a matter of public record, the commitments are also a matter of public record and it could be construed that by publishing these commitments, DOE has led the public to believe that the commitments, even those conditionally stated, will be honored. If not honoreds DOE's creditibility in terms of environmental protection and enhancement could be jeopardized. As a consequence, the concept of an EIS Follow-up pilot program was put forth with the aim of determining the feasibility of establishing and implementing a program for tracking commitments made in EISs and RODs. Early consideration of an EIS Follow-up pilot program centered around the concern for DOE having made commitments in FEISs, how many, what kind, and what was the real or implied need to assure satisfaction of these com* The observations made in this paper are those of the contractors and are not necessarily supported by the DOE.
333
mitments. In the beginning, the investigators had little feeling for the number of EISs published, let alone kinds and numbers of commitments. The first order of business was to obtain a list of EISs that DOE had published. A consolidated list was finally obtained from a commercial source. A listing of about 120 EISs from DOE and associated predecessor agencies was prepared and scheduled for updating annually. From the list of EISs, a sample of eight were ultimately chosen to be examined in detail for commitments. The first lot of six EISs yielded a broad spectrum of commitments, some significant, others trivial, yet others that were simply not understandable. Emphasis at that time was on what the commitments were and how might a tracking system be established. It became clear that some screening needed to be performed to separate commitments that warranted tracking from those that did not. As a consequence, a set of criteria were developed against which the commitments were to be tested. These criteria were used 1 to examine FEISs and RODs for the Waste Isolation Pilot Plant and the Rocky Flats Plant Site.2 In the process of examining these two latter EISs, it was concluded that, although a large number of commitments for the protection and enhancement of the environment had been made, that very few were significant enough, in the investigator's opinion, to warrant tracking to assure satisfaction. The further into the process of commitment extraction and comparison with criteria, the more it became apparent that an overall appraisal for DOE of the degree of coverage and the balance of legal and moral obligation became necessary as a basis for determining feasibility. As a consequence, a proposed level for a DOE EIS Follow-up Program and alternatives with respect to that proposal" were developed based on findings to date. REGULATORY OR LEGAL BASIS How far an EIS Follow-up Program should be carried will be governed in part by what is required under the law and by DOE Orders. As a consequence, a review of the regulatory or legal basis for honoring commitments for mitigation of adverse environmental impacts was made. From the standpoint of regulatory or legal* obligation, there is some question whether commitments made in EISs prior to July 1, 1979 need be honored at all. A case in point came before the Fifth Circuit Court of Appeals in Atlanta, Georgia in Noe versus Metropolitan Atlanta Rapid Transit Authority (5th Cir. May 4, 1981), wherein it was concluded, among other things, that 42 U.S.C. § 4332 requires only the filing of an environmental impact statement, not compliance therewith. Thus it could be construed by extension that compliance with any commitments made was not mandated. It * Conclusions and recommendations are those of the authors and should not be construed as representing legal opinion.
334
may also be construed that the NEPA requires only disclosure of predicted impacts associated with a proposed major federal action and of alternatives to the proposed action and their predicted impacts, but does not in itself require commitments to mitigating actions. The FEIS in this case was published in the early seventies and long before the CEQ regulations that became effective July 1, 1979. As a consequence, the finding referred to in this case may be relevant to FEISs published by AEC and ERDA, but only of uncertain significance to those published by DOE and irrelevant after July 1, 1979. On the other hand, Regulations for Implementing the Procedural Provisions of the National Environmental Policy Act by the Council on Environmental Quality (CEQ) 49 CFR Parts 1500-1508 provides a regulatory basis for commitment to mitigatory action and monitoring to determine compliance and establishes the Record of Decision as a part of the NEPA process. These regulations became effective July 1, 1979. Some pertinent passages are as follows: "§ 1505.2 ...each agency shall prepare a concise public Record of Decision," "§ 1505.2(c) State whether all practicable means to avoid or minimize environmental harm from the alternative selected have been adopted, and if not, why they were not. A monitoring and enforcement program shall be adopted and summarized where applicable for any mitigation." "§ 1505.3 Mitigation [§ 1505.2(c)] and other conditions established in the environmental impact statement or during its review and committed as part of the decision shall be implemented by the lead agency or other appropriate consenting agency." Moreover, DOE Order 5440.1A of October 20, 1980, paragraph 5C describing responsibilities of Responsible Supervisory Officials states (9) "Specify in final environmental impact statements which mitigating measures they are committed to implement in connection with the proposed action, ..." (10) "Monitor and prepare, where appropriate, periodic reports on the status of post-final environmental impact statement program or project implementation, particularly with respect to any mitigating measures included in the program or project." In February 1981, DOE published its comprehensive Environmental Compliance Guide DOE/EV-0132 which has as the end point of Flow II provision for an action-entitled Conduct NEPA Follow-up which calls for "Verifying That Action is in Accordance with Commitments Made in FEIS through Mitigation Monitoring." On March 23, 1982, 46 FR 18026, the CEQ published a memorandum to agencies containing answers to 40 most asked questions on NEPA regulations. As stated by CEQ, "These answers, of course, do not impose any additional requirements beyond those of the NEPA regulations." The memorandum does, however, indicate CEQ's intent with respect to the use of the earlier regulations 40 CFR 1500-1508. For example, question 34(c) asks "What provisions should Records of Decision contain pertaining to mitigation and monitoring?" In part, the question is answered, "The Record of Decision
335
must identify the mitigation measures and monitoring enforcement programs that have been selected and plainly indicate that they are adopted as part of the agency's decision." Also, question 34(d), "What is the enforceability of a Record of Decision?" The answer to which is, in part, "... agencies will be held accountable for preparing Records of Decision that conform to the decisions actually made and for carrying out the actions set forth in the Records of Decision... A Record of Decision can be used to compel compliance with or execution of the mitigation measures identified therein." In summary, for commitments made prior to July 1, 1979, obligation for satisfaction would appear to have no legal basis since the NEPA itself apparently did not require fulfillment of commitments for mitigating environmental impacts. It would also appear that commitments for mitigation of environmental impacts and programs to monitor satisfaction of these commitments has had the force of federal regulation since July 1, 1979 and that DOE formally recognized an obligation in terms of these regulations through publication of guidelines for implementation on March 20, 1980. On the other hand, it could be construed that DOE has a moral obligation to affect mitigation to which it has committed itself for the reduction of adverse environmental impacts even in the absence of regulation. The extent to which an EIS Follow-up program might be established was analyzed considering both regulatory and moral bases for fulfilling commitments together with a sampling of the commitments to be found in various DOE EISs and RODs. It was concluded chat first priority should be given to providing guidance to the authors of EISs and RODs such that fewer and more clearly made commitments to mitigate significant environmental impacts would be made in the future. The development of a system for extracting commitments from existing EISs and RODs was given second priority and the development of a tracking system followed by implementation of the system was given third priority. A proposed level to which the program should be carried and alternatives to that proposal follows: A PROPOSED LEVEL FOR AN EIS FOLLOW-UP PROGRAM Based on the belief that circumstances will arise in the course of preparing EISs where mitigation of impacts will appear to be necessary, that commitments regarding mitigation of impacts will be made, and that once made are to be honored. It is, therefore, proposed I.
that DOE* establish guidelines for authors of its EISs and RODs to include commitments for mitigating environmental impacts therein in accordance with the following criteria:
* Actual guidelines for authors would be expected to be provided by the staff of the Office of Environmental Compliance. The criteria shown are offered as "strawmen". They resulted principally during deliberation on development of criteria for commitments warranting tracking. 336
1. Commitments must address specific impacts for which mitigating action is to be undertaken and must be clearly stated; e.g., Meets:
(Water quality impacts) "All drilling fluid, saltpile runoff, and wash water will be held within diked areas."
Fails:
"No runoff laden with high levels of dissolved salts will be discharged from the plant area" (objection -high levels is not specific). "Local, regional and national cultural resources will be preserved, and adverse impacts upon them minimized" (objection -- resource not identified nor the impacts that are to be minimized -- all sweeping statement). "Mitigation measures are available to minimize the consequence of coal pile runoff. Collection of runoff and conventional primary treatment could render impacts of substances minimal..." (objection -- vague, addresses what could be done but not. what would be done). "Install improved mitigation measures as they become available" (objection —-does not address specific significant impact or commitments).
2.
Commitments for mitigation of impacts are to be made only for those impacts that contain at least a presumption of significant detriment to the environment; e.g., Meets:
(Water quality) "No fly ash, bottom ash or dredged material from the Brayton Point Generating Station will be placed on or in salt marshes."
Fails:
"After construction, all temporary buildings will be removed" (objection — trivial in terms of the environment) . "Mitigation measures to be employed will address the problem of the effects of radionuclide contamination in radiocarbon dating of archeological sites" (objection — appears to be throw away commitment, basis for any effect on carbon dating not established). "Microbial processes in terrestrial and aquatic communities are being studied to determine primary productivity and to assess what impact these processes may have in radionuclide mobilization and demobilization" (objection — assuming the ultimate impact is radio-
337
logical in nature, fail to see significance; moreover, there is no basis for the presence of radionuclides to mobilize or demobilize). 3. Commitments must be such that failure to fulfill the commitments must produce a significant effect that is measurable or reasonably predictable; e.g.,
4.
Meets:
"Mined rock will be protected from runoff by a ditch." If the mined rock is salt, uncontrolled runoff could cause serious pollution of the environment.
Fails:
"The plant site, the mined rock pile, the evaporation pond and sewage treatment plant will be enclosed by fences restricting access to ponded water by wild life" (objection -- lack of basis that ponded water at WIPP would be harmful to wildlife, and as a consequence the existence of a significant environmental impact is not reasonably predictable).
Reasonable assurance exists that the commitment will be effective in reducing or mitigating the impact for which it was made. Meets:
"All effluents will be neutralized to a pH of 6-8 prior to discharge to the holding ponds or in the holding pond to obtain the optimum settling of insolubles and to provide a neutral effluent from the pond."
Fails:
"Specific ecological changes in vegetation patterns will be determined by assessment of species diversity" (objection -- studies have not shown environmental monitoring of this sort to be efficacious in determining changes in the environs of major plants -- it sounds good but would probably not yield desired result.
5. Commitments must be such that responsibility for their satisfaction is assignable (Headquarters, field office, contractor, or other agency); 6. Commitments must be such that their status can be determined by site visit, contact with field office, or contractor or by environmental measurement; that is they must be trackable. II.
that DOE instruct that commitments be clearly written and be collected in one place in the ROD and that a tracking system be developed and employed to follow the commitments to satisfaction, and
338
III.
that DOE* examine existing DOE RODs (EISs without RODs are excluded) according to the following criteria for commitments that should be tracked to satisfaction. These criteria have been developed based on detailed review of commitments in several recent DOE EISs and RODs. For the most part, the criteria are the same as those proposed for governing the making of a commitment in EISs and RODs. 1. The commitment as stated in the ROD and/or FEIS must be understandable. (Statements that appear to be commitments but are not understandable should be listed but not tracked. Depending on the significance of implications in the statement, resolution of uncertainty may be appropriate. Listing all commitments assures reviewers that all have been considered even if some are not tracked). 2. A presumption of significance of the commitment must exist. (This is a judgemental criterion but must be applied since there are many commitments of marginal or no significance that would not warrant tracking). 3. The status of a commitment must be verifiable; by site visit, contact with field office or contractor or by environmental measurement. (Commitments conditioned by such phrases as "if possible" should be listed but need not be included among those tracked). 4.
Failure to fulfill a commitment must produce an effect that is measurable or reasonably predictable.
5.
Responsibility for satisfaction of the commitment must be assignable. (Headquarters, field office, contractor)
6.
The commitment in question is not or other federal agencies. (Such but need not be tracked except to bility to track has been accepted
monitored by local, state, commitments should be listed confirm that the responsiby the appropriate agency).
7. Commitments that apply only to impacts measurable during the construction phase of a completed project should not be included in the tracking system. 8. Commitments that are made that are called for in conformance with non-NEPA oriented DOE Orders should not be tracked. (Such commitments should be listed but not tracked; e.g., a commitment to develop an emergency preparedness plan). * Examination of EISs and/or RODs for commitments warranting tracking and the tracking of these commitments would be expected to be performed within the Office of Operational Safety. 339
IV. that DOE develop a system to verify that the commitments determined to warrant tracking are tracked and that the commitment is satisfied. It is the intent of this proposed level of EIS follow-up to minimize the number of commitments made in future EISs and RODs, restricting them to only those associated with mitigation of truly significant environmental impacts and to track commitments of significance in published RODs to assure that the commitment has been honored. In a few instances, determination of the efficacy of commitments in mitigating significant environmental impacts may also be warranted; that is, did the action produce the effect that it was supposed to. By adhering to the previous criteria, it is expected that the number of commitments to be tracked would be small. The main attraction of this proposed program is that commitments in future EISs would be reduced markedly if criteria restricting commitments to those truly significant were applied. At present, only a small number of EISs have been followed by RODs and any tracking system developed for these and future commitments would likely be smaller and lower in cost than for a system scaled on present practice. A detraction of this proposal would be the uncertainty in the level of significance of commitments being ignored, and thus the degree to which DOE's credibility as an agency dedicated to the protection of the environment is jeopardized. In not providing follow-up on all EIS commitments, DOE is taking an unquantifiable risk of future litigation and of adverse public sentiment. ALTERNATIVES TO PROPOSED EIS FOLLOW-UP PROGRAM 1.
Implement the foregoing proposed EIS Follow-up Program and in addition screen DOE EISs not having RODs (pre July 1, 1979 EISs) for commitments warranting tracking to satisfaction. This alternative is one step beyond that believed to be required by existing regulations; thus, DOE commitments for which satisfaction is morally rather than legally required is proposed. The attraction of this alternative is that then all of the commitments made in DOE EISs and RODs will be subject to the screening and tracking system and DOE is in a position of less vulnerability of criticism. The detraction of this alternative is that it will add significantly to the load of EISs to be reviewed and to the additional commitments that would require tracking and thus ultimately to the cost of the program.
2.
Make no provision for EIS Follow-up; that is, continue as at present and ignore the fact that commitments have been made, do not follow-up to see whether commitments have been made and do not change DOE guidance relating to mitigation of environmental impacts. The attraction for this alternative is that it obviates the need for pursuing preparation of additional guidance to authors or cost of a tracking system.
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3.
Provide guidance for the future, but provide no follow-up for commitments incorporated in EISs or RODs published prior to effective application of the proposed guidance. The attraction for this alternative would be a minimal effort in commitment tracking; that from only significant commitments in the future. The detraction to this alternative is again some risk of litigation and loss of credibility, and also lack of accordance with regulations regarding commitments made in RODs.
4.
Provide guidance for the future but provide for no follow-up for commitments made either in the past or in the future. The attraction for this alternative would be that future commitments would be restricted to a few significant impacts and no tracking system would be developed. The detraction is again in the risk of litigation and loss of credibility and does not provide for compliance with regulations regarding commitments made in RODs at any future time.
CONCLUSIONS The proposed level of an EIS Follow-up program is believed to be feasible and that it can and should be implemented. Guidance to authors should result in fewer, but more important, commitments for mitigating adverse environmental impacts Selecting the significant commitments from RODs published since July 1, 1979 for tracking to satisfaction should result in conformance with regulations, orders, and the intent of the NEPA.
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REFERENCES 1.
Final Environmental Impact Statement — Waste I s o l a t i o n P i l o t P l a n t , Southeastern New Mexico, DOE/EIS-0026, October 1980. 2 volumes.
2.
Final Environmental Impact Statement — Rocky F l a t s P l a n t S i t e , Golden, J e f f e r s o n County, Colorado, DOE/EIS-0064, A p r i l 1980. 3 volumes.
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SESSION SEVEN ENVIRONMENTAL MONITORING II
7A
A METHODOLOGY FOR MAKING ENVIRONMENTAL AS LOW AS REASONABLY ACHIEVABLE (ALARA) DETERMINATIONS Roger C. Brown Dwayne R. Speer Health, Safety and Environment Rockwell Hanford Operations Richland, Washington 99352
ABSTRACT An overall evaluation concept for use in making differential cost-benefit analyses in environmental as low as reasonably achievable (ALARA) determinations is being implemented by Rockwell Hanford Operations. This evaluation includes consideration of seven categories: 1) capital costs; 2) operating costs; 3) state of the art; 4) safety; 5) accident or upset consequences; 6) reliability, operability, and maintainability; and 7) decommissionability. Appropriate weighting factors for each of these categories are under development so that ALARA determinations can be made by comparing "scores" of alternative proposals for facility design, operations, and upgrade. This method of evaluation circumvents the traditional basis or a stated monetary sum per person-rem of dose commitment. This alternative was generated by advice from legal counsel who advised against formally pursuing this avenue of approach to ALARA for environmental and occupational dose commitments.
BACKGROUND Existing radiation protection programs are based primarily on maintaining personnel exposure and dose to the general public below federally established limits (and in many cases below somewhat more restrictive control levels established by employers). In addition, because ALARA is not a totally new concept, many radiation protection programs are already aimed at reducing occupational and environmental doses to as low as reasonably achievable. However, the phrase "as low as reasonably achievable" contains the philosophical concept of "reasonable" which is difficult to quantify. Currently ALARA determinations
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are the result of a complex and largely subjective evaluation process. A system or mechanism for making ALARA determinations in an orderly manner using qualitatively and quantitatively defensible bases is needed. For this type of assessment, most existing ALARA evaluation practices are impractical or inadequate. Determination that a particular action is ALARA requires three distinct decisions: first, a decision must be made as to which situations or designs are to be analyzed for possible investment to reduce exposure; second, a decision must be made as to which alternative to recommend for each situation or design concept analyzed; and third, management must decide which alternative to implement. The first decision can be made by developing criteria and standards which pertain to situations requiring ALARA determinations. Then the analyses must be performed. It is at this point that ALARA analyses can be used for evaluation of each of the possible alternatives. Recommendations as to the apparent best course(s) of action can then be made, but final decisions must await review of the entire project or situation with respect to parameters other than simple dose reduction. A defensible point value approach is the recommended method of arriving at a final decision. INTRODUCTION A methodology for doing ALARA evaluations for releases of radioactive materials to the environment, based upon a point value system is presented herein. This methodology goes beyond the more traditional cost of reduction per man-rem saved approach and factors in nontraditionai costs such as the cost of construction in expected accidental deaths. Also considered is a projected impact to the environment from the postulated environmental release which is not directly related to calculated radiation doses to people. The point value system developed is based upon establishing categories and assessing the overall value of that category in terms of expected costs or expected benefits. Categories initially included are: Capital Cost Operating Cost Safety Accident Consequences State of the Art Reliability
346
• • 0
Operability Maintainability Decoimnissionability
This evaluation scheme is used to compare the total cost of doing a project with the projected total impact on the environment if the project is not done. Project costs include both monetary and industrial safety considerations and environmental impacts include calculated population doses and calculated impacts not directly related to human radiation doses. MONETARY CONSIDERATIONS Costs associated with a facility and operation exhibit a significant portion of the evaluations undertaken in any decision-making process whether by an individual or within a large, complex organization. This ALARA protocol relys heavily on costs because of the above reality. Cost assignments are made on the basis of best available estimates as follows: Total Cost = Capital Cost + (Annual Operating Cost x Service Life) A point assignment is made on the basis of one point per $1,000 (See Appendix for rationale). Capital cost estimates are required at an early stage of a design development. Costs of modifications, upgrades, retrofits, etc., can usually be estimated from previous experience if more definitive data are not available. Operating cost may be more difficult to determine, especially over the service life of the facility. Components of operating cost would include: • • • •
Maintenance Personnel Utilities Raw materials
Service life also presents a problem in that in many cases estimates will be required. Professional experts should be consulted for those cases where precedent or explicit statements of the service life are missing.
347
Decommissioning costs could also be very significant if the Initial design did not consider ease of decommissioning. However, since estimates of these costs are subject to large errors due to time and technology constraints, they are not addressed directly here but rather they are indirectly addressed later as a portion of the total cost. SAFETY CONSIDERATIONS Safety and accident consequences are quantified on a basis of 100,000 points per death regardless of whether that death can be attributed to an industrial accident or radiation dose (see Part 1 of the Appendix). Industrial Safety If a project being evaluated involves significant work with heavy equipment or hazardous materials, then an estimate of the expected probable deaths from the work should be made. For example, statistics kept by the National Safety Council (NSC81) show that 0.07 deaths can be expected from one million man-hours of construction activities. Therefore, if the project being evaluated will require 10,000 hours then statistically speaking 0.0007 deaths can be expected for a point assignment of 70 at 100,000 points per death. Environmental Radiation Safety Assessments for environmental safety are also made at the rate of 100,000 points per death however in this case the conversion is made to points per person-rem by applying a projected mortality rate of 8 x 10~5 early cancer deaths per person-rem for an average population as developed from epidemiological and statistical studies and reported by the National Research Council (Alexander 82, BEIR 80). For calculational convenience this is rounded up to a probability of 10~4 early cancer deaths per person-rem which is equivalent to 10 points per person-rem. Environmental radiation safety assessments are made using a two tiered approach. First, The direct calculated dose committments are determined using models such as OACRIN, GRONK, FOOD, and ARRRG (Houston 74, Soldat 74, Napier 80). Second, an assessment to estimate indirect detriments to the environment is made. This second assessment promotes an awareness of environmental detriment even though dose assessment models indicate a limited impact.
348
The assessment of indirect detriments is done in two steps. Determination of the maximum possible cumulative dose equivalent commitment to a hypothetical population is the first step. This calculation ignores any pathway analysis and assumes that all radioactive material is incorporated into the population until it has decayed. Equation (1) is a formulation of this parameter.
"e-xitdt / where: Ppc = population cumulative dose equivalent commitment k I
= summation for "k" radionuclides released
Ai
= total activity of radionuclide "i" released
Bi
= dose equivalent conversion factor for radionuclide "1"
qi
= maximum permissible body burden for radionuclide "i"
Xi
= the mean life of radionuclide "i"
t
= time
In the case of noble gases which do not become incorporated into the metabolic processes of the body, it will be necessary to perform analogous computations to evaluate the equivalent PQQ. Point assignments of 10 points per person-rem are made based on the above calculation. The second step involves evaluation of the effluents in order to infer a fraction of the above PQQ point assignment. This quasianalytical evaluation assesses the impact of the environmental detriment imposed by the releases postulated. Four assessments are performed: 1.
The "bubble concept" is utilized to determine the impact of these effluents on the total bubble from the site*.
See appendix for explanation of bubble concept and total decay concept.
349
".
A pathway factor is used to qualitatively assess the degree of difficulty for these effluents to be a part of a population dose equivalent.
3.
A time to decay factor which assesses the total time these materials can be expected to remain radioactive*.
4.
The concentration of the radionuclide in the effluent is compared to the concentration guides for unrestricted release (DOE 81).
Table 1 contains suggested values for these assessments.
Table 1. Effluent Evaluation Factors Parameter
Range
Factor Value
Bubble Concept Size(s)
Increase Stable Decrease
1 2 3
Pathway Availability(a)
High Medium Low
1 2 3
Time to Total Decay (p=0.95) T>1Q6 104
Time (T)
years <106 years years
1 2 3
Radionuclide Concentration Concentration*(c)
>Table >Table
I II and I II
1 2 3
*Note: Table I and Table II refer to the appropriate concentration guides in DOE 5480.1A. See appendix for explanation of bubble concept and total decay concept.
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The summation of the factor values is converted to a negative power of ten (i.e. 10-s+a+t+c) w n ich is then multiplied with P[)C point total. This point total is added to the total from the environmental dose assessment. For example, suppose a facility were to release 20,000 Ci of $H in an airborne effluent and the p DC point total is calculated to 2x10^ points (Eq 1). Suppose the bubble will be increased, pathway availability is high, total decay time is less than 10 4 years, and the concentration is at Table 1 concentration guides. The summation of the factor values in this case is 6; therefore the multiplier is 10~6, and as a result, 2000 points would be assessed in this situation. The summation of all of the points from the safety assessment will then be carried forward with cost assignments. ACCIDENT CONSIDERATIONS Safety analysis reviews provide accident scenarios and consequences in terms of frequency and severity. The appropriate parameters and values can then be utilized to assign point values for costs incurred. These costs would include monetary losses and expenditures and personnel and population exposure to radiation. In addition the public relations aspect must be considered along with other identified accident consequences. A quasi-analytical approach may serve to estimate accident consequences where safety analysis review data are not available and intangibles require incorporation. Table 2 indicates such a methodology. The sum of the factor values presented in Table 2 provides a scale factor which determines the point assignment. That assignment is made by considering the sum as a percentage value and multiplying it by the total of the cost and safety considerations developed to that point.
351
TABLE 2. Accident/Upset Conditions Consequence Evaluation Factors. Range
Factor Value
Accident Severity
High Medium Low
3 2 1
Accident Frequency
High Medium Low
3 2 1
Effectiveness of Mitigating Actions or Designs
High Med i um Low None
1 2 3 4
Public Relations Impact
High Low
2 1
Cleanup Costs
High Medium Low
3 2 1
Uncertainty of Measurements
High Low
2 1
Parameter
STATE OF THE ART State of the art is a generic term used in this context to denote the ability to change procedures, design, equipment, and facilities to reflect the current level of sophistication of technological improvements in all fields of endeavor related to waste management and fuels reprocessing. A failure to apply recognized state of the art in any aspect of operations involving materials or agents hazardous to health may cause the perpetrator to be held liable for damages in tort actions (Tressler 69, Hutton 66). Another purpose of a viable ALARA program is to force advances in effluents control techniques. On this basis, state of the art takes on added significance (BEIR 77). These considerations add impetus to improving facilities and operations with the objective of reducing detriments, both occupational and environmental. Status quo is not a viable ALARA option. Table 3 contains suggestions for assessing state of the art.
352
TABLE 3. State of the Art Factors. Degree of State of the Art
Factor Value
High Medium Low
0 5 10
The state of the art factor value selected is treated as a percentage to be multiplied by the sum of the costs and safety. This product is then considered the state of the art contribution to the total score. RELIABILITY - OPERABILITY - MAINTAINABILITY These three categories are grouped together because they are interrelated. The successful implementation of these categories into a facility design can reduce occupational and environmental detriments to values much less than those experienced with less than adequate reliability, operability, and maintainability. Again a qualitative approach is used in lieu of more detailed information. Table 4 lists suggested values. TABLE 4. Reliability-OperabilityMaintainability Factors. Parameter
Range
Factor values
Reliability (R)
High (Redundancy) Medium Low
0 5 10
Operability (0)
High (Human Factors) Medium Low
0 5 10
Maintainability (M)
High (East of Maintenance) Med i um Low
0 5 10
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The sum of the three factor values for a specific case are treated as a percentage of the total cost (Cj) plus safety (S) point score. The product of this multiplication is then added as the reliability, operability, and maintainability portion of the total point score (i.e., R+O+M = X%, *% x (Cy+S) = value). DEC0MMIS5I0NABILITY Decommissioning, or the ease of decommissioning, must be given proper consideration in order to do an ALARA analysis. Facilities which are not designed with decommissioning in mind can lead to unnecessary costs and exposure of personnel and the public to radiation in the future (Hinson 80, MacDonald 80). In evaluating the decommissioning of a facility or design we must consider inventories of radioactive materials, ease of decommissioning in terms of removal and/or renovation efforts, and the facility design flexibility to accomodate other activities at the end of the normal service life. Other works (Hinson 80, Manion 80, and MacDonald 80) provide basic considerations. Three basic alternatives are available for decommissioning (Manion 80). These are: 1. Permanent in situ protective storage of all or part of all residual inventory of radionuclides. 2. Temporary protective storage followed by removal of all hazardous residual radionuclides to an approved storage/burial facility and release of the site for unrestricted use. 3. Temporary protective storage followed by removal of most of the hazardous residual radionuclides to an approved storage/burial facility and release of the site for limited use as a controlled facility. Suggested values for qualitative decommissioning factors are presented in Table 5. These can be used in lieu of more precise analysis when not available. TABLE 5. Decommissionability Factor Values. Range
Remarks
Factor Value
High Medium Low
No Design Effort Some Design Effort Extensive Design Effort
0 5 10
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This factor value is treated as a percentage of the cost plus safety point score. The product of this operation is added as the decommissioning portion of the total score. EXAMPLE EVALUATION As an example let us define a hypothetical situation where a facility is processing radioactive materials and discharging water contaminated with strontium-90 at Table I concentrations at a rate of 1000 gallons (3.8 X 10 3 liters) per day. This water is discharged to a subsurface disposal facility and the plant is located in an arid region fifty miles from the nearest ground water user. We want to know if it would be reasonable to install a water treatment device to clean the water to drinking water quality. Intuitively we believe that it will not be cost effective since the discharge is below ground (no airborne transport pathway) and the nearest drinking water well is fifty miles away (insignificant ground water transport). Cost Considerations Let us assume the treatment plant will cost one million dollars to build and 50 thousand dollars per year to operate for a 20 year expected lifetime. Construction will require 10 thousand man-hours to complete and operation will require 50 hours per year of construction type activities. Therefore the total projected cost is given an evaluation point total of 2000 for monetary costs. Expected deaths from industrial accidents yield a point total of 77 for a total of 2077 (Table 6 ) . TABLE 6. Project Implementation Costs Item
Cost
Point value
Construction
$1,000 ,000 10,000 man hours
1000 70
Operation (20 years)
$1,000 ,000 1,000 man hours
1000 7 Total 2077
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Safety Considerations The environmental inpact of such a release would normallly be addressed in a facility Safety Analysis Report or similar documentation. That documentation should define the population and maximum individual dose commitments according to established dose models. Let us assume that the calculated population dose commitment is 2.3 person-rem. The assessment for indirect enviromental detriment is done by calculating the maximum possible population dose using equation 1 and modifying that dose according to the factors given in Table 1. The absolute maximum possible population dose for 20 years of plant operation at Table I concentrations is 1.73 x 10^ person-rem. That dose is modified according to Table 1 as follows: •
The total bubble size is expected to increase. value equals 1.
Factor
9
The pathway analysis indicates that the availability to the environment is low due to subsurface disposal and no near water wells. Factor value is 3.
•
Time to total decay is calculated to be 1920 years. Factor value is 2.
•
Release concentration is at the Table I value. value is 1.
Factor
These factor values total to 7 for a total calculated indirect environmental impact of 17.3 person-rem (1.73 x 10 8 x 1 x 10" 7 ). Therefore the total environmental impact calculated for this evaluation is approximately 19.6 person-rem. The point value for the potential benefit of cleaning up this effluent to drinking water quality is 196 points thus far. The safety considerations relative to industrial accidents during construction and operation were addressed earlier under the cost considerations heading.
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Accident Considerations Accident considerations for this situation being evaluated are such that operation of the plant without the proposed modification should be safer than with the modification. If we assume that the Safety Analysis Report indicates that the factors for accident conditions are as listed in Table 7, then we can also assign applicable values for the proposed change as also shown in Table 7. Table 7. Accident/Upset Factors for Exampl e Evaluation. Factor values for Parameter
Status Quo
Proposed Modification
Accident Severity
1
2
Accident Frequency
1
1
Effectiveness of Mitigating Actions or Designs
2
2
Public Relations Impact
1
1
Cleanup Costs
1
2
Uncertainty of Measurements
1
1
7
9
Totals
The totals shown in Table 7 will be converted to percentages and used to assign point values later. State of The Art We must assume that the existing effluent treatment system is not equivalent to the current level of sophistication available in the industry. Therefore we will assign a State of the Art factor value of 10 as indicated earlier in Table 3. The new equipment proposed is assumed to be the State of the Art therefore a factor value of zero is assigned.
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ReHabi 1 ity-Operabi 1 ity-Maintainabi 1 ity These categories are evaluated as shown in Table 8. The proposed new system is assumed to be more complex and therefore less reliable. Table 8. Reliability, Operability, and Maintainability Factors for Example Evaluation. Factor values for Parameter Status quo
Proposed modification
Reliability
0
5
Operability
0
0
Maintainability
5
5
Decommissionabi1ity Decommissioning of the existing disposal system is assumed to be via deactivation and insitu disposal with no significant removal and handling of radioactive materials. Therefore no design effort is required and a factor value of 0 is assigned. The new system is assumed to require some design effort for decommissioning and a factor value of 5 is assigned. Final Evaluation Finally all the factors and values discussed are drawn together as shown in Table 9 and the total point value for the status quo is considerably lower than the value for the proposed modification. Therefore the modification is not reasonable to persue. CONCLUSIONS The preceeding sections have outlined a methodology by which differential cost-benefit analyses in environmental ALARA determinations can be carried out in a systematic fashion. This methodology is not intended to be the final word on performing these analyses, but it is rather a mechanism by which to initiate ALARA considerations at an early stage of the design process. Continuing refinement with use should improve this methodology as more experience is gained.
358
Table <).
Example Evaluation. Factor values for
Parameter Status quo
Proposed Modification
Construction Costs
0
loon
Construciton Accidents
0
70
Operation Costs
0
1000
Operation Accidents
0
7
196
0
Environmental Impact Accident Consequences
14 (7%)
187 (9%)
Reliability
0 {0%)
104 (5%)
Operability
0 (0*)
0 (0%)
Maintainabi1ity
10 (5%)
104 (5%)
State of the Art
20 (10%)
Decommissionability Totals
0 (0%) 240
0 (0%) 104 (5%) 2576
ACKNOWLEDGEMENT This paper reflects part of the work currently being done at Rockwell Hanford Operations to develop methodologies for doing ALARA evaluations which optimizes benefits to radiation workers, the public, and the environment. That greater work is being headed by G. C. Strickland, K. M. Tominey and ourselves. REFERENCES Alexander, R. E., 1982, "Instructions in Radiation Risk Calculations",prepared for a continuing education course offered by the Health Physics Society at the 1982 Annual Meeting, Las Vegas, Nevada.
359
Committee on Biological Effects of Ionizing Radiations, 1977. Considerations of Health Benefit-Cost Analysis for Activities involving Ionizing Radiation Exposure and Alternatives, EPA 520/4-77-003, U.S. Environmental Protection Agency, Washington, D.C. Committee on Biological Effects of Ionizing Radiations, 1980. The Effects on Populations of Exposure to Low Levels of Ionizing Radiation: 1980, (BEIR III), National Research Council, National Academy Press, Washington, D.C. Department of Energy, 1981. Environmental Protection, Safety, and Health Protection Program for DOE Operations, DOE Order No. 5480.1a, Washington, D.C. Hinson, Charles S. and Thomas D. Murphy. 1980. "Occupational Exposure and ALARA," in Decontamination and Decommissioning of Nuclear Facilities, ed. by Marilyn M. Osterhout, Plenum Press, New York. Houston, J. R., D. L. Strenge, and E. C. Watson. 1974. DACRIN Computer Program for Calculating Organ Dose from Acute or Chronic Radionuclide Inhalation, BNWL-B-389, Pacific Northwest Laboratory, Richland, Washington. Hutton, Gerald L. 1966. Legal Considerations on Ionizing Radiation, Bannerstone House, Springfield Illinois. McDonald, Richard R. 1980. "Evaluating Decommissioning Costs for Nuclear Power Plants," in Decontamination and Decommissioning of Nuclear Facilities, ed. by Marilyn M. Osterhout, Plenum Press, New York. Manion, William J. and Thomas S. LaGuardia. 1980. Decommissioning Handbook, DOE/EV/10108-1, Nuclear Energy Services, Inc., Danbury, Connecticut. Napier, B. A., R. L. Roswell, W. E. Kennedy, Jr., and D. L. Strenge. 1980. ARRRG and FOOD - Computer Programs for Calculating Radiation Doses to Man from Radionuclides in the Environment, PNL-3180, Pacific Northwest Laboratory, Richland, Washington. National Safety Council. 1981. Accident Facts, Chicago, Illinois.
360
Soldat, Jo R., N. M. Robinson, and D. A. Baker. 1974. Models and Computer Codes for Evaluating Environmental Radiation Doses, BNWL-1754, Pacific Northwest Laboratory, Richland, Washington. Tressler, David L. 1969. "Tort Liability in Radiation Cases," in Medical Radiation Information for Litigation, DMRE 69-3, ed. by S. C. Bushong, J. L. Cox, V. P. Collins, J. 8. Neibel, and 6. 8. Murphy, U. S. Department of Health Education and Welfare, Rockviile, Maryland.
361
APPENDIX 1.
POINT VALUE RELATIONSHIPS Kathren 80 developed current values of a person-rem for use in ALARA evaluations and concluded that if dose reduction can be achieved at a cost of £$2000 per person-rem, then it should be done. Similarly he concluded that if it would cost >.$60,000 per person-rem, then it probably should not be done. We have chosen to use $10,000 per person rem since our evaluation techniques are somewhat conservative. We achieve this by assigning one point per $1,000 and 10 points per person-rem and, since the early cancer death risk from radiation is about lxlO~4 per person-rem, we assign 10^ points per death from nonradiation attributed causes.
2.
THE BUBBLE CONCEPT The U.S. Environmental Protection Agency (EPA) routinely applies a multiple source or bubble concept in determining the allowable effluent release levels in an area with more than one producer of such effluent. This concept is applied by calculating the madximum allowable ambient concentration of apollutant (and thereby, the areal environmental detriment), using that value to determine the maximum allowable release levels, and allocating individual relase limits to the various pollutant sources within the area or bubble. A modification of that concept can be applied to radioactive effluent releases from an area of multiple release points. That modification could be applied by calculating the amount of air or water that would be required to dilute the total quantity of airborne or waterborne effluents to DOE Order 5480.1, Chapter XI, Table II concentrations. That calculation could be based on total effluents released, total decayed effluents, or any subportion thereof. For example, historical airborne effluent release records indicate that the current worldwide inventory of 239pu and 9 "Sr which are the result of Hanford Site 200-Area airborne releses may be as high as 1.36 curies and 14.9 curies respectively. If all this material had been retained in the 14 3 local air, it would take 8 x 10 ft of air to dilute the 239 Pu and 9 x 10 + 9 ft 3 of air to dilute the 9 0 Sr to Table II concentrations. The 9®Sr bubble size is a small percentage of the total and does not affect the radius of the equivalent hemispherical bubble.
362
3.
TIME FOR TOTAL DECAY OF RADIOACTIVE MATERIAL The classical treatment of the decay of a radioactive material predicates exponential decay behavior with the passage of time and is valid for a large number of excited nuclei which will undergo radioactive decay (Evans, 1955). The exponential decay description does not lend itself to predictions of the time interval necessary for all of the excited nuclei of a radioactive species to decay. An equally rigorous approach assumes a population of excited nuclei with its inherent constant half-life and is formulated to determine the amount of time required for all members of the population to decay with a specified degree of confidence (Jackson, 1965). In the limiting condition of requiring absolute certainty that all members of the population have decayed, this approach predicts the passage of an infinite period of time which is in agreement with the traditional treatment. In the following paragraphs the latter approach is used to provide estimates at the 95% confidence level of the time required for decay of given radionu-lide populations. The number of atoms in a population of a given radionuclide is directly proportional to the product of the half-life and the activity. A unit activity of a radionuclide with a half-life of one year has fewer parent nuclides than does the same activity of a radionuclide with a half-life of one million years. Therefore, the time interval in terms of the number of half-lives will increase with increasing half-life for the same initial activity at the reference time. Radioactive materials discharged to the environment in effluents can be assumed to be radiologically detrimental during the period of time in which any radioactivity remains. An estimate of the period of time during which a given activity, A, (in microcuries) of radioactive material with half-life, T 1/2 (in years) will require to completely decay at a specified confidence level (i.e., probability that all atoms have decayed is 95%) can be made in the following manner (Jackson 65). This estimate is of use in comparing the duration of effects of discharge of different radionuclides to the environment. The number of atoms, N, present is: N = 1.68X1012AT 1/2
(1)
363
where: 1.68xlO12 = constant to convert uCi to disintegrations per second, year to seconds, and the conversion from decay constant to half-life. A = activity of the radionuclide in uCi. T 1/2 = half-life of the radionuclide in years. Since a particular radionuclide has either decayed or not decayed, the probability, Po, of all N atoms decaying in n half-lives is: Po = [l-(l/2)n]N
(2)
Equation 2 can be rewritten as: an Po = Nan [l-(l/2)n ]
(3)
An approximation of the natural logarithm on the right hand side of Equation 3:
Substituting into Equation 3 SlnPo = -N(l/2)"
(5)
and solving for n n = S,nN - S,n[Jln(^)]
(6)
fi,n2
If the probability of total decay is assumed to be 0.95, Equation 6 can be evaluated as: N = 44.9 + 1.44 Jin AT1/2
364
(7)
For example, 1.0 uCi of 2 3 9 Pu, T 1/2 = 24.390 years, will require 59.5 half-lives (1.45 x 10 6 years) to have a probability of 0.95 that all the plutonium has decayed. There will, of course be " 5 u daughters in existence at this time. Table 1 is a listing of some radionuciides of concern in effluents and the time required for all of the activity to decay with a probability of 0.95. TABLE 1. Time to Total Decay (p = 0.95) of 1.0 uCi of Various Radionculides. Radionuclide
3H 60co 85 K r
9°Sr 106 Ru 134 Cs 137 C s 154 Eu 155 Eu 239 P u 241Am
No. of Half-Lives 48.5 47.3 48,3 49.7 44.9 45.9 49.8 48.0 47.1 59.5 53.7
Total Time (Years) 6.0xl02 2.5xlO2 5.2xlO2 1.4x103 4.5X101 9.5X101 1.5xl03 4.1x103 2.2xlO2 1.5xlO6 2.3xlO4
The use of this concept is best illustrated by an example. Assume that a fuels reprocessing plant will emit an estimated 1x10^ Ci/yr of 85|
365
Jackson, Herbert L. 1965. "A Discussion of the Approximate Nature of the Exponential Decay Law as Applied to Radioactive Nuclides and Gamma-Ray Absorption," Health Physics (IIIK pp 179-183. Kathren, R. L, R. C. Yoder, A. E. Desrosiers, N. P. Nisick, E. E. Oscarson, 0. R. Mulhern, W. P. Howell, and D. A. Waite. 1980. A Guide to Reducing Radiation Exposure to As Low As ReasonabTy~Achievab1e (ALARA), D0E/EV/1830-T5, U.S. Department of Energy, Washington, D.C.
366
7B ALARA BEYOND DOLLARS PER PERSON-REM
D. Waite and W. Harper Office of Nuclear Waste Isolation Columbus, Ohio
ABSTRACT The trend in radiological assessments over the past decade is to extract more and more information from each bit of data. In the 1960's it was sufficient to calculate doses to the maximum individual and populations from exposure to critical nuclides in critical pathways. The 1970's brought the UNSCEAR and BEIR Reports and made the extension of calculated doses to predicted health effects fashionable. The 1980's may see data analysis extended still further to the use of an index of lost productivity. Advantages of these types of uses of dose calculations include the derivation of further insight into the system being analyzed and the attainment of a greater measure of comparability with impacts from agents other than radiological. The question is, whfre does the uncertainty in the analytical result overwhelm any possibility of its usefulness? This presentation addresses this question in the context of a performance analysis of a high-level nuclear waste repository. Estimates of uncertainty in dose calculations, health effects predictions and loss of productivity predictions are derived from examination of the open literature and the use of propagation of uncertainty techniques. The expected values and uncertainties are compounded as they would be in an ALARA analysis and the outcomes are analyzed. The advantages and disadvantages of truncating the analysis at each stage of extension are discussed.
INTRODUCTION The trend in radiological assessments over the past decade is to extract more and more information from each bit of data. In the 1960's it was sufficient to calculate doses to the maximum individual and populations from exposure to critical nuclides in critical pathways. The 1970's
367
broughtthe UNSCEAR^ U N S C E A R ' 1 9 7 2 ) and B E I R ( B E I R ' 1 9 8 ° ) reports and made the extension of calculated doses to predicted health effects fashionable. The 1980's may see data analyses extended still further to the use of an index of lost productivity. The objective of this paper is to estimate the price, if any, in uncertainty one must pay in ALARA (As Low As Reasonably Achievable) analyses for the additional insight one may derive from lost productivity versus dollars per person-rem values. The forementioned objective is pursued in this paper by first presenting the most common approach to ALARA calculations (dollars per person-rem) as a base case. This case is analyzed on the basis of (1) benefits to be derived from the approach, (2) common variations in application of the base case methodology, and (3) common criticisms of the technique. Potential alternatives for extending the dollars per person-rem method are examined to derive (1) the motivation for the particular extention and (2) the data needs introduced into the calculation by extending the base case. Each of the potential extension methods and the base case are analyzed to determine the uncertainty inherent in each approach. These analyses are derived from examination of the open literature and the use of propagation-of-uncertainty techniques. The expected values and uncertainties are compounded as they would be in an ALARA analysis and the outcomes are analyzed. Finally, the "information worth" at each extension step is evaluated, based on the additional insight derived and the uncertainty involved. BASE CASE The ICRP recommends that the acceptability of levels of occupational or population exposures to radiation proposed for a given activity should' be determined by a process of cost-benefit analysis(ICRP-22). The techniques of cost-benefit analysis can be highly sophisticated, but for the purposes of this discussion, the benefits of interest are broadly defined as including all benefits to society (including but not limited to, specific benefits to individuals and groups of individuals). Such benefits are likely to be both tangible (i.e., specifically identifiable and, theoretically at least, subject to quantification in terms of monetary or other units) and intangible (i.e., recognized as contributing to the satisfaction of human desires but not subject to formal quantification). Costs are broadly defined as the sum total of all negative aspects of a given operation, including the value of all goods and services used in constructing, operating, maintaining, inspecting, replacing and terminating the proposed activity and all other expenses, losses, liabilities and induced adverse effects (including damage to health and damage to the environment), whether tangible or intangible (including any effects contributing to human unhappiness.
368
Many of the factors enumerated above will commonly not require consideration in making the ALARA determination. In many instances, the question is whether or not the activity is being carried out at a sufficiently low level of exposure, and thus of detriment, so that any further reduction in exposure would not be considered to justify the incremental cost required to accomplish it. In making this determination, the costbenefit analysis shifts from a consideration of the total btnu^AX of the activity to a consideration of the dLL^eAnntial bznz^-ot that might be involved in requiring the activity to be carried out at one level of exposure rather than another. The net benefit, B, of an operation involving exposure to ionizing radiation can be expressed by the following equation: B-V-P-S-D where V is the gross value, P is the basic production cost, S is the cost, including social costs, of achieving a selected level of safety, and D is a term representing the total detriment. For a particular operation involving radiation exposure, we can consider V and P constant, and we then wish to make a choice of S and D with the aim of maximizing the benefit, B. This situation is reached when the sum, S + D, is a minimum. It is then convenient to apply two constraints to the analysis. The first of these is to require that the dose to any individual should be below the relevant maximum permissible dose or dose limit. This constraint limits the detriment to any single individual and this is necessary if benefits and detriments are to be summed over groups of individuals when the benefits and detriments may affect different groups. The second constraint is to require the benefit B to be positive, so that there is a net gain from the proposed operation. In the general process of cost-benefit analysis, the next stage is that of optimizing the benefit over a choice of different operations in which V and P are also variables. However, here we are concerned principally with the interpretation of the phrase, "as low as is readily achievable, economic and social considerations being taken into account", and thus deal only with the stage of optimizing the benefit for a given operation or group of operations. If the operations are changed in such a way as to change a variable reflecting the exposure of people to radiation, E, the optimum benefit is given when dE
= 0, assuming B is a concave function, or
dS _
dD
where V and P have been assumed constant for this purpose.
369
In more conventional terms, this condition is reached when the cost of achieving the next element of reduction in E is larger than the value of the resulting reduction in detriment. This point is clearly that at which "any further reduction in risk would not be considered to justify the effort required to accomplish it"uCRP-22, 1973). if the fully economic and social costs are adequately reflected in the numerical values of both S and D, this point also provides a definition of the phrase, "as low as is readily achievable, economic and social considerations being taken into account", since any further reductions in dose add economic or social penalties in excess of the reduction in detriment from the radiation. Figure 1 shows the relationships discussed above. If E is expressed in person-rems, then D can be expressed in monetary terms. The constraint to keep individual doses below the relevant maximum permissible dose or dose limit can be applied explicitly and doses close to this limit can be discouraged by artificially increasing the monetary equivalent applied to person-rems accumulated at high levels of individual dose. The second constraint, that of ensuring that B is positive, will usually be applied at the national level, while the process of optimization achieved by the process of differential cost-benefit analysis will be carried out either explicitly or implicitly on a case-by-case basis. In the control of occupational exposures, the differential process is generally implicit in day-to-day decisions( Katnren > 1980).
s and 0
Flo. 1 Differential cosi-benelit analysis. £ = a variable reflecting the exposure, possible in man-rems. S = total cost of achieving a value of £. D = total cost of detrimeni associated with a value of £
In applying this type of analysis to an ALARA program, the difficulty of determining how to measure costs and benefits becomes immediately obvious. Effects of radiation (leukemia, for example) are not measured with the same units as the benefits derived from the performance of the radiation work. The problem can take on enormous dimensions, including abstract considerations of societal benefits, losses, and obligations.
370
Some investigators have attempted to monetize the effects of radiation exposure to personnel, comparing the values thus obtained with the cost of reducing the exposure. The value of a person-rem reported in the literature ranges from a few tens of dollars to a few thousands of dollars, a range of about two orders of magnitude(Kathren, 1980). The most commonly used value in the U.S. is $1,000 per person-rem and is based on guidance given in Appendix I to 10 CFR 50, which deals with licensing of power reactors. Other values reported in the literature are somewhat lower, ranging from $10 to $980(Cohen, 1970; Dunster, 1970; Hedgron, 1970; ICRP-26, 1977; Otway, 1970; Sagans 1972). it is this dollars per personrem approach that has been adopted as the base case for this discussion. On first examination, using dollars as the common denominator for costs and benefits of incurred radiation exposure appears to neatly solve the problem of cost-benefit analysis. The approach is conceptually simple and straightforward, and can be applied readily and universally to radiation exposures in toto. Moreover, the concept is applicable to other hazards, including those from exposure to nonradiological hazards. And, there is a precedent, for monetization of risk and cost is commonly done in setting pay scales and in settling industrial accident and other claims. As explained in DOE's guide to ALARA implementation, in application, the user of the monetized cost-benefit comparison usually assumes that linear extrapolation of radiation effect versus dosage holds at all levels down to zero. Thus, it is possible to express the number of effects per unit dose or, more commonly, per cumulative dose. For example, at a level of 100 rem, a particular effect might occur, on the average, once in 10,000 exposures. By simple linear extrapolation, this effect would occur once in 1,000,000 exposures if the dose were only 1 rem. Thus, the risk (R) for this effect can be expressed as 1 per 10° person-rem, or expressed mathematically, R = P in which P is the number of effects, or the probability of producing an effect, and zdi represents the summation of the doses to a population of i people. Suppose the effect considered is an early death from cancer, and that risk is 1 per 10^ person-rem(BEIR, 1980), The next step is to assign a value to that loss of life. The value used for industrial accident fatalities or some other arbitrary value might be used; monetary values set for industrial fatalities are ordinarily in the range of $50,000 to a few hundred thousand dollars. Taking $250,000 as the value, with a risk of 1 per 10^ person-rem, produces a value of only $25 per person-rem, a value in consonance with those produced by other investigators. Even considering $1,000,000 as the cost or value of any effect of significance such as death, injury, or illness, and a total risk of 4 per 10^ person-rem (a fairly high value) gives a cost equivalent per person-rem of only $25. This is still within the range of values ($10-$980) put forth by those who have attempted to equate dose in terms of dollars.
371
The inadequacies of such an approach are many; they include the perception and acceptance of risk (by the individual and the society), the morality of knowingly exposing employees and the general public to an easily and cheaply controlled risk, and, more basically, the validity of the basic data and assumptions on which the monetization is based. Other questions of a fundamental nature might also be raised: Who benefits and who is at risk? If those who benefit and those who are at risk are not of the same group, how might the costs and benefits be equitably applied to both groups? If the cost is very high, why spend so much more to reduce dose? On the other hand, what values can one realistically place on a human life? Or on a disabling illness or injury? Or on aesthetics or morality? As has been mentioned above, attempts to quantify the rem (or personrem) in units of dollars commonly range from about $10 (i.e., "a few pounds sterling") to $1000's. Most of these studies were done in the early 1970's and hence the dollar values should be corrected for inflation. Assuming an inflation rate of 7% yr~' and assuming that the previous studies were based on 1970 dollars, the values approximately double in terms of 1980 dollars. Thus, the value of a rem could be considered to range from $20 to $2,000. In this range then the reduction is costbeneficial and the expenditure should be made. Conservatism would dictate that the highest value be used; thus, as a direct and positive statement of guidance, the following is often given: if dose reduction can be achieved at a cost of - $2,000 per person-rem, then it is cost-beneficial and should always be done. These types of decisions were much more difficult, if not impossible, to make in terms of dose calculations alone, thus the major benefit of the dollar per person-rem procedure. As mentioned by the ICRP, one may wish to improve on the fixed dollars per person-rem concept by allowing the exchange rate to vary as one gets closer or further away from an established dose limit. An excellent discussion of this slight modification of our base case methodology is to be found in a document prepared in the United Kingdom by the National Radiation Protection Board(Clark, 1981). This document is based on the premise that additional costs should be allocated to collective doses made up of annual individual doses which approach the dose limit, over and above that allocated to all collective doses. The procedure has two fundamental features. The first is the pecuniary cost associated with statistical health effects in the irradiated population, which are assessed only from the collective dose. This cost is related to lost output, health care, etc. It provides a minimum value of the cost to society of the health detriment and is the only cost assigned by the Board to collective doses made up of very low individual doses.
372
The second feature is that as individual doses approach the dose limits, there are many reasons including those of equity, risk aversion, and legislative requirements for allocating more resources to their reduction; this is reflected in the procedure by assigning a higher cost to collective doses made up of higher individual doses. The additional cost is a matter of judgment and the total given in the table below represents the present opinion of the Board. This system should lead to concentrating resources on the reduction of radiation exposure received by people at a higher level of individual dose, and is thus consistent with the emphasis given for some time in radiological protection to doses to "critical groups". It must be emphasized that, while the principles may have more general application, the numerical values are specific to the irradiation of members of the public from routine releases and depend on the dose level at which the limit for members of the public has been set. Using risk factors of 1.25 x 10" 2 S v 1 (1.25 x 10~ 4 renr 1 ) for induction of fatal cancers in the exposed population and 0.80 x 10~ 2 Sv"1 (0.80 x 10"4 rem-1) for the induction of hereditary effects in all subsequent generations, the approximate allocation to avert a single health or hereditary effect ranges from E 100,000 for very low individual doses, to E 2,500,000 for individual doses near the dose limit, as indicated in Table 1. Table 1. Cost of Collective Dose
Individual annual dose equivalent^-' band ( S v ) W
< 5.1O"5 5.1O"5 - 5-1O~U S.io"u - 5.1O'3
Percentage of annual dose equivalent limit for members of the public 5 mSv (500 mrem)
2000
<1 1 -
(c) Costv ' of unit collective dose equivalent made up of individual dose equivalents within the band (£/man Sv)
10
10000
10 - 1OO
50000
The sum of the weighted dose equivalents in a year from external radiation plus the sum of the weighted committed dose equivalents from intakes in the same year, as defined in ASP2. 1 Sv = 100 ran In 198O HZ prices. For other base years, the cost of unit collective dose equivalent should be adjusted, for exanple, in accordance with price indices derived from Index of Retail Prices (all items) published by the Departaen-S of 2aployment.
373
EXTENSION OF BASE CASE Beyond multiplying a calculated dose in person-rem times a dollars per person-rem conversion factor to monetize a radiological detriment (the foundation of the base case), two distinct extension techniques are considered here. The first entails the expansion of the normal scope of the health effect risk concept. In many cases where health effects estimates are made, "health effect" and "statistically possible fatality" can be interpreted to be synonyms. This was the assumption made in the base case description. A logical modification of this limited definition of health effect would be to include not only statistically possible fatalities, but nonfatal detriment as well. The difference between this methodology and that of the base case can be captured by either increasing the health effect per person-rem value or by increasing the dollar per person-rem value. The uncertainties associated with both options will be addressed in the next section of this paper. The second method of extending the base case methodology is to transform the health effects estimate, be it either fatalities alone or fatalities plus injuries as described earlier, to an index of lost productivity. One such index commonly used in other areas is person-days lost. This technique introduces the need to estimate the person-days lost per health effect and the value in dollars of each person-day lost. Advantages of these types of uses of dose calculations include the derivation of further insight into the system being analyzed and the attainment of a creater measure of comparability with impacts from agents other than radiological. The question is, where, if ever, does the uncertainty in the analytical result overwhelm any possibility of its usefulness? The next section of this paper addresses this question. INTRODUCTION TO KEY VARIABLES AND ASSUMPTIONS Analysis of the base case and its two suggested extensions require the development of statistical information. At present there is no universally agreed on data base for such items as dollars/person-rem. A literature search provided the estimates used in the quantification of variables used in this analysis. Data for the variables.of interest here show a wide spread in this published literature. Ideally, a unique probability distribution function would be obvious from the data base for each of these variables. However, present data for the variables of interest here are such that any of several probability distributions may be hypothesized; therefore, we have chosen probability distributions that we consider not only reasonable representations of the data, but also mathematically tractable. All calculations made in this discussion are derived from the following five variables:
374
X-, = X« = X3 = X4 = Xc =
fatal ities/person-rem $/person-rem lost life in years/fatality $/lost year incidence/fatality, where incidence = fatality + nonfatality
Before these are used to study the three cases, it is important that the development of our statistical data base and the assumptions underlying it be discussed. It is important to recognize the assumptions made, because the calculations that result rest on them. While different assumptions could have been made, it is hoped that the results are clear-cut and robust enough as not to be greatly impacted by changed assumptions. The first assumption is that the above five factors are approximated by lognormal probability distribution functions. A lognormal distribution is well suited for non-negative data skewed to the high side as the data of interest are. In addition, lognormal distributions are ideal for the propagation of uncertainty given the multiplicative nature of the calculations. Thus, we believe the lognormal is a reasonable choice for this analysis. Details on the mechanics of the propagation of uncertainty methodology utilized in the paper are given later. After deciding that lognormal distributions were reasonable, the parameters characterizing this distribution were estimated. Estimates were made of the 50th (the median) and 95th percentiles of the distribution, and are designated as X.5Q and Xgr, respectively. Since the log transform of the lognormal distribution results in the normal distribution to be used in later propagation of uncertainty techniques, the following calculations provide the needed mean and standard deviation of the normal distribution. v' = In (X i 5 0 ) °'
= 1n
~ 1n 1.645
(X
.50 } '
where u 1 = mean of normal distribution a1 = standard deviation of normal distribution In = natural logarithm Due to a lack of better information^ these five variables were assumed to be statistically independent. It is probable, however, that some positive correlations exist between these factors. The assumption of independence results in somewhat lower uncertainty in the results for the three cases
375
than would be indicated if positive correlation between variables were taken into account. DETERMINATION OF DISTRIBUTION FOR VARIABLES In this section, X # 5 Q and X.95, the median and 95th percentile of each lognormal distribution are given. With skewed data, it is common practice to work with medians. A median, which represents the point on the distribution where 50% falls below and 50% is above, is easier to relate to and work with since it is not so easily impacted by large outlying values as is the arithmetic mean of the raw data. For example, the U.S. Government reports median income as opposed to average income. Since no well-accepted data bases exist, evaluation of X 50 and X t g5 for the variables of interest here were often judgment calls of what looked reasonable from the scant data. In particular, X 9 5 was hard to estimate due to the lack of information in the tails of the data. Conservative choices were made when doubt existed; therefore, variability in these distributions may be somewhat overestimated resulting in increased estimates of uncertainty. For the variables introduced below, refer also to Table 2. Fatalities/person-rem, Xi, was derived from the BEIR 8o( B E I R > 1 9 8 °) and UNSCEARlUNSCEAR, 1972) reports as well as other sources(Alexander, 1982; Cohen, 1980: Cohen, 1979; GEIS, 1980). j n particular, Table E.I.I (page E.4) of GEIS, and Table V-4 (page 195) of BEIR 80 were extremely valuable. These sources led to a choice of X.so = 100 fatalities/10^ person-rem and X,95 = 500 fatalities/10^ person-rem. As emphasized in these documents, somewhat different assumptions underlie all of the presented numbers. The assumptions include such things as the form of the dose-response model (linear, linear-quadratic, etc.) and the choice of the projection model (absolute-risk vs. relative risk). Dollars/person-rem, X 2 . values( C o h e n ' 1 9 8 ° ; Cohen, 1970; Dunster, 1970; Hedgron, 1970; ICRP-22, 1973; ICRP-26, 1977; Otway, 1970; Sagan, 1972) found in the literature were found to be highly variable. Questions exist on how to adjust these for inflation. Since a conservatively high value, $1000/person-rem, published in 10 CFR 50, App I, is still used as the official value by NRC, the decision was made not to adjust the distribution of these for inflation. However, conservatism was used in the selection of the 95th percentile, such that inflated values would still be covered by our distribution. This resulted in X.50 = $200/person-rem, and X,g5 = $2000/person-rem. Lost years of life per fatality(Alexander, 1982; Cohen, 1980; Hodgson, 1982), X3, was used to evaluate the loss of productivity due to radiation induced cancer. This is a conservative estimate since it is assumed that all years lost would be equally productive regardless of the age at which the individual dies. For example, in terms of national productivity, ones retirement years are not usually as economically valuable
376
as ones working years. Estimates of X.50 = 17.32 years lost/fatality, and X 9 5 = 30 years lost/fatality, were obtained. Dollars per lost year of life, X4, was obtained through both sources of published salary surveys(Christensen, 1982), and through discussions with individuals in the personnel field. We believe that the use of a charge-out rate would be the best indicator for estimates of productivity. Typical conversions from salary to charge-out rate involve multiplying the former by roughly 2.5. The resulting estimates are X.50 = $73,500/year, and X.95 = $125,000/year. To be able to include non-fatal occurrences of cancer, a distribution was developed for the ratio of incidences of cancer to fatal cancers, X5. Incidence includes both fatalities and non-fatalities caused by radiation induced cancer. The distribution adopted here is based solely on Table V-15 of BEIR 80. No attempt was made to obtain separate distributions based on sex, even though such data are= presented in Table V-15. Estimates for the ratio are X.so = 1.77 and X.95 6. THEORY OF PROPAGATION OF UNCERTAINTY The appropriate method to use in the propagation of uncertainty depends on the following: (1) Model relating desired response to inputs (X], Xp, . . ., X5 in our case) (2) State of knowledge concerning the inputs. The mathematical model relating the response of interest to the input can be extremely complex and can be approximated by a long-running computer code. In cases such as these, techniques such as the adjoint sensitivity method or latin hypercube sampling may be desirable. For less complex computer models, Monte Carlo simulation techniques may be used to propagate input uncertainties through to the output (response). When an explicit equation relating the variables exists, it is often possible to use analytical techniques. The choice of the technique to be used depends on the form of the equation and our knowledge of the input variables (Harper, 1982). i n many cases, one can estimate the response mean and variance (uncertainty) by various forms of a Taylor's series expansion. For the models (equations) to be studied here, there is a simple relationship of the following form: Y = l\Lz . . . Z k where Y = desired result
377
Z-f = input or intermediate result, i = 1, 2 . . ., k Taking logarithm of both sides results in the following: Y1 = In Y = z In Zi = sZ,' where the prime denotes the log transformed variable. The Zj's 'ised here will all be lognormally distributed; therefore, Y' will be normally distributed. Since the Zi's are assumed to be statistically independent, My' = Z VZ\ and cf23yi = z a^z'
i
where u y ' = mean of In Y a2yi = variance of In Y y . = mean of In Z-j o?z = variance of In Z-j Thus, based on the given assumptions, the distribution of the response and its defining parameters can be calculated. The variability or uncertainty of the output is now obtainable and, if desired, any desired percentiles of its distribution can be calculated. ANALYSIS OF INFORMATION WORTH As previously stated, the question is, where does uncertainty in the result of the extended base case method overwhelm any possibility of its usefulness? There is no definitive measure to test this. The intent of this paper is to compare the uncertainty in the base case with the two suggested extensions. Given the uncertainty that exists in the base case, our proposed criterion will examine any increase/decrease in uncertainty as we move away from it. If large increases in uncertainty result as this progression occurs, then the additional insight gained by the extension is being overwhelmed by the noise (uncertainty) in new variables being introduced. In statistical terms, the uncertainty is represented by the variance; however, the variance (or its square root--the standard deviation) are not quantities easily related to. This is especially true for skewed distributions such as the lognormal. The criterion we propose for use in evaluating the "information worth" is the following uncertainty ratio:
378
Y.95 .5O where Y.95 = 95th percentile of the response Y
distribution, ^.50
=
^Oth percentile of the response distribu-
tion (median). This ratio increases as the variance increases. distribution, y
Indeed, for the lognormal
•y|jj- = exp (1.645 a 1 ) where a 1 is the standard deviation of In Y. In addition to being a direct function of the variance, this uncertainty ratio provides sound information that is intuitively appealing. Besides being concerned with a measure of central tendency (the median in our case), another concern is about the upper tail of the distribution (how bad can things get?). Provided the underlying assumptions are correct, 95% of the time we can expect to be within this multiplicative factor (the above ratio) of the median. When the uncertainty and thus this ratio increases, the spread of possible outcomes can become such that the information given by the median is swamped by the noice, i.e., the signal/noise ratio has diminished to a level that the output is useless. This measure is used for comparison between the base case and its two proposed alternatives. So besides a median comparison, an uncertainty comparison is given. COMPARISON OF ALTERNATIVES All three methods are normalized to person-rem so that the end result is in dollars. The necessary calculations are briefly described as well as the new variables that are created in the process. The major results are given in Tables 2 and 3. The intermediate variables created are as follows: W1 = $/fatality C = conversion factor for converting all incidences into fatalities Wp = nonfatalities/fatality The three cases to be considered are given below: Base case: Y-, = X-j W, Extension 1:
Y 2 = X-j W-j C
Extension 2:
Y3 = X] C X3 X4
where Y-j, Yp, Y 3 represent the results for the three cases.
379
For the base case and its first proposed extension, a new distribution must be derived. It represents dollars per fatality and will be represented by W ] . It is found by dividing dollars/person-rem, X2, by fatalities/ person-rem, X-]. Since X] and Xg are both lognormal, Wi will likewise be lognormally distributed. Its major parameters are seen in Table 2. The result for the base case, Y ] , is obtained through the following relationship:
Y! =
XT
H]
Normalizing to per person-rem results in the median of Y-j = $200. The proposed information worth ratio is approximately equal to 25.5. The distribution of Y] thus has a very wide spread ranging from a median of $200 to a 95th percentile of $5,095. Realizing this, one is immediately hit with the tremendous uncertainty in the base case alone. One could pause here and ask if the uncertainty in the base case rules out any possible worth in the result; however, the goal here is not to study the base case, but to compare it with possible extensions. The first extension to the base case involves expanding the detrimental effect of radiation induced cancers to include non-fatalities as well as fatalities. To do this, a conversion factor was developed to convert cancer incidences (both fatal and nonfatal) into fatalities. Then W ] , dollars/fatality, could be used to end with a result in dollars. It was assumed that a nonfatal occurrence of cancer was ^qual to 0.1 times the cost of a fatality. This conversion factor, C, is derived as follows: C = (1 + .1W2) where W2 = nonfatalities/fatality. The needed information for W2 is gotten easily from X5 (cancer incidences/ fatalities). The distribution for C is assumed to be well approximated by a lognormal distribution with parameters as indicated in Table 2. The distribution for this extension is thus seen to be the product of C with the base case. The resulting answers as seen in Table 3 are a median value of $215, and a ratio of the 95th to 50th percentiles = 25.9. Thus, based on the given assumptions, little additional uncertainty is added over the base case. The second extension involves an examination of lost time per radiation-induced cancer and the dollars per lost time. The same conversion factor, C, as used above, is again used here to incorporate nonfatalities. The resulting distribution is thus derived from the product of X ] , C, X3 and X4. The median result for this approach, $137, is lower than before. In aadition, the uncertainty ratio is only 6.1 as compared to the previous values of 25.5 and 25.9. Contrary to intuition, by examining lost time and its dollar value in terms of productivity, additional insight has been gained and yet a large reduction in uncertainty has resulted. (Thus, a large increase in information worth.) This can be
380
traced to the variable driving the uncertainty in the two previous results. That variable is X2, tne dollars per person-rem, and is derived from widely varying data. Most of these data are based on non-quantitative judgment calls, and may incorporate a fear of radiation. Table 2. Variables Used Median 95th Percentile fatalities/person-rem $/person-rem
1 x 10' 4 200
a1
y'
5 x 10' 4
-9.210
.9784
2000
5.298
1 .3997
lost life/fatality (in years)
17.32
30
2.852
.3339
$/lost years
73500
125000
11.205
.3228
1.77
6
.571
.7421
14.509
1 .7078
incidence/fatality
2 x 10'6
$/fatality
.77
5
-.261
1.1317
1.08
1.5
.074
.2014
nonfatalities/fatality conversion factor
3.316 x 10 7
Table 3. Results Normalized to Person-Rem 95th Percentile
Uncertainty Ratio
$200
$5,095
25.5
Extension 1
215
5,581
25.9
Extension 2
137
840
Median Base
6.1
CONCLUSIONS The previous discussion has been developed independent of the context in which the ALARA analysis is exercised. Since this type cf analysis, as applied to occupational exposures in an operating high-level nuclear waste repository, is of special interest to the authors, the conclusions to be drawn will be couched in terms of the occupational dose assessment
381
strategy developed for repositories. This strategy contains the following major steps: •
Development of a detailed work operations list
•
Performance of a dose assessment on an operation-by operation basis
•
Comparison of assessed dose from an operation to a screening dose limit
•
Performance of ALARA analysis
•
Comparison of total assessed dose to regulatory dose limit.
Since the ALARA analysis is preceded in this assessment strategy only by a calculation yielding an individual occupational dose, none of the ALARA approaches are precluded by the strategy itself. On the contrary, the dose calculation step would provide the base data set upon which ALARA, on the basis of any of the approaches would be founded. Furthermore, the ALARA analysis being only an additional screening mechanism in the strategy will not hinder the completion of the overall strategy, i.e., the combining of assessed doses from all operations and the subsequent comparison to dose limits. These facts established, one can comfortably establish a preferred ALARA method exclusively on the basis of the data shown in Tables 2 and 3. The uncertainty reflected by the 25.5, 25.9 and 6.1 values in Table 3 for the base case, extension 1 and extension 2, clearly indicate a preference should be shown to extension 2. Data shown in Table 2 indicate why the base case and extension are so close in uncertainty and why extension 2 is so clearly superior. Examination of the a 1 entries for the variables shown in Table 2 yields an indication of the disproportionate contribution that the dollars per person-rem variable makes to any result which it impacts. Having a value of 1.4 times the next most uncertain of the 5 original variables and 4 times the uncertainty of several others, incentive is given for avoiding any methodology which contains dollars per person-rem directly. When this avoidance is heeded in an ALARA method, 'as it is in extension 2, the reduction in uncertainty (from « 25 to - 6) is significant. Therefore, somewhat contrary to intuition, the further insight that was described early in this discussion can be derived from a process with actually less uncertainty than the simpler approach presently being used. One might ask why the introduction of monetary units into the calculation is so much more certain when achieved in terms of dollars per lost day as opposed to dollars per person-rem. The answer is related to two
382
major points. First, all dollar per person-rem estimates have been based on subjective judgment on the basis of little or no data. On the contrary, lost productivity cost data are collected routinely on extremely large populations under many diverse situations resulting in actuarial data of excellent quality. Some difficulty was encountered in this study in organizing a data base for the nonfatality components of extension 1 and 2. It is a commonly held perception by individuals familiar with such types of data that the data do exist but because of the lack of radiation protection utilization in the past, such data have not been organized and made widely availble in a useful form. Therefore, there should be no disincentive for adopting the extension 2 methodology and abandoning the present dollars per person-rem method which, relative to viable alternatives, provides little insight and much uncertainty.
383
REFERENCES Alexander, R. E. 1982 (June).
Instructions in Radiation Risk Calculations
BEIR. 1980. The Effects on Populations of Exposure to Low Levels of Ionizing Radiation, Committee on the Biological Effects of Ionizing Radiations, National Academy of Sciences, Washington, DC. Christensen, Ralph C. 1982. "Trends in Health Physics Salaries and Employment: 1980", Health Physics. 42: 107-117. Clark, M. J., Fleishman, A. B., and Webb, G. A. M. 1981. Optimization of the Radiological Protection of the Public, NRPB-R120. Code of Federal Regulations, Title 10, Part 50, Appendix I. Cohen, B. L. 1980. "Society's Valuation of Life Saving in Radiation Protection and Other Contexts". Health Physics. 38: 33-51. Cohen, B. L. 1980. "Perspective on Occupational Mortality Risks". Health Physics. 40: 703-724. Cohen, B. L., and Lee, I-Sing. Physics. 36: 707-722.
1979. "A Catalog of Risks". Health
Cohen, J. J. 1970. "Plowshare: New Challenge for the Health Physicist". Health Physics. 19:633. Dunster, H. J., and McLean, A. S. 1970. "The Use of Risk Estimates in Setting and Using Basic Radiation Protection Standards". Health Physics. 19:121. GEIS. 1980. Management of Commercially Generated Radioactive Waste, Vol. 2, Final Environmental Impact Statement. D0E/EIS-0046F. Harper, W. V., and Waite, D. A. 1982. Long-term Performance Objective/ Measure for Repositories. ONWI-398. Hedgron, A., and Lindell, B. Health Physics. 19:121.
1970. "PQR--A Special Way of Thinking".
Hodgson, T. A., and Rice, D. P. 1982. "Economic Impact of Cancer in the United States", Chapter 12 of Cancer Epidemiology and Prevention, Schoffenfeld, D., and Fraumeni, J. F. Jr., eds. W. B. Sanders Co., Philadelphia, Pennsylvania. ICRP-9. 1966. Recommendations of the International Commission on Radiological Protection. ICRP Publication 9. Pergamon Press, Oxford, England.
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ICRP-22. 1973. Implications of Commission Recommendation Jlv* Doses Be Kept as Low as Readily Achievable. ICRP Publicatior \ .^rgarnon Press, Oxford, England. ICRP-26. 1977. Recommendations of the Internationa ienw-ision on Radiological Protection. ICRP Publication 26. Perganibn Press, Oxford, England. Kathren, R. L , Selby, J. M., and Vallario, E. J. 1980. A Guide to Reducing Radiation Exposures to as Low as Reasonably Achievable T^ARA) D0E/EU/1830-T5. Battelle Pacific Northwest Laboratories, Richland, Washington. Otway, H. J., Burnham, J. B., and Sohrding, R. K. 1970. "Economic vs. Biological Risk as Reactor Design Criteria". IEEE Nuclear Science Symposium. Sagan, L. A. 1972. "Human Costs of Nuclear Power". Science^ 177: 487-493 UNSCEAR. 1972. Ionizing Radiation: Levels and Effects - Volume II: Effects. United Nations, New York.
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7C
IMPROVEMENTS TO ENVIRONMENTAL SURVEILLANCE AT THE RADIOACTIVE WASTE MANAGEMENT COMPLEX (RWMC) OF THE IDAHO NATIONAL ENGINEERING LABORATORY
K. S. Moor, S. K. Rope, G. B. Wiersma, T. H. Smith EG&G Idaho, Inc. Idaho National Engineering Laboratory Idaho Falls, Idaho
ABSTRACT Environmental monitoring has been conducted throughout the 30 years of RWMC operations at the INEL. Monitoring data collected during this time provide a record of the effect of the RWMC on the surrounding environment. A Monitoring Activities Review (MAR) is being conducted: (1) to review the methods used in the current program and upgrade them as required to state-of-the-art, and (2) to review the data accumulated to date in order to detect trends and determine whether revised sampling designs are needed. Results of the MAR will be used to modify the Program. Critical environmental pathways are being identified using a comprehensive pathways approach. The use of systems analysis, including simple pathway models based on first order kinetics, is an excellent way to design monitoring networks and analyze monitoring data. This allows investigators and administrators to consider interactions that may be occurring in the system, and provide guidance in determining the relative importance of collecting and analyzing various media. The desirability of achieving certain confidence levels versus cost can then be evaluated. This systematic approach is described in this paper.
387
INTRODUCTION The INEL and the RWMC The Idaho National Engineering Laboratory (INEL) was established in 1949 by the U.S. Atomic Energy Commission for the operation and testing of nuclear facilities, reactors and equipment. The 230,000-ha INEL is now the responsibility of the U.S. Department of Energy (DOE). The Radioactive Waste Management Complex (RWMC) was established near the southwestern corner of the INEL in 1952 as a controlled area for the management of solid radioactive waste. The RWMC is operated for the DOE by EG&G Idaho, Inc. The RWMC encompasses 58 ha. Buried waste is contained in the Subsurface Disposal Area (SDA). Waste stored above ground is in the Transuranic Storage Area (TSA), Figure 1. The closest population center is Idaho Falls, approximately 80 km east. Solid waste from national defense programs and research activities is stored or buried at the RWMC. The waste is of two types, transuranic (TRU) or low level, based upon the type of radioactive contamination. Between 1954 and 1970, approximately 110,000 cubic meters of low-level and TRU waste were buried in pits and trenches in the SDA. (Until 1964, the TRU waste was often intermixed with the low-level waste,) Some of the original waste containers have deteriorated, allowing some of the surrounding soil to become contaminated. Since 1970, an additional 57,000 cubic meters of low-level waste have been buried in the SDA. Since 1977, containers of low-level waste with high radiation levels have been placed in soil vaults. The vaults are unlined holes, augered to various depths, filled with drums, and covered with soil. Burial of TRU waste ended in 1970. Since 1970, nearly 57,000 cubic meters of TRU waste have been placed in interim storage on asphalt pads at the TSA, pending permanent disposal. The History of Environmental Surveillance at the RWMC Individual monitoring activities at the RWMC began in 1960, but were limited to efforts in support of operational activities. In 1973, a plan was developed for a comprehensive program. The program was implemented on a routine basis in 1975. Routine monitoring consists of air, surface water, and surface soil sampling for radionuclide concentrations; penetrating radiation surveys; and visual inspections. Subsequent monitoring activities have been added to investigate soil moisture exclusion and relative humidity conditions in stored waste enclosures. Improvements have been made in routine monitoring from 1975 through the present. In addition to monitoring by EG&G Idaho, Inc., the U.S. Geological Survey (USGS) routinely samples subsurface water from monitoring wells in and adjacent to the RWMC. The Radiological and Environmental Sciences Laboratory (RESL) of the DOE uses thermoluminescent dosimeters to routinely measure direct penetrating radiation at the RWMC boundary. 388
XXX
Waste Management Facility WMF-601 Transuranic Disposal Area
Transuranic Storage Area
Subsurface Disposal Area Soil vault area
Intermediate Level Transuranic Storage Facility INEL 2 0623 XXX
Figure 1. General layout of the RWMC. 389
Besides routine monitoring, special environmental studies are occasionally conducted at the RWMC. For example, five drilling projects have been conducted to measure the concentrations of radionuclides in subsurface beneath the RWMC. Also, the RESL has conducted several studies investigating potential radionuclide transport via biota at the RWMC. The results of routine monitoring are summarized in annual reports, (i.e., Janke and Zahn 1982), with detailed data being archived separately. Special studies are reported in individual technical reports. Results have indicated that the effects of the RWMC on the surrounding environment are very small and within the limits prescribed by pertinent regulations.
OVERVIEW OF PROGRAM IMPROVEMENTS The RWMC Environmental Surveillance Program is adequate to measure the effects of the RWMC on the environment, in accordance with DOE requirements. However, the Program is continually undergoing improvement. A major improvement effort which began in 1982 is being implemented. This paper describes several of those improvements, which are directed primarily at enhancing the use of monitoring data. Figure 2 depicts the flow of information in the improved program. The figure is modified from a similar figure in (Corley et al. 1981). In the modified program, monitoring data are first examined for validity and consistency ("data analysis" on Figure 2 ) , then entered into a data management system. The data are archived and printed out for reports as necessary. Statistical methods are used to identify patterns in the data over time or location. This activity is termed "trend analysis." Through analyses of data trends, projections are made of near-future concentrations of radionuclides at various locations. (The reliability of those projections is estimated.) A "simplified mathematical model" is used to project the transport of radionuclides to man through a set of pathways. The model uses data from monitoring activities, special studies, and other sources. Individual pathways resulting in important dose contributions are then evaluated further using "detailed mathematical models" as necessary. The results of all the activities just described are continually reviewed in an activity called "assessment of results." Regulatory requirements and benchmark radionuclide concentrations at the RWMC are included. From those assessments, revisions to the monitoring activities and special studies are formulated and implemented. Thus, the information flow completes the loop as shown in Figure 2. Only part of the activities shown have been implemented to date. These activities are discussed in the following sections starting with the Monitoring Activities Review (MAR) of monitoring design, procedures, and data. 390
Revisions^
Revisions y Monitoring activities
Special studies
J
L
Data analysis (subsection 12.1) Data archival (subsection 12.1)
Enter into data management system Do^ data cover" ^extended time'
Data reports
JYes Trend analysis (subsection 12.2) Short-term projections Simplified mathemetical model of pathways (subsection 12.3)
Data from other sources
No
Detailed mathematical model of important pathway (subsection 12.4)
Long-term projections
Assessment of results of monitoring data, trend analysis, & modeling (subsection 4.2)
J Figure 2.
Regulatory requirements and benchmark concentrations INEL 2 0627
Proposed n o w of information f o r the improved RWMC Environmental Surveillance Program
391
MONITORING ACTIVITIES REVIEW The objectives of the Monitoring Activities Review (MAR) include: (1) to improve the conduct of ongoing RWMC monitoring activities, and (2) to improve our understanding and use of the data. The expected products of the MAR are (1) improved monitoring procedures; (2) possible revisions in the locations, sampling frequencies, number of samples, and media selected for analysis; (3) improved justification for all the preceding items; (4) improved quality control; (5) improved understanding of the accuracy, trends, and utility of the data; (6) improved reporting of the data; and (7) enhanced use of the data to guide decisions. The scope of the review generally covered the following, for each monitoring activity: (1) objective of the monitoring activity; (2) sampling rationale (what to sample, where, how often, how many); (3) sample collection (how); (4) sample storage and transfer (how); (5) sample analysis (how and for what); (6) data generation, reduction, and evaluation; and (7) data reporting. Modeling and experimental studies not considered routine were excluded from the MAR. A team of management and technical experts reviewed the procedures and data in detail. The reviewers became familiar with available material concerning the various monitoring activities by briefly reviewing relevant portions of the RWMC Environmental Handbook, annual reports, and monitoring system reports where the data are achieved. The reviewers also evaluated recommended practices for monitoring, including portions of Corley et al. (1981), the ANL Generic Handbook on Environmental Monitoring for Low-Level Waste Disposal Sites, Health Physics Society (1980),' DOE Program Review Sheets, and other widely circulated guides for DOE monitoring. Some reviewers also visited other DOE sites early in the MAR to observe DOE's state-of-the-art monitoring methods. Those sites included the Hanford, Nevada, and Los Alamos sites. Based on the document review and visits, a list of concerns was drafted by the review team which included specific problems, identified and suspected, for each monitoring activity, Recommendations for improving the monitoring activities were prepared and are being reviewed.
SIMPLIFIED MODEL OF PATHWAYS A comprehensive, simplified model is being developed which describes potential transport pathways to man from the RWMC. This model is intended to be part of the information flow for the Program (Figure 2). The conceptual model for transport of radionuclides to offsite individuals is shown in Figure 3. The boxes represent environmental "compartments" which may contain radionuclides. For example, the surface soil above the burial ground may be defined as a compartment for which the model is designed to predict the concentration.
392
Offsite man or population
External exposure
Ingestion
Inhalation Offsite air Wet, dry deposition K1
a o
•a
Leaf fall weathering Offsite soil
o S Q. 07
Onsite air
K2
K1
Onsite surface soil
CD CM CM CVJ
c
cc
Irrigation
(a)
Buried waste
Unsaturated zone storage
Leaching
Snake River Plain Aquifer
(a)studied in more detail in "onsite" model. K1 = Resuspension K2 = Deposition INEL2 4130
Figure 3. Pathways diagram for transport of radioactivity from the RWMC to offsite individuals. 393
Transfer from one compartment to another is described by first-order rate constants, having the units of inverse time. Since transfer is described by a single rate constant, the model is called "simplified." This is in contrast to models which include a calculation of fundamental transport processes within the model itself. Arrows on Figure 3 represent "pathways," or possible routes between compartments. The processes which can result in transfer of radionuciides from one compartment to another are shown beside the pathway arrows. Several processes may act to move radionuclides along the same pathway. For example, migration and intrusion (plants, animals, or man) are separate processes which can move radionuclides from buried waste to surface soil, Figure 1. In the early stages of development, the simplified model will include all feasible pathways, with no regard to their relative importance. The model may be revised when insignificant pathways are identified and eliminated. Different scenarios are handled by varying the rate constants. The model will consider three phases of surveillance and control for the burial site. The first phase is continuing operations, with full control and surveillance. The second phase is post-closure, with minimal control and surveillance for 100 years. The third phase is no control or surveillance. The simplified pathways model will be used as part of the technical framework for the Program. This systematic approach is an excellent way to design monitoring networks and analyze transport processes (Wiersma et. al. 1979). As a management tool, it allows investigators and administrators to consider interactions that may be occuring in the system. As a calculational tool, it will provide fast, approximate results at a reasonable cost. If mere precise answers are desired, more detailed models, some of which currently exist at the INEL, may be used. Some of those detailed models are expensive to run and may take several days before answers are available. The simplified model will help provide guidance on when further detailed modeling development is necessary, Figure 2. A computer program has been set up on the INEL computer system which calculates the inventory of radionuclides in each compartment with time. The program solves a set of simultaneous linear differential equations (one equation for each compartment) by matrix inversion. The model was modified from the Savannah River Laboratory DOSTOMAN model, which has been used at that site to project dose to man from buried radioactive waste (Fauth and Wilhite 1980). The initial amount of a radionuclide in each compartment and the first-order rate constants are required inputs. Initial estimates of rate constants are being obtained from data collected at the RWMC or the INEL, when possible. Literature values from other sites will also be used initially when INEL-specific data are not available. Tentative fiscal-year 1983 tasks include implementing a plotting capability and interactive mode.
394
USING THE SIMPLIFIED PATHWAYS MODEL AS PART OF THE MONITORING ACTIVITIES REVIEW (MAR) The simplified pathways model is being used in the MAR of the Program. Objectives of the simplified model, with regard to the MAR, are listed and discussed below: o
To identify environmental pathways which should be monitored by the Program
o
To identify significant transport pathways and processes during different time periods
o
To identify key input parameters which need investigation by special studies
o
To revise monitoring activities procedures, if necessary, to provide input data in correct form for models.
The simplified pathways model may identify pathways which should be routinely monitored by the Program. For example, there has been no routine biotic monitoring at the RWMC. An analysis of data collected from past field studies will give an estimate of the relative importance of some of those biotic pathways to determine the need for routine monitoring in the future. The simplified model can also be used to identify significant transport pathways and processes during different time periods. This information may be useful in management of the waste and in environmental monitoring design. Because of radionuclide decay and ingrowth, a different set of radionucl ides will be present at different times in the future. The projected inventories of 27 high-hazard nuclides have been estimated through the year 3093 using the ORIGEN computer code and burial inventories and projections. Since each radionuclide can behave differently in the environment, different pathways and processes may dominate at different time pariods. For example, the fission products Cs, Sr, and Co decay relatively quickly and also are relatively mobile in food chains. The longer-lived components of low-level and TRU waste tend to be less mobile, and their movement is mainly determined by the movement of soil to which they are adsorbed. Key input parameters which are missing or are highly variable will be identified during the simplified modeling development. Special studies may be recommended to obtain additional data for those parameters in the model. Some present monitoring techniques may be revised or augmented to make the data useful for modeling, rather than simply archived and used for compliance purposes. For example, in addition to collecting concentrations of radionuclides in intermittent runoff at the burial ground, an estimate could be made of the total annual volume of surface water exiting the site.. This information could then be used to estimate
395
the total contaminant flow via a surface water pathway. Also, data on the size distribution of resuspended particles could help to, estimate the dispersion potential of the measured airborne radionuclides. After the critical pathways have been identified, the monitoring activities are analyzed for cost-benefit. Errors '-e propagated (NBS 1980), and the number and volume of samples needed to achieve a desired level of confidence in the data are determined. The costs associated with a desired level of sampling effort can then be determined. From this information, decisions can be made on the degree of confidence desired for data from each monitoring activity.
SUMMARY The results of the cost-benefit analyses, trend analysis, and modeling, along with the monitoring data, will be used to assess many management issues: o
Compliance with regulations
o
Revisions to current monitoring activities
o
Identification of special studies to define Key parameters for the models
o
Adequacy of current practices (waste acceptance criteria, packaging, confinement, etc.) to limit present and future radiation doses to acceptable levels
o
Reliability of current monitoring data
o
Functional relationships between contamination levels in different environmental media
o
Identification of key radionuclides and pathways for dose to man
o
Guidance to help direct program funds where the greatest need exists
o
Comparisons of projected dose from waste radionuclides with dose from natural and fallout sources
o
Projections of the future concentrations of various radionuclides in air, soil, water, and biota
o
Required steps for RWMC closure
o
Desired post-closure monitoring, if any.
396
Several steps will be required to make the Program fully operational, Figure 2, where the flow of monitoring data and modeling results guide design. At present, many of the activities shown in Figure 2 do not exist. Other activities are being conducted, but not as part of a systematic design effort.
397
REFERENCES Corley, J. P. et al. 1981. A Guide for: Environmental Radiological Surveillance at U.S. Department of Energy Installations, DQE/EP-0023. Fauth, D. J. and E. L. Wilhite. 1980. "Prediction of Radionuclide Migration from Savannah River Plant's Buried Waste," in Transactions of the American Nuclear Society, 34: 119. Janke, D. H. and T. P. Zahn, 1982. Annual Report 1981: Environmental Surveillance for the INEL Radioactive Waste Management Complex, EG&G Idaho, Inc., EGG-2209. NBS (National Bureau of Standards) Special Publication. 1980. A Guide and Recommendations for Reporting of Environmental Radiation Measurements Oata, Ad Hoc Subcommittee on Data Reporting. U.S. Health Physics Society. 1980, Upgrading Environmental Radiation Data, Committee Report, J. E. Watson, Chairman, Office of Radiation Programs, U.S. EPA 520/1-80-012. Wiersma, G. B.t et al. 1979. Kinetic and Exposure Commitment Analyses of Lead Behavior in a Biosphere Reserve, Chelsea College, University of London Monitoring and Assessment Research Center Report Number 15.
398
7D INTERNAL REVIEW SYSTEM FOR ENVIRONMENTAL PROTECTION PROGRAMS OF THE DEPARTMENT OF ENERGY FACILITIES OPERATED BY UNION CARBIDE CORPORATION NUCLEAR DIVISION
H. H. Abee and M. E. Mitchell Union Carbide Corporation Nuclear Division Oak Ridge, Tennessee
ABSTRACT In order to ensure satisfactory environmental protection programs for the four Department of Energy (DOE) facilities operated by Union Carbide Corporation Nuclear Division (UCCND), a comprehensive internal review system has been established. This system provides a highly structured mechanism for evaluating and documenting all aspects of the environmental protection programs of the four UCCND operated facilities, and for communicating this information to the appropriate levels of UCCND management. The program establishes a specific team of professionals for conducting the review, a specific process for conducting the review, and a definitive set of measures of performance. Each facility is then reviewed in the same manner and evaluated by the same criteria. The results of the review are discussed with the facility manager and subsequently documented and transmitted to all appropriate levels of management.
INTRODUCTION The documented assurance of a technically-sound, cost-effective environmental protection program, including compliance with all applicable regulations, is a task facing many, if not all, environmental managers.
Based on work performed at DOE Facilities at Oak Ridge, Tennessee, and Paducah, Kentucky, operated by the Union Carbide Corporation Nuclear Division. 399
-This task is often mandated through corporate policies and, in the case of DOE contractors, it is required by DOE Order 5482.1A. One of the most effective tools for implementing such a quality assurance effort is a wellorganized internal environmental review program that allows for systematic evaluation of a facility's success in meeting environmental regulations as well as corporate policies. To be successful, this type of program should contain the following essential elements: explicit senior management support; a review or audit manager and/or team independent of the facility being reviewed; qualified review personnel; a structured review program with written procedures or guidelines; a consistent, formalized report system; and an effective, corrective action program. In order to enhance the assurance of a high quality environmental protection program, as well as to meet corporate requirements and DOE Order 5482.1A, Union Carbide Corporation Nuclear Division has developed a comprehensive internal environmental review program for the four facilities it operates for DOE in Oak Ridge, Tennessee, and Paducah, Kentucky. The program was implemented during 1982 and all four facilities have been subjected to one review. THE REVIEW PROGRAM The UCCND internal review program is coordinated by the UCCND Office of Health, Safety, and Environmental Affairs--a central staff group that has general oversight responsibilities. The major component of the program is the detailed on-site review of each individual facility by a team of experts selected by this staff group. For any individual facility review, the review team will usually consist of a member of the Office of Health, Safety, and Environmental Affairs.--who serves as the chairperson, the UCCND Environmental Sample Analysis Coordinator, and two of the environmental coordinators from UCCND operated facilities not being reviewed. In order to ensure a comprehensive, consistent review of each of the facilities, a precise set of objectives and review topics has been established. The objectives can generally be categorized as follows; 1. Determine the adequacy of the scope of each facility's environmental program; 2.
Determine the adequacy of the quality of each facility's environmental program;
3. Determine the adequacy of each facility's management system to effectively implement their environmental program; and 4.
Determine the overall effectiveness of each facility's environmental program relative to Union Carbide and DOE environmental protection requirements.
400
Given these four primary objectives, a detailed list of specific criteria has been developed as measures of performance (Table 1 ) . The actual review process thus becomes a detailed evaluation through discussion, observation, and inspection of the environmental program relative to the criteria. The frequency of the facility reviews is determined by the maturity and overall quality of the individual programs, but generally does not exceed every two years. THE REVIEW PROCESS Each facility review is preceded by a letter from the UCCND Office of Health, Safety, and Environmental Affairs to the appropriate facility manager. The letter generally summarizes the UCCND and DOE policy for internal reviews and establishes a date for the specific facility review. It also provides a general listing of topics to be addressed and requests that the facility environmental coordinator prepare a detailed agenda for the review. Where specific facilities and equipment need to be observed and/or specific people need to be talked with, the letter will make such requests. The review process begins with a discussion between the review team, the facility environmental coordinator, and the facility manager or his designated representative. The objectives of the review are again discussed and any changes in the review process are effected at this time. Following these discussions, the review process commences in accordance with the predetermined agenda, with the review team utilizing the criteria listed in Table 1 as a guide for obtaining the appropriate information. Gathering this information involves discussions with the environmental coordinator and his staff, appropriate line organization managers, and hourly employees and first-hand observation of selected facilities and equipment. After all needed information is gathered by the review team and preliminary evaluations have been made, a closeout discussion is held with the facility manager and environmental coordinator. At this meeting, the review team communicates preliminary findings and recommendations. The final effort of the review team involves the preparation of the final review report which is transmitted to the appropriate facility manager with a cover letter requesting that prudent actions be taken to address the review team's recommendations. This report then becomes a working document for the facility environmental coordinator to use in improving his facility's environmental program as well as a reference document for subsequent reviews. SUMMARY AND CONCLUSIONS In summary, Union Carbide Nuclear Division has developed an internal review program that it believes to be a valuable tool in maintaining a high quality, cost-effective environmental program for each of the facilities it operates for DOE. To date, each of the four DOE facilities has been reviewed once and each environmental coordinator has been assigned to two review teams. In addition to the obvious benefits reaped by the facility
401
being reviewed, the review team members always find that such an in-depth review of another facility's program provides them valuable information to be taken back to their respective sites for use in their own programs. Based on the experience afforded by four separate reviews, it is apparent that the success of this review system has been and will continue to be dependent upon explicit senior management support, an independent review team, technically-qualified reviewers, written criteria for the highly-structured review process, a formal reporting system, and an effective corrective action program.
402
TABLE 1 UCCND ENVIRONMENTAL MANAGEMENT PROGRAM PERFORMANCE CRITERIA 1. Organization and Commitment a. Concern for and support of environmental matters are demonstrated at all levels of management. b. Management's concern and support are visible to all employees. c.
Environmental responsibilities and accountabilities are well documented, understood, and accepted by all levels of line management as well as by functional groups who impact environmental performance.
d.
Environmental objectives are established and documented annually for appropriate managers.
e. Adequate resources (including personnel) are available to implement an effective environmental program. f. All environmental complaints are documented and treated with concern and, where necessary, are followed up with plans for corrective actions. 2.
Procedures a. All necessary procedures pertaining to the facility environmental program are in place. b. An adequate system for new procedure review and approval is in place. c. An adequate system for updating existing procedures is in place. d. All employees are knowledgeable of environmental procedures and of their implementing responsibilities.
3.
Training a. All necessary environmental training programs are in place and functioning as required. b.
Environmental training responsibilities are well defined.
c.
Employee training activities are documented.
d.
Evidence of follow-up and retraining is available.
403
4.
Communications a. Environmental responsibilities are communicated to employees as necessary and evidence of such communication is available. b. There is adequate horizontal and vertical two-way communication on environmental matters at all levels of management. c. Abnormal environmental events are effectively communicated to facility, UCCND management, and DOE. The communication responsibilities, including time restraints, are documented and understood by the appropriate personnel.
5. Facilities and Equipment a. All projects involving new facilities and modification to existing facilities receive adequate review from design to construction by qualified environmental personnel. b. All appropriate permits are applied for and obtained on schedule. c. All potential sources of environmental pollution, as well as all pollution abatement equipment, are routinely inspected and tested to ensure proper operation so as to minimize releases.
6.
d.
All abnormal environmental incidents are thoroughly investigated and the appropriate documentation is prepared.
e.
Internal audits of pollution abatement procedures are conducted, documented, and corrective actions are taken where necessary.
NPDES Program a. All liquid effluents are characterized and the characterization is documented. b. All liquid effluents are in compliance with their respective NPDES 1 imitations. c. All flow measuring, monitoring, and sampling equipment are properly installed, maintained, and calibrated. d.
All maintenance and calibration activities are adequately documented.
e. All applicable treatment, monitoring, and sampling equipment are equipped with redundant power sources. f. An adequate procedure exists for ensuring that all new liquid effluents are reviewed for NPDES permits.
404
7. Laboratory Procedures a. EPA-approved analytical methods are used. b. A comprehensive QA program is employed (including the use of control and/or spiked samples) and documented. c. Adequate records of analyses and data are maintained. d. Minimum detectable reporting limits are maintained consistent. e. Analytical data are reported in acceptable and consistent units. f. "Turnaround" time for samples is satisfactory. g. Anomalies in data are handled in a satisfactory manner. 8. Airborne Emissions a. All airborne emissions are characterized and the characterization is documented. b. All airborne effluents are in compliance with applicable regulations. c. An adequate procedure,exists for ensuring that all new and modified processes are reviewed for applicable airborne effluent permits. 9.
RCRA Program a. Chemical and physical analyses for all hazardous wastes are completed and documented. b. A written waste-analysis plan is available. c. A procedure for off-site hazardous waste shipments is in place including provisions for proper packaging and labeling. d. All off-site shipment manifests are retained in an accessible file. e.
Written schedules for inspection of storage, treatment, disposal, monitoring, security, and emergency response equipment are in place and are being implemented.
f. A site contingency plan is available. g. A written closure plan for storage treatment and disposal facilities is in place. h. Groundwater sampling plans are in place and are being implemented.
405
i. A post-closure plan is available for each hazardous waste facility. j. 10.
Adequate records of all hazardous waste disposals are in place including written evaluation reports of commercial disposers.
PCB Program a. Storage containers and storage areas for PCB wastes are properly maintained and labeled. b.
PCB equipment and waste storage facilities are routinely inspected and the inspections are documented.
c. An up-to-date inventory of all PCBs is in place. d>
All PCB articles are properly labeled.
e. All disposal activities are with firms meeting all UCCND criteria and facility evaluation reports are documented. f. All disposal activities are well documented. 11.
Environmental Monitoring Program a. The environmental monitoring program is well documented including specific criteria for the program objectives. b. The monitoring program includes methodologies for determining off-site radiological and chemical doses. c.
The monitoring program is sufficient to provide data for satisfactorily determining off-site environmental impacts.
d. The program includes a system for routinely evaluating off-site impacts and for generating action plans to correct any problems. 12.
Items to be Discussed With Line Organization Managers a. Discuss and observe specific measures of performance dealing with environmental concerns. b.
Discuss and observe documentation pertaining to specific employee training activities dealing with environmental concerns.
c. Discuss availability of and familiarity with environmental procedures. d.
Discuss specific environmentally-related inspection programs.
e. Determine opinion of the facility environmental program effectiveness.
406
f. Discuss and inspect any specific environmental action plans. 13.
Items to be Discussed With Hourly Employees a. Discuss specific orientation and training programs they have attended. b. Discuss their access to and familiarity with environmental rules and regulations applicable to their specific work assignments. c. Discuss their evaluation of the benefits of environmental training they have received. d. Determine any suggestion for improving specific parts or all of the environmental program.
7E SITE MONITORING FROM SOIL SAMPLE ANALYSIS
C. T. Illsley Rockwell International Golden, Colorado
ABSTRACT Soil samples have been collected for the past three years as part of a long range monitoring program. The program is designed to provide information on possible migration of Plutonium in soil and to provide data for comparison with the EPA proposed guidance on transuranium elements in the environment. The program was initiated in 1979 and will continue at least through 1983. Samples have been collected at six locations west of Indiana Street within the eastern boundaries of the Rocky Flats Plant site. The EPA comparison study has been performed at five sites and the plutonium migration study is underway at the sixth site. The data on plutonium analyses will be compared to the EPA screening level of 0.20 yCi/m2 (74 X 10 8 Bq/km2) in the five boundary sites. Possible migration trends will be examined for the plutonium data on soils from the other site.
INTRODUCTION It has been known since 1970 (Krey, 1970) that soil on and around Rocky Flats contains plutonium at concentrations exceeding that from worldwide fallout from weapons' tests. Krey and others (Krey, 1976) indicate that releases from past operations have amounted to about 11.4 curies of plutonium, about 99% of which was leakage from steel drums containing contaminated cutting oil between 1959 and 1969. The HASL data suggest that of the 11.4 Ci, 8.6 Ci are on site. The remaining 2.8 Ci are dispersed at distances from the Plant at levels equal to or below the level of plutonium from fallout (0.0015 uCi/m 2 ). From the 8.6 Ci included on site, the HASL data estimate that about 1.7 Ci are in the area that is now covered with an asphalt pad.
409
The only Federal standards for plutonium in the environment are for levels of plutonium in air and drinking water; consequently there are no official Federal standards with which to compare plutonium levels in soil. In 1977, the EPA released a draft guidance document, "Proposed Guidance on Dose Limits for Persons Exposed to Transuranium Elements in the General Environment" (EPA, 1977). The guidance provides screening levels for air of 1 fCi/m3 and for soil of 0.20 yCi/m^. These screening levels are values which indicate a need for further testing to determine whether dose limits are exceeded in areas having uncontrolled access. The EPA specifies that the soil testing method should be adequate to analyze soil to a one centimeter depth and having a soil particle size less than two millimeters. Extensive data taken in areas surrounding the plant site indicate that there are no offsite areas which have concentrations of plutonium in soil higher than the EPA screening level. The current study is an attempt to investigate the plutonium concentrations in soil immediately inside the Plant's east boundary and to compare them to the EPA proposed guidance. The HASL data indicate plutonium levels in the range between 0.05 and 0.50 pCi/m2 f o r the soil in the area near the Plant's eastern boundary. Access to this area is not open to the general public and is controlled by a barbed wire fence and locked gates. SAMPLE COLLECTION AND ANALYSIS Soil samples have been collected at six locations west of Indiana Street within the eastern boundaries of the Rocky Flats Plant site. The sites are shown on Figure 1 as numbers 13, 16, 17, 21, 25 and 28. Sites 13 and 21 were sampled in 1979, sites 16 and 28 were sampled in 1980 and the other sites were sampled in 1981. The EPA comparison study has been performed at sites 13, 17, 21, 25 and 28 and will be continued at three additional sites in the future. The plutonium migration study is underway at site 16. Nine composite samples, each composed of nine subsamples, were collected at each of the five EPA comparison sites. Collection was done according to published procedures (AEC, 1974 and Harley, 1972). Each set of nine subsamples was collected on a spacing of 20 meters and composited to yield one of the final samples. The geometry of each subsample was controlled by use of a 10 X 10 X 1 cm cutting tool. The soil contained within the tool cavity was removed and placed in a metal paint can. Two series of samples for the plutonium migration study have been taken at site 16. Sixty samples, made up of five composites each, were taken at 30 locations. These locations were selected on a random basis from a grid of 64 squares (2 meters on each side of a square) separated by alleys 1 meter wide. The subsamples were taken from the four corners and the center of each square. The remaining squares will be sampled in subsequent years to determine surface and depth changes in plutonium concentrations with time. 410
Colorado 128
(not to seal*)
FIGURE 1. LOCATION OF SOIL SAMPLING PLOTS
The samples from each square consisted of surface and core samples. Surface material was collected by means of a 10 X 10 X 5 cm cutting tool. The core samples were taken from the same sites as the surface samples using an orchard auger measuring 8.3 cm in diameter. The depth of the cores was from 5 down to 20 cm. Surface and core samples were retained as individual samples but received identical preparation and analysis. Sample preparation and analysis were performed according to published procedures (EPA, 1979 and Harley, 1972). The entire soil sample was dried, sieved through a 10 mesh sieve, weighed and the fine portion was ball milled. A 100 gram aliquot was taken for additional pulverizing in a stainless steel shear. A 10 gram portion of the pulverized soil was then analyzed for plutonium. Blank soil and synthetic standard soils were batched with the field samples and replicates in a random manner as quality controls. Chemical recovery from the radiochemical procedure was determined by the addition of plutonium-236 for plutonium separations. Plutonium was extracted from the soil with a mixture of nitric, hydrofluoric, and hydrochloric acids. Plutonium was isolated by hydroxide precipitation, and purified by anion exchange. The purified material was then electrodeposited on a polished stainless steel disk for determination by alpha spectrometry. The analytical results were reported by the laboratory in units of disintegrations per minute per gram of dry soil fines. These values were converted to picocuries per gram (pCi/g) of 2 dry soil (less than 10 mesh) and microcuries.per square meter (yCi/m ). The latter surface values were derived by multiplying the sample concentration by the sample weight and dividing by the sample area. COMPARISON WITH EPA PROPOSED GUIDANCE Plutonium concentrations in soil samples at sites 13, 17, 21, 25 and 28 are presented in Table 1. As shown in the table, the values at site 13 range from 1.77 to 3.52 pCi/g with the median of 2.54 pCi/g. The median in surface units, 0.015 yCi/m? is 7.5% of the EPA proposed guideline in soil (0.20 yCi/m^). Higher values were measured in soil samples from site 17. The range is between 5.89 and 15.0 pCi/g with the median at 7.96. The median in surface units, 0.108 yCi/m2 i S 54% of the guideline and the maximum value of 0.252 is 126% of the guideline. This particular site of about 10 acres apparently received a greater deposition of windblown plutonium than the other sites. It is more like the deposition found at site 16, as will be demonstrated later. The amounts of plutonium in the soil from the other three sites are similar to those at site 13. The soil from site 21 was found to contain plutonium in the range between 2.02 and 4.01 pCi/g with the median at 2.46 pCi/g- The median in surface units, 0.020 yCi/m2, is only 10% of the EPA guideline. Concentrations at site 25 were in the range from 0.95 pCi/g
412
TABLE 1. Plutonium Concentrations in Surface Soil Samples at the East Boundary of the Rocky Flats Plant (1979 and 1980) Location
uCi/m2
fiCi/g_a
Location
fip.i/ga
uCi/m2
13-1
2.84 + 0 .24
0.019
17-7
6.17 ± 0..59
0.108
13-2
1.77 • 0 .10
0.013
17-8
7.94 + 0..77
0.091
13-3
2.34 • 0..21
0.014
17-9
5.89 + 0..52
0.080
13-4
3.52 ' 0,.19
0.026
Mean
8.40 • 0. 72
0.119
13-5
3.01 ' 0..42
0.014
Median
7.96
0.108
13-6
2.44 » 0. 19
0.013
RSDC
13-7
2.35 ' 0. 10
0.017
21-1
4.01
0. 36
0.022
13-8
2.54
•
0.
15
0.015
21-2
2.13 ' 0. 14
0.017
13-9
3.27
•
0 . 22
0.018
21-3
3.22 • 0. 44
0.019
Mean
2.68 • 0 . 07
0.017
21-4
2.08
*- o30 .
0.020
Median
2.54
0.015
21-5
3.21 ' 0. 31
0.021
21-6
2.32 • 0. 18
0.019
RSD C
20"<
25%
32%
43% 1
17-1
9.21 • o r 97
0.112
21-7
2.02 • 0. 21
0.021
17-2
8.48 • 0 . 70
0.112
21-8
2.46 + 0. 15
0.016
17-3
8.16 » 0 . 73
0.103
21-9
3 . 3 . * 0. 22
0.022
78
0.252
Mean
2.75
0.020
2.46
17-4
15.0
• +
0.
17-5
6.80
0 . 43
0.103
Median
17-6
7.96 » 0 . 72
0.112
RSDC
25%
+• 0 . 09
0.020
m
a. Concentrations are for the fraction of soil measuring less than 2 millimeters in size. b. Samples were collected to a depth of 1 cm. c. Percent relative standard deviation of the mean.
413
TABLE 1. Plutonium Concentrations in Surface Soil Samples at the East Boundary of the Rocky Flats Plant (1979 and 1980)(Cont'd) pCi/ga
Location
yCi/m2
Location
pCi/ga
yC_i/m£
25-1
1.44 t 0.10
0.026
28-1
1..03
t 0.07
0.016
25-2
2,,46
0.041
28-2
0.75
• 0.05
0.010
25-3
2.,38 ± 0.09
0.039
28-3
1.09 :-
25-4
2.,59 + 0.19
0.045
28-4
0.93
25-5
1.46 i
o.n
0.034
28-5
1.22 J 1
0.08
0.016
25-6
1.63 + 0.10
0.025
28-6
0.87 J• 0.07
0.014
25-7
1.74 + 0.11
0.021
28-7
0.88 < 0.06
0.009
25-8
0.95 ± 0.06
0.020
28-8
1.30
i 0.08
0.013
25-9
1.76 ± 0.11
0.037
28-9
1.49
•
0.10
0.019
Mean
1.82 f 0.11
0.032
Mean
1.06 < : 0.02
0.014
Median
1.74
0.034
Median
1.03
0.014
RSDC
±
0.14
30%
RSD C
29%
22%
0.07
0.016
> 0.07
0.013
2
a. Concentrations are forthe fraction of soil measuring less than 2 millimeters in size. b. Samples were colllected to a depth of 1 cm. c. Percent relative standard deviation of the mean
414
Colorado 128
0.020 Qj)
en
0.014 Smart Ditch (not to tcalt)
V
\
FIGURE 2.
MEDIAN VALUES OF PLUTONIUM IN SOIL (yCi/m 2 )
to 2.59 pCi/g with the median at 1.74 pCi/g. Compared to the EPA guide the median (0.034 yCi/m^) is 17% of that value. The lowest concentrations of plutonium were found at site 28 where the range was between 0.75 and 1.49 pCiVg^with the median at 1.03 pCi/g. The median in surface units, 0.014 uCi/m > is 7% of the EPA guidance value. The median values at the five sites are shown on Figure 2 for perspective. It appears that the site with the highest plutonium deposition is almost due east of the location of the former drum storage area. Other sites are either southeast or northeast of the 903 area and slightly more distant. Although the prevailing winds blow from Rocky Flats to the southeast, resuspension of plutonium from the 903 area probably occurred over a relatively few days instead of over a long period of time. The wind direction on these few days could have been easterly instead of towards the southeast. MIGRATION STUDIES Results in Table 2. is from 6.16 plutonium in PCi/g.
of analyses for plutonium in soil from site 16 are presented As shown in the table, the range of values in surface samples to 22.8 pCi/g with a mean of 9.72 pCi/g. Core samples contain the range between 0.24 and 2.08 pCi/g and the mean is 1.08
The median value for surface samples collected in 1980 is 8.68 pCi/g and 8.96 pCi/g for 1981. These data indicate that there has been no change in the surface deposition. The median concentration for core samples taken in 1980 is 1.15 pCi/g and 0.90 pCi/g for 1981. From these limited comparisons, it is concluded that there has been no vertical migration of plutonium at this site. Continued evaluation will be performed with data collected in 1982 and 1983.
416
TABLE 2. Plutonium Concentrations in Soil Samples From Within the Eastern Boundaries of the Rocky Flats Plant in 1980 and 1981 Location
0
Surface** (pCi/g) d
Core 5 J (pCi/g) d
Location 0
Surface a (pCi/g) d
16-5
6.16 + 0.37
0.72 ± 0.06
16-4
16-10
8.68 ± 0.46
0.70 ± 0.05
16-7
16-12
8.98 t 0.45
0.98 + 0.04
16-11
22.8
± 2.3
0.90 ± 0.04
16-13
7.40 t 0.45
1.03 ± 0.07
16-22
10.8 ± 0.7
1.18 ± 0.10
16-21
9.59 ± 0.55
1.42 + 0.11
16-26
15.3
1.01 ± 0.08
16-28
8.66 t 0.46
0.84 + 0.06
16-27
8.96 t 0.62
1.00 ± 0.08
16-34
7.34 ± 0.39
0.60 ± 0.05
16-29
9.33 ± 0.69
1.03 ± 0.10
16-40
8.40 ' 0.29
1.00 ' 0.03
16-31
16-49
12.40 + 0.70
1.46 + 0.10
16-35
8.35 t 0.52
0.80 ± 0.06
16-36
7.89 ± 0.48
1.05 ± 0.07
16-50
7.52 +. 0.21
1.46 \ 0.09
13.0
' 1.3
Core D _. (pCi/g) g
9.24 ± 0.76
12.8
±1.1
+ 0.8
0.24 ± 0.03 0.55 ± 0.06
0.80 ± 0.05
16-51
11.10 i 0.72
1.60 + 0.04
16-42
8.54 .t 0.50
0.65 ± 0.04
16-55
9.92 ± 0.25
1.61 * 0.08
16-45
7.84 ± 0.44
0.89 ± 0.07
16-56
10.80 + 0.70
1.73 ± 0.04
16-52
7.10 ± 0.52
1.88 + 0.14
16-59
6.96 t 0.15
1.15 +. 0.04
16-58
7.10 ± 0.44
0.82 ± 0.07
16-64
10.00 ± 0.23
2.08 + 0.06
16-62
8.72 ± 0.53
1.18 ± 0.09
Mean
8.93 + 0.13
1.22 ± 0.02
Mean
10.52 ± 0.78
0.93 ± 0.07
Median
8.68
1.15
Median
RSD e
19?.
35%
RSD e
8.96 39%
0.90 38%
a. Samples to a depth of 5 centimeters (2 inches). b. Sampled from 5 to 20 centimeters (2 to 8 inches). c. The first number of each location refers to site 16 as shown in Figure 1. The second number is the sample location onthe grid at site 16. d. Concentrations are for less than 2 millimeter size fraction of soil. e. Percent relative standard deviation. 417
-REFERENCES "Acid Dissolution Method for the Analysis of Plutonium in Soil," EPA 600/7-79-081, March, 1979. Harley, J. M., Ed., Procedures Manual and Supplements "i-4, Health and Safety Laboratory, U.S. Atomic Energy Commission, 1972. Krey, P. W. and E. P. Hardy, Plutonium in Soil Around the Rocky Flats Plant. HASL-235. U.S. Atomic Energy Commission, Health and Safety Laboratory, New York, New York. August, 1970. Krey, P. W., E. P. Hardy, H. Volchok, J. Toonkel, R. Knuth, M. Cooper and T. Tamura. Plutonium and Americium Contamination in Rocky Flats Soil-1973. HASL-304. Energy Research and Development Administration, Health and Safety Laboratory, New York, New York. March, 1976. "Measurements of Radionuclides in the Environment: Sampling and Analysis of Plutonium in Soil," U.S. Atomic Energy Commission Regulatory Guide 4.5, May, 1974. "Proposed Guidance on Dose Limits for Persons Exposed to Transuranium Elements in the General Fnvironment," EPA 520/4-77-016, September, 1977.
418
7F
SUMMARY AND RESULTS OF THE COMPREHENSIVE ENVIRONMENTAL MONITORING PROGRAM AT THE INEL'S RAFT RIVER GEOTHERMAL SITE
R. A. Mayes, T. L. Thurow,a and L. S. Cahn b EG&G Idaho, Inc. Earth and Life Sciences Idaho National Engineering Laboratory Idaho Falls, Idaho
ABSTRACT
The Raft River Geothermal Program at the Idaho National Engineering Laboratory (INEL) was a research project designed to demonstrate that moderate temperature (~150°C) geothermal fluids could be used to generate electricity and provide an alternate energy source for direct-use applications. Development of the geothermal reservoir began in 1975 and the environmental program was initiated soon after drilling began. The major elements of the monitoring program were continued during the construction and experimental testing of the 5-MW(e) power plant. The monitoring studies established pre-development baseline conditions of and assessed changes in the physical, biological, and human environment. The Physical Environmental Monitoring Program collected baseline data on geology, subsidence, seismicity, meteorology and air quality. The Biological Environmental Monitoring Program collected baseline data on the flora and fauna of the terrestrial ecosystem, studied raptor disturbances, and surveyed the aquatic communities of the Raft River. The Human Environmental Monitoring Program surveyed historic and archaeological sites,considered the socioeconomic
a. Current address: Department of Range Science, Texas A & M University, College Station, TX 77843 b.
Current address: 373 4th Street, Idaho Falls, ID 83401
419
environment, and documented incidences of fluorosis in the Raft River Valley. In addition to the environmental monitoring programs, research on biological direct applications using geothermal water was conducted at Raft River. Areas of research included biomass production of wetland and tree species, aquaculture, agricultural irrigation, and the use of wetlands as a treatment or pretreatment system for geothermal effluents.
INTRODUCTION Geothermal energy may be used either for the production of electricity or for direct-heat applications. Compared to some other energy technologies, the environmental effects of geothermal energy production are generally considered small. There may however be measurable impacts associated with the exploration, testing, and production phases of a geothermal resource. Several overviews of potential environmental impacts resulting from geothermal development have been published (Pimentel, 1978; Strojan and Romney, 1979; Suter, 1978; Tucker and Tanner 1978; O'Banion and Layton, 1981; Layton et aj., 1981; and Spencer et a]_., 1979). Among the environmental concerns identified are: loss or modification of fish and wildlife habitat; potential socioeconomic impacts (particularly in sparsely populated areas of the western U.S. where many geothermal resources are located); gaseous emissions (primarily carbon dioxide and hydrogen sulfide and, to a lesser extent, mercury and radon); water use; land subsidence; induced seismicity; discharges to surface waters of geothermal fluids high in total dissolved solids; and effects of accidental spills and blowouts. These general issues and other site specific concerns were considered during the environmental monitoring program at the Raft River Geothermal Site. The monitoring program was designed to: (1) provide baseline data with which to assess future impacts of development; (2) collect monitoring data during operations; and (3) serve as an example of how an environmental program associated with a larger-scale geothermal development might be conducted. The last item was an important objective of the environmental program.
FACILITY DESCRIPTION AND HISTORY The Raft River Geothermal Site is located in southeastern Idaho in the Raft River Valley (Figure 1). The presence of a moderate temperature (~150°C) resource in this area had been reported by the U.S. Geological Survey (USGS). A joint effort between the U.S. Atomic Energy Commission (a predecessor to the Department of Energy), the Idaho Department of Watet Resources, and the Raft River Rural Electric Co-op, was undertaken to develop this resource. Between 1975 and 1978, 5 production wells (depths from 1497m to 1994m) and two injection wells (depths of 1176m and 1185m) were drilled. 420
79 a 7 / RRGP-4 I— RRGP-5
Figure 1.
Location of Raft River Geothermal Site and exploration (RRGE), . production (RRGP), and injection (RRGI) wells.
421
The Raft River Basin was declared a critical ground-water area by the Idaho Department of Water Resources in 1963 and the area was closed to further ground-water development. Because there was concern that injection of geothermal waters in the intermediate depth aquifer could affect the quality of water in the shallow aquifer used for irrigation and culinary water, seven monitor wells were drilled. Monitoring results from those and existing wells in the area are reported by Allman et a j . , 1982. Changes in ground-water quality observed were negligible. Short transient pressure responses were noted as a result of geothermal production and injection. The geology, geophysics, hydrology, and geochemistry of the Raft River Geothermal Site have also been described (Dolenc et aJL, 1981; Tullis and Dolenc, 1982; Russell, 1982; and Hull, 1982). A major portion of the Idaho National Engineering Laboratory (INEL) work at Raft River was directed toward the design and construction of a binary cycle pilot plant with a nominal rating of 5MW(e). The principal objective of the pilot plant was to demonstrate the technical feasibility of generating electric power from a moderate temperature geothermal resource in an environmentally acceptable manner. In the binary cycle plant, geothermal fluids are used to heat a secondary working fluid (isobutane) which expands through a turbine-generator to produce electricity. The working fluid is then condensed and reheated in a closed system. Heat is removed from the isobutane by circulating cold, treated geothermal water through a heat exchanger. This heat is then dissipated in a wet evaporative cooling tower. Plant startup testing occurred during August to November, 1981. During the week of October 28, the plant was brought up to its full thermal power of 45MW(t). Testing and operation of the 5MW(e) facility continued through mid-June, 1982. More complete descriptions of the 5MW(e) plant and its water treatment system are provided by Whitbeck and Stiger, 1982 and Suciu et aj. 1982. ENVIRONMENTAL MONITORING AND RESEARCH PROGRAM The environmental monitoring and research activities at the Raft River Geothermal Site were conducted during the period from 1975 to 1982. The monitoring program was established to characterize the existing environment prior to development and to measure changes in environmental parameters as a result of geothermal development. In addition, research regarding biological direct applications of geothermal fluids was conducted to identify other potentially beneficial uses associated with geothermal development. These monitoring and research studies required the cooperation and participation of not only Department of Energy and EG&G Idaho personnel, but also individuals from numerous local groups, State and Federal agencies, universities, and private contractors. The many studies which made up the environmental program are summarized in detail (with the exception of ground-water monitoring which is reported separately) in a final report (Thurow and Cahn, 1982). That report forms the basis for this summary and should be consulted for details.
422
PHYSICAL ENVIRONMENT Physical aspects of the environment may affect entire ecosystems. A physical environmental monitoring program was established to detect changes in the physical environment and to indicate potentially adverse results from geothermal development. Geology The Known Geothermal Resource Area (KGRA) is located at the southern end of the Raft River Valley. The valley, 60 km long and 20 to 24 km wide, is bounded by the Black Pine Mountains, the Jim Sage Mountains, the Raft River Range, and the Snake River Plain (Figure 2 ) . The Raft River meanders northward through the basin from the southern end of the Jim Sage Mountains. The geologic structure of the Raft River basin near the KGRA has been studied extensively and is described in detail in other reports (Dolenc et. a j . , 1981; Thurow and Cahn, 1982). The geothermal reservoir in the KGRA occurs near the Horse Well and the Bridge fault. A USGS analysis of the thermal fluids in the reservoir suggests the fluid is at least 60 to 70 years old. Static water levels in the thermal reservoir are about 100m above the land surface. Sei smicity The possibility of inducing earthquakes as a result of fluid withdrawal or injection is of concern during geothermal development. At Raft River, a seismic network was established (initially in July, 1974) to collect baseline data and monitor microseismic activity during geothermal field testing, production, and injection. The low level of background seismicity found in the vicinity of the KGRA indicates a low-stress environment. It is unlikely that earthquakes would be triggered by geothermal activities in the low-stress environment near the Raft River facility and, to date, no increase in seismic activity has been detected. Subsidence Excessive groundwater withdrawals from unconsolidated or poorly consolidated aquifers may cause land subsidence and fracturing. Areas in the northern Raft River Valley (about 40 km from the geothermal site) have subsided over 0.9m in the last 20 years because of excessive groundwater pumping for irrigation. A detailed surveying grid was established in 1975 at the Raft River geothermal well field to monitor potential subsidence caused by geothermal fluid withdrawal. The grid was surveyed again in 1978 and 1980. The grid was expanded in 1979 after completion of all wells to periodically include elevation checks at specific wells during production and injection tests.
423
Defined Faults Inferred Faults Down thrown side of fault
o Burley
10 J Kilometers
Figure 2. Major structural features of the southern Raft River Valley.
424
With the exception of five points located in cultivated fields where elevation changes were due to farming activities, all changes in elevation measured during the subsidence surveys were within the experimental error. No detectable changes in elevation occurred as a result of geothermal development at Raft River. Meteorology Meteorological data are important to a monitoring program because wind speed and direction, temperature, and precipitation may have direct effects on components of the ecosystem and influence air quality data. A weather station was established near the geothermal site in 1975 to monitor wind velocity and direction, precipitation, ambient air temperature, and dewpoint temperature. The mean annual temperature in the valley is 8°C, and the extremes are -30°C and 40°C. Rapid cooling during clear evenings creates night time inversions, but winds and morning heating of the ground usually clear the inversions before afternoon. Because of a high frequency of windy days, dispersion characteristics at the site are good. Precipitation data are quite variable from year to year and month to month. The annual average precipitation during the period of geothermal development was 255 mm. Air Quality Monitoring Based on experience from the Geysers development in California, one of the environmental concerns at Raft River was the emission of H~S. It should be noted that the Raft River plant [5MW(e)] is small compared to the generating capacity at the Geysers project [>900 MW(e)]. Particulate emissions from construction and operation activities, and from the drying of mists emitted from the cooling towers were identified as concerns also. Original plans called for the use of a chromate corrosion inhibitor in the cooling tower, but a phosphate and zinc sulfate mixture was used instead. All measurements were made in accordance with standard Environmental Protection Agency (EPA) reference methods (40CFR 50). The baseline air monitoring program was initiated in 1975 and expanded in 1980. Total suspended particulate (TSP) data were collected at four locations around the geothermal site after 1980. On April 28 and 29, 1982, the emissions from the cooling tower were sampled to determine fluoride, sulfate, hydrogen sulfide, zinc, and particulate emissions. The annual average TSP concentrations are well below the primary National Ambient Air Quality Standard (NAAQS) of 75 yg/m 3 and the secondary NAAQS of 60 yg/m (Table 1). All samples except those taken on June 30, 1976 and June 19, 1981 were also below the primary 24-h standard (260 jig/m ) and the secondary 24-h standard (150 pg/m ). The cause for the 1976 exception was wind-raised dust from bare fields near the sample station. The 1981 exception was caused by a road maintenance crew working
425
TABLE 1. GEOMETRIC MEAN TOTAL SUSPENDED PARTICULATE (TSP) CONCENTRATIONS
Sampling Period
a A
B
C
D
10/80 to 12/80 01/81 to 03/81 04/81 to 06/81
10.7 10. 7 4.77 4. 11.3 11. 3
14.5 14.5 6.7 6.7 19.5 19.5
12.7 6.7 17.2
16.4 7.6 16.1
07/81 to 08/81 10/81 to 12/81 01/82 to 03/82 04/82
34. 6 34.6 6.4 6. 4 5.55 5. 1. 5 T.5
41.1 41.1 7.9 7.9 6.0 6.0 13.5 13.5
45.7 6.2 5.5 12.0
45.1 6.7 5.2 12.2
a. Station A is the upwind or background location.
on an unpaved road near Station B. TSP concentrations were higher during the dry summer months and lower during the wetter periods of fall and winter. Values were also lower when a snow cover existed. The largest 3 number of TSP values were in the range of 0 to 10 yg/m . The TSP concentrations during the 1980-82 sampling period ranged from 0.3 to 3 3 388.8 yg/m with the average being about 19 yg/m . TSP values at Station A (background) were generally 30-50% lower than sites located near unpaved site access roads and plant construction activities. The cooling tower emmissions tests revealed that most of the measured parameters (sulfate, fluoride, phosphate, suspended solids, total particulates and hydrogen sulfide) were below the limits of detection for the methods used. Zinc was present at an average of only 11 yg/m 3 . The U.S. EPA conducted tests of radon gas emissions in March 1976. Radon-222 concentrations in the geothermal fluids were about 390pCi/L, a relatively low concentration. It was concluded that TSP levels at the site result mainly from agricultural activities and from construction and traffic at the geothermal facility. Ambient levels were well below standards. The impacts on air quality of site activities, plant operation, and cooling tower emissions were minimal. BIOLOGICAL ENVIRONMENT Effective environmental management requires an understanding of the diversity and population interactions of the biotic community. At the
426
Raft River Valley, baseline data on aquatic and terrestrial flora and fauna have been collected. These baseline studies have significantly improved the understanding of the ecology of the valley. Due to programmatic changes, the power plant did not operate as long as originally intended. Therefore, no ecological impacts due to operations were observed. However, these studies provided much-needed data about plant and animal populations and their natural variations. This information will be useful for assessing impacts of any future development in the valley and provides a good example of the type of information needed to assess impacts. The raptor studies demonstrate how such information can be used to minimize ecological impacts. Raptor Ecology and Disturbance Raptors are important biological indicators of environmental perturbations and as such may reflect changes occurring within an ecosystem. Data from this study provide baseline information for south central Idaho which can be used as a reference for similar habitats typical of the Great Basin. Twenty-one raptor species were present in the Raft River Valley. Golden eagles, Swainson's hawks, ferruginous hawks, and several species of owls were most common. A total of 181 active raptor nests were found in the valley during 1978 and 1979. The limited land disturbance and increase in human activity associated with the Raft River geothermal development did not have an observable effect on raptor populations. Observed declines in large raptor nesting success (i.e., golden eagle and ferruginous hawk) were associated with the natural cyclic trend in the jackrabbit population. Jackrabbit population cycles were studied extensively as part of this program, and without those data, the change in raptor nesting success would have been more difficult to interpret. The ferruginous hawk, the largest hawk in North America, is prone to nest desertion from disturbance and its numbers are apparently declining nationwide. The Raft River Valley has one of the most stable ferruginous hawk populations remaining in the country (Thurow et aj., 1980). During 1978, 1979, and 1980, nests were disturbed by several means to simulate noises common to development of a geothermal site (e.g. vehicles, small gasoline engines, investigator approaching on foot). Flushing distance and fledging rates were used as measures of response to the disturbance. The study concluded that nesting success'of the ferruginous hawk in Raft River Valley was not impaired by geothermal development and associated human activity as long as buffer zones (approximately 0.6 km) were not violated. HUMAN AND CULTURAL MONITORING PROGRAM Developing the geothermal resources of the Raft River Valley could provide local residents with many benefits and opportunities; however, some undesirable alterations could also result. The high fluoride levels
427
sometimes associated with geothermal development were of concern in the Raft River Valley. A program was established to identify potential socioeconomic changes that could accompany development of the geothermal resource. A survey was conducted in the Raft River Valley to document the existence of historic and archaeological sites. The development had no impact on known sites and no undiscovered sites were located during construction activities. A socioeconomic evaluation of Cassia County and potential impacts that could be associated with development at the Raft River Site was conducted from 1976 through 1980. Many benefits resulted from development of the geothermal site. Many locals were employed during development and wages at the site were generally higher than average wages in the county. Geothermal development increased tax revenues and reduced unemployment. There were no significant land use impacts. Water with a high fluoride content can cause chronic fluoride poisoning (fluorosis) in humans and animals. Because fluorosis in a nearby community had been linked to high fluoride levels in the water supply, an investigation of the incidence of denta' fluorosis in the Raft River Valley was undertaken. The incidence of dental fluorosis in Valley residents appeared to be abnormally high. However, fluoride levels in drinking water were low and no cause for the higher than normal incidence of fluorosis was found. The fluorosis was not associated with development of the geothermal resource. BIOLOGICAL DIRECT APPLICATION RESEARCH Research designed to test the feasibility of using energy expended geothermal fluid for beneficial biological uses was conducted at the Raft River site. The most common method of disposal of energy expended geothermal water is injection. However, in the arid west where water supplies are limited, some geothermal waters might be suitable for irrigation or other purposes. Survivability and productivity of various agricultural, aquacultural, rangeland, and tree species were tested at the Raft River facility. Studies were also conducted to assess the potential of biological systems such as wetlands for water purification. These studies are described by Thurow and Cahn, 1982 and by Breckenridge et al. 1982.
428
REFERENCES Allman, D. W., J. A. Tullis, M. R. Dolenc, T. L. Thurow, and P. A. Skiba. 1982. Raft River monitor well potentiometric head responses and water quality as related to the conceptual groundwater flow system. Report No. EGG-2215, Volume II, EG&G Idaho Inc., Idaho National Engineering Laboratory, Idaho Falls, Idaho. Breckenridge, R. P., L. S. Cahn, and T. L. Thurow. 1982. "Biological treatments and uses*of geothermai water as alternatives to injection." Geothermai Res. Council Bull. ll(4):3-6. Dolenc, M. R., L. C. Hull, S. A. Mizell, J. A. Strawn, and J. A. Tullis. 1981. study. Report No. EGG-2125, Volume I, National Engineering Laboratory, Idaho
B. F. Russell, P. A. Skiba, Raft River geoscience case EG&G Idaho, Inc., Idaho Falls, Idaho.
Hull, L.C. 1982. "Changes in fluid chemistry during injection, Raft River KGRA." Geothermai Res. Council Bull. 11(4):11-14. Layton, D. W., L. R. Anspaugh, and K. D. O'Banion. 1981. Health and environmental effects document on geothermai energy—1981. Report No. UCRL-53232, Lawrence Livermore National Laboratory, Livermore, California. O'Banion, K., and D. Layton. 1981. Direct use of hydrothermal energy: review of environmental aspects. Report No. UCRL-53196, Lawrence Livermore National Laboratory, Livermore, California. Pimentel, K. D. 1978. An environmental overview of geothermai development: the Geysers-Calistoga KGRA. Report No UCRL 524-96, Lawrence Livermore National Laboratory, Livermore, California. Russell, B. F. 1982. "Raft River wellfield testing and analysis." Geothemai Res. Council Bull. 11(4) :6-10. Spencer, S. 6., B. F. Russell, and J. F. Sullivan. 1979. Potential use of geotherma'l resources in the Snake River basin: An environmental overview. Report No. EGG-2001, EG&G Idaho, Inc., Idaho National Engineering Laboratory, Idaho Falls, Idaho. Strojan, C. L., and E. M. Romney. 1979. An environmental overview of geothermai development: the Mono-Long Valley KGRA. Laboratory of Nuclear Medicine and Radiation Biology, University of California, Los Angeles, California. Suciu, D. F., P. M. Wikoff, and R. D. Sanders. 1982. "5MW(e) water treatment plant design and systems performance." Geothermai Res. Council Bull. 11(2):14-16.
429
Suter, B. W. II. 1978. Effects of geothermal energy development on fish and wildlife. U.S. Fish and Wildlife Service, Topical Brief: Fish and Wildlife Resources and Electric Power Generation, No. 6. Report No. FWS/OBS-76/20.6. Thurow, T. L., and L. S. Cahn. 1982. Final environmental report: INEL geothermal environmental program. Report No. EGG-2215, Volume I, EG&G Idaho Inc., Idaho National Engineering Laboratory, Idaho Falls, Idaho. Thurow, T. L., C. M. White, R. P. Howard, and J. F. Sullivan. 1980. Raptor ecology of Raft River Valley, Idaho. Report No. EGG-2054, EG&G Idaho, Inc., Idaho National Engineering Laboratory, Idaho Falls, Idaho. Tucker, F. L., and L. R. Tanner, eds. 1978. Proceedings, Geothermal Environmental Seminar—78. Environmental Systems and Services, Kelseyville, California. Tullis, J. A. and M. R. Dolenc. 1982. "Geoscience interpretations of the Raft River resource." Geothermal Res. Council Bull. 11(2):6-9. Whitbeck, J. F., and R. R. Stiger. 1982. "5MW(e) Raft River Pilot Plant description and operation." Geothermal Res. Council Bull. 11(2):9-14.
430
7G
ENVIRONMENTAL PROGRAM AT THE MHD COAL FIRED FLOW FACILITY T.P. Lynch, J.A. Cooper, J.H. Hansen, R.C. Attig The University of Tennessee Space Institute Tullahoma, Tennessee
ABSTRACT The environmental program at the University of Tennessee Space Institute (UTSI) Coal Fired Flow Facility (CFFF) integrates emissions monitoring, ambient air monitoring, water quality monitoring, and terrestrial monitoring into a program designed to follow the MHD technology from the raw coal processing to the ambient environment. After approximately a two year monitoring period the environmental effort was aimed more intensely at complementing the process gas and effluent monitoring with ambient monitoring. With baselines complete, studies are now being done to assess potential impacts of full scale MHD plants and to develop methods for controlling effluents before they reach conventional downstream pollution control devices. The trends observed thus far from monitoring the MHD gas stream and water effluent indicate high temperature, high sulfur coal combustion to be very feasible energy technology for the future.
INTRODUCTION The U.S. Department of Energy (DOE) Magnetohydrodynamics (MHD) Coal-Fired Flow Facility (CFFF) is located on Woods Reservoir at the University of Tennessee Space Institute (UTSI), Tullahoma, Tennessee. The University of Tennessee Space Institute has been engaged in the development of coal-fired MHD electrical power generation over the past decade (Dicks 1978). Construction and activation of the CFFF was completed in the Fall of 1981 and it became fully operational in early 1982. Operation of this experimental test facility provides an opportunity to obtain environmental data associated with MHD technology without any anticipated negative impact on the surrounding environment. Part of the role of UTSI, as participants in the DOE programs is to document environmental aspects of coal-fired MHD. The main objectives of the environmental program at UTSI are impact assessment and regulatory compliance. The emphasis of the program is
431
two-fold: ambient data documentation and source monitoring from CFFF operation. In an effort to document MHD power generation as an environmentally acceptable energy technology, water quality, terrestrial ecology, and air quality/meteorological studies are conducted. An effort has been made to generate an adequate data base in each of these areas to permit environmental impact assessment after the CFFF has been in operation for a sufficient period of time. The overall objective of the current test series at the CFFF is to provide definitive and conclusive experimental confirmation that sustained control of N0 x and SO? effluent levels in a coal fired MHD plant can be achieved. Tests in this series are normally short, with coal firing periods of about four hours or so, but the final test in the series will be substantially longer. To achieve this major objective, several subordinate objectives must be realized as well. These subordinate objectives include the following: 1. Conduct warm tests at low flow rates to establish the flow train gas dynamics and verify procedures and instrumentation for safe operation of previously untested hardware components. 2. Operate with coal/No. 2 fuel oil mixtures over a range of conditions to evaluate instrumentation and identify component and system responses. 3. Operate the test train at the nominal design conditions to verify predicted performance and establish suitable flow conditions for MHD power production. 4. Demonstrate combustor operation suitable for power production and verify short term responses consequential to the MHD interaction. 5. Evaluate overall system and individual component performance under repeated and extended operation. 6. Evaluate the effects of varying primary combustor mass flow and stoichiometry on N0 x formation and relaxation. ENVIRONMENTAL PROGRAM The environmental effort at UTSI is directed at understanding factors affecting the formation and decomposition of N0 x , and measuring the levels of N0 x , SOg and particulates in the CFFF effluent. This effort was initiated during testing of 7.5 and 15MWt components at the UTSI Energy Conversion Facility and some results on emissions were described in earlier papers (Strom 1978; Attig 1980). One purpose of the experimental CFFF test program is to develop and evaluate pollution control techniques and to determine environmental impacts from effluents, particularly N0 x , SO2 and particulates under variable conditions. The experimental data will then be used to predict expected results in larger scale equipment using
432
appropriate models. Comparison of N0 x data from the CFFF and the smaller equipment in the Energy Conversion Facility has shown good agreement. The MHD program at UTSI also included establishing baseline conditions for ambient air, water and terrestrial quality. This program was developed in cooperation uith personnel from DOE and the Montana Energy and MHD Research Institute (Hall et al. 1980). A summary of the program is shown below. Summary of Environmental Study Comments Emissions Monitoring
One port in stack is used for compliance emission monitoring. Others are used to monitor process gas.
Ambient Air Monitoring SOg N0 x Particulates Meteorological Data
Three stations - one for base conditions, two stations to establish conditions in prevailing wind directions.
Water Quality Monitoring
Monitor water chemistry from three stream sites (Woods Reservoir), CFFF effluent,& holding ponds. NPDES monitoring.
Terrestrial Monitoring
Six terrestrial plots-two of which are control sites. Monitoring of soil conditions, vegetation, etc.
Air, water and terrestrial monitoring data are being accumulated so that it will be possible to assess the potential impact of the CFFF and larger MHD systems on the environment. Timely identification of potential problems and their resolution will allow the development of the MHD technology without harmful effects to the environment. Emissions Monitoring The major pollution control effort for the UTSI program has been directed toward nitrogen oxides. The extremely high combustion temperatures lead to formation of high N0 x levels, and thus it has been essential to develop techniques for reducing these levels to acceptable limits. Several control approaches were considered and the one selected early in the program was two-stage combustion. It allows sufficient time
433
in the first stage, or fuel rich zone, to relax the N0 x sufficiently to permit subsequent completion of combustion while maintaining low N0 x levels. For the current test series, the Low Mass Flow (LMF-1) test train was used. As shown on Figure 1, the components in LMF-1 include a vitiation heater in which the oxidant (O2 mixed with nitrogen) is preheated, the combustor, a nozzle, diagnostic channel, diffuser, radiant furnace, and a secondary combustor. The combustion gases then flow through a materials test module to a water quench. Particulates are removed in a wet venturi scrubber and a cyclone after which the gases pass through an induced draft fan to the stack. To assess N0 x formation and decomposition and provide sufficient information to model the process, eight M0 x analyzers are located along the low mass flow train, six in the fuel rich zone and two following completion of combustion. Since the formation and relaxation of N0 x are very temperature dependent, temperatures are also measured along the flow train. The low mass flow train cannot duplicate commercial conditions since there is no air heater installed in the flow train. However, results have substantiated previous smaller scale tests (Strom 1978; Attig 1980) and have been encouraging thus far. Results for one set of operating conditions are shown on Figure 2. In general, measured N0 x levels are close to levels predicted by The PROF Kinetics Code, developed for the U.S. Environmental Protection Agency. Measured sulfur dioxide levels are typically between 50 and 100 ppm at the stack; well below EPA guidelines. The control of SOg results from reaction of SO2 with the potassium used to produce the ionized gas in the MHD generator. If potassium in the form of carbonate is blended with the pulverized coal, it dissociates and subsequently reacts with the SO2 to form potassium sulfate which is recovered in the solid state. This reaction provides an effective means of SO2 control. Commercial MHD systems will require recycling of the potassium for economic reasons. If the potassium is to be reused for SO2 control, it must be regenerated to remove the sulfur. It should be noted that the combustor has operated with a coal/No. 2 fuel oil mixture with capability to go to 100% coal. The injection of fuel oil into the combustor is necessary to simulate full size power plant conditions at less than 100% slag carry-over. Since a high oxygen enrichment is used to achieve MHD temperatures, operation with 100% coal as the primary combustor fuel could result in sulfur and ash concentrations in the flue gases that exceed the full power plant design conditions. By varying the coal to fuel oil ratio, the CFFF has the capability to evaluate the effect of varying ash and sulfur concentration on system performance and N0 x and SOg levels. Ambient Air In order to adequately describe impact assessment from the CFFF, nitrogen 434
STACK SAMPLE
SECONDARY COMBUSTOR
H P TEST SECTIONSRADIANT FURNACE —
to CJ7
COAL & SEED DIAGNOSTIC CHANNEL
°2 +N 2
A
OIL' 1
NOZZLE ACOMBUSTOR
MATERIALS TEST MODULE ASH/SEED HOPPER
DIFFUSER
' S L A G SCREEN MV
CYCLONE
/
VITIATION HEATER INDUCED DRAFT FAN CIRCLED NUMBERS DESIGNATE LOCATION OF NO/NOX SAMPLING POINTS
FIGURE 1. LMF-1 Test Train Configuration
MOM PPM
8000-
7000LEGEND
° Etptrimtntol nmptnturas • Cokuhttd nmptnhm O Exp*im«*ol NOx Vokm • CW«*W»tf /VOr I'o/i/i'i (Equilbrim) * Co/cuhM Kinttic KQ? \
6000
K P3000
5000-
2500
4000-
1-2000 c
3000
KI500
o 2000
M0O0
-p. CO
1000
H500 •
o
o
o
Radiant Furnxi
20
FIGURE 2.
30
40
50 60 70 80 90 100 110 120 Dittanct along flow train. Ft Time 1720hours, N/O 70S, Stoichkmetry .84, ThermalInput 23.9 MWt
N0x and Temperature along LMF1 Flow Train during LMF-C3 Test, September 10, 1982
350
oxides, sulfur dioxide, particulate and meteorological data baselines have been established at three locations surrounding the CFFF stack. Location of each site was based primarily on meteorological conditions with reference to the CFFF stack. The Major Impact Site, shown in Figure 3, was located in the most prevailing wind direction, NNE of the stack. The Second Impact Site was placed in the second most prevailing wind direction, SSE of the CFFF stack, approximately 246 meters from the lake shore. The Base Site, was located approximately 11.3 kilometers from the CFFF stack in a northeastern direction to have minimal effects from CFFF operation. The air monitoring t r a i l e r s are similarly equipped except that the major Impact t r a i l e r also houses acoustic radar instrumentation for acquisition of vertical atmospheric s t a b i l i t y data. All three t r a i l e r s are equipped to maintain relatively constant temperatures for s t a b i l i t y of electronic monitoring equipment and contain the following equipment: . Two General Metal Works Model GMW 2000 H High Volume Samplers for collection of total suspended particulate (TSP) and inhalable part i c u l a t e (IP) . Thermo-Electron Corporation Model 14B/E Continuous Analyzer for nitrogen oxides (N0-N02-N0x) . Thermo-Electron Corporation Model 43 SO2 Continuous Analyzer . TechEcology, Inc. Wind Speed Sensor . TechEcology, Inc. Wind Vane Horizontal Wind Direction Sensor . Yellow Springs Instrument Co., Inc. Model 2014-24.3/31.5-HA-3-WH Barometric Pressure Sensor Campbell Scientific Corporation SO;?/NOX Data Logging System .
Campbell Scientific Corporation Meteorological Data Logging System
.
Met One Incorporated Dew Point Sensor
All parameters except the high-volume particulate samplers are measured and stored automatically on paper and magnetic tape for retrieval on the UTSI VAX-11/780 computer. The data is then reduced and plotted as shown in Figures 4 through 7. Data for the baseline period (August 1980-July 1981) showed a l l values below EPA standards, as expected. Meteorological data taken verified horizontal wind speed and direction to be predominately from SSW with secondary wind directions being NNW and NE. .Water_Qua_l_i_ty Program
The water quality environmental program is designed to assess any 437
AEOC
Access Boa*
•MtftraMTUiMkMtOW
Second Impact Sit*
WATER QUALITY
~*"AIR QUALITY * TERRESTRIAL QUALITY + COAL-FIRED FLOW FACILITY HOLDING PONDS
CAPITOL HILL TENNESSEE N3515-waBOO/7 LR d/BS
FIGURE 3. CFFF Environmental Monitoring Sites
438
3 HR MOVING AVERAGE CHEMISTRV DATA SITE 1 (IMPACT S I T E )
150
SEP 0CT NOU PEC UAN FEB MAR APR (NO; 24 MR STANDARD)
iJUN
NITROUS OXIDES (PPB) 100-
50 —
(AAM)
W<
I
•ES ^
0 FIGURE 4.
NOX Baseline Data for August 1980 through July 1981
HOURLY AUERAGED CHEMISTRY DATA SITE 1 (IMPACT SITE) 150
*UG SEP OCT NOU PEC tfAN FEB MAR APR MAY vJUH >JUL "(24 HR)
SULFUR
DIOXIDE (PPB)
—
r
-
"(AA i
0 FIGURE 5.
¥i
?f : i r1 •
SO2 Baseline Data for August 1980 through July 1981
439
TOTAL SUSPENDED PARTICULATES MONTHLY AVERAGES
100. M I C R 0 G R A M S
80' S T A T I 0 N
C
u
0 6
B I C M E T E R 0 AUG
OCT
DEC
FEB
APR
JUN
1980-1981
FIGURE 6. Typical Total Suspended Particuiate (TSP) Plot for August 1980 through July 1981
440
AUG
UINDROSE SEP 80
0 TO 8 PI/SEC X TIME 44.510
NORTH UEST-
8 TO 16 fi/SEC % TIME 38.281
•EAST
SOUTH
16 TO 34 M/SEC % TIME 13.368 Figure 7.
24 TO 38 M/SEC X TIME 3*841
Windrose for Major Impact S i t e . Wind Direction Prevalence in Percent of Time for Given Speed Intervals January 1981
potential impact of the CFFF effluent on Woods Reservoir and to comply with environmental legislation applicable to MHD technology. Overall objectives of the program are to: l)maintain existing stream conditions, 2) characterize and monitor wastewater from the CFFF, and 3) assure regulatory compliance with water quality standards applicable to stream usage. Water samples were initially collected weekly to establish a baseline and are now collected monthly at selected monitoring sites (Figure 3) and analyzed in the CFFF Chemistry Laboratory for the parameters listed in Table I. Dissolved oxygen, pH, temperature, and conductivity are measured in the field with a Hydrolab unit at all monitoring sites. A one year baseline documentation has been completed and a report entitled "MHD Coal Fired Flow Facility Baseline Water Quality Study" has been issued (Cooper 1980). Water samples are taken from an upstream control site (station #4728G4) to characterize the water quality of the stream before it reaches the CFFF discharge site (station #4728G1). Below the discharge site samples are taken in the Rollins Creek Bay area (station #4728G6) to assess the water quality immediately below the entrance of the CFFF effluent. By establishing water quality conditions upstream prior to the entry of the CFFF effluent, data comparisons can be made at and below the entrance of the effluent and any potential impact of CFFF effluents on the stream can be assessed. Samples are collected at station #4728G7, which is a sufficient distance downstream from the discharge entry to allow mixing and dilution of the effluent in Woods Reservoir. All wastewater except sanitary wastewater from the CFFF enters a holding pond system consisting of two concrete ponds that are approximately 15m x 30m x 1.2m, located in tandem in a ravine area sloping toward Rollins Creek (Figure 8 ) . This retention pond arrangement has a total capacity of 1135.5m3 and allows for chemical treatment of the wastewater and settling of solids to regulated levels. A second pond is available as a reservoir in case of heavy rains, emergency spills or maintenance. During operational periods at the CFFF, samples from the holding ponds are also analyzed for the parameters listed in Table I prior to and following the test runs. These pond samples are compared with monthly analyses to assess any potential facility impact on the aquatic environment and to alert the environmental staff of potential non-compliance parameters. Terrestrial Monitoring The terrestrial monitoring program has been designed to determine if operation of the CFFF has an impact on the terrestrial area surrounding the facility. The program consists of the following elements: (1) chemical analysis of leaf and soil samples, (2) observation and evaluation of plant health and, (3) measurement of tree diameters and heights. The study is limited to vegetation and soil effects. No data on animals or birds living in or passing through the area has been or is planned to be collected. To obtain the data for the terrestrial program, six plots measuring 100 meters by 40 meters have been established (Figure 3 ) . Four are located
442
Table I. Characteristics of Wastewater From the CFFF Following Holding Pond Retention Effluent (Site 4 728G1), mg/i£ November 81
Parameters
September 82
Temperature (Air)
18.2
24.0
Temperature (Water)
15.0
25.0
Conductivity
80.0
140.0
Dissolved Og
7.0
6.8
PH
7.4
7.3
Hardness
98.0
73.0
Alkalinity (Total)
90.0
60.0
5.6
3.5
Fluor1de
0.12
0.30
Sulfate
8.7
4.0
Nitrogen (NH3-N)
0.05
0.40
0.02
0.07
Settleable Solids
1.0
0.01
Suspended Solids0
2.7
0.00
Chloride b
Phosphate b
137.0
Total Solids 0
0.03
Arsenic
15.0
Calcium
109.0 0.003 13.85
Iron
0.55
0.33
Lead0
<0.01
0.009
Magnesium
2.26
2.05
Potassium
0.63
1.62
Silicon
0.06
0.27
Sodium
7.33
1.81
* All measurements In milligrams per liter (mg/£) except for temperature (°C); Conductivity (iimhos/cm); and pH. 0
Not Included in original baseline measurements.
443
Coo! handling •yttcmrun off
/I
I,
Natural run off own.-95% .continuoutly racyctad Pump houM
Storm run Off
,o/ 0 _ 5 0 gpm
Aimotphir*
North not toscole
Make 2 5 0 f l M L _ v f * .OUTi.lT »'«llCTJ«t
T) Rollini Lake
I
Coolina 1&wer
I
Lake pump house
FIGURE 8. CFFF Wastewater Control
n
along radial lines of the CFFF stack in the most prevalent and second most prevalent wind directions. Two control plots are located nearly 10 kilometers from the f a c i l i t y . On each of the six plots, diameter and height measurements of a l l trees greater than 15 cm in diameter are recorded at a point 1.3 meters above ground l e v e l . Leaf samples are taken and color determined with reference to the Munsell Color System. Disease signs and symptoms, i f any, are noted. A representative number of leaves are collected from trees from each plot for chemical analyses of the following elements: total fluoride, total sulfur, arsenic, cadmium, chromium, mercury, potassium and lead. Soil samples are taken from each plot and analyzed for the following elements and compounds: sulfate, n i t r a t e , arsenic, cadmium, chromium, mercury, potassium, lead, and phosphate. Soil and veget a t i o n sampling is accomplished in a design which f a c i l i t a t e s plot to plot and year to year s t a t i s t i c a l comparisons of results.
445
REFERENCES CITED Attig, R.C., et al. 1980. Air Pollution Control with a Coal-Fired Magnetohydrodynamic System. Presented at th"e Second Conference on Air Quality Management in the Electric Power Industry, Austin, TX. Cooper, J., 1980. MHD Coal Fired Flow Facility Baseline Water Quality Study—Woods Reservoir. D0E/ET/10815-T3, U.S. "Department of Energy. Dicks, J.B., et al. 1978. Coal Fired MHD Power Generation-Including Balance of Plant. Presented at Fourth US-USSR Colloquium on MHD Electrical Power Generation, Washington, D.C. Hall, L., et al. 1980. Recommended Environmental Monitoring Program for The MHD Coal-Fired Flow Facility. DE-AC07-79-ID-12036, U.S. Department of Energy. Prepared by the Montana Energy & MHD Research and Development Institute, Inc., Butte, Montana. Strom, S.S., et al. 1978. Controlling N0 x from a Coal-Fired MHD Process. Presented at 13th Intersociety~Conference for Energy Conversion, San Diego, CA.
446
7H MEASUREMENTS AND MODELING OF GAMMA ABSORBED DOSES DUE TO RELEASES FROM A LINEAR PROTON ACCELERATOR: EXPERIMENTAL DESIGN AND PRELIMINARY RESULTS
Brent M. Bowen, Thomas E. Buhl, Jean M. Dewart, Wayne R. Hansen Daniel Talley, Anita I. Chen, William A. Olsen, Donald M. Van Etten Environmental Surveillance Group (H-8) Los Alamos National Laboratory Los Alamos, New Mexico
ABSTRACT External radiation levels due to positron annihilation radiation from U C , l 3 N, and 1 5 0 released by the 800 MeV linear proton accelerator at the Los Alamos Meson Physics Facility (LAMPF) have been monitored at a fence-line location both by thennoluminescent dosimeters (TLDs) and high pressure ionization chambers (HPICs). The accelerator is located in irregular terrain consisting of mesas and canyons. Fifteenminute, accumulated external radiation levels were recorded with the HPICs. Instruments on a nearby meteorological tower concurrently measured wind speed and direction at three levels, temperature at two levels, solar radiation, and rainfall. Real-time radionuclide release rates and stack velocities were measured at the release point with in-stack monitors. This paper presents analyses of short-term radiation levels using HPICs and long-term levels using TLDs. Work being done to develop a computer model to predict external radiation levels based on meteorological data is also discussed. I.
INTRODUCTION
As part of the Environmental Surveillance Program at the Los Alamos National Laboratory, external gamma radiation levels due to releases of short-lived radioactive gases from a linear proton accelerator have been monitored routinely for the past six years. The Los Alamos Meson Physics Facility (LAMPF), an 800 million electron volt (MeV) linear proton accelerator, emits radioactive air activation products. The highest observed fence-line dose equivalent resulting from Laboratory operations in 1981 was due to emissions from LAMPF (Env. Surv. Group, 1982). The maximum annual fence-line dose in 1981 due to the LAMPF operations amounted to 17 mrem, or 15% of the natural background. The 447
maximum individual dose equivalent was 4.8 mrem, 4% of the natural background, and 1% of the Department of Energy (DOE) Radiation Protection Standard. During the summer of 1982, three portable, high-pressure ionization chambers (HPICs) were placed in the field to measure short-term gamma radiation levels caused by the LAMPF plume. This was in addition to the thermoluminescent dosimeter (TLD) network that routinely measures longterm gamma levels. Instruments on a nearby meteorological tower concurrently measured wind speed and direction at three levels, temperature at two levels, solar radiation and rainfall. The purpose of this paper is to present the preliminary results of the modeling of gamma doses* from airborne radionuclides emitted by LAMPF. Airborne radiation levels predicted from source terms and onsite meteorological data for short-term (15 min to 1 day) and long-term (5 month) time periods are compared with measured results in this paper. In addition, the experimental design used for the study and future plans are presented. II.
BACKGROUND
The Los Alamos National Laboratory is located on the Pajarito Plateau on the eastern flanks of the Jemez Mountains. The Sangre de Cristo Mountains are nearly 70 kilometers (km) to the east. The Plateau slopes from the base of the Jemez "•fountains [~2500 meters above sea level (MSL)] east-southeastward down to the Rio Grande (~1700 MSL) over a distance of 25 km. There are numerous alternating "finger" mesas and canyons running along the slopeline of the Plateau. The canyons are 50100-meters (m) deep and 100-200-m wide, while the mesas vary from 200600 in in width. LAMPF is located on a rnesa top at the southern edge of Los Alamos Canyon. The LAMPF release stack is located about 700 in from the closest fence-line location. Three short-lived, gaseous, radioactive air activation products, 1 5 0, l l C , and 13 N, are principally emitted from the stack. Because of the annihilation of positrons from the decay of these radionuclides, gamma radiation is emitted from the plume of gases as i t travels downwind from the stack. The amount of radiation is affected by the downwind distance and the h a l f - l i f e of the isotopes. Some characteristics of these three radioactive gases are given in Table 1. Note that although less n C is emitted than 1 5 0, the h a l f - l i f e of n C is nearly 10 times that of 1 5 0. This means that for a 3.5 meters per second (m/s) wind, more of n C remains at the fence-line distance (~700 in) than 1 5 0 . The contribution of 13N to the total gamma radiation is small because i t comprises only a small percentage of the releases and has a r e l a t i v e l y short h a l f - l i f e . *For convenience, a l l doses reported here have been calculated as absorbed dose in tissue unless otherwise indicated. 448
TABLE 1. Source Terms and Amount of Gaseous Activation Products Remaining at 700 m Downwind (Q )
C;ASKOI;S KMISSION DATA AT 700M (m/s)
GASEOUS RAT!-:
ACTI VA'l'ION
(Ci. h)
PRODUCTS
III. A.
(s)
3.5
7.0
10.0
is0
57
36.4
123
32
57
67
"C
36
23.2
1230
89
95
96
7
41
600
79
89
92
METHODOLOGY Siting and Instrumentation
1. Source Terms The experimental design is shown in Fig. 1. The stack is 30-rn high and has a diameter at the top of 0.9 m. A Kanne-type air-ionization instrument is used to monitor the radioactive gases in the stack discharge. A voltage is recorded, which is correlated to the radio3 nuclide air concentrations in Curies per cubic meter (Ci/m ) in the stack gas. The percentage composition of 1 5 0 , U C , and 13 N in the stack gas has been determined previously by analyzing bag samples of the stack gas over a few minutes. Stack velocity is measured by an anemometer placed halfway up the stack. 2.
Meteorology
One meteorological tower is about 100 in from the stack. A threeway propeller system measures the horizontal and vertical winds at a height of 23 m. Another meteorological tower is 725 m across Los Alamos Canyon from the stack at a bearing of 12°. This is the closest fence-line location to the stack. This tower has three-way propeller systems at heights of 1.3, 4.0 and 12.0 m, which measure horizontal and vertical winds. Also, thermistors equipped with blowers measure air temperature at 1.3 and 12.0 m. Solar radiation and rainfall are also recorded at this site.
449
LEGENth
T
HET TO»ER
H
HPIC
X 0
TLD 100
T
o 200
STACK
SCALE (METERS)
Figure 1. Map of Experimental Design 3_. Gamma Radiation Monitoring A series of 12 lithium fluoride (LiF) TLDs monitors long-term gamma doses on a routine basis. They are 70 m apart, subtending an angle of 5° from the stack. Another TLD network unaffected by Laboratory operations provided the background absorbed dose level for external radi at ion. Three HPICs were installed in June, 1982, to measure real-time gamma exposure rates (converted to absorbed dose rates in tissue for comparison with the model and TLD results). One is located at the tower, with the other two located 90 m on each side. This corresponds to an angle of 7° between each HPIC as viewed from the stack. Unfortunately, due to severe weather and electrical malfunctions, all three HPICs did not operate simultaneously for any length of time. A severe lightning storm in July, 1982, damaged the two HPICs away from the tower. B. Data All data were collected by microprocessors, except for the TLD data. One microprocessor processed the source-term data and meteorological data collected near the stack. Another processed the meteorological data and the HPIC data from across the canyon. Both microprocessors provided 15-min averages and standard deviations of all wind data. 450
C. Modeling A Gaussian-type atmospheric dispersion model was used to predict absorbed doses in tissue from gamma radiation. The following equation from Meteorology and Atomic Energy was used for short-term predictions (Healy, 1968): 1
here
D(x,y,O)
=
2-JLi A £_ , (l) u D is the dose (rads), y is the t o t a l gamma-ray absorption coefficient (1/m), u, is the energy-absorption coefficient for gamma rays _ (Vm), E is the average gamma energy emitted at each d i s integration (MeV/Dis), Qx is the amount of source material remaining in a plume after a travel distance x (Curies), I j + k l ^ ( ° r I j ) is an integral accounting for radiation __ away "from the plume and u is the average wind speed (m/s).
A s l i g h t l y d i f f e r e n t equation was used to predict long-term radionuclide concentrations and resulting doses for 22.5° sectors. Since the three gases have d i f f e r e n t h a l f - l i v e s and release rates, the equations were used to calculate a gamma dose for each isotope. The individual gamna doses were then summed to obtain a total gamma dose. The I j inteqrals for both short- and long-term periods were determined from nomograms (Healy, 1968). The standard deviations of the d i s t r i b u t i o n of qases in the plume, ay (horizontal) and az ( v e r t i cal) were determined d i r e c t l y from the onsite meteorological data using a method suggested by Draxler (Draxler, 1976). The equations used are:
oy = aQx fx (17^)
(2)
a = a £ - f (T/T.) (3) z wu 2 i where a^ is the standard deviation of trie horizontal wind direction (radians), a w is the standard deviation of the vertical wind speed (m/s), x is the downwind distance, u is the average horizontal wind speed, and fj and fo are functions of downwind distance and thermal stability (dimensionless). Since the nomonrams of Ip for short term cases require a sinqle a (i.e., an isotropic plume), a was approximated by the geometric average of a y and a z : 0 =
451
IV. MODELING RESULTS AND DISCUSSION A. Long Term Absorbed doses from gamma radiation were measured and predicted for a five-month period (4 June - 3 November, 1982). The frequencies of wind direction and speed for this period at the tower across Los Alamos Canyon (State Road 4) are shown in a wind rose in Fig. 2. The prevailing winds are from the SSW and SW for this period. These two directions occur nearly 33% of the time. Much of the high frequency of these two wind directions is caused by a strong mountain-valley wind up the Rio Grande Valley (Bowen, 1981). Also, note the much smaller frequencies of S and SSE winds. It would be expected then that higher gamma radiation levels due to LAMPF airborne emissions would occur on the east side of the TLD network. Figure 3 shows the measured and predicted gamma absorbed doses for the NNW, N, and NNE sectors at the fence-line location. Both measurement and modeling show higher absorbed doses do indeed occur on the east side of the TLD network. The measured external radiation contribution from LAMPF was calculated by subtracting the background external radiation dose (measured by another TLD network unaffected by Lab operations) from measured external radiation doses in each of the three sectors along the fence-line. However, a -2.8 mrad measured contribution is shown for the NNW sector. This negative number may be due to the high degree of spatial variability of the background external radiation dose. (The standard deviation of the background dose was about 8 mrad, about equal to the absorbed dose expected fron LAMPF operations.) Subtracting a large background from a large total absorbed dose, both with large variances due to spacial variability, probably causes this negative number. Note that the highest absorbed dose is predicted for the NE sector. Installation of TLDs in this sector is planned in the future. B.
Short-Term Modeling
The short-term spatial and temporal variability of gamma radiation levels can be shown with the use of HPICs. Figure 4 shows a daily plot of 15-minute averages of the east and west HPIC external radiation exposure readings (the tower HPIC was inoperative) along with temperature, vertical wind and horizontal wind direction and speed. Note the striking peak of 140 micro Roentgens per hour (pR/h) at the west HPIC without a corresponding peak at the east HPIC. The westward movement of the plume (i.e., higher concentrations) can be seen as the wind shifts counterclockwise at 0700 and 1600. Note that these two monitors are 180 m apart or 14° apart as viewed from the stack. Both monitors show a background exposure level of about 11 yR/h at the end of the day when the wind becomes westerly.
452
STATE ROAD 4 4 JUNE - 3 NOV 82 1-25 2.5-5 si-
2. Wind Rose for 5-Month Period at Fence-Line Location (State Road 4) Site PREDICTED VS MEASURED RADIATION DOSE (MREM) DUE TO LAMPF EMISSIONS BY SECTOR FOR PERIOD 4 JUNE TO 3 NOVEMBER, 1982
LEGEND 0.0
MEASURED RADIATION DOSE
(.AMPF
(0.0) PREDICED RAD'ATION DOSE
STACK
X
TI.D LOCATION
453
10MZ1NG RflOIHTIOh AT 1.0 rCTEB TH. IN 1ICR0 tOCNTGOiS; 6'.6 100 1 D M I D O T i 6 f»T I0HERIMLIDI S2S
100 "
HCSTIWSHI
Figure 4. Typical Day on Which Gamma Radiation is Received Due to the Presence of the LAMPF Plume Hourly gaimna absorbed doses were modeled on four days for the tower HPIC. Comparison of the predicted and measured doses is shown in Fiq. 5. Only a weak positive correlation with large scatter is evident. However, the highest predicted values are close to the highest measured values. This may indicate that a prediction over such a short ti.-.ie period (1 h) is only successful for the general area, and not successful for a fixed point. Daily model predictions were made for 1.1 days on which all necessary data was available and measurable absorbed dose was obtained. Figure 6 shows the comparison of predicted and measured daily gamma doses. There is a good correlation between predicted vs measured value. Note that the model underpredicts in 9 of the 11 cases, so that some refinement of the model may be necessary for application in the irregular terrain at Los Alamos. It appears that daily doses can be rather accurately predicted based on the limited amount of data presented here. V. SUMMARY AND CONCLUSIONS A network of monitors measuring gamma radiation for short- and long-time periods were used with meteorological and source-term data to predict gamma radiation emitted by air activation products released by a linear proton accelerator. Long-term predictions tend to agree with measurements. However, the spatial variability of background 454
PREDICTED VS MEASURED HOURLY GAMMA DOSES (,uRAD) 60-
•
50 H
40^
*T •* • * •
•I.
10-
1
0
I0
'
i
^0
:30
I
r
-10 50 MEAWRED
-1
t
1
60
70
80
Figure 5.
PREDICTED VS MEASURED DA1LV GAMMA DOSES ( M RAD) 350-
200-j
-150-
I °-
•*
'
.
^
50
100
150
Y=0.54X+51
R=0.82 P=99%
; too -i
200 MEASURED
250
300
350
Figure 6. external penetrating radiation is larger than the contribution of the source, thereby castinq some doubt on the results. Hourly predictions show much scatter compared with measured values, but show sinilar peaks as the measured values. The predicted daily values are shown to be strongly correlated with measured daily values. It is hoped that
455
modeling results from the LAMPF releases inay assist in modeling potential releases from other facilities at the Los Alamos National Laboratory. VI. FUTURE PLANS AND RECOMMENDATIONS New electronics are being developed for the HPICs to improve their operation and reliability. With long-term reliability, the HPICs can be used to determine more precisely the background external radiation doses. This will allow a better determination of the accuracy of longterm modeling. A larger data base for making daily absorbed dose predictions will also then be available. TLDs will be located farther east on the fence-line where long-term modeling has indicted slightly larger values. A model is currently being developed that will compute Ij values, relieving the tedious and slow 'work of deriving them from nomograms. Finally, further analyses will be performed on the stack gases to better estimate the ratios of the three radioisotopes: n C , s 13 * 0, and N. ACKNOWLEDGMENTS We wish to thank Jerry Miller and Bob Dvorak of the Health Physics Group (H-l) of Los Alamos for their cooperation and help with the source term information. Thanks also to Alan Stoker, Deputy Group Leader of the Environmental Surveillance Group (H-8), for his helpful suggestions and encouragement. Finally, special thanks also go to Mary Lou Keigher and Lois Schneider for the skillful and speedy preparation of this manuscript. This work was- done under the auspices of the US Department of Energy and the University of California. REFERENCES Draxler, R. R. Parameters."
1976. "Determination of Atmospheric Diffusion At_. Env_. 10:99-105.
Environmental Surveillance Group. April 1982. "Environmental Surveillance at Los Alamos During 1981." Los Alamos National Laboratory report LA-9349-ENV. Bowen, B. M., .1. M. Dewart, and F. G. Fernald. 1980. "A Study of the Nocturnal Drainage Flow Over a Sloping Plateau in North-Central New Mexico." Proceedings of Second Conference on Mountain Meteorology, November 9-12, 1981. pp. 225-232. Am. Met. Soc. Healy, J. W. and R. E. Baker. 1968. "Radioactive Cloud-dose Calculations" in Meteorology and Atomic Energy. US Atomic Energy Commission.
456
SESSION EIGHT GROUND-WATERING MONITORING AND ASSESSMENTS
8A
MONITORING FOR ORGANIC CONTAMINANTS IN GROUND WATER
D. A. Myers* and J. M. Meuser* Pacific Northwest Laboratory Richland, Washington 99352
ABSTRACT Ground-water monitoring at DOE sites has been predominantly oriented toward the collection and interpretation of radiological data. In addition, there is cause for concern over the potential for ground-water contamination by organic chemical constituents entering the environment from energyrelated activities. Many of these constituents are potential carcinogens that exhibit low threshold limits, and have halflives that approach infinity. Monitoring for these chemicals requires special tools, techniques, and procedures to assure the quality of samples. Well structures must be nonreactive with the monitored species. Sampling procedures must minimize air contact. Requirements for preserving samples may range from simple refrigeration to the use of sealed septum bottles. Because the volatile nature of some of these contaminants, the sample may require special handling in the laboratory as well. The sample must be analyzed as soon after collection as possible. Laboratories certified for the necessary analyses are few and are often located far from the area of sample collection. Pacific Northwest Laboratory (PNL) has developed strategies that assure the proper handling of samples from collection to completion of analyses. These plans have been used for a number of Department of Defense studies and can easily be used for other similar efforts. INTRODUCTION Ground-water monitoring is entering a new age. No longer can we be content with relatively simple analyses for inorganic materials found in nature or even with relatively complex analyses for radionuclides (naturally •Senior Research Scientist and Research Scientist
459
occurring or anthropogenic). Today we must try to monitor various chemical species that are totally manmade, many of which are suspected of causing cancer from exposure to concentrations as low as a few parts per billion (micrograms per liter). The unfortunate difference between these compounds and the much-feared radionuclides is that the manmade compounds do not go away as time passes. THE SAMPLING POINT Special techniques and methods are necessary to monitor these manmade chemicals compounds. A great deal of effort must be expended before the first crew is sent out into the field to collect samples for laboratory analyses. Wells The most common means of obtaining a ground-water sample is from a well. However, wells drilled for sampling inorganic compounds and radionuclides may not provide good samples for trace analyses of organic species. In addition, the use of mild steel casings and galvanized screens, or, worse yet, perforated casings, can result in anomalous analytical results if the contaminants are stripped or if the contaminants react with the metal casing, causing changes in the contaminants before analysis. Thus, to assure analytical integrity, inert or nonreactive materials must be used in well construction. Of the materials available on the market today, the most appropriate materials are less expensive than mild or stainless steel. At PNL we have studied several hazardous waste sites for the U.S. Environmental Protection Agency (EPA) and for the U.S. Army Toxic and Hazardous Materials Agency (USATHAMA). Both of these agencies approve polyvinyl chloride (PVC) for well materials and linear polyvinyl chloride (CPVC) may be approved in the near future. We recommend that well materials made of these compounds be joined mechanically rather than by the solvent/ cement methods commonly used for household plumbing, as the solvents can bias the analytical results. Numerous manufacturers now offer flush joint, screw-together casing and screens for monitoring well installation. Commonly used casing and screen diameters range from 2 to 6 in. When these rather fragile materials are used, an oversize hole must be drilled and the screen and casing placed after drilling; this makes use of cable tool or percussion drilling methods difficult. Rotary drills and large-diameter, hollow-stem augers, where available, have proved most satisfactory. Well completion is as important, if not more so, than the method of drilling. Wells must be completed in a manner which eliminates them as potential contaminant pathways to the ground water. At the same time, valid samples from the aquifer being tapped must be attainable. An engineered gravel or sand pack around a preselected screen size is the most economical means of assuring the best well completion. A pure silica packing material is recommended. These materials should be rounded to
460
subrounded grains in order to install readily with minimal bridging of the annular space between the casing and the well bore. This silica packing is best installed to a height approximately 3 to 5 ft above the top of the screen. Immediately over this sand pack, a bentonite clay seal should be installed with a minimum thickness of 5 ft. The remaining annular space should be grouted with neat cement or a cement/bentonite mixture via a tremie pipe until the grout reaches the surface (Figure 1 ) . If the depth of the well is much greater than 100-150 ft, then casing materials other than plastic should be considered. This is due to limited collapse strength inherent to PVC or CPVC casings.U) Because of the brittle nature of plastic well casings, a protective outer casing made of steel should be set around the well while the cement is still workable. Further protection can be provided by installing 3 or 4 wood or steel posts about 3 ft from the casing. After the well is installed, it should be developed by pumping, surging, blowing, or other appropriate means until the water is clear and clean of sediment and drilling fluids. A development technique should cause no more than minimal changes in the ground-water chemistry, and should be delayed until the cement grout has set. In some instances this last step is difficult to complete within reasonable periods of time. In aquifers of low permeability, well development may be restricted to the time required to pump or bail the well to dryness several times over a period of days. Tight-fitting caps should be placed over the sampling well. A cap that is threaded to match the casing threads is ideal because it will seal the well and prevent foreign matter from entering the well between sampling trips. SAMPLING An analysis is only as good as the sample on which it is made. This statement is particularly true in the case of volatile organic carbon (VOC) compounds. Because these compounds are susceptible to removal as a result of partial pressure differences present in an open borehole, the sample must be collected from water drawn into the well immediately before actual sample collection. Environmental Protection Agency guidelines call for the removal of 3 to 5 equivalent volumes (EV) of water from a well before sampling to minimize the influence of contact with air and casings. This guide is particularly valid when sampling for VOCs. All collection methods should cause as little disturbance of the sample as possible and direct airlift sampling should never be used. Appropriate sampling methods use: bailers; swabs; electric submersible, centrifugal, or turbine pumps; noncontact airlift pumps; and specially designed jet pumps. Pump lubricants must not be allowed to contaminate samples. Vacuum pumps should be avoided because they will degas the sample, which could cause analyte loss or secondary chemical changes. Ideally, individual sampling services for each well are used to avoid cross-contamination. All equipment must be thoroughly washed and rinsed with distilled water between sampling points.( 2 )
461
Containers for organic samples must ba cleaned to eliminate outside contamination. Bottles, as received from suppliers often have an oily coating that could interfere with chemical analyses. At PNL, we use modification of the EPA-recommended cleaning procedure for all environmental samples, including quality control (QC) samples created in the laboratory. The containers are washed with detergent, rinsed with distilled water, serially rinsed with solvents of decreasing polarity, and finally heated in a muffle oven (see Table 1 ) . Careful planning is thus necessary to ensure that sufficient containers are available during sampling. No longer is it possible to pick up canning jars or milk bottles as they are needed. Two types of containers are used for organic samples: amber glass bottles for most analytes, and 40-ml septum-cap vials for VOCs. Plastic (polyethylene) bottles are not suitable because a variety of organic residuals may contaminate the sample. At the very least, phthalate esters (plasticizers listed as U. S. EPA "Priority Pollutants") will cause sample contamination. Teflon-lined caps for the bottles are preferable to foil lining because they can provide a better seal. A tight seal will prevent possible sample contamination by organics in the air (e.g., vehicle exhaust, pesticide spraying) and sample loss or mixing during shipment. After the requisite 5 EV are removed from the well, the sample is placed in the appropriate containers; sufficient volumes are obtained for all analyses of interest. Samples should not be exposed to the atmosphere or agitated any more than is necessary; this is especially critical for VOCs. The sample vial must be filled to overflowing, without aerating the sample, and capped so that no bubbles are trapped within the vial. Obtaining a nonaerated sample is especially difficult when commonly available pumps are used. All samples should be preserved immediately as appropriate for the analytes of interest and placed in cool containers. Ice chests containing a coolant are easily handled in the field. A partial listing of preservation techniques and holding times is presented in Table 2. VOC samples must be stored in an inverted position, out of any standing water in the ice chest. Once again, careful planning is necessary to ensure that sufficient samples are obtained for all analytes of interest. CHEMICAL ANALYSIS After the cooled samples are received at the laboratory, they must be prepared/analyzed as rapidly as possible. Recommended holding times generally are untested guidelines designed to reduce the likelihood of secondary chemical reactions or analyte degradation. When guidelines are not provided, the time from sampling until preparation/analyses should not exceed seven days. The laboratory selected to perform the chemical analyses should be thoroughly familiar with the approved techniques and methods for environmental samples. When monitoring for organics, tested and approved methods
462
TABLE 1. Recommended Procedures for Cleaning Sample Containers a.
b.
c.
Amber bottles 1)
Soak bottles in detergent for one day.
2)
Scrub to remove deposits of foreign materials.
3)
Rinse with copious amounts of distilled water.
4)
Rinse with acetone.
5)
Rinse with methylene chloride (nanograde).
6)
Rinse with hexane (nanograde).
7)
Heat to 200°C.
8)
Allow to cool.
9)
Cap with clean caps with teflon liners.
Bottle caps 1)
Remove paper liners from caps.
2)
Wash with detergent.
3)
Rinse with distilled water.
4)
Dry at 40°C.
Teflon liners (avoid contact with fingers) 1)
Wash with detergent.
2)
Rinse with distilled water.
3)
Rinse with acetone.
4)
Rinse with hexane (nanograde).
5)
Air dry.
463
TABLE 2. Containers, Preservation, and Holding Times* Maximum(c) a
Measurement
Container( )
P P P
Acidity Alkalinity
Ammonia
Preservative(b) Cool, 4°C Cool, 4°C
Cool, 4°C
Holding Time 14 days 14 days 28 days
H2SO4 to pH 2 Biochemical oxygen demand Biochemical oxygen demand,
P P
carbonaceous Bromide
Chemical oxygen demand
Cool, 4°C Cool, 4°C
48 hours 48 hours
P
None required
P
Cool, 4°C H2SO4 to pH 2
28 days 28 days
Chloride Chlorine, total residual
P
None required Determine onsite
Color Cyanide, total and amenable
P P
Cool, 4°C
P
Cool, 4°C
28 2 48 14
days hours hours days
NaOH to pH 12
to chlorination
0.008% Na2S203(d) Dissolved oxygen
Probe Winkler Fluoride Hardness
Hydrogen ion (pH) Kjeldahl and organic nitrogen
G bottle & top G bottle & top
P P P P
Determine onsite Fix onsite None required HNO3 to pH 2 Determine onsite
Cool, 4°C
1 hour 8 28 6 2 28
hours days months hours days
H2SO4 to pH 2
Metals(e) Mercury
P P
Metals except above
P
HNO3 to pH 2
Nitrate
P P
Cool, 4°C Cool, 4°C
Oil and Grease
P G
Cool, 4°C
Organic Carbon
G
Chromium VI
Nitrate-nitrite
Cool, 4°C HNO3 to pH 2 n D R % kofrofiT
48 hours 28 days 6 months
48 hours 28 days
H2SO4 to pH 2 Nitrite
Cool, 4O(; H2SO4 to pH 2 Cool, 4°C H2SO4 to pH 2
464
48 hours 28 days 28 days
TABLE 2. Continued
Measurement
Container
Organic Compounds(f) Extractables (including phthalates, nitrosarnines organochlorine pesticides, PCB's, nitroaromatics, isophorone, polynuclear aromatic hydrocarbons, haloethers, chlorinated hydrocarbons and TCDD)
G, teflonlined cap
Extractables (phenols)
G, teflonlined cap
Purgeables (halocarbons, aromatics, Acrolein, and Acrylonitrile)
G, teflonlined septum
Orthophosphate Pesticides
Phenols
G, teflonlined cap
G
Phosphorus (elemental) Phosphorus, total
Residue, Residue, Residue, Residue, Residue, Silica Specific Sulfate Sulfide
P
total filterable nonfilterable settleable volatile
G P,G
P
P
P P P P
conductance
P P P
465
Preservative Cool, 4°C 0.008% Na2S20 3 ( d )
Maximum Holding Time 7 days (until extraction) 30 days (after extraction)
Cool, 4°C
7 days (until extraction) 0.008% Na 2 S203( d!) 30 days (after extraction) Cool, 4°C 14 days 0.008% Na2S?03(d )
Filter onsite Cool, 4°C Cool, 4°C
48 hours
Cool, 4°C Cool, 4°C Cool, 4°C Cool, 4OC Cool, 4°C Cool, 4°C Cool, 4°C Cool, 4°C Cool, 4°C Zinc Acetate
14 14 7 7 7 28
7 days (until extraction) 0.008% Na2S203(d ) 30 days (after extraction) Cool, 4°C 28 days H 2 S0 4 to pH 2 Cool, 4°C 48 hours Cool, 4°C 28 hours H2SO4 to pH 2 days days days days days days 28 days 28 days 28 days
TABLE 2. Continued
Measurement
Container
Preservative
Sulfite Surfactants Temperature Turbidity
P P P P
Cool, 4°C Cool, 4°C Determine onsite Cool, 4°C
Maximum Holding Time 48 hours 48 hours Immediately 48 hours
(a) Polyethylene (P) or Glass (G) * ' Samples should be preserved immediately after they are collected. For composite samples each aliquot should be preserved at the time of collection. When use of an automatic sampler makes preserving each aliquot impossible, samples may be preserved at 4°C until compositing and sample splitting is completed. (c) Samples should be analyzed as soon as possible after collection. The times listed are the maximum times that samples may be held before analysis and still be considered valid. Samples may be held for longer periods only if the laboratory has data on file to show that the specific types of samples under study are stable for a longer time. Some samples may not be stable for the maximum time period given in the table. A laboratory must only hold a sample for a time period within which the sample is known to maintain its integrity. (d) This should only be used in the presence of residual chlorine. (e) Samples should be filtered immediately onsite before preservation for dissolved metals. (f) Guidance applies to samples to be analyzed by Gc, Lc, or GC/MS for specific organic compounds. (Compounds not found in Table 2 should be preserved at 4°C; storage:
1 week.)
*Adapted from Federal Register, December 18, 1979. Vol. 44, No. 244, pp. 75050-75052.
466
are most desirable. Occasionally, analysis methods may have to be developed for unusual analytes, but such activities should be minimized. Usually, the analytes of interest will be compounds for which National Primary Drinking Water, Water Quality, or Human Health Standards exist. Sources of approved methods include ASTM, APHA-AWWA-WPCF, Standard Methods for the Examination of Water and Wastewater.w)4 and U. S. EPA, Federal Register, December 3, 1979, 600 Series Methods,( ) and Methods for Chemical Analysis of Water and Wastes.(5) In addition, all analyses must must be performed under conditions of strict quality control. The nature of the site investigation or monitoring will dictate whether formal chain-of-custody procedures must be followed. However, in all cases sampling activities must be documented in a bound log book. Each sample should be assigned a unique identification number and be permanently labeled with sample number, date, time, sampler, location, preservative, and analyte(s) of interest. The log book should contain these data along with a description of unusual sample, site, or weather conditions; field measurements such as pH, conductivity or temperature; and equipment or instruments used. At the laboratory, all samples are logged into a bound notebook, divided into analytical lots, and assigned a unique number which includes a reference to the analytical lot and the position within the lot. An analytical lot is that number of samples that can be handled simultaneously through the entire procedure. At least one quality control (QC) sample must be included in each lot. This may be a duplicate of a real sample, a method blank, or a spiked sample of natural or distilled water. Each type of QC sample is useful for obtaining information about method performance, but none provides all types of QC information. When possible, more than one QC sample should be included in a lot. All QC samples are prepared by an individual not involved or responsible for analyses and are sent to the analyst blind. This person receives all analytical results and evaluates the data from the QC samples to determine if the lot of samples was analyzed under controlled conditions. Statistical methods are used to construct control charts for spike recovery, duplicates, and method blanks. If an analysis is deemed to be out-ofcontrol, the cause must be identified, corrected, samples reanalyzed, and all actions documented. CONCLUSIONS The monitoring of organics in ground water is complex. Each activitiy can affect the analytical results and the accuracy with which environmental samples reflect ground-water conditions. Although the problem is complex, the path that must be taken to assure defensible results is well traveled. Laboratories capable of performing the necessary analyses are available but must be scrutinized before being used. Our experience has shown that it is cost effective to ship samples across the country by air freight to achieve the quality of analysis that these efforts demand.
467
STEEL CAP
|—y. ALIGNED LOCKING - ~ » . _ / RINGS S
I
THREADED SCHEDULE ^ - " 4 0 PVC CAP V%" VENT (SLOT) ON BOTH PVC AND STEEL CASING 2' DIAMETER CEMENT GROUT BASE
WELL
SCHEDULE 40 4 " DIAMETER PVC PIPE
B0RE CEMENT AND BENTONITE SEAL
SCHEDULE 40 4 " DIAMETER PVC PIPE
CEMENT AND BENTONITE GROUT
SILT AND — § = : CLAY SAND AND OR GRAVEL
SCHEDULE 40 4 " DIAMETER WELL SCREEN SAND PACK BENTONITE SEAL ^ s r - " AQUIFERSATURATED SAND AND/OR &&%•& GRAVEL.SILT.AND ^ f ^
ipr \ " T ~r
SAND PACK •
I
3'-8'
EXAGGERATED FRACTURES
FIGURE 1.
Well Construction Detail
468
NOT TO SCALE
REFERENCES Johnson et al. 1980. "Experimental Determination of Thermoplastic Casing Collapse Pressure." Ground Water. 18(4). Plumb, R. H., Jr. 1981. Procedures for Handling and Chemical Analysis of Sediment and Mater Samples. Greenberg, C , and F. Jenkins. 1980. Standard Methods for the Examination of Hater and Wastewater. 15th Edition, APHA-AWWA-WPCF. U.S. EPA. 1979. Federal Register, December 3, 1979, 600 Series Methods. U.S. EPA. 1979. Methods for Chemical Analysis of Water and Hastes.
46914-70
8B
GROUNDWATER MONITORING PROGRAM AT THE ROCKY FLATS PLANT
N. D. Hoffman Rockwell International Golden, Colorado
ABSTRACT Groundwater at the Rocky Flats Plant is periodically monitored through the use of 56 monitoring wells. Samples taken from the groundwater monitoring wells are analyzed for both radiological and nonradiological parameters. The radiological parameters include: total alpha, total beta, plutonium, uranium, americium, and tritium. Nonradiological parameters include pH, nitrate, total dissolved solids, and approximately 43 other elements. Samples collected from most of the groundwater monitoring wells indicate that there is no significant contamination of the groundwater. Low concentrations of uranium, tritium and nitrates above the local background concentrations have been found in monitoring wells in the vicinity of the plant site solar evaporation ponds, indicating some seepage from the ponds. These solar evaporation ponds have been used to store process wastewater prior to treatment. Sampling techniques for the collection of groundwater samples, interpretation of selected data, and the quality control program at Rocky Flats will also be discussed.
INTRODUCTION The Rocky Flats Plant routinely samples 56 groundwater monitoring wells. Analyses are conducted to determine if there is any movement of chemical or radioactive materials of plant origin into water-bearing strata underlying the site. Groundwater monitoring also provides historical data and insures compliance with DOE monitoring guidelines. The groundwater monitoring program at the Rocky Flats Plant and a discussion of selected data are presented in this report. The locations of the groundwater monitoring wells are shown in Figure 1. The figures are shown at the end of this text. Five of the monitoring wells (1-66, 2-66, 3-66, 21-74, and 22-74) are located west of 471
the west security fence, northeast of the solar ponds, east of the solar ponds, east of the east security fence and near the south security fence, respectively. These monitoring wells range from 140 to 320 feet (43 to 96 meters) in depth and provide information concerning water movement in gravel and bedrock formations. The 60-series monitoring wells were drilled in 1960 to check for movement of materials from solar evaporation ponds. These ponds are used to store process wastewater prior to treatment. The ponds are hydrologically upgradient from the monitoring wells. These monitoring wells (1-60, 2-60, 3-60, 4-60, 5-60, and 6-60) are approximately 30 feet (9 meters) deep. The 71-series test wells were drilled in 1971 to determine if significant migration of radioactivity was occurring from the Plant's holding ponds. They range from 22 to 30 feet (7 to 9 meters) in depth. Prior to 1980 the holding ponds were used to store laundry water and treated sanitary wastewater. Currently, the holding ponds are used to store surface runoff water and treated sanitary wastewater from the Plant reverse osmosis facility. Two of the 71-series monitoring wells (1-71 and 2-71) are located in a drainage field below an asphalt pad (Figure 1, southeast area of Plant site). The asphalt pad covers an area that was used to store drums containing machining lubricants and chlorinated solvents contaminated with plutonium. The drums were removed in 1968 and the area was coated with asphalt in 1969. There is one well at each corner of the asphalt pad. These four wells are referred to as the 68-series wells and are about four feet (one meter) deep. The remaining wells range from 3 to 100 feet (1 to 30 meters) deep and are generally located near drainage areas, holding ponds and old waste burial sites. The casing type of the monitoring wells drilled in the sixties is galvanized iron. The monitoring wells drilled in 1971 have casing made of steel. The monitoring wells drilled in 1974 have casing made of plastic and the remaining monitoring wells have casing constructed of polyvinylchloride. SAMPLING METHODS AND ANALYSES The groundwater from the monitoring wells is sampled and analyzed quarterly by onsite laboratories. The equipment used for sampling includes an All Terrain Vehicle (ATV) for easy access to remote locations, submersible pump, power generator, bailers, and a four wheel drive vehicle. The ATV is equipped with a winch which is used to lower the pump into the wells. The pump is used to pump the wells dry one week prior to sampling. The wells are sampled using the bailer method. At the time of sampling, the depth to groundwater is measured using a tape measure. The groundwater samples are collected in eight liter bottles and delivered to the laboratory where the samples are filtered twice. The
472
first filter is a 5-10 micron Whatman No. 2V filter which removes large pieces of soil and particles. The second filter is a 0.45 micron Whatman EPM 1000 g^ade filter. After the samples are filtered, they are split and a part of the sample is sent to another onsite laboratory. The two laboratories perform different analyses on the samples. Prior to 1982, sampling of the groundwater monitoring wells was performed at approximate five-month intervals. In 1982, quarterly sampling of the groundwater monitoring wells began and additional new parameters were analyzed in order to comply with the Department of Energy (DOE) intent to provide a groundwater monitoring program equivalent to that which is required under the Resource Conservation and Recovery Act (RCRA). The radiological analyses on the groundwater samples include gross alpha, gross beta, gamma-emitting isotopes, plutonium, uranium, americium, and tritium. The chemical and biological parameters that are analyzed include pH, total dissolved solids (TDS), conductivity, alkalinity, hardness, nitrate, and sulfate. There are approximately 43 elements analyzed by emission spectroscopy or atomic absorption techniques. Analyses of several organic materials are also performed on the groundwater samples. QUALITY CONTROL Analytical results are reviewed by an engineer and a geologist in the enviromental monitoring group at Rocky Flats. When internally established control guides are exceeded, data reliability and contaminant sources are investigated. Quality control measures include (1) visual inspection of the monitoring wells, (2) well maintenance and (3) periodic observations by environmental monitoring group personnel during sampling. Item (3) includes a checklist for the inspector to note any deficiencies and corrective actions required during sampling of the monitoring wells. In addition to the above procedures, the Colorado Department of Health (CDH) samples five of the monitoring wells (4-60, 1-66, 3-71, 15-74, and 17-74) concurrently with the Rocky Flats laboratories. The quality control program for the laboratory includes the following elements: 1.
Development, evaluation, improvement, modification, and documentation of analytical procedures
2.
Scheduled instrument calibration, control charting, and preventive maintenance
3.
Participation in inter!aboratory quality comparison programs
4.
Intralaboratory quality control programs.
473
EVALUATION OF SELECTED GROUNDWATER DATA The median plutonium (Pu-239,-240) americium (Am-241), uranium (U-233, -234, and -238), tritium, and nitrate concentrations in water from the monitoring wells from 1975 through 1981 are presented in Figures 2 through 6. The figures show the median concentration from all monitoring wells for each sampling period. The plutonium concentrations in water from all monitoring wells have ranged from less than 0.01 to 2.7 X 910" 9 pCi/ml. Background levels are commonly between 0.02 and 0.1 X 10"" pCi/ml.(') The Radioactivity Concentration Guides (RCG's) are published by the DOE and the CDH. There are no applicable RCG's for groundwater. The RCG for plutcnitm in wcter ef-Huwnt to the general population is 1667 X 10~ 9 yCi/ml.(2,3) The plutonium concentrations in the 60-series wells and monitoring well 6-71 ranged from 0.01 to 0.67 X 10~ 9 yCi/ml. The median plutonium concentrations have all been less than or equal to 0.1 X 10-" yCi/ml. Monitoring wells 1-66 and 3-66 had plutonium concentrations less than or equal to background level. Monitoring9 well 2-66 had plutonium concentrations ranging from 0.01 to 1.04 X 10" pCi/ml. The value of 1.04 X 10~ 9 yCi/ml reported in 1975 is questionable. All subsequent samples from monitoring well 2-66 had plutonium concentrations that were equal to or slightly above background levels. Although this is east of the solar evaporation ponds, the other seven monitoring wells in the vicinity of 9 the solar ponds had plutonium concentrations that were less than 0.4 X 10~ yCi/ml. In 1977 and 1978 three samples from monitoring well i-71 (located southeast of the asphalt pad) contained plutonium concentrations of 2.1, 2.7, and 1.14 X 10"" yCi/ml. It appears that windblown dust from the pad area may have entered the well through an improperly sealed well head. An improved well closure was installed, and the 1979, 1980, and 1981 data indicated plutonium concentrations at background levels. Americium concentrations in water from groundwater monitoring wells ranged from less than 0.01 to 1.0 X 10" 9 yCi/ml,9 The median americium9 concentrations ranged from less than 0.01 X 10" yCi/ml to 0.14 X 10~ yCi/ml. Background concentrations in nearby water are in the range from 0.05 to 0.5 X 10~ 9 yCi/ml.l ' only two samples from monitoring wells had americium concentrations above background levels. Both of these monitoring wells are located in9 the vicinity of the solar ponds. None of the values exceeded 1330 X 10" yCi/ml, the DOE and CDH RCG for surface discharges to uncontrolled areas. The concentration of uranium in water samples from the monitoring 9 wells ranged from 0.05 to 156.23 X 10~ yCi/ml. The median uranium concentrations ranged from 2.1 X 10" 9 yCi/ml to 9.2 X 10" 9 9 yCi/mi. Background levels of uranium typically range from 5 to 15 X 10" yCi/ml.") The most restrictive RCG for uranium-233, -234, and -238 in water is 200 X 10" 9
474
yCi/ml.v(2)' The anomalous uranium levels were found for the most part in water from wells east of the solar ponds. Anomalous uranium levels have also been found in monitoring wells 9-74, 10-74, and 15-74 (all southeast of the Plant site) and monitoring wells 17-74 and 18-74 (both near holding ponds). Uranium concentrations found in these monitoring wells may be natural and not of Plant origin. Small pockets of low grade uranium ore are not uncommon in the Arapahoe bedrock formation, which underlies the Plant. In the spring of 1981, a new well was drilled east of monitoring well 18-74. It is anticipated that determining the uranium levels at well 3-81, downstream of wells 17-74 and 18-74, will further define the flow pattern and source of previous anomalies in this area. The more recent data indicate that samples from this well (3-81) contain uranium concentrations slightly above background levels (ranging from 3 to 9 X 10" 9 uCi/ml). Monitoring wells 9-74 and 10-74 are located in the vicinity of an old burial area where soil and asphalt which may have been contaminated were deposited in 1969. The median tritium concentrations have ranged from 160 X 10" 9 yCi/ml 9 to 1846 X 10~ vCi/ml. There has been only one sample in which the tritium concentration exceeded the Environmental Protection Agency (EPA) drinking water standard of 20,000 X 10~ 9 gCi/ml. One of the many chemical parameters analyzed on the groundwater samples is nitrate (N0 3 as N ) . Most of the samples from the groundwater monitoring wells had concentrations of nitrate less than the EPA drinking water standard (10 mg/1 N0o as N ) . The median nitrate concentrations ranged from 1 to 5.3 mg/1 {NO3/N). The nitrate concentrations in monitoring wells near the solar ponds have ranged from less than 1 mg/1 to 3108 mg/1. Relatively high concentrations of nitrate (4 to 64 mg/1) were found in monitoring wells 9-74 and 10-74. SUMMARY AND CONCLUSIONS The historical data collected on the groundwater monitoring wells from 1975 through 1981 support the following conclusions: 1.
The plutonium and americium concentrations in water from the monitoring wells were generally at background levels. There are no applicable RCG's for groundwater; however for perspective, the concentrations of plutonium, uranium, americium, and tritium on all groundwater samples were well below the DOE and CDH RCG's for water discharged to uncontrolled areas.
2.
Uranium concentrations above background levels were found in monitoring wells near the solar ponds. Relatively high uranium concentrations were also found in two monitoring wells near holding ponds, and two wells southeast of the Plant. Monitoring wells 1-82 and 2-82 were drilled to determine if the high uranium and nitrate levels found in samples from monitoring wells 9-74 and 10-74 were migrating in the groundwater southeast of the Plant. Some exploratory
475
drilling was done in the vicinity of these wells during 1982 to see if there were any physical debris buried in the area. The results of the drilling showed that nothing of significance was buried in the area. Careful monitoring of wells 9-74 and 10-74 and the monitoring wells drilled in 1982 (1-82 and 2-82) will continue. 3.
Tritium concentrations were for the most9 part well below the EPA drinking water standard of 20,000 X 10" uCi/ml.
4.
Monitoring wells in the vicinity of solar ponds (1-60, 2-60, 3-60, 4-60, and 6-71) had groundwater concentrations of nitrate that ranged from <1 mg/1 to 3108 mg/1. Relatively high concentrations of nitrate also occurred in monitoring wells 9-74 and 10-74.
5.
Through the groundwater monitoring program, it has been determined that some seepage has occurred from the solar ponds. However, downgradient monitoring indicates that the seepage is localized. To correct this, process wastewater is no longer discharged to the solar ponds. Also, three of the solar ponds have been cleaned out and the cleaning of the largest pond will probably start during 1983.
REFERENCES T-
Environmental Impact Statement, Rocky Flats Plant Site, DOE/EIS-0064 U.S. Department of Energy, Washington, D.C., April 1980.
2.
Rules and Regulations Pertaining to Radiation Control, Part IV, Colorado Department of Health, 1978.
3.
"Standards for Radiation Protection," DOE Order 5480.1, Change 6, Chapter XI, Department of Energy, Washington, D.C., August 13, 1981.
476
LEGEND • MONITORING WELL GREATER THAN 30 METERS A MONITORING WELL LESS THAN 15 METERS
Figure 1. Locations of Groundwater Monitoring Wells at Rocky Flats
.12 -
U a. f
a. .02 1975
1976
1977
1978
1979
1980
1981
Figure 2. Median Pu Concentrations in Monitoring Wells
u a. m O X
jT <
-06 h
.04 .02 i
1975 Figure 3.
l
l
1976
1977
1978
1979
1980
Median Am Concentrations in Monitoring Wells
478
1981
o a. o 4 2 " 197S
1976
1977
1978
1979
1980
1981
Figure 4. Median U Concentrations in Monitoring Wells 2000
1975
1976
1977
1978
1979
1980
Figure 5. Median Tritium Concentrations in Monitoring Wells
479
1981
1975
1976
1977
1978
1979
1980
1981
Figure 6. Median Nitrate Concentrations in Monitoring Wells
480
8C OAK RIDGE Y-12 PLANT GROUNDWATER MONITORING AND ASSESSMENT PROGRAM
Paula M. Pritz Oak Ridge Y-12 Plant* Union Carbide Corporation, Nuclear Division
ABSTRACT The Oak Ridge Y-12 Plant is located in Eastern Tennessee in the Ridge and Valley physiographic province. It is situated in Bear Creek Valley, south of the City of Oak Ridge. The principal disposal facilities for the Y-12 Plant (located in Bear Creek Valley approximately two miles west of the center of the Plant) are underlain by the Conasauga Group--a Cambrian-aged formation of shale, siltstone, and limestone. The Conasauga Group is a poor aquifier that has a relatively high groundwater table with residuum materials usually less than 30 feet in thickness. The Y-12 Plant is currently monitoring 16 Resource Conservation and Recovery Act wells on a quarterly basis and 18 additional wells on an annual basis. These monitoring wells are located both upgradient and downgradient of the disposal facilities. Sampling is performed by the Environmental Monitoring Group using procedures established by the Department of Energy and Federal and State regulations. The samples are analyzed in the Y-12 Plant Laboratory, which is certified by the Environmental Protection Agency (EPA). A program (still in the planning and development stages) to locate, identify, and catalog approximately 100 wells in Bear Creek Valley and the areas surrounding the Y-12 Plant, will hopefully expand and enhance our knowledge of the groundwater characteristics in this area. Plans include establishing high and low groundwater elevation tables, flow directions and velocities, and mineralization characteristics, and acquiring supplemental shallow seismic and resistivity measurements. Subsequently, the location of wells used for monitoring data will be reevaluated and revised where necessary to establish what we hope will be one of the best monitoring data collections for this area of Eastern Tennessee. * Operated for the U.S. Department of Energy by Union Carbide C poration, Nuclear Division, under Contract W-7405-eng-26.
481
INTRODUCTION Over the years, environmental laws and regulations have continually stressed the importance of being able to detect any groundwater contamination by developing adequate monitoring programs. The promulgation of the Resource Conservation and Recovery Act (RCRA), along with the recent amendments to RCRA which become effective on January 26, 1983, have once again emphasized the need for a competent groundwater monitoring and assessment program. In light of these recent developments, the Environmental Affairs Department of the Oak Ridge Y-12 Plant decided to review the groundwater monitoring program to assure that it was adequate for compliance with the requirements of RCRA and other applicable Federal and State waste management regulations. While the Y-12 Plant had been monitoring several wells in the past, there was no guarantee that the same wells were sampled each time. This mix-up developed through the years due to a variety of factors. Perhaps the biggest factors were the lack of a uniform coding system and an up-to-date location map identifying all wells which were to be sampled. Since this had not been done, the technician sampling the well would simply describe the location using landmarks. If two wells were in the same general location and a different technician was taking the sample during the following period, it would be easy to sample the wrong well. Therefore, shortly into our investigation it was already evident that we needed to simply and better organize the groundwater monitoring program. COMPLIANCE ASSURANCE PLAN Our major concern and the principal objective of the study was the answer to the question of whether or not our groundwater monitoring program was adequate for compliance with monitoring requirements set forth in the various waste disposal regulations. After reviewing these applicable regulations, it was determined that there were five areas which warranted a more detailed investigation. These areas are listed in Figure 1 as the Compliance Assurance Plan. Wells The first task was to identify the specific wells which were to be used as monitoring sites at each waste management area. The criteria governing our decision were the number of wells and their position with respect to the hydraulic gradient. We wanted to make sure that at least one well was hydraulically upgradient to represent the background water quality, and at least three wells were downgradient at the limit of the waste management area.
482
These wells were plotted on a topographic map and given identification numbers (Figure 2 ) . Each well was then located in the field and labeled with its given number. Copies of the location map were distributed to those technicians responsible for groundwater monitoring. Hopefully, this will eliminate any future programs of sampling an "incorrect" well. Sampling Technique and Frequency The investigation revealed that our method of sampling was in accordance with Environmental Protection Agency (EPA) or American Society for Testing Materials (ASTM) guidelines. Sample preservation was conducted when necessary and all samples were transported to the laboratory in a timely manner. One area which we are striving to improve is the time elapsed between when the well is pumped and the rample is taken. This is basically a manpower coordinating problem and still needs some improvement. Since we are basically concerned with establishing initial background concentrations for groundwater constituents, it was necessary to conduct sampling on a quarterly basis for the first year. After the first year, we plan to go to a semiannual or annual system wherever possible. Parameters and Analytical Techniques Reviewing our past disposal practices and all applicable regulations, it was not difficult to compile a list of 53 possible testing parameters. Fortunately, some of these parameters, such as various insecticides and pesticides, could be removed from the list since we do not and never have used those substances. The next step was to decide which parameter to test for in each sample. For the most part, this was dependent on the type of waste management area (e.g., sanitary of low-level radioactive) and our past and present disposal practices. Such parameters were included on the list as a precautionary measure until sufficient background information could be compiled, and subsequently eliminated from the list when analyses showed them to be nonexistent as a groundwater constituent. Figure 3 is a partial listing of the parameters. All of the water samples are analyzed by the Y-12 Plant Laboratory. The procedures used are those set up by EPA or ASTM, and fully comply with the regulations. In addition, the Plant Laboratory is an EPA Certified laboratory.
483
Hydroiogic Information A subject which had never been comprehensively studied was the hydro!ogic characteristics of the principal waste management areas. As a result of the new RCRA requirements, it was deemed appropriate to conduct such a study. The information obtained will enable us to evaluate our groundwater data more accurately and recognize trends in groundwater constituent levels earlier. This project will most likely be conducted in the Spring of 1983 to allow adequate time for preparation of statistical evaluation techniques. Statistical Evaluation One of the reasons for collecting groundwater information is to be able to determine any statistically significant changes in groundwater constituents that migrate from a waste management area. Up to now, limited statistical evaluations have been performed. We are currently investigating our statistical capabilities and increasing them where necessary to conform with new requirements and to enhance our prediction capabilities. Wherever possible, analytical data and statistical programs will be combined to evaluate trends in constituent levels and further our understanding of the migration of grounwater constituents. SUMMARY As a result of the study conducted (Figure 4 ) , we have completely evaluated our "primary" groundwater monitoring program and have eliminated any doubts in the quality of the sampling and analytical techniques. We are collecting appropriate background measurements for the revised monitoring program and have made advances in statistical approaches for evaluating the data obtained. The hydrologic investigation, which will be conducted in the Spring of 1983, will allow further understanding of the groundwater characteristics which will enhance our capability to detect trends in groundwater constituent levels.
484
1. Review number and location of monitoring wells. 2. Review sampling techniques and frequency. 3. Compile parameters list and review analytical methods. 4. Conduct hydrologic investigations, 5. Upgrade statistical evaluation programs.
Figure 1. Compliance Assurance Plan
485
"\
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OAK RIDGE Y-12 PLANT GROUNDWATER PONITORING
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
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X
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X
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Burial Grounds - Besr Creek Road
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3.
Classified Burial Grounds - Chestnut Ridge
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h.
United Nuclear Site
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X
X
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Mew Sanitary Landfill - Chestnut Ridge
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X
6.
y.cu Hope Pond Sludge Basin
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X
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Old Sanitary Landfill Area
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X
X
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Coal Yard - Special Sample
S-3 Ponds
2.
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—I
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1.
X
X
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CO
FOOTNOTES:
TGX A B C D E F -
X
X
= Total Organic Halogens; TOC = Total Organic Carbon RCRA Regulations - Constituents for groundwater protection Tennessee State Sanitary Landfill Regulations EPA Interim Drinking Water Standards Y-12 Plant recommended Hazardous Waste Regulations - Parameter for establishing groundwater quality Hazardous Waste Regulations - Indicator parameter for groundwater contamination
Kioure 3. Oak Ridge Y-12 Plant Groundwater Monitoring
X
X
X
X
1. Proper location and identification of wells. 2. Confidence in sampling and analytical techniques. 3. Determination of appropriate parameters and sampling frequency. 4. Hydrological investigation to be conducted in Spring of 1983. 5. Additional statistical evaluation capabilities.
Figure 4. Accomplishments
488
8D GROUND-WATER MONITORING PROGRAMS AT THE HANFORD SITE, WASHINGTON STATE PARTS I AND II 0. S. Wilbur Rockwell Hanford Operations Richland, Washington PART III L. S. Prater and P. A. Eddy Pacific Northwest Laboratory Richland, Washington ABSTRACT The U.S. Department of Energy Hanford Site has been the location of nuclear activities since 1943. Site operations have resulted in the production of large volumes of low-level contaminated cooling water and low-level, liquid radioactive wastes. Ground-water monitoring is necessary because these wastes are discharged to the ground. Two ground-water monitoring programs are currently in operation: Rockwell Hanford Operations conducts the site-specific ground-water monitoring program for the Separations Area, and Pacific Northwest Laboratory conducts the site-wide ground-water surveillance program. These two ground-water monitoring programs are conducted separately, but activities of the programs are coordinated and data are exchanged actively. PARTS I AND II The principal objectives of the Separations Area ground-water monitoring program are to assess the performance of disposal facilities and to determine the impact of liquid waste discharges on the ground water. The Separations Area is located near the center of the Hanford Site. The area contains uranium fuels processing facilities, plutonium separations facilities, and the major radioactive waste storage and disposal facilities. From the plutonium separations process, low-level radioactive wastes are discharged to the ground via surface and subsurface structures. 489
The Separations Area ground-water monitoring network consists of 105 wells. Samples from these wells are collected at frequencies ranging from monthly to semiannually. Samples are selectively analyzed for radioactive and nonradioactive constituents. The impact of liquid waste discharge on the quality and flow of ground water is determined from established histories and concentration guides. All aspects of the ground-water monitoring program are founded on an understanding of the ground-water flow system. PART H I The principal objectives of the site-wide, ground-water surveillance program are to measure the concentration and distribution of radionuclides in the ground water and to evaluate the impact of ground-water contamination at Hanford on people and the environment. Ground-water surveillance was initiated at the Hanford Site in the early 1950s to monitor conditions in the unconfined aquifer. Over 2,200 wells have been drilled on the site to obtain geohydrological and radiological information. About 850 of these wells are in the sampling network for the Ground-Water Surveillance Program. Each year, samples are collected from about 300 wells. A variety of radioactive and nonradioactive analyses are performed on the samples. The data obtained are used to monitor the growth of contaminant plumes.
INTRODUCTION The U.S. Department of Energy (DOE) Hanford Site has served as an integrated nuclear facility since 1943. The Site occupies 1,476 square kilometers (500 square miles) in semiarid, south-central Washington State (Figure 1 ) . The Site was established as part of the Manhattan Project during World War II. At present, the main activities include reactor operation, fuel fabrication and reproc2ssinq, waste management, and research and development. Operations at the Hanford Site have resulted in the production of liquid wastes. Beginning in the mid-1940s, small volumes of selected low-level, radioactive liquid wastes were discharged to the ground through subsurface structures. The rate of discharge of low-level liquid waste to the ground increased steadily and reached a peak in 1955. Thereafter, the discharge of low-level wastes decreased; ground disposal of radionuclides was minimized through process improvements and shutdown of plants in response to cutbacks in defense needs. In addition to the low-level wastes, large volumes of cooling water and 490
steam condensate from chemical processing facilities have been discharged to surface ponds and ditches. From 1944 through 1980, more than 1 x 10l2 liters of liquid waste have been discharged to the ground.
•"••••"-•-..„„
'••"
J
OABIE MOUNTAIN
1 1 1
L.
SEPARATIONS AREA
|
RCP82O536
Figure 1. Separations Area Location
Rockwell Hanford Operations (Rockwell) is a prime contractor to DOE at the Hanford Site. Rockwell operates and maintains the irradiated uranium fuels processing facility, plutonium separations facility, and major waste storage and disposal facilities within the Separations Area (Figure 1). As part of the Rockwell waste management effort, a comprehensive ground-water network is maintained for the Separations Area. The Separations Area Ground-Water Monitoring Program satisifies DOE requirements that all Rockwell discharges be monitored. These requirements are stated in Chapter XI of DOE Order 5480.1 and Chapter XII of DOE Order 5484.1. 491
In addition to the Separations Area Ground-Water Monitoring Program, Pacific Northwest Laboratory (PNL) maintains the Hanford Site Ground-Water Surveillance Program. The purpose of that program is to estimate and evaluate the impact of the unconfined ground-water contamination from the Hanford Site on the general public. Ground water sampling and analysis for the two programs are coordinated and data are actively exchanged.
PART I. HYDROGEOLOGY OF THE HANFORD SITE Geology The Hanford Site is located within the Pasco Basin, a structural and topographical basin in south-central Washington State. The boundaries of the Pasco Basin are defined by anticlinal structures of basalt. There are three main geologic formations beneath the Hanford Site. They are, in ascending order, the Columbia River Basalt Group, the Ringold Formation, and glaciofluvial sediments. The Columbia River Basalt Group is a thick sequence of basalt flows extruded from fissures during the Miocene epoch. The Ringold Formation, a Pliocene fluvial sedimentary unit, overlies the Columbia River Basalts except in areas where erosion has removed these sediments. The glaciofluvial sediments, informally named the Hanford formation, were deposited during the Pleistocene epoch on top of the Columbia River Basalt Group and the Ringold Formation. Hydrology The aquifer affected by liquid waste disposal at Hanford is the unconfined aquifer. The unconfined aquifer is contained within the Ringold Formation and the younger, overlying glaciofluvial sediments of the Hanford formation. The bottom of the unconfined aquifer is the basal"-, surface or, in some areas, a clay zone in the Ringold Formation. The unconfined aquifer at the Hanford Site is more than 70 meters (230 feet) thick in some areas and is absent along the flanks of basalt structures (Graham, 1981). Natural recharge to the unconfined aquifer under the Hanford Site occurs at the area of high relief west of the Site. From the recharge areas, ground water flows downgradient to di= harge areas in the east, primarily the Columbia River. This general flow pattern is interrupted locally by the ground-water mounds and basalt outcrops in the Separations Area. These mounds are the result of artificial recharge from waste disposal in the Separations Area and have had a profound effect on the flow system of the unconfined aquifer (Graham, 1981). The unconfined aquifer beneath the Separations Area receives artificial recharge from liquid disposal facilities at a rate approximately ten times that of natural recharge. Cooling water 492
disposed of to the ponds began to recharge the unconfined aquifer artificially at a rate that exceeds the ability of the sediments to transmit this water. This formed ground-water mounds under the high-volume disposal sites: U-Pond, B-Pond, and Gable Mountain Pond (Figure 2). The water table has risen 26 meters under U-Pond and 9 meters under B-Pond compared with pre-Hanford conditions (Newcomb et al., 1972). Ground-water flow lines are drawn perpendicular to the water-table contour lines to indicate the direction of flow (Figure 2 ) . These ground-water mounds have affected ground-water flow patterns. The movement of ground water from the west to U-Pond (under 200 West Area) is redirected around the mound. Flow from U-Pond is primarily toward the east; gradients in this area are extremely steep, greater than 10 meters/kilometer (50 feet/mile). Between the 200 East and 200 West Areas, the gradient abruptly flattens out. This is attributed to an increase in the hydraulic conductivity of the sediments and a drop of the basalt surface which results in an increase of nearly 40% in the thickness of the saturated zone. The flow system in 200 East Area is complex due to the change in aquifer thickness and hydraulic properties, the influence of B-Pond, and the basalt structures of the Gable Mountain/Gable Butte area. Ground water flows out of the 200 East Area to the north, through Gable Gap, and also south along a ground-water trench defined by the 122.5-meter (402-foot) contour. Flow from B-Pond is radially outward to the east and combines with the flow paths out of 200 East Area (Graham et al. } 1981).
PART II. SEPARATIONS AREA GROUND-WATER MONITORING PROGRAM The monitoring program for assessing radionuclide ground-water contamination under low-level, radioactive disposal facilities is based upon the understanding of the hydrology of this area. Knowledge of the ground-water flow system predicates well location, well completion, and sampling frequency (Wilbur and Graham, 1982). In 1981, there were 105 wells in the Separations Area ground-water network. The wells are either 15 centimeters (6 inches) or 20 centimeters (8 inches) in diameter. The larger-diameter casings allow the wells to be fitted with submersible pumps and permit access with hydrologic testing equipment. The well renovation program and the well maintenance program are designed to ensure the integrity of the wells and the quality of the samples. Well Renovation To ensure the integrity of wells, the well renovation program continued in 1981. The well renovation program addresses wells drilled without surface seals within 100 meters (300 feet) of a liquid waste disposal site. All wells drilled after 1978 are fitted with a surface 493
|WATER TABLE MAP WITHIN THE |SEPARATIONS AREA '981 DATA
Figure 2.
Ground-Water Map, 1981
seal that meets Washington State water well standards. Without the protection of a surface seal, soil may be eroded away from a well casing, thus creating voids around the well. These voids provide a possible pathway for surface and/or subsurface contamination to reach the ground water. Wells are renovated by first perforating the existing casing. Then a liner is set in the well and grouted in place. The perforations in the old casing allow grout to flow into any voids surrounding the well (Figure 3 ) . Well Maintenance A well maintenance program, established in 1981, is designed to ensure that representative ground-water samples are collected. The major task of this program in 1981 was the installation of 87 submersible sample pumps in existing network wells. Pumps cannot be hung in some monitoring wells, as the casing is cut flush to the ground and cemented in place. The pumps are attached to a pitless adapter for sample collection (Figure 4 ) . An external waterproof electrical plug is housed under a bolting well cap. A portable power source outlet can be easily attached to the external plug. The bolting cap prevents foreign material from entering the well.
i)
Figure 4. Well Maintenance
Figure 3. Well Renovation 495
In the well maintenance program, well cleaning systems were tested for their ability to remove mineral and corrosion buildup in network wells. Television inspection of test wells proved the Sonar Jet*) development system to be effective in removing incrustation from the inside of the well casing. High-frequency sound waves from the Sonrr Jet removed material from the intercasing surface. The well cleaning task has become a part of the annual well maintenance program. Sampling The following criteria determine sampling frequencies: •
Wells monitoring sampled monthly.
active
liquid waste
disposal
sites
ire
•
Wells monitoring inactive liquid waste disposal sites that contain contaminants with a high potential of bei^r) remobilized are sampled monthly.
•
Wells (other than specific sites mentioned above) indicat j contamination above background concentration levels are sampled bimonthly or quarterly depending upon the level of contamination and the stability of the concentration trend.
•
Wells yielding samples indicating background are sampled semi annually.
contamination
Standard field procedures are used to collect the ground-water samples. The samples are prepared in the field for subsequent analyses. Water-Level Measurements Maps depicting water-level elevations are produced semiannually for the Hanford Site. Sitewide, 225 water-level measurements are taken to the nearest three millimeters (0.01 foot) with a steel tape. Water-table maps and depth to water-table maps are produced annually for the 200 East and 200 West Areas. In addition, a water-table map for the Separations Area has been produced for this report (Figure 2 ) . Analyses The analytical schedules are based on the inventory of the sites being monitored and the concentration histories of the wells. The samples are analyzed selectively for the following radionuclides: total alpha, total beta, 6 0 Co, ^ R u , 1 3 7 C S , 9°Sr, tritium, and uranium. Nitrate is also present in the process waste streams and i§ used routinely at the Hanford Site to define contamination in the ground water. a) Registered trade name 496
Data Interpretation, Reporting, and Storage Data are interpreted promptly, usually the same day that the results are obtained from the laboratory. Data are interpreted to assess the performance of the active liquid waste disposal facilities and to determine general contamination patterns. In assessing the performance of the disposal sites, data are compared with the DOE concentration guides for water in an uncontrolled area. If groundwater contamination, with the exception of tritium, from an active site exceeds 10% of the concentration guidelines given in Table II of DOE Order 5480.1, Chapter XI for the long-lived radionuclides, the disposal facility is removed from service.13) The 10% standard is adopted to ensure that the concentration guides are not exceeded by subsequent drainage after shutoff. Data from the Separations Area ground-water monitoring network are stored on the Comprehensive Information Retrieval and Model Input Sequence computer data base system maintained by PNL. Reports are issued annually and quarterly (Wilbur and Graham 1982). Contamination Plume Maps For the Separations Area, contamination plume maps are drawn annually for total beta, nitrate, and tritium concentrations. These maps are constructed using the average concentrations of the constituents from samples taken from wells in the Separations Area. Contamination plumes indicate general patterns of ground-water contamination. The collection of annual contamination concentration data and the construction of plume maps allows for the tracing of plume movement over time. Tritium contamination plumes (Figure 5) for the Separations Area show general patterns, which are primarily the result of past disposal practices. Tritium concentrations are divided into four ranges: less than 30, 31 through 300, 301 through 3,000 and greater than 3,000 pCi/mL (Figure 5 ) . The tritium contamination has migrated away from various liquid waste disposal facilities in a general west-to-east direction along major flow lines (Graham, 1981). Tritium concentrations from 200 West Area have moved approximately five kilometers in 35 years. At that velocity, the tritium plume out of 200 West Area will reach 200 East Area in approximately 11 years (1993). This is in agreement with previous estimates (Graham et al., 1981). The tritium plumes from 200 East Area disposal sites have migrated farther than tritium from 200 West Area because the aquifer is located in the more transmissive sediments. Contamination plumes indicated general patterns of groundwater contamination. Analysis of data from the ground-water monitoring network indicates that contamination levels in ground water migrating out of the Separations Area are less than DOE concentration guidelines for water in an uncontrolled area (Wilbur and Graham, 1982). b)
Tritium concentrations are allowed to reach Table II guidelines in the ground water because no further increase in concentration is expected after shutoff. 497
498
PART III. THE HANFORD GROUND-WATER SURVEILLANCE PROGRAM The Ground-Water Surveillance Program for the Hanford Site is conducted by PNL for DOE. This program is part of the Hanford Environmental Surveillance Program, which is designed to evaluate existing and potential pathways of exposure to radioactivity from site operations. The objectives of the Ground-Water Surveillance Program are: 1) to measure and report the concentration and distribution of contaminants in the ground water, 2) to determine the movement of contaminants with time, and 3) to evaluate the impact of ground-water contamination at Hanford on people and the environment. Ground-water surveillance was initiated at Hanford in the 1950s following the disposal of liquid effluents to the ground. These effluents percolated downward through the unsaturated zone toward the water table. As the liquids percolated downward, some radionuclides, such as 90Sr, 137cs, and 239p U; were effectively removed from the wastes .via sorption reactions, while other contaminants, such as l°6Ru, 60co, 129it 9 9 J C J 3H, and nitrate, were not easily adsorbed and eventually reached the ground water (Eddy, Cline, and Prater, 1982). The contaminants that entered the ground water have moved in the general direction of ground-water flow and at a rate nearly equal to its velocity. Their concentrations in the ground water have decreased with time by radioactive decay, ion exchange, diffusion, and hydrodynamic dispersion. One of the primary activities of the Ground-Water Surveillance Program has been to monitor the changes in concentration and distribution of these contaminants. To assist in this effort, an extensive well network has been developed. Well Network Since the beginning of the waste disposal operations at Hanford, over 2,200 wells have been drilled on the site to obtain hydrogeological and radiological information. About 850 of these wells have been prepared for use as ground-water monitoring wells and are included in the sampling network for the Ground-Water Surveillance Program (McGhan and Damschen, 1979). Locations of selected sampling wells are shown in Figure 6. Most of the monitoring wells are six or eight inches in diameter and have steel casings. The wells are usually perforated along the upper portion of the saturated zone of the aquifer.
499
UI'H
ESTIMATED BASALT OUTCROP ABOVE WATER TABLE
MARCH 1982
Figure 6. Locations of Selected Wells on the Hanford Site
500
A routine maintenance program has been established for the wells. Under this program, wells that have deteriorated are repaired and new wells are brought into conformance with standards to ensure that representative samples can be collected. Maintenance of the wells may include: 1) brushing or swabbing the inside surface of the casing, 2) bailing or pumping the well, and 3) perforating the casing or installing screens. Sampling and Analysis Each year samples are collected from about 300 wells* Groundwater samples are collected on a monthly, quarterly, semiannual or annual basis, depending on the location of the well and the constituents to be analyzed. Submersible sampling pumps have been installed in most of the wells to allow collection of more representative ground-water samples. Installation of submersible pumps prevents the possibility of crosscontamination that may result from the use of a bailer to collect water samples from several wells. The ground-water samples are analyzed for a variety of radioactive and nonradioactive parameters. Many of the samples are analyzed for tritium and nitrate because these two constituents indicate the movement of contaminated ground water. In addition, samples from selected wells are routinely analyzed at less frequent intervals for the radionuclides 90sr, 137r,s, and 6 0 Co. Gross alpha activity is also determined. Selected samples are analyzed by gamma spectrometry to identify the mixture of gamma-emitting radionuclides present. Standard radiometric and chemical methods are used to analyze the routine ground-water samples (American Society for Testing and Materials, 1979). Other radionuclides, such as 129j a n d 99Tc, have been detected in the ground water beneath the Hanford Site. Therefore, selected samples are analyzed for these contaminants. Analyses for 238u, fluoride, and chromium are also made on ground-water samples taken from areas where these substances are present. Organic chemicals are not presently included in the list of routine analyses performed on the samples. However, a few special samples have been collected and analyzed for organics in an attempt to determine if there are substantial quantities of these contaminants in the ground water beneath the Hanford Site. Depending on the results of this investigation, monitoring of organic contaminants may become part of routine ground-water surveillance at Hanford. Supporting Tasks A number of tasks are conducted in support of the routine GroundWater Surveillance Program. These tasks contribute to the program by providing additional data on the hydrogeologic system at Hanford. 501
There are two types of tasks: field support tasks and computerassisted, data assessment tasks. Field tasks currently being conducted include geophysical logging, aquifer testing, and a study of interactions between the unconfined aquifer and the Columbia River. Data assessment tasks include geophysical log interpretation and development of various types of hydrologic models. Evaluation of Ground-Water Surveillance Data Radionuclide concentrations in the ground water beneath the Hanford Site are evaluated in terms of their respective Concentration Guides (CGs) (DOE, 1981). Radioactive materials are also compared with drinking water standards of the Environmental Protection Agency (EPA 40 CFR 141) which have been adopted by the State of Washington. The comparison between the actual concentration and th* CGs provides a conservative method for evaluating the potential significance of most water-borne materials. The CGs used are those that apply to uncontrolled areas (Eddy, Cline, and Prater, 1982). Each year data obtained by the Ground-Water Surveillance Program are used to evaluate the status of the ground water beneath the Hanford Site. A document containing this evaluation is produced annually. Included in these annual reports are maps that show the distribution of contaminants in the ground water (Figure 7 ) . A discussion of the radiological impact of the contaminated ground water is also included. Although in recent years the tritium plume maps indicate that the plume has reached the river, its contribution to the river has not been distinguished from other sources at this time (Eddy, Cline, and Prater, 1982).
CONCLUSION The two ground-water monitoring programs at the Hanford Site have been designed to effectively monitor and report specific conditions in the unconfined aquifer beneath the site. These programs have been specially developed for the hydrogeologic and operational conditions at Hanford. The effectiveness of the programs is evaluated on a continuing basis so that any necessary improvements can be implemented. Both of the programs are ongoing and will continue to monitor and document conditions in the hydrologic system at the Hanford Site.
502
U9 16" 22°30'
THUE NOF TH j MAGNETIC MOUTH
TRITIUM (pCi/ml) 1981
ESTIMATED BASALT OUTCROP ABOVE WATER TABLE
0
1
2
3
«
KILOMETERS MARCH 1992
Figure 7. Distribution of Tritium in the Unconfined Ground Water
503
46 40'
REFERENCES American Society for Testing and Materials (ASTM). 1979. 1979 Annual Book of ASTM Standard, Part 31: Water. American Society for testing and Materials, Philadelphia, Pennsylvannia. Eddy, P. A., C. S. Cline, and L. S. Prater. 1982. Radiological Status of the Ground-Water Beneath the Hanford Site, JanuaryDecember, 1981. PNL-4237, Pacific Northwest Laboratory, Richland, Washington. Graham, M. J. 1981. The Radionuclide Ground-Water Monitoring Program for the Separations Area, Hanford Site, Washington State. RHO-SA-216, Rockwell Hanford Operations, Rich land, Washington. Granam, M. J., M. D. Hall, S. R. Strait, and W. R. Brown. 1981. Hydrology of the Separations Area. RHO-ST-42, Rockwell Hanford Operations, Richland, Washington. McGhan, V. L. and D. W. Damschen. 1979. Hanford Wells. Pacific Northwest Laboratory, Richland, Washington.
PNL-2894,
Newcomb, R. C , J. R. Strand, and F. 0. Frank. 1972. Geology and Ground-Water Characteristics of the Hanford Reservation of the U.S. Atomic Energy Commission, Washington. Geological Survey Professional Paper 717, U.S. Government Printing Office, Washington, D.C. U.S. Department of Energy. 1981. Environmental Protection Safety, and Health Protection Program for DOE Operations. DOE Order 5480.1, Chapter XI, U.S. Department of Energy, Washington, D.C. U.S. Department of Energy. 1981. Environmental Protection Safety, and Health Protection Information Reporting Requirements. DOE Order 5484.1. U.S. Department of Energy, Washington, D.C. Wilbur, J. S. and M. J. Graham. 1982. Results of the Separations Area Ground-Water Monitoring Network for 1981. RH0-HS-SR-5P, Rockwell Hanford Operations, Richland, Washington.
504
SESSION NINE EMERGENCY ASSESSMENTS
9A
ACCIDENT ANALYSIS AND DOE CRITERIA
Joseph M. Graf John C. Elder Los Alamos National Laboratory Los Alamos, New Mexico
ABSTRACT
In analyzing the radiological consequences of major accidents at DOE facilities one finds that many f a c i l i t i e s fall so far below the limits of DOE Order 6430 that compliance i s easily demonstrated by simple analysis. For those cases where the amount of dispersive
energy
available
are
enough
radioactive material for
accident
and the
consequences
approach the limits, the models and assumptions used become c r i t i c a l .
to In
some cases the models themselves are the difference between meeting the c r i t e r i a or not meeting them.
Further, in one case, we found that not
only did the selection of models determine compliance but the selection of applicable criteria
from different
chapters of Order 6430 also made
the difference. DOE has recognized the problem of different
criteria
in
different
chapters applying to one facility, and has proceeded to make changes for the sake of consistency. needed
in
an
accident
We have proposed to outline the specific steps analysis
parameters, and assumptions.
and
suggest
appropriate
models,
As a result we feel DOE siting and design
criteria will be more fairly and consistently applied.
507
INTRODUCTION Emergency planning, reactors
and nonreactor
preparedness nuclear
Energy (DOE) Complex are outlined
and response
facilities
within
programs the
for
Department
both of
in DOE Order 5500.3 (DOE 5500 1981).
This Order as well as DOE Orders 6430 (DOE 6430 1981) and 5480.1A (DOE 5480 1981) Chapter V refer
to the Nuclear Regulatory Commission's (NRC)
10 CFR 100 guidelines (NRC Title 10) for establishing protective response recommendations and for accident consequence evaluation. reactors
are
somewhat
different
from
large
Although DOE
commercial
light-water
reactors, application of 10 CFR 100 to reactors i s relatively forward compared to i t s application to nonreactor f a c i l i t i e s .
straight
10 CFR 100
i s unique in the U.S. as a regulation promulgating numerical guidelines for
evaluating
the radiological
dose consequences of major
accidents.
Because of i t s key importance in accident evaluation and emergency planning 10 CFR 100 must be explained with regard to i t s intended use and adapted for use at DOE f a c i l i t i e s .
CRITERIA IN DOE ORDERS Several DOE Orders contain criteria to be used in evaluating the consequences of major accidents at nuclear f a c i l i t i e s .
The criteria are
typically general, but refer to 10 CFR 100 for situations where numerical guides are needed.
Radiological Emergencies Requirements for the development of DOE s i t e
emergency plans and
procedures for radiological emergencies are given in DOE Order 5500.3. In establishing four levels of response to emergencies the Order refers to
Protective
Response
Recommendations
numerical radiation doses to individuals
508
(PRRs).
These
are
in the population
projected which may
trigger protective response.
The Order recognizes the need to establish
site specific planning zones and PRRs because of the wide variation in types of DOE nuclear facilities and potential radiological emergencies. An attachment to DOE 5500.3, entitled "Plan Criteria,"
establishes
acceptable criteria for the development of site specific emergency plans and procedures.
Again i t
totally applicable to all
recognizes that no single criterion DOE s i t e s ,
can be
but suggests in a footnote
that
existing site criteria or 10 CFR 100 guidelines can be used. Nuclear Facility Safety The "Plan Criteria51 referred
to above state
that existing
safety
analysis reports (SARs) should be reviewed for appropriateness and used where applicable.
SARs are required by Chapter V of DOE Order 5480.1A as
a basic program element.
The basic requirement i s for "An independent
safety analysis review process that includes a formal documented system for
the
identification
and
control
of
risks
through
preparation,
independent review, and approval of safety analysis." Specific numerical guidance is not given in Chapter V but one (the first)
stated
purpose is to assure that nuclear
facilities
are
sited,
designed, constructed, modified, operated, maintained, and decommissioned in accordance with generally uniform standards, guides and codes that are consistent with those applied to comparable licensed nuclear
facilities.
In context this could only refer to the NRC's regulations, Title 10, and Regulatory Guides.
General Design Criteria
DOE f a c i l i t i e s are to be sited and designed in accordance with the Department's "General Design Criteria," DOE Order 6430 (draft). I of these criteria i s rather general and calls for
Chapter
a safety
analysis
that includes measures taken to achieve accepted levels of risk.
Chapter
509
XXI of the Order specifies
the use of the
10 CFR 100 approach
to
evaluation of accidental releases from a siting standpoint and requires use of lifetime dose commitments and dose guidelines "comparable to those of 10 CFR 100."
Chapter XXII of the Order spscifies particular 1-year
dose commitments for design of bays for uncased HE-Plutonium a c t i v i t i e s .
Use Of NRC Criteria Applying NRC criteria
as suggested by the above DOE Orders might
appear to be a simple matter.
However, as soon as one attempts to use 10
CFR 100, several facts become clear: 1)
10 CFR 100 was written for
a specific
accident at a specific
type of plant; that i s , a major loss of coolant accident at a light-water reactor,
2)
The dose limits (25 rem whole body, 300 rem thyroid) are stated for the organs of interest in the event of a release of
fresh
fission products, 3)
The time intervals over which doses are to be evaluated those appropriate for
are
slow leakage of radionuclides from a
containment structure, and 4)
The d e f i n i t i o n s of
s i t e boundary, low population
population center distance are those appropriate
for
zone and a large
central-station power plant. Because of these facts 10 CFR 100 really doesn't apply to the majority of DOE f a c i l i t i e s .
Except for
a few DOE r e a c t o r s ,
accidents at DOE
f a c i l i t i e s directly comparable to a major loss of coolant accident for a reactor have not been defined.
Dose commitments to organs other than
thyroid and whole body must be evaluated. government-owned
Distances to the boundary of
land are typically very large compared to commercial
510
reactor
site-boundary
distances,
yet
many
facilities
are
in
close
proximity to other f a c i l i t i e s that are quite unrelated programmatically or that are administrative f a c i l i t i e s .
Most f a c i l i t i e s to which the DOE
c r i t e r i a above are applied have an inventory of radionuclides much less than that of a large nuclear power plant.
Application To DOE Facilities Because
of
the
relatively
low
inventory
of
dispersive energy available in most DOE f a c i l i t i e s
radionuclides
and
compared to nuclear
power plants, details of accident postulation and consequence evaluation have not been c r i t i c a l .
However, for f a c i l i t i e s where there are large
dispersive forces such as weapons assembly f a c i l i t i e s and f a c i l i t i e s with large inventories of tritium or transuranic elements the details of the calculations and the selection of criteria can become very important.
An example of this i s given in Figure 1, in terms of the amount of plutonium that would have to be inhaled to just meet the stated Values range from 0.13 ygm to 43 »igm.
limit.
The c r i t e r i a themselves vary over
two orders of magnitude depending on which Chapter of DOE 6430 selected.
is
Choosing between the computer codes DACRIN (Houston 1976) and
INREM (Killough 1976) produces a variation of a factor of three in some cases.
Assumption of compound class Y or class W adds another order of
magnitude to the variation.
MODELS For any standard specified.
or l i m i t a t i o n ,
the way i t
No comparison can be valid if
parameters are not the same in each case.
i s to be used must be
the models, assumptions and The NRC has addressed this for
10 CFR 100 through the report, TID-14844 (DiNunno 1962).
511
This report,
DOE CRITERIA
(r«f: DOE ORDER 6430 XXI 6.b.(2).(o)) l i m i t s lor Expotur* of Public: 1 Vaor Lung Ooaa 1.5 R«m 1 Vaor Bon* Dot* 3.0 Ram Totol Body Dot* 2.9 Ram
PLUTONIUM FACILITIES ( r a t : DOE OREDER 6430 XXI 6.0.(1)) Ltmlta for Exposura of Public (10 CFR 100): SO Yaar Sona Data
ISO R«m
Totol Body Deta 25 Ram 50 Yaar Lung Doaa 75 Ram
0.01
0.1
1
10
100
AMOUNT Or PLUTONIUM INHALED TO MEET LIMIT ( p g m )
Fi%ur° 1.
C o m m - M son of 00? criteria in tsrms
thf .-•.m^unt-
Plutonium tin at would have to be j.nb-il«d t-> just meet tlr= c r i t e r i o n . Bar.?, r-epresent the v a r i a b i l i t y of trhs calculated value for different input param°tf?rs.
512
which i s
specifically
referenced
in
10 CFR 100 as further
guidance,
provides d e t a i l s that include: 1)
The
amounts of
fission
products
released
into
the
reactor
containment, 2)
The fraction of the iodine removed by absorption or plate-out,
3)
The containment leak r a t e , and
4)
The atmospheric dispersion type to be used.
Given the reactor factors,
power level,
TID-148H1 provides a l l
breathing rates and other
the
fission
product
conversion
parameters needed to calculate
prescribed distances based on inhalation of from
the
cloud.
the
iodines and external dose
Without
such
prescriptions
the
individual analyses could well lead to interminable discussion.
We have looked at three pertinent parameters in dose calculation to evaluate
the
variations
that
might be
found
in
each;
meteorological
dispersion, dose models and cancer induction risk factors. Meteorological Dispersion
The Gaussian plume model with Pasquill s t a b i l i t y classes i s widely used for atmospheric dispersion calculations. evaluated at
For a ground level release
1 km the relative concentration varies from 3 x 10~
class A to 7 x 10
for class F, a factor of two hundred.
that selecting class F usually would be prudent. disadvantages.
First,
it
might
be
unrealistically high concentrations.
too
I t might seem
However, there are two and
indicate
Second, in the case of
elevated
r e l e a s e s c l a s s A r e s u l t s in doses higher locations.
for
conservative
than c l a s s F a t
close-in
The NRC recommends collecting s i t e specific data (NRC 1.115-
1979) and evaluating the probability of occurrence of a given dispersion
513
condition.
Although clearly more r e a l i s t i c than an arbitrarily selected
value, even t h i s method can lead to a range of values spanning more than an order of magnitude if
the differences
between the 5% and the 50%
probable values are considered.
In s i t u a t i o n s where a l a r g e amount energy i s released accident,
as
in
an
explosion,
calculation
of
dispersion
in
the
may
be
complicated further by the elevation of the material in a quickly rising cloud.
Special computer codes such as DIFOUT (Luna 1969) t h a t
greatly different
from those for
reactor
plumes must be used.
are
These
models may predict concentrations two or three orders of magnitude different than simple Gaussian plume models. Dose Models Calculation of doses from external exposure to radionuclides may be subject
to
some
semi-infinite
discussion
but
many
cloud models described
(Slade 1960).
analysts
simply
accept
the
in Meteorology and Atomic Energy
These are supported by their
use in several
Regulatory
Guides. Calculating doses to internal organs from inhaled radionuclides may provide more choices.
The NRC's Regulatory Guide 1.109
conversion factors
radionuclides important for
for
reactor
but factors for transurantc elements are not included. calculating International
doses
from
inhaled
Commission on
transuranics
Radiological
important revisions in the l a s t decade.
provides dose evaluation,
The models for
recommended
Protection
(ICRP)
by
the
have
seen
Computer codes based on these
models can produce dose estimates differing magnitude depending on the input parameters.
by more than an order of The variations shown in
Figure 1 are mostly due to the selection of clasj W compounds as opposed to class Y.
But differences in the computer codes (DACRIN vs. INREM) can
produce variations of a factor of three.
514
Risk Factors for Cancer Induction 10 CFR 100 s p e c i f i e s dose l i m i t a t i o n s for whole body dose and thyroid dose only. comparable
to
the
In p r a c t i c e the NRC has used numbers roughly 25 rem whole body number
evaluating doses to other organs.
in
terms of
risk
for
Comparable risk has been determined by
the ratio of the annual dose limit
for
the organ and the annual dose
limit for the whole body. Several recent publications (ICRP 1977, NAS 1980, UNSCEAR 1979) have addressed the relative risk of cancer mortality as a function of organ dose.
As a result we can make new estimates of organ doses that give a
cancer risk comparable to 25 rem whole body.
TABLE 1.
Organ Doses Yielding
These are summarized below:
Comparable Risk of
Mortality as a 25 rem Whole Body Dose (Rem)
BEIR III ORGAN Whole Body
ICRP 26
UNSCEAR
APP. A
25
25
25
Thyroid
500
500
Lungs
125
100
130
Bone
500
1000
4000
Note:
Where a range of risk values was
cited in the reference,
the value of the
upper end of the range was used here.
515
Cancer
Some of the values d i f f e r markedly from the values now commonly used (150 rem bone, 75 rem a l l other o r g a n s ) . Considering t h a t a l l of t h e above factors a r e multiplied together in c a l c u l a t i n g dose and cancer r i s k ,
i t i s easy t o see how v a r i a t i o n of
three or four orders of magnitude can r e s u l t in the f i n a l v a l u e s . ACCIDENT ANALYSIS GUIDE The sections above on DOE c r i t e r i a and models a r e intended to show t h a t there i s plenty of room t o use c r e a t i v e a n a l y s i s in demonstrating compliance with c r i t e r i a . d i s t i n c t disadvantages:
This has c e r t a i n advantages but also has two 1) each analysis has to stand alone,
selecting
or defining the c r i t e r i a t h a t w i l l be used and j u s t i f y i n g a l l parameters and assumptions used, and 2) i t i n v i t e s disagreement within DOE as a r e s u l t of d i f f e r e n t i f well intended analyses.
We have proposed and have
s t a r t e d t o d r a f t a guideline document to permit a more common technical and
philosophical
basis
for
radiological
accident
analysis
a t DOE
facilities. The guide as envisioned will proceed step by step through the analysis for representative DOE facilities.
NRC's use of 10 CFR 100 as
described in the Regulatory Guides and the Standard Review Plan will be outlined.
Accidents for DOE facilities comparable to a major loss of
coolant accident will be postulated and analyzed.
Specific items we plan
to address include: 1)
Site boundary location; comparability to site boundaries for licensed facilities
2)
Credit to be given for engineered safety features
3)
Meteorological analysis for rare occurrences
516
i|) Dose calculation methods 5)
Risk to individual organs from a given dose
6) Other DOE facilities located nearby 7)
Environmental contamination calculated doses, and
and
its
impact
relative
to
8) Limits for accidents more credible than the one comparable to a loss of coolant accident. Such specific guidance would clearly need thorough peer review. individuals
within
the
DOE
complex
are
using
approaches
that
Many are
perfectly valid in context but differ in detail from those used by others.
We are making an effort to seek out persons interested in the
guide and to obtain their critical review. CONCLUSION Several DOE Orders including the ones for emergency planning and response, nuclear facility safety and general design refer to 10 CFR 100 for numerical guidance relating to the consequences of major radiological accidents.
10 CFR 100 is specific for determining site suitability in
the event of a loss of coolant accident at a large nuclear power plant. It therefore needs translation and adaptation for use in the many other cases to which it is applied.
To this end DOE has started a
project to draft guidance for analyzing major accidents at DOE nonreactor nuclear facilities.
517
REFERENCES
DiNunno, J . J . , F . D. Anderson, R. E. Baker, and R. L. W a t e r f i e l d . 1962. C a l c u l a t i o n of Pi stance Factors for Power a nd T e s t Reao t o r Sites. TID-1 '4844, U.S." Atomic E n e r g y C o m m i s s i o n , D i v i s i o n o f L i c e n s i n g and R e g u l a t i o n . Department of Energy. 1981. Reactor and Nonreactor Nuclear F a c i l i t y Emergency P l a n n i n g , P r e p a r e d n e s s and Response Program for Department of Energy O p e r a t i o n s . DOE Order 5 5 0 0 . 3 . Department of Energy. 1981. Environmental P r o t e c t i o n t S a f e t y , and Health P r o t e c t i o n Program for DOE O p e r a t i o n s . DOE Order 5480.1 A. Department of Energy. 1981. General Design C r i t e r i a of Energy F a c i l i t i e s . DOE Order 6430 ( D r a f t ) .
for
Department
Houston, J . R., D. L. S t r e n g e and E. C. Watson. 1976. DACRIN - A Computer Program for C a l c u l a t i n g Organ Dose from Acute or Chronic
Radionuclide I n h a l a t i o n . BNWL-B-389.
Pacific
Northwest Laboratories
report,
1CRP. 1977. Recommendations of the I n t e r n a t i o n a l Commission on Radiological Protection. ICRP Publication 26, Pergamon Press, Mew York. K i l l o u g h , G. G., D. E. Dunning, J r . , and J . C. P l e a s a n t . 1978. INREM Il.y. A Computer Implementation of Recent Models for Estimating the Dose Equivalent to Organs of Man from an Inhaled or Ingested Radionuclide. WJREG/CR-0114, ORNL/NUREG/TM-84 .~ Luna, R. E . , and H. W. Church. 1969. DIFOUT:. A Model for Computation of Aerosol Transport and Diffusion in the Atmosphere. Sandia National Laboratory Report. SC-RR-68-555. National Academy of Sciences. 1980. The Effects on Populations of Exposure to Low Levels of Ionizing Radiation: 19.80. Committee on the Biological Effects of Ionizing Radiation, National Academy P r e s s , Washington, 1980. Nuclear Regulatory Commission. 1979. Atmospheric Dispersion Models for P o t e n t i a l Accident Consequence Assessments at Nuclear Power P l a n t s . Regulatory Guide 1.145. Nuclear R e g u l a t o r y Commission. Reactor S i t e C r i t e r i a . Federal Regulations, T i t l e 10 Part 100. S l a d e , D. H. 1968. Washington, DC.
Code of
Meteorology and Atomic Energy Commission.
United Nations S c i e n t i f i c Committee on the Effects Radiation. 1979. Sources and E f f e c t s of I o n i z i n g United Nations, New York. 518
of Atomic Radiation.
9B EMERGENCY PREPAREDNESS FOR NONRADIOLOGICAL INCIDENTS AT HANFORD
R. D. Gilmore Hanfor-1 Ei*v ironmental Health Foundation B. 1. McMurray and K. R. Heid Pe.ific Northwest Laboratory* Richland, Washington
ABSTRACT The multiple contractors at Hanford are committed to a common mission for the safe, secure, environmentally sound, and cost effective performance of Hanford's assigned energy and defense programs. This has resulted in utilization of site support service organizations for providing many essential emergency, security, and technical skills. These organizations support all Hanford contractors to avoid unnecessary duplication of resources. Because of Hanford's unique multiple contractor operations in common facilities, special coordination and communication requirements exist. This paper describes Hanford's preparedness for the effective utilization of available resources during nonradiological emergencies.
INTRODUCTION Hanford, like any other major industrial facility, is not immune from the risk of experiencing an emergency incident. There is an abundance of potentially hazardous materials used on a daily basis. On August 13, 1981, the U.S. Department of Energy (DOE) issued management order 5500.2 entitled "Emergency Planning, Preparedness, and Response for Operations." This document included in the purpose statement requirements to provide planning, preparedness, and response to operational emergencies involving toxic or other hazardous material. The receipt of this directive prompted a review of emergency preparedness for nonradiological incidents at Hanford. The current emergency preparedness programs for nonradiological incidents are not dissimilar to those long established at Hanford. The procedures utilized for activation of, and communication between, emergency * Pacific Northwest Laboratory Department of Energy.
is operated
519
by Battelle
for the U.S.
resources and the responsible event facility management have been proven effective. The Hanford program could be adapted to meet similar needs in other large, complex, industrial facilities. BACKGROUND The U.S. Department of Energy's Hanford site is located on a 570 square mile site in south-central Washington State (Figure 1). Approximately 11,500 contractor employees and 350 Department of Energy employees perform the actual research, development and demonstration, production, chemical processing and site support service functions at Hanford. The Hanford site also includes complementary commercial operations on site leased land and immediately adjacent properties. Examples of on-site commercial operations include a steam generating plant, nuclear power production reactors, and radioactive waste disposal facilities. Adjacent operations include a nuclear fuel fabrication facility and private research laboratories. EMERGENCY ORGANIZATIONS Hanford has adopted a practice of using normal channels for obtaining assistance whenever possible during an emergency. When feasible this includes the utilization of existing alarms, communications, equipment, experience, facilities, organizations, and personnel. This results in common, or at least similar, plans for most activities during emergencies. A review of the Hanford emergency organizations can be categorized as (1) the event contractor; (2) facility, area, and management emergency teams; and (3) support services such as fire, patrol, medical, industrial hygiene, or special radiological assistance capabilities. The Department of Energy operates the Emergency Action Coordination Team (EACT). Under emergency conditions it is the responsibility of the event contractor's facility emergency team to provide the initial alarms and notifications to assure the safety of the building occupants; to rescue injured or trapped individuals; to plan for the stabilization of the event and for recovery of the facilities. The area emergency teams are responsible for the safety of nearby occupants; for notifying other facilities in the area; and for providing support requested by the facility emergency team. The management emergency team provides the necessary notifications to the Department of Energy, primarily through the DOE Emergency Action Coordinating Team, and provides any necessary support to the facility or area emergency teams. The management emergency team assumes responsibility for meeting public relations needs and provides support and direction for recovery operations. Support service organizations include fire and patrol, medical, meteorology, radiological assistance, environmental protection, and industrial hygiene as examples. The Industrial Hygiene organization
520
C
rown sumv smsw OfMMTON I.A.IONB HANKWO OmMTIONS ONTIIMSKMUS4 TRAMTOKTATION
AKUUMNC
FIGURE 1. U.S. Department of Energy Hanford Site 521
of the Hanford Environmental Health Foundation (HEHF), as an illustration, provides specialized industrial safety and hygiene skill resources as required during an emergency. Examples may include information and initiation of action in regards to toxic materials; fire, explosion, or reactivity hazards; hazardous working conditions or other risks to personnel during emergency response or recovery operations. They record, analyze, and prepare data as well as advise the appropriate parties of critical or missing items as observed during the handling of an incident. A function such as this becomes especially important when unique potential hazards, not previously anticipated, are created by an incident. Assistance from support organizations in an emergency is provided directly to the event site contractor (Figure 2 ) . The support organization provides services based on input from the event contractor and provides independent evaluations and recommendations to the personnel responsible for managing the incident. The Department of Energy's Emergency Action Coordinating Team (EACT) assures the availability of adequate support resouces such as fire, medical or patrol to the event contractor. It also provides contact with other non-Hanford agencies as required.
PNLPERSONNEL DOSIMETRY (RADIOLOGICAL INCIDENT) PNL ENVIRONMENTAL EVALUATION (RADIOLOGICAL INCIDENT)
EVENT SITE CONTRACTOR PNL METEOROLOGY (AIRBORNE RELEASE INCIDENT)
INPUT EVALUATION
HEHF INDUSTRIAL HYGIENE (IMONRADIOLOGICAL INCIDENT)
FIGURE 2. Normal Assistance from Support Organizations When Uniform Dose Assessment Center not Activated 522
UNIFORM DOSE ASSESSMENT CENTER (UDAC) In a major incident coordination between all impacted Hanford parties is essential. The Uniform Dose Assessment Center (UDAC), a fully equipped and operating emergency coordination facility, has been established at Hanford. It is operated for DOE by Battelle, and can be mobilized and staffed with resource personnel from both Battelle ana Hanford Environmental Health Foundation. The purpose of UDAC is to provide technical assistance to the Department of Energy and its contractors; to assess the consequences of any incident; and to recommend appropriate actions. To achieve this, UDAC has been staffed with technical expertise from support service organizations in the areas of radiological, environmental, meteorological, industrial and biological safety. Only staff members that can be totally dedicated to UDAC during an emergency are utilized. UDAC is not intended to replace normal sources of assistance available to contractor organizations. Any incident of sufficient potential to have an impact on adjacent facilities, operating areas, or off-site involves activation of UDAC. UDAC can also be activated at the request of any Hanford site organization. The primary functions of UDAC are to provide the necessary interaction with the Department of Energy; to provide environmental and personnel exposure evaluations; to provide data plotting to anticipate future consequences of any incident; and to provide consultation on industrial health and safety, meteorology, biolological safety, environmental protection, and radiation monitoring. UDAC provides information and support as requested by the event contractor. UDAC'also keeps the assigned representative from the Washington State Emergency Team appropriately informed of the status of all events. When UDAC is activiated, the original communication lines for the event site contractor are not disrupted (Figure 3 ) . UHAC establishes a coordinating body to effectively utilize personnel skills and related resources. UDAC is normally staffed by senior personnel from the technical operating components of the represented support services organizations. During an emergency, this coordinating organization brings together in one facility technical skills with the necessary administrative, communication, and information handling support required. UDAC serves a coordinating role for both the Department of Energy and the event site contractor. It is important to emphasize that in case of an incident, the normal lines of communication always remain open with the technical service organizations. It is optional if the event contractor and UDAC choose to communicate directly. FACILITY EMERGENCY PREPAREDNESS Hanford contractors have prepared facility emergency plans where appropriate. These documents provide brief, clear, and concise procedures
523
COORDINATOR
EACT
t PNL PERSONNEL DOS1METRY (RADIOLOGICAL INCIDENT) PNL ENVIRONMENTAL EVALUATION (RADIOLOGICAL INCIDENT)
DIRECTOR 1i
ADMIN. SUPPORT
EVENT SITE CONTRACTOR PNL METEOROLOGY [AIRBORNE RELEASE INCIDENT)
COMMUNICATIONS
PLOTTING — ^ INPUT *•- —
EVALUATION
\^
HEHF INDUSTRIAL HYGIENE (NONRADIOLOGICAL INCIDENT)
RECORDER
FIGURE 3. Normal Assistance When Uniform Dose Assessment Center Activated for effectively managing credible potential emergency incidents, including those of a nonradiological nature. Such plans include a description of facility emergency organizations, designation of responsibilities, identification of available resources, and notification requirements. Facility emergency preparedness programs include training of personnel in the appropriate actions to be taken in credible incidents. Reviews are regularly made of the adequacy of emergency preparedness of all Hanford facilities. A program has been implemented to identify those facilities where potentially hazardous nonradiological materials are procured, used, or stored in greater than 500-unit lots (e.g. 500 kilograms). The 500-unit criterion was arbitrarily selected to minimize the inappropriate review or development of emergency plans for minor laboratory inventories of materials. FUTURE PLANS Specific development needs have been identified for emergency preparedness for nonradiological incidents. Included are additional dispersion modeling for selected nonradiological potential release candidates,
524
additional training of emergency response teams in dealing with nonradiological incidents, training additional field personnel in the use of direct reading instruments, and identifying in-place monitoring requirements. Additional training exercises involving nonradiological contaminants are planned. Special emphasis will be placed on addressing the adequacy of resources in the areas of personnel, training, and instrumentation. Monitoring capabilities for personnel and environmental sample collections must be evaluated. A need for establishing action levels for a variety of nonradiological contaminants is also recognized. Ct CLUSION The multiple contractors at Hanford are committed to a common mission for the safe, secure, environmentally sound, and cost effective performance of Hanford's assigned energy and defense programs. This has resulted in utilization of site support service organizations for providing many essential technical skills, these organizations support all Hanford contractors and avoid unnecessary duplication of resources. Experience has shown that site support service groups provide a mechanism of cost-effective utilization of specialized technical skills during emergencies. Because of Hanford's unique multiple contractor operations in common facilities, special coordination and communication requirements exist. UDAC is effective in providing coordination of resources and consultation capabilities for dealing with nonradiological incidents at Hanford. ACKNOWLEDGEMENTS This work was sponsored by the U. S. Department of Energy under contracts DE-AC06-76RL0-1830 and -1837.
525/53^
9C DOSE PROJECTION CONSIDERATIONS FOR EMERGENCY CONDITIONS AT NUCLEAR POWER PLANTS
G. A. Stoetzel, R. W. Poeton, J. V. Ramsdell, 0. C. Powell, and A. E. Desrosiers Pacific Northwest Laboratory Rich!and, Washington
ABSTRACT This paper will review the problems and issues associated with making environmental radiation dose projections during emergencies at nuclear power plants. The review will be divided into three areas including source term development, characterization of atmospheric dispersion and selection of appropriate dispersion models, and development of dosimetry calculations for determining thyroid dose and whole body dose for ground level and elevated releases. Uncertainties associated with these calculations will also be discussed.
INTRODUCTION Offsite dose assessment during nuclear power plant emergencies is required of a l l licensees by T i t l e 10 Code of Federal Regulations, Part 50.47(b)(9), which requires that "adequate methods, systems and equipment for assessment and monitoring actual or potential o f f s i t e consequences of a radiological emergency are in use." Currently, there is no federal guidance on real-time dose assessment for use in emergency response f a c i l i ties (ERFs) during emergency situations. The only available federal guidance is specific to pre-operational (licensing) review or routine effluent releases. The purpose of this NRC-funded study was to review problems and issues associated with o f f s i t e dose assessment in emergency situations. Dose assessment can be performed at any one of three locations during an emergency including the control room, technical support center (TSC), or emergency operations f a c i l i t y (EOF). Calculations made during the i n i t i a l hour or two of an emergency would generally be performed in the control room, and are referred to as rapid methods. Once the TSC and EOF are established, dose assessment functions are transferred to either f a c i l i t y from the control room. Calculations done in the TSC or EOF are commonly referred to as intermediate methods. The rapid methods are generally less sophisticated than the intermediate methods used in the TSC or EOF, where the highest level of expertise in evaluating the consequences of environmental releases is expected to be found. 527
There are three basic phases in performing offsite dose projections; source term development, characterization of atmospheric dispersion and selection of appropriate dispersion models, and development of dosimetry calculations for determining thyroid dose and whole body dose for ground level and elevated releases. The remainder of this paper will provide a brief review of these phases. The discussion will concentrate on calculating doses from the passing plume. The audience is referred to a recently published report (Stoetzel et al. 1982) for a more detailed discussion. SOURCE TERM The first step in assessing the radiological consequences of an accident is to determine the actual or potential source term. This includes the quantity and radionuclide makeup of the radioactive material released to the environment as well as its temporal variation. During the initial and intermediate phases of an accident, involving hours rather than days, the releases of radioactive material to the atmosphere are most important. Noble gases and halogens dominate the initial assessments of the radiological impacts, although other radionuclides may be important for intermediate or long-term considerations, depending on the severity of the accident. There would be several different situations where source term estimates may be needed during an emergency, including the following: •
Monitored release - release is occurring and release pathway is monitored.
•
Unmonitored release - release is occurring and release pathway is unmonitored.
•
Potential release - reactor conditions indicate the potential for a release.
Monitored Release Ideally, monitoring of releases during m accident would provide immediate direct information on the quantities, forms, and identities of radionucl ides being released. Some effluent monitors incorporate gamma ray spectrometers. However, in general, the information available is more limited. In some cases only gross dose rate measurements of atmospheric releases are available. Typically, monitored systems are capable of providing measurements of total noble gas, radioiodine, and particulate releases as specified by NUREG-O737 (U.S. NRC 1980a). Separate absorbers or filters are used to remove the radioiodines and particulates, snd the noble gases are monitored in realtime. Such a system might be found in the facility's main exhaust stack. The following potential sources of error need to be considered in evaluating such a system Tjr estimating accident source terms:
528
1) Assumptions about the mix of radionuclides in each of the separate fractions, particularly iodines and noble gases, should be stated explicitly. The relationship between noble gas and radioiodine release fractions becomes important as it determines the scope of necessary protective actions. Since the thyroid Protective Action Guides are five times higher than those for whole body exposure (U.S. EPA 1975) and one Ci/m3 of radioiodine will produce a dose to the limiting organ (child thyroid) which is 1000 times the gamma dose from one Ci/m3 of noble gases at one hour past shutdown, thyroid doses would dominate the protective action decisions for any release to the environment where the ratio of iodine to noble gas release fraction is greater than 200. 2) Particulate activity has the potential for considerable impact on whole body exposures. This effect may be relatively more important when releases of radioiodines are reduced. Significant errors may be introduced by failure to consider important groups of radionuclides in accident assessments. For example, soft metals (i.e., Sb, Te, Rb, Sr, Ba) would contribute 94% and 97%, respectively, to the bone marrow and lung dose for an accident with the following source term: noble gases - 100%, halogens - 5%, cesium - 2.5%, and soft metals - 1.5% of the core inventory released. 3) Any necessary corrections for radionucl ide mixes, other than those for which the system is calibrated, should be provided. Corrections may be needed for time after shutdown and for the type of accident (e.g., gap release vs. core meltdown). For example, assuming all noble gas activity to be 1 3 3 Xe will underestimate dose by a factor of between 16 and five during the first 12 hours post-shutdown. Considering all radioiodine activity to be 1 3 1 I will overestimate dose by a factor ranging from five to two over the first 24 hours after shutdown. Figures 1 and 2 are graphs of correction factors which can be used to 133 correct measurements made assuming 1all noble gas activity to be Xe and all radioiodine activity to be 3 1 I . Initial release estimates should be based on available plant parameter information. Field measurements would provide important feedback relative to these estimates, but they must be used with caution. Field measurements are subject to errors in sampling techniques, and even measurements properly made can be affected by local and temporal variations in concentrations. Integrated, rather than instantaneous, radiation measurements should be used wherever possible. Field measurement data should be examined for internal consistency and compared with known atmospheric conditions before being used to modify release estimates. Unmonitored and Potential Releases Unmonitored and potential releases should be estimated on the basis of previously calculated accidents, plant specific information, and known plant conditions. For instance, in the case of release of coolant without 529
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FIGURE 1 .
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Correction Factor f o r Noble Gas Measurements Assuming a l l A c t i v i t y i s 1 3 3 Xe
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FIGURE
Correction Factor for Radioiodine Measurements Assuming a l l A c t i v i t y i s 1 3 1 I
fuel damage, the most recent coolant analysis could be used to estimate the r e l e a s e . In some cases, d i r e c t i n - p l a n t surveys ( e . g . e f f l u e n t l i n e s or containment entrances) may provide information on which to base release e s t i m a t e s . Shielding assessments and other information necessary f o r the a p p l i c a t i o n of such methods should be a n t i c i p a t e d and made a v a i l a b l e in emergency procedures. Accident c l a s s i f i c a t o n s i n the Reactor Safety Study (U.S. NRC 1975) and Draft NUREG-0771 (U.S. NRC 1982a) could be used to estimate possible 530
releases to the environment. Uncertainties are large for a source term developed generically and applied to a specific accident. For whole body dose, an uncertainty of a factor of 100 to 10,000 can be expected for a core melt accident. Sampling and analysis capability as specified by NUREG-0737 (U.S. NRC 1980a) should be in place to provide radionuclide data for input into source term estimates. Sampling results should be available within three hours following the decision to take a sample. Methods should be provided for modifying source term estimates based on radionucl ide mix information from sample analysis* Environmental measurements could also be used to assist in the source term estimate for an unmonitored release. ATMOSPHERIC DISPERSION Many meteorological aspects of the emergency response decision-making process at nuclear facilities involve estimating the transport, diffusion, transformation and depletion of material released, or projected to be released, to the atmosphere. Problems associated with making these estimates have been outlined by Van der Hoven (1981) and discussed at length in Nuclear Regulatory Commission sponsored workshops (Dabherdt et al. 1982 and Watson et al. 1982). Although no single set of procedures for estimating transport and diffusion of airborne material is appropriate for all nuclear facilities, several models have been developed under federally funded programs that may provide a starting point for the development of site specific procedures for a nuclear facility. Potential model features and the importance of the features of site topography and release height will be discussed. Pi spersion Model s Appendix 2 of Regulatory Guide 1.101 (NUREG-0654) (U.S. NRC 1980b) discusses assessment of potential and actual exposures and doses. Two types of atmospheric models are defined for making these estimates. Class A models are intended for initial, rapid dose estimates. These estimates are expected to be available within 15 minutes of the determination that an emergency condition exists, and are for use within the plume exposure Emergency Planning Zone (EPZ). Class B models are intended for use in the larger ingestion EPZ. Both Class A and B models are to be site specific in that they should use local meteorological data and account for the effects of topography in making diffusion estimates. Other than the difference in areal extent of the EPZs to be covered, the primary distinction between the model classes is in treatment of the time variations of atmospheric conditions. Class A models generally treat time variations by assuming a sequence of steady state conditions without
531
considering the transition from one state to the next. By contrast, Class B models represent nonstationary and inhomogeneous conditions explicitly. In addition to the basic difference between model types, there are differences in the features that can be expected to be included in the models. Fewer atmospheric processes will be represented in Class A models than in Class B models. Specifically, Class B models must treat deposition processes and forecast meteorological conditions. Regulatory Guide 1.101 (NUREG-0654) provides for use of Class A models for the entire ingest ion EPZ until emergency response facilities are fully operational if compensating measures are taken to account for potential errors at longer transport distances. When a starting point has been established, it is necessary to select capabilities and features to be incorporated in the ultimate Class A and B models. Clearly, the importance of features and capabilities is a function of the specific design of the facility and its topographic setting. For example, treatment of transport and diffusion from an elevated source may not be important if a facility does not have a significant potential for an elevated release. Similarly, explicit terrain representation is not important if the topography within the ingestion pathway EPZ is essentially flat and there are no major features that could affect air flow. Table 1 lists a number of potential model capabilities and features and suggests a relative importance for each as a function of general topographic setting and release height. However, the table should not be interpreted as setting Class A or Class B model requirements. Given an adaptable starting point, it should be possible to incorporate most of the features and capabilities listed in Table 1 within the current state-ofthe-art in transport and diffusion modeling. In many cases, the mathematical algorithms required are standard model components that have been used for years. Model Uncertainity Basically, atmospheric models predict averages. A range of variability around model estimates must be accepted as a natural consequence of our inability to completely define the state of the atmosphere at any time. Uncertainty estimates are generally for maximum ground-level concentrations and consider only diffusion. They do not consider other processes affecting concentrations, such as radioactive decay, dry deposition, and washout by precipitation. It is also important to note that thf\y are estimates of the uncertainty at the plume centerline, they are not estimates of uncertainty at specific points in space. This difference is significant. Estimates of uncertainty at a point in space must include uncertainty in plume position.
532
TABLE 1 .
Capabilities and Features for Emergency Response Atmospheric Transport and Diffusion Models Topographic Setting Coast Va M ev Releasei Heiqht Elevated Ground Elevated (iround Eievateo
Plains Model Feature
wound
Transport transport
1
1
!
1
1
1
Output i n i t i a l transport spaed
1
1
1
I
1
1
Temporal 1 / varying wind field
2
2
1
1
1
1
Horizontally varying wind field
--
2
2
2
2
2
Output i n i t i a l direction
2
2
2
field --
• -
--
-
2
Appropriate release he i gilt
1
i
1
1
1
I
Appropriate diffusion coefficients
1
1
1
1
1
1
Explicit terrain representation Diffusion
Hi«ing layer depth
1
Fumigation
•
Explicit terrain representation
-
—
1
1
I
1
1
2
--
7
--
2
1
—
I
—
1
Key
Level of Importance
1 7
Should be in all models Should be in models for EOF use
In 1977, the American Meteorological Society, Committee on Atmospheric Turbulence and Diffusion (American Meteorological Society 1978) estimated the upper l i m i t of accuracy for diffusion models to be about a factor of two under conditions where the meteorological parameters controlling d i f f u sion are measured near the point of release and where topographic r e l i e f is minimal. Under exceptional circumstances, which include rough t e r r a i n , wake flows, over water t r a j e c t o r i e s , extremes of atmospheric s t a b i l i t y or i n s t a b i l i t y , and ranges of more than ten km; uncertainties may be as great as an order-of-magnitude i f standard models are used. Use of site specific models may decrease the uncertainty. Working Group C at an NRC sponsored Workshop (Dabberdt 1982) estimated the uncertainty of site specific models to be about a factor of three under optimum daytime conditions with r e l a t i v e l y strong winds. Greater uncertainty, approaching a factor of ten, was estimated for conditions when the wind speed decreases toward calm. A similar range of uncertainty (factor of three to ten) was given for night time, ground-level releases. The uncertainty range was extended to a factor of 100 for complex t e r r a i n . No level of uncertainty was estimated for nonstationary meteorological conditions, such as sea breezes. 533
The uncertainty associated with model estimates is also a function of averaging time. In general, uncertainty decreases with increasing averaging time. Estimates given above are associated with hourly averages. If the averaging interval is reduced to five or ten minutes, an additional factor of two can be added to the uncertainty {Ramsdell and Hinds 1971). Uncertainties associated with instantaneous spot estimates and observations are even larger. DOSIMETRY The dosimetry review w i l l consider methods for calculating o f f s i t e dose rates and doses from known radionuclide concentrations at downwind receptor locations. The review concentrates on doses received in the plume emergency planning zone (EPZ), including external gamma radiation dose to the whole body and inhalation dose to the thyroid from the passing plume. Compared to estimates of the source term and atmospheric dispersion, dosimetry calculations are r e l a t i v e l y insensitive to the problems of projecting from a partial data base and the need for timeliness. Two caveats are necessary, however. The dosimetry code must be capable of handling the various radioelements potentially in a release and the method of plume shine calculation must coincide with the spatial model of atmospheric plume dispersion. Otherwise, the uncertainties associated with radiation transport and the pre-accident composition and d i s t r i b u t i o n of the o f f s i t e population should not cause uncertainties of the maximum o f f s i t e impact to exceed a factor of two. Intermediate Dose Calculation Methods Intermediate dose calculation methods would be used in the EOF and TSC. These methods require more rigorous modeling and the capability to include a broader release spectrum of radionuclides compared to a rapid dose calculation method. In addition, use of an appropriate computer probably would be required. External Gamma Radiation Dose from Plume Methods for calculating external gamma radiation dose from a passing plume are based on the type of release. For a ground-level release, a semi-infinite plume model is generally recommended. For an elevated release, a f i n i t e plume model is recommended. A f i n i t e plume model also could be used for a ground-level release. Using a semi-infinite plume model, gamma radiation dose rate is proportional to the average radionuclide concentraton in air and can be calculated by multiplying the average concentration by a dose conversion factor. Dose conversion factors have been calculated by many sources including but not limited to U.S. NRC (1977); U.S. NRC (1975); Kocher (1980); and Scherpelz, Borst and Hoenes (1980). For noble gas radionucl ides, dose conversion factors generally agree within a factor of two for the above sources. 534
Using a semi-infinite plume model for calculating external gamma doses from a ground-level release w i l l overestimate dose near the release point particularly under stable atmospheric conditions. This is because the model assumes that the concentration of radioactive material at ground level is constant out to i n f i n i t y in a l l directions in the hemisphere above the ground. However, in r e a l i t y , the concentration of a plume decreases at distances away from the plume centerline. Therefore, the model accounts for more radioactive material than would actually be present. Correction factors should be developed for use under these conditions. Several factors that need to be considered in order to obtain a more accurate method of calculating external gamma dose include: •
calculation of five-cm depth dose which attenuates gammas in the f i r s t f i v e cm of tissue, resulting in a whole-body dose,
•
calculation of radioactive decay during plume t r a n s i t , and
•
calculation of daughter product ingrowth during plume t r a n s i t .
The calculation of the five-cm depth dose should always be done when calculating external gamma dose from a plume. However, there are situations when ignoring radioactive decay and daughter product ingrowth during plume t r a n s i t w i l l not s i g n i f i c a n t l y affect results. Radioactive decay during plume t r a n s i t would not need to be considered for a noble gas release in which there was a long holdup time (>24 hr) prior to an environmental release. In t h i s situation short-lived noble gas nuclides (85m K r > 87 K r > 88 K r > i35 m Xe ) would decay t~> insignificant levels with the main contributor to external dose being 133Xe ( t 1 , 2 = 127 h r ) , which would decay i n s i g n i f i c a n t l y over a t r a n s i t time of several hours. As previously mentioned, f i n i t e plume models are used for elevated releases, since a semi-infinite plume model would tend to underestimate dose as i t doesn't consider shine from the elevated plume. Finite plume calculations can be complicated because spatial integration over complex plume geometries is required. Since consideration is given to the spatial d i s t r i b u t i o n of radionuclide concentrations when using a f i n i t e plume model, the interface with the meteorological transport and diffusion model is important. These calculations usually require the use of computers. Most f i n i t e plume models involve a numerical integration technique. Usage of numerical integration techniques, such as described in Regulatory Guide 1.109, Garrett and Murphy (1981), Strenge, Watson and Houston (1975), Scherpelz, Burst and Hoenes (1980), and Lahti, Hubner and Golden (1981) w i l l give the most exact results for f i n i t e plume calculations. However, i f computer a v a i l a b i l i t y or operating cost is prohibitive for the rather timely numerical integration calculations, alternate methods can be used with reasonable accuracy. Alternate methods include using f i n i t e plume correction factors (Healy and Baker 1968, U.S. NRC 1975, and U.S. EPA 1975) and the nomograms developed by U.S. NRC (1982b). 535
The WASH-1400 study (U.S. NRC 1975) and EPA (1975) used f i n i t e plume correction factors developed by Healy and Baker (1968) for calculating external gamma dose from elevated releases. The correction factors are a ratio of dose from a f i n i t e plume to dose from a semi-infinite plume of uniform concentration. The meteorological model for the f i n i t e plume portion of the calculation considers an isotropic cloud made up of a single puff release. The NRC has developed precalculated relationships for 8 5 Kr, 8 7 Kr, Kr, S9Kr, I33x e , I35x e , I37x e , I38xe and 41 Ar using the f i n i t e plume methodology in Regulatory Guide 1.109 (U.S. NRC 1977). The relationships are presented as nomograms and regression equations whereby knowing wind speed, s t a b i l i t y class, and distance to the receptor location, dose in mrem/Ci released can be determined. These nomograms and equations are discussed b r i e f l y in Dabberdt et a l . (1982) and are documented in U.S. NRC (1982b). 88
Inhalation Dose to Child Thyroid Inhalation of radioiodines and particulates are of major concern during plume passage because inhaled or ingested radioiodines concentrate in the thyroid. Several factors need to be known to compute the inhalation dose to the child thyroid (most sensitive member of population). These factors include: 1) air concentration of radioactive material at a downwind receptor location in Ci/m 3 ; 2) breathing rate in m 3 /hr; and 3) dose factor in rem/Ci inhaled. The rate at which inhalation dose commitment is received is a product of these three factors. The a i r concentration term should be corrected for radioactive decay and daughter ingrowth during plume t r a n s i t . When calculating thyroid dose, daughter ingrowth need not be considered i f only radioiodines are in the release. However, i f tellurium (Te) is present, ingrowth should be considered. Breathing rates are specified in ICRP Publication 23 (1975). Dose factors for radionuclides of interest can be calculated using lung deposition and clearance models in ICRP Publication 2 (1959) or Morrow (1966) and ICRP (1972)--Task Group Lung Model (TGLM). Inhalation dose factors are dependent on age. Age-dependent parameters include thyroid mass, fractional uptake of iodine by the thyroid, effective h a l f - l i f e in the t h y r o i d , breathing rate, and energy absorbed in the thyroid per disintegration of radioiodine. NUREG-0172 (Hoenes and Soldat 1977) and Regulatory Guide 1.109 used the i n i t i a l ICRP lung model, while U.S. EPA (1975) and WASH-1400 (U.S. NRC 1975) used the TGLM to calculate dose factors. Comparison of dose factors calculated from the above references generally agreed within a factor of two for equivalent age groups. Rapid Dose Calculation Methods Rapid dose calculation methods are intended for use in the control room during approximately the f i r s t hour of an accident until the TSC and EOF are activated. As outlined by Dabberdt et al . (1982), rapid methods should consider external gamma doses and inhalation doses to the thyroid 536
from the passing plume. Such c a l c u l a t i o n a l methods should be s i m p l e , t o allow usage by i n d i v i d u a l s w i t h a minimum knowledge of dose assessment. T y p i c a l l y , c a l c u l a t i o n s would not i n v o l v e c o n s i d e r a t i o n of as many radionucl ides as i n t e r m e d i a t e methods. Examples of these methods i n c l u d e nomograms, i s o p l e t h s , desk-top c a l c u l a t o r s , microcomputers, and s i m p l i f i e d hand c a l c u l a t i o n s . An example of a computerized r a p i d c a l c u l a t i o n method i s the I n t e r a c t i v e Rapid Dose Assessment Model (IRDAM) code (Poeton et a l . 198?). SUMMARY
In summary, the priorities for emergency dose assessment are: • a capability to develop an accident-specific source term; •
a characterization of atmospheric transport and diffusion and selection of appropriate models;
•
a development of dosimetry calculations for determining thyroid dose and whole body dose for ground level and/or elevated releases;
•
a verification that the ensemble of models for determining source term, atmospheric dispersion, and dosimetry calculations are coordinated and produce reasonable results; and
•
realization that physical properties of atmospheric transport and dispersion will limit inherent accuracy of models for monitored releases.
REFERENCES American Meteorological Society, 1973. Rull. Am. Met. Soc., 59, 1025-1026.
"Accuracy of Dispersion Models,"
Dabberdt, W. F., et a l . 1982. Workshop on Meteorological Aspects of Emergency Response Plans for Nuclear Power Plants: Proceedings and Recommendations. Final Report, SRI Project 3689, SRT International, Menlo Park, California. Garrett, A. J. and C. E. Murphy, J r . 1981. A Puff-Pi ume Atmospheric Deposition Model for Use at SRP in Emergency Response Situations. EP1595, E. I . duPont de Nemours & Co., Savannah River Laboratory, ~Aiken, South Carolina. Healy, J. W. and R. E. Baker. 1968. "Radioactive Cloud-Dose Calculations." In Meteorology and Atomic Energy 1968. Ed. 0. H. Slade, U.S. Atomic Energy Commission, Washington, Q.C. Hoenes, G. R. and J . K. Soldat. 1977. Age-Specific Radiation Dose Commitment Factors for a One-Year Chronic Intake. NUREG-0172. Pacific Northwest Laboratory, Richland, Washington. 537
International Commission on Radiological Protection. 1959. Report of Committee II on Permissible Dose for Internal Radiation. Publication 2, Pergamon Press. International Commission on Radiological Protection. 1972. The Metabol ism of Compounds of Plutonium and Other Actinides. Publication 19, Pergamon Press, New York, New York. International Commission on Radiological Protection. 1975. Report of the Task Group on Reference Man. Publication 23, Pergamon Press, Oxford, United Kingdom. Kocher, 0. C. 1980, "Dose Rate Conversion Factors for External Exposure to Photon and Electron Radiation from Radionudides Occurring in Routine Releases from Nuclear Fuel Cycle Facilities." Health Physics P., Vol. 38, pp. 543-621. Lahti, G. P., R. S. Hubner, and J. C. Golden. 1981. "Assessment of X-Ray Exposures Due to Finite Plumes." Health Physic J., Vol. 41, pp. 319-340. Morrow, P. E. 1966. "Deposition and Retention Models for Internal Dosimetry of the Human Respiratory Tract." Health Physics P., Vol. 12, p. 173. Poeton, R. W., M. P. Moeller, G. J. Laughlin and A. E. Oesrosiers. 1982. Interactive Rapid Hose Assessment Model (IRDAM) for Reactor Accidents. NUREG/CR-3012, PNL-4511, Pacific Northwest Laboratory, Richland, Washington (In Publication). Ramsdel1 , P. V. and W. T. Hinds, 1971. "Concentration Fluctuations and Peak-to-Mean Concentration Ratios in Plumes from a Ground-Level Continuous Point Source." Atmos. Environ., 5, pp. 483-495. Scherpelz, R. I., F. J. Borst, and G. R. Hoenes. 1980. WRAITH - A Computer Code which Calculates Internal and External Doses Resulting from Atmospheric Releases of Radioactive Material. PNL-3382, Pacific Northwest Laboratory, Richland, Washington. Stoetzel, G. A., P. V. Ramsdel1, R. W. Poeton, n. C. Powell, and A. E. Desrosiers, 1982. Pose Projection Considerations for Emergency Conditions at Nuclear Power Plants. NUREG/CR-3011. Pacific Northwest Laboratory, Richland, Washington. (In publication.) Strenge, D. L., E. C. Watson, and P. R. Houston. 1975. SUBDOSA - A Computer Program for Calculating External Doses from Accidental Atmospheric Release of Radionuci ides. BNWL'-B-SBl, Pacific Northwest Laboratory, Richland, Washington.
538
U.S. Environmental Protection Agency. 1975. Manual of Protective Action Guides and Protective Actions for Nuclear Incidents. EPA-520/1-75-001, Revised June 1979 and February 1980, Environmental Protection Agency, Washington, O.C. U.S. Nuclear Regulatory Commission. 1975. Reactor Safety Study - An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants. WASH-1400. NUREG-75/014, Washington, O.C. U.S. Nuclear Regulatory Commission. 1977. Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I . Regulatory Guide 1.109, Washington, O.C. U.S. Nuclear Regulatory Commission. 1980a. C l a r i f i c a t i o n of TMI Action Plan Requirements. NUREG-0737, Washington, n.C. U.S. Nuclear Regulatory Commission, 1980b. Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, Regulatory Guide 1.101 (NUREG-0654), Rev. 1), Washington, O.C. U.S. Nuclear Regulatory Commission. 1982a. Reactor Accident Source Term Assumptions. Washington, n.C.
Regulatory Impact of Nuclear (Draft) NUREG-0771,
U.S. Nuclear Regulatory Commission. 1982b. Nomograms for Evaluation of noses from a Finite Cloud. NUREG-0851, Washington, O.C. Van der Hoven, I . A., 1981. Meteorological Considerations in the Development of a Real-Time Atmospheric Oispersion Model for Reactor Effluent Exposure Pathway, NUREG/CR-2584, USNRC, Washington, O.C. Watson, E. C , ed., 1982. Proceedings of the Workshop on Environmental Assessment, Sponsored by the U.S. Nuclear Regulatory Commission, NUREG/CR-0025, Washington, O.C.
539
9D IRDAM - INTERACTIVE RAPID HOSE ASSESSMENT MODEL
R. W. Poeton, M. P. Moeller, G. J. Laughlin A. E. Desrosiers, and J. B. Martin Pacific Northwest Laboratory Rich!and, Washington
ABSTRACT As part of the continuing emphasis on emergency preparedness, the U.S. Nuclear Regulatory Commission (NRC) sponsored the development of a rapid dose assessment system by Pacific Northwest Laboratory (PNL). IRDAM, an acronym for Interactive Rapid Dose Assessment Model, is a microcomputer based program for rapidly assessing the radiological impact of an accident at a nuclear power plant. IRDAM can handle a variety of accidental release scenarios. Information on meteorological conditions, source term mixture, and time periods for decay and plume passage are all incorporated into the calculations. The activity or quantity of a release can be identified in several ways including isotopically, grossly, within containment, or from leaking coolant. An elevated release is modelled as a concentric set of cylinders collapsed to a line source. The resulting flux is used to calculate dose rates and projected doses at fixed downwind distances from the release point. The IRDAM system is highly interactive and assumes an operator has little or no computer experience. Ry answering questions prompted by IRDAM, an operator is guided through the input requirements for the calculations. The model has site-specific information and defaults programmed which allows the operator to rapidly make radiological dose estimates with a minimum of information.
INTRODUCTION An Interactive Rapid Dose Assessment Model (IRDAM) was developed by Pacific Northwest Laboratory (PNL) for the U.S. Nuclear Regulatory Commission (NRC). IRDAM is intended for rapid dose assessment during the first few hours after an accident at a nuclear power plant. It is assumed that more sophisticated computers, programs, and data will be available after the first several hours. During the early period, however, IRDAM can be a valuable tool to aid the utility, county, state, or NRC emergency directors in protective action decision making.
541
The IRDAM code is designed to accommodate a wide variety of accidental release scenarios. During the early stages of an accident, only minimal information n.ay be available on meteorological conditions and the source and magnitude of a radiological release. IRDAM can accept these inputs, provide default values for parameters where information is lacking, and rapidly estimate doses at downwind locations. Later, when post-accident sampling data or more complete information on plant conditions become available, IRDAM can perform more refined dose assessment calculations. IRDAM is a "user friendly" code which assumes that an operator has expertise in radiation protection, but little or no computer experience. At the beginning of the program, IRDAM offers the operator three levels of "help." With "more help," IRDAM presents more prompts and opportunities to recheck inputs. As the operator becomes more experienced, "less help" can be requested. These features of IRDAM may prove quite valuable during the stressful early period of a real accident. RELEASE SOURCE TERM IRDAM provides four basic options for the input of information regarding the relea.se source term. These options provide a range of specificity in describing the release and also provide for the input of source term information from different sources (e.g., plant parameter information vs. stack monitor data). The four release options are described below. 1. Radionuclide Release In choosing this option, the user is provided with the data in Table 1. Table 1.
Release Radionuclides
Kr- 83m Kr- 85m Kr- 85 Kr- 87 Kr- 88 Kr- 89 Xe- 131m Xe- 133m (9) Xe- 133 (10) Xe- 135m (1) (2) (3) (4) (5) (6)
(11) (12) (13) (14) (15) (16) (17) (18) (19) (20)
\l\
Xe-135 Xe-137 Xe-138 Cs-134 Cs-137 1-131 1-132 1-133 1-134 1-135
The user provides information on a particular radionuclide by specifying its corresponding number and an associated release rate in curies/ secondo Any or all of the radionuclides may be specified in any order desired. This is the only option which includes consideration of cesium isotopes.
542
IRDAM contains s e m i - i n f i n i t e cloud dose conversion factors from U.S. NRC Regulatory Guide 1.109 (U.S. NRC 1977) corresponding to the f i r s t 13 radionuclides in the IRDAM l i b r a r y , as well as dose conversion factors derived from Kocher (Kocher 1980) for isotopes of cesium, as shown in Table 2. Inhalation dose conversion factors from Regulatory Guide 1.109 are provided in Table 3 for the f i v e radioiodines, based on infant thyroid dose.
Table 2.
Semi-Infinite Cloud nose Conversion Factors
Radionuclide 83m Kr 85i _ 85
88
Kr
Xe 133m Xe 133 Xe 135m Xe 135 Xe 137Xe 138Xe 134Cs 137Cs
Table 3.
rem-m3/Ci-hr
Inhalation Dose Conversion Factors
Radionuclide 131, 132 133, 134 135
8.62E-3 1.33E+2 1.84E0 6.75E+2 1.68E+3 1.89E+3 1.04E+1 2.89E+1 3.36E+1 3.56E+2 2.06E+2 1.62E+2 1.01E+3 9.66E+2 3.70E+2
rem-m3/Ci-hr 1.06E+7 1.21E+5 2.54E+6 3.18E+4 4.97E+5
The isotopic release completely describes the makeup and activity of the source term and is probably not representative of the type of option that would be chosen in the initial phases of an accident. It is useful for comparison purposes, however, or for making estimates based on single radionuclides. 2.
Gross Release
Much less specific information is required for input into the gross release option. The only essential piece of information for this method is a gross release rate in Ci/s. Information on the relative fraction of iodine and noble gas radionuclides can be input to further characterize the gross release. 3.
Containment Leakage
This method does not require input of information on release rate of radioactivity. Instead, this information is generated from input on the containment leak rate and the type of accident involved. It assumes that the source term is bottled in containment and is leaking at the specified
543
rate. Three source terms based on accident type are available. They are fuel melt, gap release, and coolant release. In a fourth case the source term in containment is calculated from containment monitor readings. The source terms based on accident type represent noble gas and iodine inventory for the given type of accident and are calculated as follows: Fuel melt:
_Ci_ = (MWe) (7.?. x 10" 2 ) (LR) s
Gap release:
£i_ = (MWe) (7.2 x 10' 4 ) (LR) s
Coolant inventory:
_CJ_ = (MWe) (7.2 x 1(T 8 ) (LR) s
where LR = fractional containment leak rate per day Note that the fuel melt represents release of 100% of the core inventory of noble gases, the gap release represents 1%, and coolant inventory
i x io~H. The containment monitor equation is based on a release of 1% of core inventory being equal to 10 4 R/hr for a PWR or BWR--Mark III and 10 3 R/hr for a BWR--Mark I or II. For source terms based on containment monitor readings, the following method is used: Ci/s = (MWe) (7.2 x 1.0"4) (CM)/A where
A = 10 43 in the case of a PWR or RWR--Mark III A = 10 in the case of a BWR--Mark I or II and CM = the containment monitor reading in R/hr.
4. Coolant Leakage For cases where coolant is leaking outside of containment (e.g. the Ginna accident), the coolant leakage option is appropriate. This method calculates an equivalent gross release based on input values for coolant activity and coolant leak rate. Source Term Adjustments With the exception of the isotopic case, all of the above options are based on noble gas and iodine source terms. IRDAM provides methods for modifying these source terms to account for such factors as relative iodine fraction present and the effectiveness of filtration. As described later, decay during holdup and transit are factored in for some, but not all cases.
544
STABILITY CLASS The atmospheric stability class may be input directly or may be calculated using one of three methods (lapse rate, Sigma theta, or theta spread). 1.
Lapse Rate
The lapse rate method for stability class calculation requires input of height and temperature differences between two points of measurement and calculates the temperature lapse rate in °C/100 m. Stability class is assigned as shown in Table 4 which is taken from Regulatory Guide 1.23 (U.S. NRC 1980). 2.
Sigma Theta
The sigma theta method assigns stability class based on the standard deviation of the wind direction (U.S. NRC 1980) as shown in Table 5. Table 5. Stability Class Assignments: Sigma Theta Method
Table 4. Stabillity Class Assign inents Lapse Rate Method AT/AZ (°C/100 m)
3. six.
< < < < < <
-1 .9 -1 .7 -1 .5 AT/AZ < -0 .5 A T / A Z < 1.5 AT/AZ < 4 AT/AZ < AT/AZ < AT/AZ <
AT/AZ
(degrees)
A 22.5 17.5 12.5 7.5 3.8 2.1
B
C 0 UJ
-1,.9 -1..7 -1,.5 -0.,5 .5 4
Class
F G
a 9 > > °9 > > > > °e > > a9 > > > >
22.5 17.5 12.5 7.5 3.8 2.1
Class
C D E F G
Theta Spread This method calculates sigma theta by dividing the theta spread by It then assigns stability class according to the sigma theta method.
Certain combinations of stability class and windspeed are incompatible. Specifically, windspeeds greater than or equal to 5 m/s are not compatible with Classes F or G and windspeeds less than 2 m/s are not compatible with class A. With the exception of the case of a user-specified stability class, IRDAM takes the windspeed and stability class into consideration and alerts the user to mismatch of input, either choosing a compatible default value or rerunning the stability class calculation. Tf the user makes a direct input of stability class which is inconsistent with the windspeed already input no warning is provided.
545
DEFAULT VALUES IRDAM provides default values (i.e., assumptions to be used in the absence of other data) for a number of cases where input is important to the calculations. In choosing a default value, IRDAM will inform the user of the nature of the default and provide an opportunity to change it if so desired. The default values given in Table 6 are built into IRDAM. GROUND LEVEL RELEASE - DOSE CALCULATIONS Calculations for the ground level release case are performed using the following equation: DR = (Q)(*/O)(CF) where 0 is the release rate, CF is the Regulatory Guide 1.109 dose conversion factor, and X/Q is the atmospheric dispersion factor. For whole body doses the conversion factors are based on a semi-infinite cloud geometry. For thyroid doses, a breathing rate for infants is included (0.25 wr/hr). Table 6. Default Values Windspeed Release Height Effective Stack Height. Stability Class
2 m/s Ground Level 50 m
Wind Direction Containment Inventory (Release Rate for Containment Leakage) Coolant Activity (for Coolant Leakage) Iodine to Noble Gas Ratio Time Between Shutdown and Release Age of Material Iodine Filter Efficiency Containment Leak Rate Duration of Release
F-For Windspeed < 5 m/s E-For Windspeed > 5 m/s No Predominant Direction Fuel Melt 10 jjCi/cc 0.02 0 hr <1 day
0.1 V d (PWR or WR--Mark III) 0.5%/d (BWR--Mark I or II) 8 hr
Values for X/Q are calculated from the matrix of X/Q values in Table 7 (in units of 1/m1^) where u is the mean windspeed. The X/Q values are c a l culated by dividing the tabulated value for a specific combination of distance and s t a b i l i t y class by the windspeed in m/s.
546
Table 7.
Normali zed
Pasquill Stability Class
Dispersion
Factors,
X/Q in
Distance
m
A
B
C
D
E
F
G
500 2.25E-5
7.35E-5
4.53E-4
8.74E-4
2.26E-3
5.O7E-3
1,000
2.38E-6
6.55E-4
1.47E-3
4.30E-7
1.28E-4
2.24E-4
5.06E-4
3,000
3.OOE-7
4.84E-5
1.21E-4
2.73E-4
8,000
1.26E-7
1.24E-5
3.17E-5
7.13E-5
20,000
5.68E-8
1.34E-4 4.59E-B 2•40E-B 5.49E-6 1.67E-6
2.78E-4
2,000
1.59E-5 2.80E-6 8.94E-7 1.66E-7 7•44E-8
1.7OE-4 4.69E-5 1.40E-5 6.77E-6 1.24E-6 2.82E-7
4.07E-6
9.27E-6
2.09E-5
For ground level release cases calculated with the isotopic release rate option, the X / Q values and Regulatory Guide 1.109 conversion factors are employed directly for each specific radionuclide under consideration. Note that since IROAM's dose conversion factor library does not include any factors for iodines (except for inhalation), no whole body dose contribution from radioiodines is explicitly calculated. In the case of other ground level releases (gross, containment leakage and coolant leakage) IRDAM compensates for the lack of an explicit calculation of whole body dose from radioiodines by assuming that the noble gas source term (for the purposes of whole body dose calculations) is equal to the noble gas activity plus the activity of any radioiodines present. For these types of release, the age of the material is taken into consideration to correct for the different relative fractions of radionuclides present in the noble gas and radioiodine components as a function of time since shutdown. If the specified age of the material is greater than or equal to 24 hr, all the noble gas activity is assumed to be * 33 Xe and all the radioiodine is assumed to be 1 3 1 I . If the age of the material is less than 24 hr, the calculated doses are still based on 1 3 3 Xe and * 3 M , but are corrected by a time variant function. This function represents the ratio of doses received from the actual radionuclide mix to those received from a single radionuclide (1JJXe or 1 3 1 I) of the same activity, as a function of time since shutdown. For noble gases with an age less than 24 hr since shutdown, the dose rate from the radioquclide mix (DR?) is related to the dose rate from an equal activity of 1JJ Xe (DRj) by tne equation:
DR2 = (OR^ 11.0 exp (-t/10)
547
where t is the interval between reactor shutdown and the initiation of the release. For radioiodines, the corresponding equation is:
DR2 = (DRj) 0.34 exp (+t/22) [RDAM's ground level release calculations include corrections for relative decay and ingrowth during holdup in containment as well as during the time since a measurement was made, but decay during plume transit is not calculated. IRDAM does not include finite plume corrections for ground level releases. For these cases, an approximate correction can be made by multiplying calculated whole body doses by the appropriate factor from Table 8, adapted from Meteorology and Atomic Energy (Slade 1968).
Table 8.
Ground Level Release Finite Plume Correction Factors Atmospheric Stability
1
A
B
C
0.43
0.30
1000 m
0.65 0.88
0.66
2000 m
0.94
3000 m 8000 m
Distance 500 m
20000 m
lass
E
F
G
0.14
0.09
0.50
0,.20 0,.33
n .25
0.16
0.06 0.11
0.85
0.70
0,.50
0.35
0.27
0.19
0.95
0.90
0.36
1.0
0,.60 0.,80
0.50
1.0 1.0
0.78 0.90
0.70
0.57
0.25 0.43
1.0
1.0
0..90
0.83
0.73
0.64
ELEVATED RELEASE - WHOLE
BODY DOSE
•)
CALCULATIONS
Whole body doses from elevated releases are calculated by modeling the plume as a set of two concentric cylinders. For each of the six distances at which doses are calculated, it is assumed that only that portion of the plume between 100 m upwind and 100 m downwind will contribute to the dose received. It is further assumed for each distance that the plume is made up of two concentric cylinders, one of radius a z and one of radius 2 a 2 , where a z is the Gaussian dispersion coefficient in the z direction. Each of these two nested cylinders is centered at the release height as shown in Figure L The total activity in the cylinders is given by S = — (200) u
548
RELEASE HEIGHT
EQUIVALENT LINE SOURCE FOR THE LARGER CYLINDER
(POSNT OF DOSE CALCULATIONS)
Figure 1. Cylindrical Source Geometry where S 0 u 200
= the cylinder source activity in curies = the release rate in Ci/s = the windspeed in m/s = the cylinder length in meters.
The activity is divided between the two nested cylinders so that 56% is in the inner cylinder and 44% is in the larger cylinder. This approximates a Gaussian distribution between the smaller cylinder and the surrounding annulus. Blizzard et al. (1968) provides methods for collapsing a cylindrical source to an equivalent line source at some distance (z) inside the cylinder. IRDAM uses these methods, employing mathematical expressions of the curves in Blizzard et. al., to calculate gamma flux at the point P from each of the two cylinders. The"equation which provides the photon flux at point P by collapsing the cylindrical source to ar) equivalent line source is: 549
F(9, b 2 )
where
= the photon flux (photons/cm2 sec) = the buildup factor = source activity per unit volume = the radius of the plume = distance from the ground to the edge of the cylinder, which is the difference between the effective stack height and the radius of the cylinder 2 = the distance from the edge of the cylinder to the equivalent line source. B Sv R a
The term F(9, 62) is the Secant Integral, defined by:
F(e, b 2 ) = /J e
-b 2 sec e1 1
de
where 6 i s 1/2 the angle shown i n Figure 1 and can be described as: e = arctan (100/a) The v a r i a b l e b? is defined as:
b 2 = t>i + n s • z
where p is the self-absorption coefficient of the source at the energy of the isotope's emitted radiation and bj is the product of p«a where p is the linear attenuation coefficient for the energy of the radiation being emitted. Since the Secant Integral is extremely difficult to solve using analytical methods, a numerical integration scheme (Simpson's Approximation) using 10 segments was used to evaluate this integral. The buildup factor, B, is determined from a family of polynomial equations generated with a least squares fit from tabulated values (Blizzard et al. 1968) which depends on the energy of the emitted radiation and bj. Likewise, the linear attenuation coefficient, \i, and the source setf absorption coefficient, v , are calculated using equations as a function of energy derived from a liast squares fit. The average energy, E for each radionuclide considered was determined using the equation:
550
n
-m (5oon)
Y E- f- e 7 '
1
whero E^ is the energy of photons emitted from radionuciide i, f^ is the frequency of emission of radiation with energy Ej, and vi- is the linear attenuation coefficient in air calculated for energy E^. The calculation of the value of z, the distance from the collapsed equivalent line source to the fading edge of the plume as shown in Figure 1, is complicated, therefore, polynomial least square fits were generated from calculated value? in order to match the calculations in IRDAM to the processing power of a micro-computer. One difference between the ground level and elevated release calculations is that the elevated case includes explicit contributions from radioiodines in the whole body dose calculation. In cases other than the isotopic release option, IROAM does this by breaking the total source term into noble gas and radioiodine components and further dividing each of these components into fractions representing the individual radionuciides. For an elevated gross release, for example, IRDAM will calculate the whole body dose due to each of the 13 noble gases and 5 radioiodines and sum the results for each distance involved until the plume touches down (see Plume Touchdown). The fractional makeup of the source term is based on relative fractions of radionuciides present at shutdown as shown in Table 9. These fractions are derived from calculations made using the computer code ISOSHLD and WASH-1400 (U.S. NRC 1975) core parameters. The fractions at shutdown are adjusted by IRDAM to account for changes due to differential decay during holdup and transit. Unlike the ground level release method, the elevated case makes no simplifying single-radionuclide approximations for times greater than 24 hr. Whole body doses continue instead to be calculated using the sum of individual radionuciide contributions as long as the plume is elevated (see Plume Touchdown.) ELEVATED RELEASE - THYROID DOSE CALCULATIONS For an elevated release, the thyroid dose is calculated in much the same way as it is for the ground level release case. The primary difference is that the X/Q values for the elevated case are calculated differently. In particular, (X/Q) E = (X/o) G . exp (-h2/2az2)
551
Table 9.
Noble Gas 83m Kr 85m Kr 85Kp
^Kr 88^ r 89^ 13lm Xe 133m Xe 135m Xe
*^Xe 138Xe TOTAL
The Relative Fractions of Radionuclides Used by IRDAM (at shutdown) Fraction of Total
Radioiodine 131r
0.013 0.038 0.000231 0.070 0.097 0.120 0.00074 0.00376 0.198 0.052 0.048 0.190 0.169 1.00
132J
133j 134j 135j TOTAL
Fraction of Total 0.116 0.164 0.222 0.271 0.227 1.00
where (X/Q)f = the ground level x/0 from the elevated release (X/Q)- = the X/Q for a ground level release n = the effective release height a z = the Gaussian atmospheric dispersion parameter in the z di rection The same breathing rate and internal dose conversion factors that are used for ground level releases are used for elevated release cases. As with the whole body dose calculation, the elevated release method deals with releases by breakdown according to radionuclide, either input (in the isotopic release option) or calculated from relative fractions at shutdown (for gross, containment leakage, and coolant leakage options). ELEVATED RELEASE - PLUME TOUCHDOWN As the plume proceeds downwind, its radius increases. At some point the radius of the outer of the two concentric cylinders at a given distance will equal or exceed the effective release height (i.e., 2a z > h ) . At this point, IRDAM considers the plume to have reached ground level. Before plume touchdown, whole body and thyroid doses are calculated as described above. After plume touchdown the calculations are performed as if the release occurred at ground level, with one important difference. Instead of using X/q values for a ground level release, values for */Q for an elevated release are used. The purpose of this change is to compensate for that fact that semi-infinite cloud dose conversion factors are being used" in circumstances that are not necessarily well modeled by the semi-infinite
552
cloud c o n d i t i o n . This method allows IROAM t o approximate f i n i t e cloud c a l c u l a t i o n s under c o n d i t i o n s t h a t are otherwise beyond the micro-computer's processing power. For gross releases, c a l c u l a t i o n s f o r distances a f t e r plume touchdown incorporate the IRDAM s i m p l i f i c a t i o n s regarding material age, radionuclide makeup, and decay during plume t r a n s i t . Copies of the IRDAM program l i s t i n g are a v a i l a b l e upon request from any of the c o - a u t h o r s . REFERENCES B l i z z a r d , E. P. et a l . 1968. Engineering Compendium on Radiation Shielding, Vol. I. IAEA, New York, S p r i n g e r - V e r l a g . Kocher, D.C., 1980. "Dose Rate Conversion Factors f o r External Exposure to Photon and Electron Radiation from Radionuclides Occurring in Routine Releases from Nuclear Fuel Cycle F a c i l i t i e s . " Health Physics 38:593. Slade, 0. H., E d i t o r . 1968. Meteorology and Atomic Energy. Energy Commission, Washington, O.C.
U.S. Atomic
U.S. Nuclear Regulatory Commission. 1975. Reactor Safety Study--An Assessment of Accident Risks in U.S. Commercial Nuclear Power P l a n t s . WASH-1400, NUREG-75/014, Washington, D.C. U.S. Nuclear Regulatory Commission. 1977. C a l c u l a t i o n of Annual Poses to Man from Routine Releases of Reactor E f f l u e n t s f o r the Purpose o f Evaluating Compliance w i t h 10 CFR Part 50, Appendix I . Regulatory Guide 1.109, Washington, O.C. U.S. Nuclear Regulatory Commission. 1980. Meteorological Programs in Support of Nuclear Power P l a n t s . Proposed Revision 1 t o Regulatory Guide 1.23, Washington, O.C. U.S. Nuclear Regulatory Commission. 1982. Nomograms f o r Evaluation of Doses from F i n i t e Noble Gas Clouds. NUREG-0851, Washington, O.C.
553/55"^
9E EMERGENCY EFFLUENT MONITORING ANn ASSESSMENT
C, 0. C o r b i t P a c i f i c Northwest Laboratory R i c h l a n d , WA 99352
ARSTRACT
Althouqh real time experience w i t h emergency e f f l u e n t m o n i t o r i n g and assessment i s a r a r e phenomena w i t h t h e Department n? Energy (HOE) f a c i l i t i e s , t h a t same t h i n g c o u l d have been said about Three M i l e I s l a n d (TMI) a few s h o r t years ago. The a t t e n t i o n t h a t
TMI received following the accident included careful review of effluent monitoring and assessment, or the lack thereof, ultimately leading to new world-wide reactor requirements for effluent monitoring and assessment. The DDE and i t s contractors have been working together to publish a DOE guide on effluent radiological measurements. This guide includes provisions for emergency effluent monitoring and assessment up to the point where effluents leave a f a c i l i t y . While the guide's main focus is c r i t e r i a that would be useful in setting up a new program or revising an old one, i t also touches on what one can do i f a l l electrical power f a i l s and none of the emergency operational effluent monitors can operate. I t stops short, however, of sending radioactive plume monitoring teams into the f i e l d for large scale radiological emergencies. This paper reviews the use of effluent monitoring equipment during radiological emergencies that may require i n i t i a t i o n of protective actions for the general public. It expands upon the effluent guide coverage, and includes methods on use of portable instruments for effluent monitoring when a l l other monitors f a i l to work, the importance of plume monitoring, and how and when i t should be used, and assessment a c t i v i t i e s including emergency action levels that would be needed to i n i t i a t e protective actions for the general publ i c .
THE NEED
From the TMI emergency, we have learned that: • •
no news is bad news (in terms of releases) a little news is worse (the hydrogen bubble) 555
• • •
the public panics easily lapse of control is nationally newsworthy unknowns during any emergency, make the emergency appear to be very complex. This paper address two objectives for emergency planning as it relates to emergency effluent monitoring and assessment: (1) to eliminate one possible unknown that could occur in an emergency; that of not knowing what quantity of radionuclides have been released to the environs, and (?) to reduce the number of calculations, during the early stage of an emergency, required to determine the quantity of radionuclides that have been released. EMERGENCY EFFLUENT MONITORING The focus of this paper is on: •
atmospheric releases of gaseous effluents containing beta-gamma radionuclide emitters,
•
a nuclear reactor orientation,
•
generalized mechanics for monitored release pathways,
•
the first 30 minutes to 2 hr when minimal manpower is available to help.
SIX STEPS TO "SIMPLIFIED" EVALUATION There are six steps that can simplify the evaluation of radionuclide affluent releases to the environs. These steps are: •
define the worst case accident release,
•
analyze "realistic" potential releases,
•
characterize measurement requirements and monitor specifications,
•
select monitors and backup monitoring systems based upon steps 1-3 above,
•
develop "simple" dose projection mechanics including instrument markings and alarm points,
•
set corrective/protective actions to fit the consequences of the environmental releases of radionuclides.
This section describes these six steps to show how they can assist in" making the emergency directors job easier during the first 30 to 120 minutes of an emergency. 556
Step I: Define the Worst Case Accident The first thing to do in this step is to estimate the maximum number of curies that could possibly become available for release. Then, assume that all of it is rapidly released to the atmosphere--which would be the worst population exposure consequence. Obtain the necessary meteorological information and use the meteorology that would cause the highest population dose. Then, using established computer proqrams, estimate population dose rates, and compare with the Environmental Protection Agency (EPA) protective action guides (PAGs).^ ' The values of importance are 1 rem, whole body and/or 5 rens, thyroid to persons in the environment. If the values obtained fall short of the PAGs, then even though measurements are still necessary, the need for completing the remaining steps to speed things up is reduced. Radionuclide releases at or above the levels requiring use of the PAGs means that it would be prudent to have the information available that the remaining steps can provide. Step II:
Analyze the Realistic Release Potential
To begin this step, it is necessary to establish the average facility inventory in curies of radionuclides that could become available for release to the environment. Then assume that the releases qo to the environs via monitored routes. These assumptions should include treatment and no treatment if treatment facilities are used. The stability classes should be reduced from seven (A through G) to three or four. This reduction of classes should be based upon study of the site meteorological data by season (i.e., winter, summer, fall and spring) and shift (i.e., days, swing and graveyard). Next, determine the likely wind speeds that will be encountered during those seasons/shi fts. The possible whole body and thyroid doses for the affected population should then be compared to the EPAs PAGs. This information will provide the basis, along with step 1, for monitor selection. If the worst and realistic inventory release estimates are used, we will be prepared for "If something can go wrong it will." Population dose estimates that have been developed by such premises serve as the prime basis for specifying the emergency monitoring equipment. Step III: Characterize Measurement System Requirements and Specifications Initially it is essential to determine the type of fixed operational/emergency monitoring equipment and measurement ranges for each monitored effluent release pathway. This determination is made by careful consideration of the data obtained in the two steps above. Additionally, if there is a location (e.g., containment dome) where radionuclides can be contained except for small leakage, monitors should be available to establish the activity level in them. It can then be determined what will be needed for accurate and reliable normal-operational monitoring. By knowing the situation that could lead to high releases of radionuclides, allows one to develop specifications that consider environmental factors such as temperature, 557
humidity and seismic criteria that the monitor and other components may be subjected to. If a special type of cable is needed to connect the monitors to the readout instruments, consider an additional supply for later use. In a few years it is possible that the cable will no longer be manufactured or procurement may require a long lead time. Such problems could prevent reliable monitoring for a significant period of time. Next define the number of alarm levels (annuciations) that will be needed to properly alert the operator to the real emergency potential. The monitoring system should have a fail safe alarm. The best emergency alarm annunciation would consider the three "upper" classification levels of emergencies: (1) alert, (2) site emergency, and (3) general emergency. These three levels are needed to supply a stimulus for corrective action (alert), onsite protective actions (site emergency) and offsite protective action (general emergency). However, if only two levels of annunciation are selected, they should be set at release rate levels that could cause alert and site area emergencies. If the worst, possible accident case release of radionucl ides could only cause a site emergency, such items as backup power, temperature, humidity and seismicity specifications for monitoring instruments need only be considered from a normal operational need. If a general emergency release of radionucl ides could occur, such specifications should include the worst environmental case considerations. Finally, if the determination has been made to monitor four environmental release paths, and a location where radionuclides can accumulate to high levels (e.g., reactor containment vessel), two or three extra monitors should be procured to allow for rapid replacement necessitated by breakdown. The number of extras would depend on the need for more than one type of monitor. If the same type of monitor, including the range of measurement capability, is used for four of the five locations only two extras would be needed, one for each type. If more than three types of monitors exist, one of each type, should be considered. Step IV: Choose Monitoring and Backup Systems Once the characterization of the requirements and specifications step has been completed, selection begins. Selection should consider the capability of other in-plant monitors to "backup" the selected operations/emergency monitors. One can always postulate conditions that could damage or otherwise render a monitor useless. If such a condition should occur during an emergency that had potential for recommending protective actions, the loss of measurement capabilty with no backup system could result in unnecessary protective action recommendations or none at all. Any facility that has the potential for the general emergency class release of radionuclides also has (or should have) other facility dose rate monitors. These monitors normally include both fixed and portable instruments. If the effluent contains beta-
558
gamma emitting radionuclides it can affect nearby fixed monitor readings. This effect is predictable and calculable. Reactor industry architecturalengineering firms have developed the worst case accident dose rate projections for all locations within a plant. This information was found to be needed but unavailable for emergency evaluations during the TMI accident. PNL uses a "personnel exposure frcn right cylindrical sources" (PERC) computer code for certain ALARA evaluations program. PERC, in the near future, will be set up to calculate dose rates for emergency events. Once the worst possible case has been developed by a computer, for dose rates throughout a facility, it is relatively easy to scale down to varying percentages thereof (e.g., 90, 80, 70, 60, etc.) Figure 1 shows a hypothetical dose rate projection for a reactor facility. While such uniformly even dose rate reductions will not be found in "real life" situations, it is sufficiently illustrative for our purpose: project the possible dose rates before the accident and use the information to set up a quick-to-use, doserate estimation system. In Figure 1, Monitor C, just outside of containment, represents the reactor vent monitor. If one uses an on-line isotopic monitor, it could easily be "swamped out" by the extremely high dose rates coming fron containment. Therefore, careful instrument placement and special shielding is required to assure good results. Assume that the A and B monitors, within the containment vessel as shown in Figure 1, read sufficient Roentgens per hour to show us that 50% of the fuel has melted and released all of the attendent radionuclides to containment. Now assume,for some reason, these two monitors fail. How could we track the amount of radioactivity within containment, recognizing that the importance of this information forms the basis to determine how much could be released. Figure 1 shows that dose rates are predictable at other fixed area radiation monitors (ARMs). These monitors can then be interpreted within the reactor control room as the backup system for inventory status. And, if the reactor effluent vent monitor was properly shielded and located far enough away from containment to allow shielding, nearby ARMs may also be useful for estimating releases. Again, the objective is to provide backups to compensate for possible failure. The final backup monitoring system is the use of portable monitors. Since it has been shown that dose rates at various ARM locations and effluent vent locations can be used for backup estimates, portable dose rate instruments can be used at those same locations to obtain the same information. Thus, even if all fixed effluent and dose rate monitors were to fail, a backup system is readily available; it if has been set up for use is calibrated, and people have been trained to use it. The final problem that can arise is; suppose that nothing is coming out of the effluent vent, but large amounts (>10% of inventory available) of radionuclides are in reactor cc .tainment. There are no monitored pathways from containment. Suppose containment integrity fails? In such a case, one could use a plume monitoring team to verify the release and direction of 559
STEP IV. B. OTHER FACILITY MONITORS AS BACKUP
o
(90) (80) (70) (60) (50) (etc.
ARMS
25
12.5 6.25
3.125 1.56 0.78
HYPOTHETICAL" R/hr FIGURE 2. Reactor Area Radiation Monitors
release. In addition, by combining meteorology, visual observation of steam releases from the site under varying meteorological conditions and varying the release rate from containment, the computer can develop dose rate relationships to curies/unit time to potential population exposure. To verify radionuclide plume releases from a reactor, one need only to get far enough away from the "event reactor" to prevent the background dose rates from causing the dose rate instrument to read above normal background. One would begin by obtaining meteorological data (to help locate the pi one) and two radiation technicians could close in from opposite sides of the plume to determine its lateral dimensions. Simple opening and closing of the beta shield would provide assurance that the monitoring team was not "inside" of the piume. Step V:
"Simple" Dose Projections
By computerizing the meteorological parameters and the effluent release rate data, dose rates can be rapidly estimated during an emergency event. But if the computer doesn't work, such calculations take more time and open the "door" for mistakes. This step summarizes the considerations needed to make simple dose projections possible at speeds comparable to a computer. In computerizing meteorological data seven stability classes are often used without regard to their applicability at the site. These seven classes (A through G) should be evaluated in relationship to the site by season and shift. The classes can easily be reduced to 3 or 4 classes by a trained meteorologist who has practical field experience. Then windspeeds (in neters/second) should be defined and A T ' S developed in °C. These values should all be based upon experienced site meteorology. Using classes that do not occur and temperatures or windspeeds that are not likely will increase the number of the X/Q tables that should be developed. Color code the tables by season, and shift and use windspeeds rounded to usable numbers. These should be developed for the general emergency class. The defined effluent radionuclide flow rate(s) are then related to specific dose rates by converting c/m, mR/hr, R/hr readings to Ci/sec values (for given mixtures of radionuclides). The data is fed into a computer to establish "dose rate tables," by instrument. The dose rate tables are then related to levels within the site and general emergency classes. Determine the worst meteorological case dose rates (for on and offsite locations) and either proceduralize or mark the monitors for the two highest emergency classes. Then the monitors can be set for the abnormal and site emergency levels, so that they will annunciate at those levels. These set points must be understood to define a rate that, if continued for a specific period of time, will cause onsite doses requiring: 1. for abnormal release alarms (but below site or general emergency levels), determining ways to stop the abnormal release, 2.
for site area release alarms (lower portion of scale) implementing corrective actions to stop the release, and 561
3.
for general emergency releases (high enough to require offsite protective actions), appropriate initiation of protective actions.
Make the general emergency class tables for each meteorological class and wind speed expected. The idea is to be able to move from meteorology X/Q tables to dose rate tables, and finally to population exposure tables quickly by simply reading instruments that are properly keyed to the tables, all in the reactor control room. Step VI: Set Corrective/Protective Actions Once one has gone through the five steps above, this step is truly simple to implement. Corrective actions are taken solely by the operator (reactor crew). These actions are taken to stop or reduce releases. Protective actions, on the other hand, are recommended by the operator. Generally, these actions would be either seek shelter or evacuate the designated sector(s). Seeking shelter would be sufficient for « and p emissions, or levels of gamma dose rate that would result in less than 1 rem whole body or 5 rem thyroid. For values equal to the above EPA PAGs, the recommended action may well be evacuate a certain segment of the population. These protective action recommendations can be made in a short period of time, even if all electrical power is lost, provided the tables described in the steps above are available to the reactor operations crew, and they have been trained in how to use them.
REFERENCES 1.
Environmental Protection Agency. 1980. Manual of Protectivie Action Guides and Protective Actions for Nuclear Incidents. EPA-520/1-75-001 Washington, D.C.
562
9F
POST ACCIDENT RADIATION MONITORS
G. J. Laughlin and R. L. Kathren Pacific Northwest Laboratory Richland, WA 99352
ABSTRACT
Under contract to the Nuclear Safety Analysis Center of the Electric Power Research Institute, technical information and specifications were obtained for commercially available radiological monitoring instrumentation designed for use as postaccident monitors. Sources of information included open literature publications, manufacturer's sales and technical literature, and discussion with technical staff members of manufacturers and users. Included in the study were high level area radiation monitors, gaseous and particulate stack effluent monitors, and liquid efflent radioactivity monitors. The information was collated and published in the NSAC Handbook of Postaccident Instrumentation (Kathren and Laughlin 1981), and included such data as range, accuracy, precision, sensitivity, and energy dependence of the detector, environmental and seismic limitations of the equipment, the testing program performed to evaluate the equipment, a list of references where the instrumentation is currently installed, and a list of features and accessories available with the monitoring systems. The information presented in this section reveals that even though a number of vendors claim to be able to meet the guidance of Regulatory Guide 1.97 (USNRC 1980), few have actually conducted tests to verify that their equipment does indeed satisfy the guidance of this Regulatory Guide, and that some of the guidance may in fact be unrealistic. INTRODUCTION The accident at Three Mile Island nuclear generating station produced considerable interest in strengthening the capability of post-accident instrumentation, including post-accident radiation monitors. Thus, directly applicable guidance for post-accident instrumentation were put forth in Nuclear Regulatory Commission Regulatory Guide 1.97 (RG 1.97), "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," which cites as its legal basis General Design Criteria 13, 19, and 64 of 10 CFR 50, Appendix A. In general, the guidance specified in RG 1.97 for post-accident radiation monitors were far more stringent than had heretofore been applied, and
563
were co a large extent derived from two other regulatory guides, RG 1.89 (USNRC 1974) and RG 1.100 (USNRC 1977), which in turn referenced two IEEE standards, 323 and 344 (IEEE 1974 and IEEE 1975). The guidance provided in the IEEE standards, at least insofar as radiation monitors are concerned, is general and vague, and not always applicable, leading to the conclusion that the drafters of these standards were not primarily concerned with radiation monitoring instrumentation. To ascertain whether commercially available radiation monitoring instrumentation had been tested or otherwise evaluated for compliance with the guidance called out in RG 1.97, a study was carried out by Battelle, Pacific Northwest Laboratories, under contract to the Nuclear Safety Advisory Center (NSAC) of the Electric Power Research Institute (EPRI). The results of this study were subsequently published as part of the NSAC Handbook of Postaccident Instrumentation (Kathren and Laughlin 1981), a comprehensive looseleaf document providing in summary form the technical capabilities of various classes of post-accident instrumentation. DATA COLLECTION A list of 88 potential manufacturers and vendors of post-accident radiation monitoring instrumentation was developed by reviewing the relevant technical literature. Heavy reliance was placed on the current (1981) Nuclear News "Buyers Guide" (ANS 1981) and on the instrumentation handbook published by the Lawrence Berkeley National Laboratory (LBL 1980). Other publications used included the Science annual review of scientific instruments, advertising in health physics and related technically based nuclear journals, and informal conversations with various users of this type of instrumentation. All 88 were contacted by telephone to determine which actually manufactured or could provide post-accident radiation monitoring instrumentation. From this universe of 8 8 , only 11 companies professed such capability, and each of these was asked to provide sales literature, technical specifications, and test results on their products. Additional information regarding product availability and technical capability was sought from a general review of the literature and from telephone and personal contacts with users. Although "word-of-mouth" and operating experience information was generously offered by the latter, only published or otherwise openly documented information was used in this study. The information obtained from the various sources was reviewed, evaluated, and collated, with special emphasis being given to response data obtained through actual testing. In general the information provided by the manufacturers was incomplete, necessitating numerous followup telephone calls. The majority of manufacturers reported that independent testing laboratories were frequently used to conduct specific tests and obtain response data, and in several cases direct contact was made with these laboratories. All information and data reported to NSAC that was ultimately published in the Handbook was reviewed for accuracy by the appropriate manufacturer.
564
RESULTS Post-accident radiation monitors were divided in the following three broad categories: 1. 2. 3.
Area Radiation monitors (ARMs or RAMs). Gaseous effluent monitors. Fission product monitors (liquids).
Technical data for each of these categories is summarized in Tables 1-3. In general, those supplying in-plant area radiation monitors reported that this instrumentation could meet the range and accuracy guidance of RG 1.97. Such, however, was not the case for area radiation monitors used in containment. The highest radiation field to which these instruments was subjected was 1 x 10° R/h, one full order of magnitude below the level of 1 x 10^ specified in RG 1.97. Extrapolation of the response data obtained at lower levels was used to estimate the upper performance range of these instruments. It should be noted that radiation fields with intensities of greater than about 5 x 10" R/h are not available for test and evaluation purposes, and indeed the need for an instrument to monitor at these levels may be entirely academic, as has been pointed out by Had lock and Kathren (1982) who noted that detection capability would be severely limited if not impossible at exposure rates of this magnitude due to heating in the detector. While the majority of area radiation monitors listed in Table 1 appear to meet the humidity and pressure requirements of IEEE Standards 323 and 344, (IEEE 1974 and IEEE 1975) less than half are able to meet the temperature requirements of these same standards. The seismic qualifications appeared easily met by those instruments that were tested. The technical data for stack effluent monitors is summarised in Table 2. Of the seven monitors listed, four could not meet the RG 1.97 range guidance. As was the case with the area radiation monitors, it is doubtful whether the detectors would be operable for any significant time period because of heat generation and other effects on the detector. For example, the upper detection limit of 1 x 10-" fJiCi/cc for Xe-133 would produce a radiation field equivalent to 2 x 10' R/h in a gas filled detector. Thus, even a very short exposure to concentrations at or near the upper level put forth by RG 1.97 could quickly debilitate the detector. Accuracy and energy dependence data was not as complete for stack monitors when compared with area monitors, largely due to the difficulty and complexity of obtaining this information for isotopic detection systems as opposed to mean level or Geiger-Muller devices designed to read out in units of exposure or dose rate. Likewise, the testing program was not as extensive as that for area radiation monitors. To some extent, this may be attributable to a greater interest in the ambient radiation monitoring systems, spurred to some extent by the problems noted with the containment monitor during the aftermath of the Three Mile Island accident. In some respects such neglect is entirely justified as the majority of the stack monitoring instruments are
565
intended to be operated outside of containment, using sample lines to extract and transport the gas samples to the detectors. This is satisfactory providing a representative sample can be transported through what is frequently an extensive sampling system and remain representative for particulates and radioiodines. Technical data for the fission product monitors is summarized in Table 3. All of the four systems listed were reported as having ranges which satisfied the guidance of RG 1.97. However, only a single system met the accuracy requirement and none was able to satisfy the energy dependence requirements. As was the case with the stack monitors, these systems performed isotopic analysis on a sample. Thus, the discussion in this regard provided for the stack monitors also applies. Only a single system was reported to satisfy the temperature, humidity and pressure guidance of RG 1.97. The remainder either had not been tested, or had been qualified at a lower level than specified by RG 1.97. Again, as is the case with the stack monitors, these instruments are intended for use out of containment, and hence would not likely be subjected to the harsh environment and high levels of radiation produced by a major accident. However, it should be noted that the upper concentration level of 1 x u.Ci/cc specified in the regulatory guide would produce a large quantity of heat in any liquid that was being monitored. If the liquid did not boil away, the heat would very quickly be expected to render the detector unuseable. A similar situation would likely occur if the liquid were permitted to boil, but in any case operability of the detector with any degree of confidence for all but the briefest interval of time is unlikely at these levels. CONCLUSIONS In reviewing the information presented in tabular form, one may well be struck with the paucity of monitoring instruments that can n-eet the guidance put forth in RG 1.97. Two conclusions can be readily drawn: 1.
The guidance of RG 1.97 is unrealistic.
2.
Currently available commercial instrumentation is inadequate to the task and needs upgrading in line with what is technologically achievable.
There is a strong element of truth in both of these conclusions. With the exception of area radiation monitors for use in containment, the environmental criteria for temperature, humidity, and pressure put forth in RG 1.97 may be unrealistic as the majority of these systems are intended to operate outside containment. Criteria for accuracy and energy dependence appear, in the main, reasonable, but the range guidance for post-accident monitors generally appear unreasonable, particularly at the upper end. Radiation fields producing ambient exposure rates to 10 megaroentgens per hour,
566
or combined beta-photon fields producing levels to 100 megarads per hour may not in fact be achievable after even the most serious of nuclear accidents, and one could even ask the question of why or of what use these measurements will be in the immediate post-accident situation. Certainly not for l i f e saving purposes! And, as for the upper limit concentrations specified for both stack monitors and fission product monitors, i t appears that these are also unreasonable, for the decay heat would most likely r e s u l t in detector malfunction, or, more l i k e l y , destruction. Finally, i t should be noted that ANSI Standard N13.1 (ANSI 1969), referenced by RG 1.97, would require expensive and complicated sampling systems, which may in fact not be necessary, to provide representative sampling. This study clearly indicates that revision of RG 1.97 is needed with an eye toward providing more r e a l i s t i c and achievable c r i t e r i a for post-accident monitoring instrumentation. The study also points out a need for actual testing at the recommended levels rather than claiming capability by extrapolation or extension of data, or by calculation. If these are realized, the goal of cost-effective r e l i a b l e post-accident instrumentation may well become r e a l i t y .
567
Summary of Reported Technical C a p a b i l i t i e s Area Radiation Monitors Manufacturer and Model
tn CT)
for
Range
Accuracy
Energy Dependence
Temperature*(PWR)
Humidity*
Seismic*
Pressure*(PWR)
RG 1.97 Guidance
Containment, 1 to 107 R/h; 0.1 to 10" R/h (in plant)
+ 1002 -50%
+.20/?, . 0 5 t o 3 MeV
120°F t o 300° F 48 C to 148 C
0 to 100% relative
0.6 g a t 33 Hz
0 to 70 psi
D i g i t a l Date Dosimetry Inc. ARMSIV
0.0 mrad/h 10" r a d / h *
±10% o f true value'"
1 5% below 100 keV 1 17. 0.1 t o 4 MeV
<20%, 10°C t o 40° C
20 to 90% without condensation
No t known
Not known
Eberline I n s t r u ments RMS 11
1 to 10" R/h*
±2% of full scale'"
±10%, MeV'"'
0.05-1.75
-23°C t o 60°C ( - 1 0 ° F t o 140°F)*
State to meet IEEE 344*
up to 15 g
Not known
Jordan Nuclear Model 1000 Ramp IV
up to 10 or 107 R/h (10 5 R/h)"
120% of true dose*
115%, MeV"
0.0-1.2
60°C(max)
Not known
30 g*
160 psi*
General Atomic RS-2A**
0.1 to 10" R/h
+100%, -50%
120%. 0 . 0 8 - 3 MeV
0°C t o 55° C (32°F t o 131° F)
Up to 100%
Not known
14.7 psi
General Atomic RS-23A or RS-23D**
1 to 10° R/h
+100%, -50%
i 20%, 0 . 1 - 3 M.-V
176°C(max) (35O°F)
0 to 1007. saturated steam
Not known
70 psi ,
Kamon Sciences KMS-HR
1 to 107 R/h
120%
+.20%, 0 . 1 - 3 MeV
2I5°C(max) (420°F)
No t known
Not known
70 psi
Nuclear Measurements Corp. High Range Monitors
10 to 10a R/h ( 1 0 6 R/h)"
Not known
300°C(max) (572°F)
No t known
Not known
Not known
Victoreen Mods 1 875
1 to 10 7 R/h (up to 106 R/h)*
135%*
±8%, 0 . 1-3 MeV -15% +P% 0 . 0 8 0.1 Me .'•••
S t a t e d t o meet IEEE 323 & 344*
Stated to meet IEEE 323 & 324*
Stated to meet IEEE 323 & 324*
Stated to meet IEEE 323 & 324*
Westinghouse High Range Containment Monitor
) to 10 8 R/h (5 x 10s R/h)*
j^O.2 of decade-
+ 20% 0 . 1 t o 3 MeV*
300 C(max) (572 F)
100%
No t known
Not known
(140°F)*
03
Not known
+IEEE-Std. 323-1974 344-1975 are i n d i r e c t l y referenced by RG 1.97 for environmental q u a l i f i c a t i o n s . '-Indicates t h a t laboratory t e s t s have been performed up to the values given. '-"-Reported that equipment has been t e s t e d , but that it is against t h e i r company policy to release t h i s
information.
S t a c k R a d i a t i o n Monitors Manufacturer and Model
Range
RG 1.97 Guidance
10- 6 to 10 5 UCi/cm
+ 100% -50%
Combustion Engineering
10" 2 to 10 s yCi/cm 3 *
±100%*
General Atomic Wide Range Gas Monitor**
1 0 - 7 t o 10 s UCi/cm3
Kaman Sciences KMG-HR
Energy Dependence
Temperature"1"
Humidity*
Seismic*
Pressure*
±20% 0.5 to 3 MeV
120°F to 300° F 49° C to 149° C
0 to 100% relative
0.6 g at 33 Hz
0 to 70 psi
Not known
4° to 54°C (40°F to 130°F)
20% to 95% relative
Not known
14.7 psi
+ 100, -50%
Not Known
49°C (120°F)
Up to 95%
No t known
15 psi
10~ 7 to 10 5 UCi/cm 3 *
Not known
Not known
-0.6°C to 49°C (30° F to 120° F ) *
15 to 95%*
No t known
14.7 psi*
Nuclear Data Automatic A n a l y s i s System
10~"to 107 uCi/cm 3 *
<.20% to 50,000 c pm*
80 keV to J . 3 3 MeV*
S t a t e d to meet RG 1.97*
Stated to meet RG
No t known
Stated to meet RG 1/97*
Science A p p l i c a t i o n Gaseous E f f l u e n t I s o t o p i c Monitoring System
10~13 to 10 2 UCi/cm3 ( p a r t i c u l a t e ) 10" 7 to 105 UCi/cm3 (noble gases)
Not known
6 keV to 3 MeV
S t a t e d to meet RG 1.97
Stated to meet RG 1.97
Stated to meet RG 1.97
Stated to meet RG 1.97
Not known
60 keV to 2 MeV
Not known
Mo t known
No t known
Not known
Not known
4°C Co 49°C (40° F to 120°F)
11° F max wet bulb
Not known
14.7 psi
Accuracy
CJ1
cr>
Science A p p l i c a t i o n s 10- 7 to 10s Reactor Coolant and UCi/cm3 Containment Atmosphere Monitoring System Westinghouse
10-3 to 105 Not UCi/cm3 (noble known gases) 10~ 3 to 102 uCi/cm3 (particulates and Iodines)*
"'Designates testing has been performed for the listed parameters. **Reported that equipment has been t°sted, but that it is against their company policy to release this information. +IEEE 323 and 344 are indirectly referenced by RG 1.97 for environmental qualifications.
TABLE 3-
Manufacturer and Model
en
s
Range
Accuracy 3
to
Summary of Reported Technical Capabilities Fission Product Monitors
Energy Dependence
RG 1.97 Guidance
10 UCi/ctn 10 Ci/cm 3
Nuclear Data Automatic Analysis System
10"" to 10 7 UCi/cm 3 *
<20% at 50 ,000 c pm*
80 keV to MeV*
Science Applications Inc. Reactor
10" 3 to 10 7 uCi/cm 3
Not
Victoreen Model 840-4P Failed Fuel Monitor
10 to 10 cpm where 10 cpm is equivalent to 107 pCi/cm3
Not known
Combustion Engineering Post Accident Sampling System
1 to 10 Ci/cm
+ 100% -50%
i 2 0 % 0.5
to 3 MeV
Temperature
+
120° to 300°F
Seismic*
Pressure*
0 to 100% relative
Up to 0.6 0 to 70 psi g at 33 Hz
Stated to satisfy RG 1.97*
Stated to satisfy RG 1.97*
Not known
Stated to satisfy RG 1.97*
60 keV to 2 MeV
Not known
Not known
Not known
Not known
80 keV to 2 MeV
Not known
Not known
Not known
300 psi
No t known
4°C to 54° C (40° F to 130° F)
20% to 9 5 % relative
Not known
14.7 psi
49° to 149°C 1.33
known
Coolant and Containment Air Monitor
±100%*
^'Designates that testing has been performed for listed parameters. 323 and 344 are indirectly referenced by RG 1.97 for environment qualifications.
REFERENCES Hadlock, E. E., and R. L. Kathren. 1982. "Thermal Limitations on Detectors in High Radiation Fields" Health Phys. 43:45. Institute of Electrical and Electronic Engineers. 1974. "IEEE Standard for Qualifying Class IE Equipment for Nuclear Power Generating Stations". IEEE Standard 323-1974. Institute of Electrical and Electronic Engineers. 1975. "IEEE Recommended Practices for Seismic Qualification of Class IE Equipment for Nuclear Power Generating Stations." IEEE Standard 344-1975. "instrumentation for Environmental Monitorin^-Radiation" LBL-1, (Vol. 3 ) . Rev. July 1980. Kathren, R. L. , and G. J. Laughlin. 1981. "Post Accident Radiation Monitors," in Post Accident Instrumentation Handbook. Nuclear Safety Advisory Center, EPRI. U.S. Nuclear Regulatory Commission. 1974. "Qualification of Class IE Equipment of Nuclear Power Plants". Regulatory Guide 1.89, Washington, D.C. U.S. Nuclear Regulatory Commission. 1977. cal Equipment for Nuclear Power Plants". D.C.
"Seismic Qualification of ElectriRegulatory Guide 1.100, Washington,
U.S. Nuclear Regulatory Commission. 1980. "instrumentation for LightWater-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions Following an Accident". Regulatory Guide 1.97. Revision 2, Washington, D.C. American Nuclear Society. 1981. No. 4, La Grange Park, IL.
"Buyers Guide".
Nuclear News, Vol. 24,
American National Standards Institute., ANSI N13.1-1969. Airborne Radioactive Materials." ANSI, New York, NY.
"Guide for Sampling
96
CAPABILITIES OF THE LOS ALAMOS NATIONAL LABORATORY'S ENVIRONMENTAL EMERGENCY RESPONSE VEHICLE
D. Van Etten, D. Talley, T. Buhl, and W. Hansen Environmental Surveillance Group, MS K490 Los Alamos National Laboratory Los Alamos, New Mexico 87544
ABSTRACT A 4-wheel drive van has been o u t f i t t e d for rapid and varied monitoring response to radiological emergencies. The vehicle's capabilities include 4-wheel drive plus auxiliary winch for accass to rugged off-road t e r r a i n . On-board equipment is powered by a 6.5 kilowatt AC generator or by external AC power where available. Monitoring systems include two multichannel analyzers; one, a 2 K portable analyzer with i n t r i n s i c germanium detector, the second, a microprocessor based 4 K analyzer with a swivel head i n t r i n s i c germanium detector. Rapid gamma searches are performed with a delta rate meter system using a chart recorder and two 4" x 4" x 16" Nal detectors. Other equipment includes portable high volume air samplers and a portable phoswich, as well as the usual portable radiation survey instruments. The construction is modular so that equipment racks, detectors, AC generator and other major structures can be removed or replaced in a natter of minutes.
INTRODUCTION The inability to perform high resolution gamma ray energy analysis in the field at Los Alamos has hampered environmental work for some time. Analysis was limited to the radiochemistry counting labs, where samples from the field were taken, delaying the time between sampling and the availability of the results. A more rapid response would be required in the event of an accident involving nuclear material in the environment. Taking a part of the counting lab into the field to give fast results could correct this deficiency. A vehicle that expands our environmental surveillance measurement capability and responds quickly to accidents was the intent of the van being assembled and operated at Los Alamos National Laboratory.
573
VEHICLE EQUIPMENT The vehicle being instrumented is a Dodge Tradesman 300 van modified for 4-wheel drive (Figure 1). The 4-wheel drive plus an auxiliary winch greatly improves its access to rugged off-road terrain. The body structure was improved with added internal framing for greater instrument support and rollover protection. The inside wall and floor were insulated with a spray foam for operation under adverse temperature conditions. Equipment installed in the van can easily be removed in minutes, depending on mission need.
Figure 1. In-situ Environmental Gamma Spectrum Measurements Electrical power is provided by a 6.5 kilowatt Onan AC generator with an electronic governor for a stable output. The \/an can be connected to an external power source if available. Heating and cooling of instruments are provided by an electric floor heater and roof mount air conditioner. The air conditioner is easily removed when height clearance is necessary. Three separate radio transceivers provide routine or emergency communication with the Los Alamos National Laboratory net, New Mexico state net, and citizen band. Additional frequencies can be added. In-situ gamma spectrum measurements are taken using an intrinsic germanium detector and a Tracor-Northern model 1710 multichannel analyzer (Figure 2). The detector is mounted in a 15 I cryostat with a 180° swivel head. The swivel head can be set as an up-looking, sidelooking, or down-looking detector. A typical measurement is made by placing the detector on a tripod with the head swiveled down at 1 m above the surface. In a spectrum taken with this configuration,
574
Fig. 2. Instrument Racks and Detector approximately 50% of the photon flux from a surface source comes from within 2 m of the detector. Profile soil sampling with 5 cm cores are taken to determine the variation of radionuclide concentration with depth. With the head oriented upwards, the soil or water samples can be counted using Marinelli beakers. The spectral data can be striped using the on-board analyzer equipped with an LSI-II computer or stored on magnetic tape for later analysis. The data can also be sent by telephone modem to a computer, for example, the PDP-11 located at the Environmental Surveillance Group facilities, for data reduction and analysis by personnel at that location. Periodic calibrations of the germanium detector for natural radionuclides are performed using the five large concrete calibration pads maintained by Bendix Corporation, which are spiked with known amounts of potassium 40 and thorium and uranium ores at Malker Field, Grand Junction, Colorado. The EML calibration method with a 114 yCi 152 Eu point source is also used. For monitoring areas out of the reach of the van, a portable intrinsic germanium detector and a portable, battery powered multichannel analyzer can be carried by one person into more remote areas. Setup is the same, and data is stored on minicassette tapes. A 50 n storage dewar carrying liquid nitrogen for the detector gives the van about a 30-day range in the field. Rapid gamma searches of large areas are performed with a delta rate monitoring system while driving the van. The system consists of a delta rate meter, single channel analyzer, stripchart recorder, high voltage bias supplies, dash read-out, and two 4" x 4" x 16" Nal detectors. The detectors are mounted upright in the rear corners of the van. The delta
575
ratemeter stores the numbers of ambient background counts accumulated in a prescribed count time. Delta thumb switch setting allows the operator to predetermine a chosen number of counts above background over the set count period. Should the radiation level exceed this delta setting, an audible alarm sounds with the frequency of the beep indicating the magnitude of the incremental change over background. The gross count rate is displayed on a strip chart recorder and a digital panel meter visible to the driver and passenger. Measurements of low energy gamma and x rays are taken with a portable phoswich detector. The portable phoswich can be used in one of three ways: as a count ratemeter for surveys of suspected contaminated areas, as a sealer, or as a detector for a multichannel analyzer. Soil samples or air filters can be counted for gross alpha and beta using a Ludlum model 2200 sealer and zinc-sulphide and plastic scintillator detectors. Soil samples are placed in petri dishes and dried before counting. Soil standards are used for calibration of the detectors. Longer term measurements of external radiation levels can be taken with Victoreen TL-15 dosimeter bulbs and read in the van. Portable high pressure ion chambers as well as the usual portable radiation survey instruments are available to the user to aid in the measurements needed. A support trailer is used to transport eight high-volume air samplers with 8" x 10" sampling heads and portable generators to power them. These air samplers" can be deployed very easily and are independent of power sources. A small meteorological tower is being added to the trailer for monitoring windspeed and direction. On-board computing capabilities are a new addition to the van. An LSI 11/23 microcomputer with memory expandable to 128 K was recently installed. The LSI was mounted inside the floppy disk rack to save space and the unit is being set up to accept data from the portable instruments as well as from the multichannel analyzer. Successful field trips using the van have been made to Amarillo, Texas, and Burlington, Iowa, in addition to routine use in the Los Alamos area. This field experience has been helpful in the testing of the design of the van and selection of instrumentation the van would carry. The van demonstrated its versatility in performing different monitoring programs under environmental conditions ranging from the December cold in Colorado to heat and humidity of July on the Mississippi River in Iowa.
576
ALPHABETICAL INDEX BY AUTHOR
ALPHABETICAL INDEX BY AUTHOR
Abee, H. H. 399 Acox, T. A. 77 Adams, S. R. 301 Alspaugh, V. L. 201 Attig, P. C. 432 Baker, ?. I. 69 Wallwark-Barber, K. M. 311 Bebon, M. J. 211 Bowen, B. h. 447 Bresson, J. F. 95 Brown, M. L. 163, 181 Brown, R. C. 291, 345 Buhl, T. E. 447, 573 Cahn, L. S. 419 Canmann, J. VI. 257 Carbaugh, E. H. 227, 257 Gate, J. L. 247 Chen, A. I. 447 Cooper, J. A. 431 Corbit, C. D. 555 Corley, J. P. 31 Crawford, T. V. 173 Davies, R. W. 1 DeCarlo, V. 95 Denham, D. H. 87 Dennison, W. 5 Desrosiers, A. E. 527, 541 Dewart, J. M. 447 Dover, S. A. 323 Dreicer, M. 247
Eddy, P. A. 489 Einina, L. C. 211 English, C. J. 87 Farber, P. S. 45 Gallegos, A. F. 311 Gil more, R. D. 519 Gladney, E. 107 Gonzalez, M. A. 247 Goode, W. E. 107 Gutierrez, J. A. 237 Hansen, J. H. 431 Hansen, W. R. 447, 573 Harper, W. 367 Hary, L. F. 141 Haynie, J. S. 237 Heid, K. R. 519 lienry, R. L. 153 Hoffman, N. D. 471 Hornbacker, D. D. 189 Huntzinger, C. J. 247 Illsley, C. T. 409 Jaquish, R. E. 31, 135, 333 Kathren, R. L. 563 Kephart, G. S. 291 Kirby, J. J. 195 Laughlin, G. J. 541, 563 Lynch, T. P. 431 Mathias, K. E. 195 Martin, J. B. 541 Hayes, R. A. 419
Index-1
McMurray, B. J. 519 Meuser, J. M. 459 Mitchell, M. E. 399 Moeller, M. P. 541 Moor, K. S. 387 Myers, D. A. 459 Naidu, J. R. 211 Nelson, I. C. 333 Oakes, T. W. 107, 119 Olsen, W. A. 447 Perrin, D. R. 107 Poeton, R. W. 527, 541 Powell, D. C. 527 Prater, L. S. 489 Price, K. R. 135 Pritz, P. M. 481 Ramsdell, J. V. 527 Rodgers, J. C. 311 Rope, S. K. 301, 387 Rueppel, D. W. 247 Schreckhise, R. G. 279 Smith, B. H. 181
Smith, T. H. 387 Soldat, J. K. 279 Speer, D. R. 345 Stas, N. J. 95 Stoetzel, G. A. 527 Talley, D. 447, 573 Thiessen, J. W. 13 Thurow, T. L. 419 Tseng, J. C. 173 Van Etten, D. M. 447, 573 Vaughan, B. E. 279 Waite, D. 367 Watson, D, G. 333 Watson, E. C. 279 Welty, C. J., Jr. 25 Wenzel, W. J. 311 Wiersna, G. B. 387 Wilbur, J. S. 489 Windt, N. F. 55 Wing, J. F. 95 Wright, S. R. 95
Index-2