XA0101477-5/I5*
IAEA-TC-389.26 LIMITED DISTRIBUTION
REACTOR TECHNOLOGY SAFETY AND SITING REPORT OF A TECHNICAL COMMITTEE MEETING ORGANIZED BY THE INTERNATIONAL ATOMIC ENERGY AGENCY AND HELD IN DIMITROVGRAD, USSR, 21-23 JUNE 1989
Reproduced by the IAEA Vienna, Austria, 1990
NOTE The material in this document has been supplied by the authors and has not been edited by the IAEA. The views expressed remain the responsibility of the named authors and do not necessarily reflect those of the governments) of the designating Member State(s). In particular, neither the IAEA nor any other organization or body sponsoring this meeting can be held responsible for any material reproduced in this document.
FOREWORD On the invitation of the Government of the Union of Soviet Socialist Republics, the Eleventh International Conference on the HTGR and the IAEA Technical Committee Meeting on Gas-Cooled Reactor Technology, Safety and Siting were held in Dimitrovgrad, USSR, on June 21-23, 1989. The meetings complemented each other. Due to the large worldwide interest, the conference attracted approximately 60 participants from 18 countries and 130 Soviet delegates. About 50 foreign participants and 100 Soviet delegates stayed over for the Technical Committee Meeting. The Technical Committee Meeting provided the Soviet delegates with an opportunity to display the breadth of their program on HTGR's to an international audience. Nearly one-half of the papers were presented by Soviet participants. Among the highlights of the meeting were the following: -
The diverse nature and large magnitude of the Soviet research and development program on high temperature gas-cooled reactors. Over 35 different research and design institutes were represented by the Soviet participants.
-
The,Government approval of the budget for the construction of the 30 MWt High Temperature Test Reactor (HTTR) in Japan. The schedule contemplates a start of construction in spring 1990 on a site at the Oarai Research Establishment and about a five year construction period. Japan also announced that a symposium on HTGR Technologies would be held on March 19-20, 1990 in Tokyo to commemorate the start of construction of the HTTR, with plans to hold such a symposium every three years.
-
Disappointment in the announced plans to shutdown both the Fort St. Vrain (FSV) plant in the United States (US) and the Thorium High Temperature Reactor (THTR-300) in Germany. These two reactors presently represent the only operating HTGRs in the world since the AVR plant in Jiilich, Germany, was also shutdown at the end of 1988.
-
The continuing negotiations between Germany and the USSR on the terms of the co-operation between the two countries for the construction of a HTR Module supplemented by joint research and development activities aimed at increasing coolant outlet temperatures from 750°C to 950°C.
-
The continued enthusiasm displayed by both the US and German representatives for the potential of the small modular designs under development in both countries and the ability for these designs to meet the stringent requirements demanded for the future expansion of nuclear power.
-
The combining of the HTGR technology interest of ABB-Atom and Siemens in Germany into a joint enterprise, HTR GmbH, in May 1989.
-
The generally favorable review of the unique safety aspects of the US MHTGR design by the US. Nuclear Regulatory Commission (NRC) in their draft Safety Evaluation Report which was issued in February, 1989.
-
The increasing interest by gradually more and more industrializing countries and even a few new industrialized countries in incorporating MHTGRs into their plans for expanded electricity production and for enhanced oil recovery and district heating.
- 2 Most of the papers provided an expansion of views or details of the development work supporting the national programs or perspectives outlined during the preceding International Conference. Apparent throughout was the gradually increasing effort taking place in the world to support the modular HTGR and the recognition that the demonstration of its unique characteristics is the logical next step in commercialization of the concept. The view expressed by industrializing countries in terms of application studies, siting and licensing aspects and by industrialized countries in terms of design and development progress, project prospects, cost analyses and long range technological growth potential into higher temperature industrial applications provided substantial evidence to this increasing recognition.
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LIST
OF
CONTENTS
MEETING REPORT
7
APPENDIX I - List of Participants
12
APPENDIX II - Table of Contents IAEA TR 312
24
APPENDIX III - Table of Contents, Special Issue of Energy, The International Journal
27
TECHNICAL COMMITTEE MEETING PAPERS SAFETY CONSIDERATIONS (Session A) Safety Concept of High-Temperature Reactors Based on the Experience with AVR and THTR W. Wachholz, W. Kroger - Federal Republic of Germany Depressurization Accident Analysis for the HTTR by the TAC-NC K. Kunitomi, I. Nishigushi, T. Takeda, M. Hishida, Y. Sudo, T. Tanaka, S. Saito - Japan
A-l A-2
Calculation Studies of Behaviour of HTGR Spherical Fuel Elements During an Accident Induced by a High-Positive Reactivity Fast Introduction A-3 A.O. Gql'tsev - USSR MHTGR Radionuclide Source Terras for Use in Siting S.B. Inamati, A.J. Neylan, F.A. Silady, L.L. Walker - USA
A-4
Transfer of Fissile Material through Shielding Coatings in Emergency Heating of HTGR Coated Particles A.N. Gudkov, S..G. Zhuravkov, M.A. Koptev, A.D. Kurepin - USSR
A-5
Analytical Method and Result of Off-Site Exposure during Normal Operation of High Temperature Engineering Test Reactor (HTTR) K. Sawa, H. Mikami, S. Saito - Japan
A-6
Advanced Gas-Cooled Reactors - Designing for Safety B.A. Keen - United Kingdom Structure Stable Alloy for Large-sized Equipment of High Temperature Gas Cooled Reactor with Coolant Temperature of 950°C Y.A. Dushin, N.N. Gribov, V.A. Ignatov, N.A. Medvedev - USSR Main Principles of Low-Power HTGR Radiation Safety Ensurance B.K. Bylkin, V.N. Grebennik, A.I. Kirjushin, N.G. Kuzovkov, N.N. Ponomarev-Stepnoi, A.A. Hruljev, I.V. Yanushevich - USSR
A-7
A-8 A-9
PROJECT PROSPECTS (Session B) Design Status of the HTR-500 Power Plant and the HTR Module Power Plant E. Arndt, R. Fisher - Federal Republic of Germany
B-l
High Temperature Technological Heat Exchangers and Steam Generators with Helical Coil Assembly Tube Bundle O.J. Korotaev, N.V. Mizonov, V.B. Nikolaevsky, E.K. Nazarov, - USSR
B-2
- 2 The Choice of Equipment Mix and Parameters for HTGR-Based Nuclear Cogeneration Plants A.L. Malevski, A.Ya. Stoliarevski, V.T. Vladimirov, E.A. Larin, V.V. Lesnykh, Yu.V. Naumov, I.L. Fedotov - USSR Prospects for Application of High-temperature Helium Reactor (HTHR) to Provide for Power Needs in Refineries and Petrochemical Plants E.A. Feigin, E.A. Raud, E.G. Romanova, P.A. Panasenko, V.N. Nikitin, - USSR Neutron-Physical Aspects of the HTR Concept with Spherical Fuel Elements N.N. Ponomarev-Stepnoi, V.N. Grebennik, E.S. Glushkov, N.E. Kucharkin - USSR
B-3
B-4
B-5
DESIGN STATUS (Session C) Improved Safety Nuclear Power - and - Heating Plant with HTGR of Modular Type R.G. Bogoyavlenskii, V.P. Vinogradov, V.P. Glebov, V.N. Grebennik N.N. Ponomarev-Stepnoi, A.A. Hruljov - USSR
C-l
Utilization of Process Heat from the HTRM in the Chemical and Related Industries M. Schad, H. Barnert, R. Candeli - Federal Republic of Germany
C-2
Safety Assessment Principles for Reactor Protection Systems in the United Kingdom W. Philp - United Kingdom
C-3
Design and Safety Consideration in the High-Temperature Engineering Test Reactor (HTTR) S. Saito, T. Tanaka, Y. Sudo, 0. Baba, S. Shiozawa, M. Okubo - Japan
C-4
The Design Status of the United States Department of Energy Modular High Temperature Gas Cooled Reactor Raymond R. Mills - USA
C-5
UK Regulatory Aspects of Prestressed Concrete Pressure Vessels for Gas-Cooled Reactor Nuclear Power Stations P.S. Watson - United Kingdom
C-6
COST AND ECONOMIC ASPECTS (Session D) Possible Applications of HTGR's in Turkey S. Metin Atak - Turkey
D-l
An Economic Assessment of U.S. MHTGR Design L. Daniel Mears - USA
D-2
The HTR, Applications, Economics and Environmental Aspects H. Barnert, M. Schad, H. Candeli - Federal Republic of Germany
D-3
Problems of Attracting Nuclear Energy Resources in order to provide Economical and Rational Consumption of Fossil Fuels E.K. Nazarov, A.T. Nikitin, N.N. Ponomarev-Stepnoi, A.N. Protsenko, A.Ya. Stolyarevskii, N.A. Doroshenko - USSR
D-4
- 3 Assessment of the Licensing Aspects of HTGR in Yugoslavia Z. Varazdinec - Yugoslavia Substantiation of Choice of the Main Physical and Thermohydraulic Parameters of Reactor Plant with small Power HTGR V.N. Afanasyev, V.F. Golovko, V.N. Zhigulsky, V.A. Karpov, A.I. Kiryushin, N.G. Kuzavkov, Yu.N. Sukharev - USSR
D-5
D-6
RESEARCH AND DEVELOPMENT (Session E) Radiation Resistance of Pyrocarbon-Bonded Fuel and Absorbing Elements for HTGR V.A. Gurin, Yu.F. Konotop, N.P. Odejchuk, S.D. Shirochenkov, V.K. Yakovlev, N.A. Aksenov, V.A. Kuprienko, I.G. Lebedev, B.V. Samsonov - USSR The Materials Programme for the High-Temperature Gas-Cooled Reactor in the Federal Republic of Germany: Status of the development of high-temperature materials, integrity concept, and design codes H. Nickel, E. Bodmann, H.J. Seehafer - Federal Republic of Germany
E-l
E-2
Research and Development Programs for HTGRs in JAERI I. Nishiguchi, S. Saito - Japan
E-3
Behaviour of HTGR Coated Fuel Particles at High-Temperature Tests A.S. Chernikov, R.A. Lyutikov, S.D. Kurbakov, V.M. Repnikov, V.V. Khromonozhkin, G.I. Soloviyov - USSR
E-4
Present Status of MHTGR Program in USA A. Millunzi - USA
E-5
LEU-HTR Critical Experiment Program for the PROTEUS Facility in Switzerland R. Brogli, K.H. Bucher, R. Chawla, K. Foskolos, H. Luchsinger, D. Mathews, G. Sarlos, R. Seiler - Switzerland
E-6
Present Status of Research and Development for HTR in China W. Dazhong, Z. Daxin, X. Yuanhul - China
E-7
The Prospects of HTR Plant in China Y. Liangcheng - China
E-8
Czechoslovak Approach to the Potential of HTGRs1 Introduction L. Jakesova, M. Podest, V. Pinkas - Czechoslovakia
E-9
Results and Future Programme of HTR's Study M. Djokolelono, S. Soentono - Indonesia
E-10
Reactor Tests and Post-Reactor Examination of HTGR Fuel Elements and Coated Particles .... E-ll Yu. G. Degaltsev, A.A. Khrulev, I.A. Mosevitskii, N.N. Ponomarev-Stepnoi, N.I. Tikhonov, V.V. Yakovlev - USSR Structural Graphite for High-Temperature Gas-Cooled Reactors Yu. S. Virgiliev, P. Ya. Avramenko, V.N. Grebennik, I.P. Kalyagina, I.G. Lebedev, V.A. Filimonov, T.N. Shurshakova - USSR 9121p/rt
6
E-12
MEETING
REPORT
The Technical Committee Meeting consisted of five half-day sessions with the papers grouped into five different categories: Safety considerations, project prospects, design status, cost and economic aspects and research and development. The list of participants of the meeting is given in Appendix I. The technical content of most of the papers presented at the meeting has been incorporated in a recently published IAEA report; Technical Report Series No. 312 Gas Cooled Reactor Design and Safety, May 1990. The Table of Contents of this report is given in Appendix II. In addition, about one-half of the papers have been included in a special issue of Energy, The International Journal dedicated to High Temperature Gas-cooled Reactors. This special issue will be published in December 1990. The Table of Contents of this special issue is given in Appendix III. This meeting report summarizes the Technical Committee Meeting in terms of the program in each country as an overview of all of the papers presented by the authors from each country. All papers presented at the TCM are also included in this report without editing. THE GAS-COOLED REACTOR PROGRAM IN THE SOVIET UNION (USSR) The design characteristics of the two HTGR-type plants which are receiving the major attention in the USSR, were presented. The 200 MWt VGM is very similar to the HTR-Module under development in Germany except for the incorporation of an auxiliary cooling loop for decay heat removal and the inclusion of an intermediate heat exchanger in the main heat transport loop. The intention is to initially operate the plant at 750°C core outlet temperature with only the steam generator and circulator in the main loop and then, in a second operational phase, to add the intermediate heat exchanger and raise the core outlet temperature to 950°C. The other design is the VG-400 plant which is rated at 1060 MWt and incorporates four main heat transport loops, each consisting of an intermediate heat exchanger followed by a steam generator/main circulator. The four loops are located in eight paired cavities in the sidewall of a cylindrical prestressed concrete reactor vessel containing the reactor core. Both of the USSR designs utilize a pebble bed core and both are designed for the ultimate application to industrial process heat/cogenerated electricity supply. The results of several development programs related to the HTGR designs were also present. Of particular interest were the reports on the development of HTGR coated particle fuel and pebble bed fuel elements. The fuel development has been a major program in the USSR for several years and the manufacturing processes are very similar to those used in Germany and the United States. The investigations have included every thorough characterization of the properties of each fuel component including the fuel kernels, coatings and the matrix graphite for pebble bed elements. Several experiments including a description of the experimental apparatus used to investigate the kinetic behaviour of pebble bed cores, the interaction between the core and the control rods and the drive mechanisms for the control rods were also discussed. Several reports were given on the development of heat exchangers, graphite and other HTGR equipment and on several HTGR application studies.
— 2— THE GAS-COOLED REACTOR PROGRAM IN GERMANY An overview of the status of electricity generation in Germany indicated that the reserve margins in Germany are presently very large and, with projected near-term growth rates in demand of 1% per year or less, very few new power plants are foreseen as needed for several years. The HTR situation in Germany, including the shutdown in power generation of the AVR in December, 1988 and the announced intention to shutdown the THTR-300 were reviewed. Germany intends to continue support for HTR technology and the maintenance of a technical basis sufficient to revive the option when needed. The operating experience and plans for the THTR-300 were presented. In September 1988, the plant was shutdown for a scheduled inspection. At that time, inspection of one of the hot gas ducts, through which hot helium passes from the core exit to the steam generator entrance, revealed some damage. Several bolt heads from the central attachment fixture on the thermal insulation cover plates had broken off. In addition, several graphite dowels that hold the lower outer graphite blocks lining the hot gas duct had been displaced. All 6 ducts were subsequently inspected and it was found that, out of about 2600 bolts, 35 bolt heads had come off. Analysis determined that the bold heads had failed due to a concentration of stresses resulting from thermal expansion of the metallic foil insulation coupled with a reduction in ductility of the bolt material as a result of neutron irradiation. The cover plates have 4 corner attachment fixtures in addition to the central attachment and further analyses concluded that the plant could still be safely operated. Late in 1988, however, a reevaluation of the risks associated with the continued operation of THTR was made. Several risk factors have changed since the initial contract for THTR was signed. These include the termination of an ongoing fuel supply by NUKEM, inability to assure spent fuel storage facilities, the possibility of additional requirements being imposed prior to obtaining a long term operating license and the increased estimated cost for decommissioning. These risks total to a figure over twice as large as originally estimated, placing a potentially large burden on the HKG consortium that owns the plant. The HKG partners asked for increased participation by both the Federal and State governments. This was not successful and, in order to limit the risks, the HKG partners gave notice of their intent to shutdown and decommission THTR. THTR generated almost 2.9 billion Kwh of electricity since the beginning of power production in late 1985 and verified many of the unique characteristics of HTGR-type plants during the operation. The three reference concepts being developed in Germany, the HTR-500, the HTR-Module and the Gas-Cooled Heating Reactor, the GHR 10, were described. The conceptual design and site-independent safety report on the HTR-500 have been completed. On the HTR-Module, a revised safety analysis report was submitted in September 1988 and a final statement on the safety concept of the HTR-Module from the Reactor Safety Committee is expected in the fall of 1989. Several safety aspects of HTGR's and the safety research being performed at KFA, Jiilich were described. The shift in emphasis in safety research of HTGR's to investigations on the efficiency of several physical mechanisms to mitigate design basis and beyond-design basis events was noted. The results of several research programs including fission product retention capability of coated fuel particles at elevated temperatures, fission product retention by graphite, the kinetics of the steam-graphite and oxygen-graphite reactions in practical configurations and the results of the recent tests in AVR on depressurized core cooldown by conduction and radiation were presented.
- 3 The behaviour of HTGR's under accident situations extending into the beyond design basis regime is largely understood and accepted by licensing authorities. A preliminary version of a film of recent AVR safety tests was also presented to illustrate the impressive safety response of small HTGR's. THE GAS-COOLED REACTOR PROGRAM IN THE UNITED STATES (US) A presentation on electricity supply and nuclear power in the US showed that, although the growth in overall energy consumption still remains low, the demand for electricity continues to expand at rates closely corresponding to that of the Gross National Product, about 2.5% per year. The need for additional capacity is becoming apparent in many areas of the country as reserve margins are becoming precariously low. The 108 operating nuclear plants in the US. are now producing about 20% of total demand but this percentage will start to decrease in the 1990's even as the few remaining plants under construction are completed. A reversal of the forthcoming downtrend will occur only if the impediments to nuclear power in the US are overcome. The MHTGR was viewed as an option whose characteristics appear to mesh well with the requirements for the revival of nuclear power. A review of the operations of the FSV plant and the plans for the future were presented. The scheduled shutdown to replace bolting material on the helium circulators which had taken place in July 1988, had extended to March 1989 due again primarily to moisture ingress into the primary system. The ingress was from a small breach in the core support section of the liner cooling system which had caused a similar problem several years previously. The plant was brought back into operation in April 1989. From the beginning of power generation, the FSV plant has generated over 5.4 billion Kwh of electricity. The financial situation regarding FSV was discussed and it was noted that extensive efforts had been undertaken to assure the continued operation of the plant. However, the costs of operation, maintenance, fuel fabrication and providing for capital improvements on what has, unfortunately, become a unique facility could not be entirely covered by the allowable 4.8 cents per Kwh return on electricity generated, with any reasonably attainable capacity factors. As a result, in early December, 1988, the PSC Board of Directors made the decision to permanently shutdown the plant. The present plan, providing that significant problems do not develop, is to continue operation at approximately 80% power for the remainder of 1989 followed by a coastdown of decreasing plant output by 5% per month during the first half of 1990. A significant program is underway to prepare for the defueling and decommissioning of the reactor. The schedule of the on-going DOE program on the MHTGR was discussed. The signing of five year contracts between DOE and the several industrial entities for the continuing design and prelicensing effort on the reference 4-module MHTGR plant was noted. The schedule for the work, which is now in the early preliminary design phase, was presented and involves a continuing extensive interaction with the US Nuclear Regulatory Commission (NRC) with the aim of obtaining a Final Design Approval in 1996. A paper covering the safety approach of the MHTGR in the US and the response of the plant to several events that were foreseen to potentially challenge the ability to retain radionuclides within the coated fuel particles
- Awas presented. The events include those that would be considered design basis as well as beyond design basis and in all cases the passive safety features prevented and mitigated radionuclide release levels to well below the limits established by the requirements. The history of interactions with the NRC was then recounted and a summary of the generally favourable conclusions from the NRC's recently released draft Safety Evaluation Report was presented. THE GAS-COOLED REACTOR PROGRAM IN JAPAN The presentations from Japan provided an update of the HTGR development program in Japan with particular emphasis on the status of the HTTR. The budget for the construction of the HTTR was approved by the Japanese Government in early 1989 and JAERI submitted the safety analysis report to the Science and Technology Agency in February. A construction permit is expected in the Spring of 1990 and completion of construction on a site at JAERI's Oarai Research Establishment is planned for 1995. The main features of the HTTR, a 30 MWt test reactor using prismatic-type fuel elements, whose major objectives will be to establish basic technologies for advanced HTGR's and to perform as an irradiation test reactor for research in high temperature technologies were described. An update of the studies by the Research Association on High Temperature Gas-Cooled Reactor Plant on the viability of HTGR concepts in Japan was presented. In 1988, the Association entered Phase II of its efforts with emphasis on assessment of the applicability of the module HTGR under the more severe seismic and extremely limited siting conditions typical of Japan. Two working groups were established. The first group initiated an examination of the specific regulatory and siting requirements in Japan with reference to the utility/user design requirements evolved during Phase I. The second working group initiated some detailed comparison studies on the differences between the US MHTGR and the German HTR-Module in responding to various transients. The group has also started to evolve various scenarios for the introduction and long term deployment of the HTGR in Japan. THE GAS-COOLED REACTOR PROGRAM IN OTHER COUNTRIES The status of the HTGR reserach and development program in the People's Republic of China (PRC) was presented. The program was initiated in 1974 with basic research in coated particle fuel, graphite technology and HTGR component development. The program is presently in its third phase and in addition to the continuing research, is performing design, safety and application studies on the modular HTGR. In addition, a project study on a 10 MWt HTR Test Module was initiated in 1988. The study is being jointly sponsored by the INET in the PRC and Seimens/Interatom and KFA in Germany. Construction of the Test Module is being contemplated at the site of the INET, northwest of Beijing. The main objective of the facility will be to verify and demonstrate some of the unique features of the modular concept. The configuration of components is identical to the German HTR-Module. A pebble bed core, smaller in dimensions than the HTR-Module, with a maximum output of 20 MWt (for later enhanced capability) is used. Application studies were presented for enhanced oil recovery in the Shanjasi section of the Shengli oil field using high temperature/high pressure steam and some remarks were given regarding the start of the cogeneration study for the Yangsham Petroleum Corporation located in southwest Beijing.
/lo
- 5 A paper discussing the feasibility study, initiated in 1988, on the construction of a MHTGR demonstration plant near the city of Zhongquing in Southwest China was presented. Zhongqing is the largest city in a very important developing district but suffers from a severe shortage of electricity supply due to the more than 40% dependence on hydropower. Dry seasons force work stoppages of 3 to 4 days each week. The construction of new coal-fired plants is difficult due to transportation problems and, although additional hydropower resources are available, the dry season type problem would not be solved. The working team has surveyed eight possible sites and selected, after a preliminary evaluation, a site about 30 km east of the city, near the Yangtze River. An evaluation of the possible supply of as much of the equipment, material and labor scope by Chinese industry has been made and the cost of two alternate technical schemes have been derived; a 2 x 350 MWt U.S. MHTGR plant with a 300 MWe turbine generator and a 4 x 200 MWt German HTR- Module plant with a 340 MWe T-G set. Even with more than 90% of the conventional part of the plant supplied by the Chinese industry the capital cost for the nuclear plants are currently estimated to be too high. Nevertheless, nuclear could be the option for the future if the participation of vendors to help defray the high introduction costs for the modular HTGR nuclear plant in China was available. An overview of the potential for the introduction of HTGR's in Czechoslovakia was presented. Nuclear-generated electricity, using the Soviet-type of pressurized water reactors presently provides about 27% of demand but the remainder is generated almost entirely by burning indigenous brown coal. The high level of pollution from these plants and from chemical industry plants, primarily located in the North Bohemian region of the country, is causing considerable concern and is being given special attention by authorities. With industry, in total, consuming more than 60% of the fossil energy resources, the need for process heat applications of the HTGR, particularly the higher temperature applications of steam-methane reforming and coal gasification, was emphasized. The Swiss industry's broad participation in several international projects and design programs on the HTGR which has extended for over 20 years was discussed. Switzerland's more recent interest in the small gas-cooled district heating reactor, the 10-20 MWt GHR-10, whose conceptual design was recently completed in conjunction with ABB in Germany was noted. The PROTEUS facility at the Paul Scherrer Institute which is available for performing ciritical experiments on small HTGR cores and the interest in international cooperation for such experiments was also discussed. The purpose would be to enable analytical code verification of some of the unique reactor physics aspects of this type of reactor. The results of the continuing studies on the application of HTGR's for heavy oil recovery in Indonesia were presented. The recent studies expanded on preliminary studies performed in 1986/87 and indicated that, with the present low prices for oil, it was not possible to show an economic advantage for the reactor application. However, with the expected future escalation in oil prices, the economic viability of the reactor application would be apparent. Various concepts for regional development in Indonesia which is very rich in several natural resources and has a strong desire to establish a domestic industrial infrastructure to exploit these resources were presented. The desire to establish an internal vendor supply industry for HTGR materials and equipment if the HTGR is deployed in Indonesia was expressed. 8974p
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Appendix I IAEA Technical Committee Meeting on Gas-Cooled Reactor Technology, Safety and Sitingf Dimitrovgrad, USSR, 21-23 June 1989 LIST
OF
PARTICIPAHTS
BANGLADESH
Mr. C.S. Karim Bangladesh Atomic Energy Commission Nuclear Power & Energy Division P.O. Box 158 Dhaka 2, Bangladesh Telex: 632203 BATOM BJ Tel. 504-383
CHINA
Mr. Wang Dazhong Institute of Nuclear Energy Technology (INET) Tsinghua University P.O. Box 1021 CN 102201 Beijing, China Facsimile: 2564-177 Telex: 22617 QHTSC CN Tel. 28-3731-294 Mr. Liangchen Ye South West Center of Reactor Engineering R & D P.O. Box 436-500 Chengdu Sichuan, Peoples' Republic of China Facsimile No.:600064-SWCRS CN Telex: 5517 Tel. 582199-206
CSFR
Mr. Frantisek Suransky Advisor to the Minister Federal Ministry of Fuel and Energy Vinohradska 8, 120 70 Prague 2, CSFR Telex: 122 954 FMPE C Tel. 260 746 Mr. Vaclav Pinkas Dept. of Nuclear Power Prague Telex: 121107 CKAE C Mr. Milan Podest Nuclear Research Institute Rez. Telex: 122626
FRANCE
/K,
Mr. Michel Lecomte Framatome, Tour Fiat F - 1405 B, Cedex 16 920 84 Paris la defense France Facsimile No:33(l) 47 96 15 09 Telex: 630 635 F Tel. 33(1) 47 96 56 73
- 2 FRANCE (Continued)
Mr. Jean-Claude Chenal Direction des Etudes et Recherches Service -RNE 1, avenue du General de Gaulle 92141 Clamart, France Facsimile No: 33(1) 47 65 34 99 Telex: 204426 F Tel. 33(1) 47 65 38 41 Mr. D.P. Bastien CEA/CEN/Saclay DEDR/CRG 91191 Gif-sur-Yvette Cedex France Facsimile No: 33(1)69 08 48 62 Telex: 604 641 F Tel. 33(1)690-83112 Mr. G.C. Chevalier CEA/CEN/Saclay DEMT 91191 Gif-sur-Yvette Cedex France Facsimile No.: 33(1) 69 08 86 84 Telex: 604 641 F Tel. 33(1) 69083010
GERMANY
Mr. Wolfgang Kroger Kernforschungsanlage Jiilich VS Postfach 1913 5170 Jiilich Mr. Claus Benedict v.d. Decken Nuclear Research Center Julich P.O. Box 1913 D-5170 Julich Telefax: 61 5327/Telex: 833 556 KFA D Tel. 02461-613074 Mr. Erwin R. Balthesen Kernforschungsanlage Jiilich GmbH Postfach 1913 D-5170 Julich Facsimile: 02461/61-5370 Tel. 02461/61-3131 Telex: 833556-80 KFA Mr. Heiko Barnert Kernforschungsanlage Julich - VS Postfach 1913 5170 Julich Facsimile: 02461/61-5327 Telex: 833556 KFA Tel. 02461/61-3233
Ab
- 3 -
GERMANY (Continued)
Mr. Ivor Kalinowski Hochtemperature-Kernkraftwerke GmbH Siebenbeckstr. 10 4700 Hamm 1 Facsimile: 02388/32-2318 Telex: 828884 KFA Tel. 02388/32-2689 Mr. Hubertus Nickel Nuclear Research Centre Julich P.O. Box 1913, 5170 Julich Tel. 02461-61-5565/Telefax: 02461-61-3699 Telex: 833556-0 KF D Mr. Kirch Norbert Kernforschungsanlage Julich GmbH Stetternicher Forst Postfach 1913 D-5170 Julich Facsimile: 02461/61-5370 Telex: 833 556-80 Tel. 02461/61-6991 Mr. Erhard Arndt Hochtemperature-Reaktorbau GmbH Gottlieb-Daimler-Str. 8 D-6800 Mannheim 1 Facsimile: 0621/451-598 Telex: 462041 Tel. 0621/151-262 Mr. M.K. Schad Lurgi gmbH Lurgi Allee 5 6 Frankfurt /M Telex: 5808-3888 Tel. 069-5808-3061
INDONESIA
Mr. Mursid Djokolelono National Atomic Energy Agency (BATAN) Jl Abdul Rohim, Kuningan Barat, Mampang, Prapatan P.O. Box 85 Kby, Jakarta 12710 Facsimile: 511110 Telex: 62354 BATAN IA Tel: 520 42 43
ITALY
Mr. Giancarlo Bolognini ENEL, Viale Regina Margherita 137 1-00198 Rome, Italy Facsimile No: 85 39 601 Telex: 610 518/610 528 Tel: 85 39 345
- 4 JAPAN
Mr. Shigehiro An Professor of Faculty of Engineering, Tokai University 2-28, Tomigaya, Shibuya-Ku Tokyo, Japan Telex.: 2224026 C/0 KRT MRU Telefax: 0463 59 3581 Tel.: 0463 58 1211 Mr. Kisamori Hiroyuki Ishikawajima-Harima Heavy Industries Co., Ltd. 1, Shin-Nakahara-Cho, Isogo-Ku, Yokohama, Kanagawa, Japan Telex: 3822386 IHI YOK J Facsimile No.: 045-752-576 Tel.: 045-751-1231 Mr. Ishii Hiroichi Kawasaki Heavy Industries Co., Ltd. 2-4-25, Minamisuna, Koto-Ku, Tokyo, Japan Facsimile.: 03-699-8587 Tel.: 03-615-5165 Mr. Okazaki Tomoaki Nuclear Power Department Babcock-Hitachi K.K. 2-6-2, Otemachi, Chiyoda-Ku Tokyo, Japan Facsimile No.: 03-242-5084 Telex.: 3502 BHK HONJU Tel.: 03-270-7365 Mr. Yamada Masao HTR Project Office Fuji Electric Co., Ltd. 1-12-1, Yurakucho, Chyoda-Ku Tokyo, Japan Telex: J22331 Facsimile No.: 03-211-7163 Tel.: 03-211-0435 Mr. Nakagawa Gun Engineering Department Mitsubishi Heavy Industries, Ltd. 2-4-1, Shibakoen 2-Chome, Minato-Ku, Tokyo, Japan Facsimile No.: 03-432-0908 Telex No. J25665 Tel.: 03-212-3111 Mr. Inoue Toyokazu Nuclear Energy Division Advanced Reactor Engineering Department Toshiba Corporation 8, Shinsugita-Cho, Isogo-Ku Kanagawa, Japan Facsimle No.: 045-756-2335 Telex.: J22587 Tel.: 045-756-2416
AS
- 5 JAPAN (Continued)
Mr. Nakaguchi Shozo Technology R & D Center Technology Research Group Toyo Engineering Corporation 12-10, Higashifunabashi 6-Chome, Funabashi-Shi, Chiba, Japan Telex.: 02983371 Telefax.: 0427-24-0529 Tel.: 0474-24-0111 Mr. Taketani Kiyoaki Fuji Electric Co,. 12-1, Yurakucho 1-Chome, Chiyoda-Ku, Tokyo, Japan Telex.: J22331 Telefax.: 03-211-7163 Tel.: 03-287-1045 Mr. Shinzou Saito Tokai-mura, Naka-gun, Ibaraki-ken 319-11, Japan Facsimile No.: 0292-82-5999 Telex.: J24596 Tel.: 0292-82-5416 Mr. Isoharu Nishiguchi JAERI, Tokai-mura, Naka-gun, Ibaraki-ken, 319-11, Japan Facsimile No.: 0292-82-5999 Telex No.: J24596 Tel.: 0292-82-6409
NETHERLANDS
Mr. J.B.M. de Haas Netherlands Energy Research Foundation ECN P.O. Box 1 1755 ZG Petten The Netherlands Facsimile No.: 31-2246-4480 Telex.: 57211 REACP Nl Tel.: 31-2246-4088
POLAND
Mr. Wielsaw Wembergier Adam Hrcyczuk/Institute of Atomic Energy 05-400 Otwock Swierk Poland, Telex: 816915 ATOM PL or 81 32 44 IBJS PL Mr. E. Obryk Institute of Nuclear Physics Radzikowskiego 152 31-342 Krakow, Poland Telefax: 012/333438 Telex: 0322461 IFJ PL Tel: 012/370222. ext. 280 Mr. Hryczuk
AG,
- 6 ROMANIA
Mr. Ionel Neamu Central Institute of Physics Bucharest Telex No.: 11397 CSEN E
SOUTH AFRICA
Mr. A. Koster Atomic Energy Corporation Pelindaba, P.O. Box 582 Pretoria, South Africa Telex: 322 948 Facsimile No: 323 7731 Tel: 012/ 316 4911
SWITZERLAND
Mr. Rudolf H. Brogli Paul Scherrer Institute CH-5232 Villigen Switzerland Facsimile No.: (056) 982327 Tel.: (056) 992693 Mr. Donald R. Mathews Paul Scherrer Institute CH-5232 Villigen Switzerland Facsimile No.: (056) 982327 Tel.: (056) 992894
TURKEY
Mr. Sevket Metin Atak Turkish Atomic Energy Authority (TAEK) Karanfil Sok. No. 67 06640 Bakanliklar-Ankara Turkey Telex No. 44455 / Tel.: 9(4) 1170661
UNITED KINGDOM
Mr. B.A. Keen National Nuclear Corporation Booths Hall Chelford Road, Knutsford, Cheshire, WA16 8QZ United Kingdom Facsimile No.: (0)565 3659 Tel. (0)565 2420/Telex.: 666000 Mr. P. S. Watson Mr. W. Philp
&
Nuclear Installations Inspectorate St. Peter's House Balliol Road, Bootle Mersenside L20 3L2 Facsimile No: 051 922 5980 Telex: 62 85 96 NUIN SPG Tel. 051-951-4933
/Pf
— 7 — UNITED KINGDOM (Continued)
Mr. R. I. Freeman Central Electricity Generating Board Division H.Q., Barnett Way Barnwood, Gloucester, U.K. Facsimile: 0452 65 2776/Telex: 43501 Tel: 0452 65 2222
UNITED STATES OF AMERICA
Mr. A. Millunzi Director of High Temp. Gas Reactor Development, Office of Nuclear Energy (NE-451) U. S. Dept. of Energy Washington, DC 20585 Facsimile: (301) 353-3870 Telex: 710-828-0475 Tel. 301-353-3405 Mr. Raymond R. Mills & Mr. Fred A.Silady General Atomics P.O. Box 85608 San Diego, CA 92138 Telex: 411 888 MEZON SU Mr. H.L. Brey Public Service Company of Colorado P.O. Box 840 Denver, CO 80201 Facsimile No. 303/480-6925 Tel. 303-480-6944 Mr. Gordon Lawrence Heins Senior Vice-President Consumers Power Company 1945 W. Parnall Road Jackson, MI 49201 Tel. (517) 788-1217/Telex: 223452(CPC0 PARN JKN) Facsimile No. (517) 788-0034 or (517) 788-00375 Mr. R. C. Liimatainen U.S. House of Representatives Energy Research and Development Subcommittee B-374 Rayburn House Office Building Washington D.C. 20515 Mr. Luther Daniel Mears
&
Ms. Susan Fielder Gas-Cooled Reactor Associates 10240 Sorrento Valley Road, Suite 300 San Diego, CA 92121 Tel. (619) 455-9500 Facsimile: (619) 452-7831
- 8 USSR
E.I. Turin State Union Design Institute, Moscow G.B. Usynin Yu. I. Anoshkin Gorky Polytechnical Institute N.G. Kuzavkov V.V. Bulygin Yu. F. Byvsev B.M. Kamashev V.A. Putyrsky V.P. Golovko Yu.P. Sukharev V.B. Chistyakov V.S. Ruchin Yu.N. Tatarsky Experimental and Design Bureau of Machinery, Gorky E.K. Nazarov V.B. Nikolaevsky State Institute of Nitrogen Industry, Moscow S.S. Keruchen'ko V.D. Roschin A.K. Denisov F.I. Novoselov State Committee of USSR Atomic Energy Yu.D. Levchenko Physical and Energetic Institute, Obninsk E.A. Raud E.G. Romanova V.M. Nikitin Research and Design Institute for Petrochemical Industry, Moscow Yu.S. Virgilyev Research Institute of Graphite Technology, Moscow K.N. Koscheev Sverdlovsk* Division of Research and Design Institute for Power Technology G.A. Sernyaev IPHM Ural Division of USSR Science Academy A.H. Breger Karpov Physical and Chemical Institute, Moscow Yu.A. Dushin A.V. Ivanov A.V. Syschikov Experimental and Industrial Enterprise "Prometey", Leningrad
- 9 USSR (Continued)
R.G. Bdgoyalevsky A.B. Anapolsky R.V. Fedchin Ail-Union Research Institute of Atomic Machinery, Moscow A.S. Skotnikov Ail-Union Research Institute of Inorganic Materials Masukevich Minyailenko Institute of Engineering Thermal Physics, Kiev Yu.M. Krakhmalov Secretariat of Counsel Mutual Economic Assistance, CMEA G.E. Soldatov Ail-Union Research Institute of Atomic Power Stations, Moscow O.I. Smirnov V.D. Buchumov Ail-Union Research-Design Institute "ATOMENERGOPROJEKT" R.S. Demeshev I.E. Pozmogova Bauman Moscow High Engineering School L.E. Gardashnikov Research Chemical Institute of Leningrad State University E.A. Larin Saratov Polytechnical Institute G.G. Zemil'gotov All-Union Research and Project Institute of Complex Power Engineering V.Za. Karakhanov D.R. Ter-Grigurov Sukhumi Physical and Engineering Institute A.V. Polikalin Union Research Institute of Instrumentation Devices V.T. Vladimirov A.L. Malevsky Yu.V. Naumov Siberian Energy Institute of Siberian Division of USSR Science Academy, Irkutsk V.A. Gurin Yu.F. Konotop V.K. Yakovlev Kharkov Physics and Engineering Institute
- 10 USSR (Continued)
S.A. Orlov V.A. Tischenko E.O. Adamov S.E. Bugaenko B.S. Rodchenko A.G. Kraev Research and Design Institute for Power Technology, Moscow V.N. Sorokin A.P. Ipatov M.V. Malko Institute of Nuclear Energy, Minsk A.S. Chernikov P.P. Oleinikov A.A. Kuznetsov S.D. Kurbakov A.I. Deryngin K.P. Vlasov Podolsk Research and Technological Institute, Podolsk A.V. Adeev A.L. Khazanov V.N. Antropov N.I. Karpunin Science and Engineering Center "Gostatomenergonadzor", Moscow V.I. Savander A.N. Gudkov A.D. Kuperin Moscow Engineering and Physics Institute, Moscow A.S. Ivashkin Experimental and Industrial Enterprise of Research and Design Institute of Commission Installation, Moscow E.Yu. Vasilieva Yu.I. Pakhunkov Ail-Union Technology Research Institute of Radiation Engineering, Moscow M.P. Nesterov S.P. Perminov S.K. Elovsky Interpreters V.I. Levchenko V.Yu. Netkachev E.A. Kobalenko Technical Service O.I. Korotaev L.F. Kulyanitsa 6.A. Luchin Experimental and Industrial Enterprise Central Boiler-Turbine Institute, Leningrad
- 11 USSR (Continued)
N.V. Mizonov Leningrad Polytechnical Institute A.M. Vorobyov Biophysical Institute of Health Ministry of USSR, Moscow E.L. Starigny Obninsk Division of Moscow Engineering and Physical Institute N.F. Garelkin Industrial Enterprise ATOMMASH, Volgodonsk V.K. Ponkratov Kharkov Division of ATOMENERGOPROJEKT
N .B. Adamova V .N. Grebennik E .S. Glushkov Yu.G . Degaltsev I.S. Mosevitsky V .M. Alekseev L .K. Malkova A .0. Goltsev V .N. Krymasov V .A. Kerpov V .A. Panteleev I..V. Kurchatov Atomic Energy Institute, Moscow
A,.Yu.
Stolyarevsky
I,.V. Khromov Directorat VTR
V.,B. Ivanov Yu.I.. Kharlanov P.,G. Averyanov Yu.V.. Chechtkin M..P. Vorobei E.,K. Yakshin E.,1. Sokolov V.,F. Masny V.,?.Pochechura V.A. Kuprienko N.A. Aksenov B.V. Samsoraov I.G. Lebedev I.A. Kungurtsev E.P. Klochkov 0.V. Skiba P.T. Orodnov A.A. Maershin V.A. Kachalin V.N. Pridachin V.A. Shipilov I.G. Kobzar A.N. Kosorukov N.G. Golub V.I. Lenin Research Institute of Atomic Reactors, Dimitrovgrad
2.2-
- 12 -
YUGOSLAVIA
Mr. Varazdinec Zlatko Institut Za Elektroprivredu 41000 Zagreb Proleterskih Brigada 37 Yugoslavia Facsimile No: 041 530 604 Telex: 22154 YU INSTEP
IAEA
Mr. J. Kupitz Scientific Secretary Wagramerstrasse - 5 P.O. Box 100 A-1400 Vienna Austria Facsimile: 43 1 234564 Tel.: 2360/2814 Telex: 1-12645 atom a Mr. A. GoodJohn Co-Scientific Secretary Wagramerstrasse - 5 P.O. Box 100 A-1400 Vienna Austria Facsimile: 43 1 234564 Tel.: 2360/2808 Telex: 1-12645 atom a
***********
9418p/rt
Appendix II Gas-Cooled Reactor Design and Safety Technical Report Series Ho. 312 TABLE
CHAPTER 1.
OF
CONTESTS
INTRODUCTION
1.1. General information 1.2. Activities of the IAEA on gas cooled reactors 1.3. Contents of this report CHAPTER 2. 2.1. 2.2. 2.3. 2.4. 2.5. 2.6. 2.7. 2.8. 2.9.
United Kingdom France Germany United States of America Japan Uniion of Soviet Socialist Republics Switzerland China Others
CHAPTER 3. 3.1. 3.2. 3.3. 3.4. 3.5. 3.6.
STATUS OF INTERNATIONAL GAS COOLED REACTOR DEVELOPMENT
GENERAL FEATURES CHARACTERISTICS
OF
GAS
COOLED
REACTORS
GAS COOLED REACTOR DESIGN AND SAFETY IN THE UNITED KINGDOM
4.1. AGR design
4.7. 4.8.
4.9.
THEIR
Genral features of Magnox reactors and AGRs Safety characteristics of Magnox reactors and AGRs General features of HTGRs Fission product retention capabilities of coated fuel particles Safety characteristics of HTGRs GCR fuel storage, reprocessing and disposal
CHAPTER 4.
4.2. 4.3. 4.4. 4.5. 4.6.
AND
Safety concept Safety analysis Design approach to safety Safety principles and guidelines Protection systems 4.6.1. Fault detection and reactor trip systems 4.6.2. Reactor shutdown systems 4.6.3. Decay heat removal systems Design approach to hazards 4.7.1. Internal hazards 4.7.2. External hazards Safety analysis 4.8.1. Basis of safety analysis 4.8.2. Fault schedule 4.8.3. Probability analysis Discussion
SAFETY
- 2 CHAPTER 5.
GAS COOLED REACTOR DESIGN AND SAFETY IN THE UNITED STATES OF AMERICA
5.1. For St. 5.1.1. 5.1.2. 5.1.3. 5.1.4. 5.2. Nodular 5.2.1. 5.2.2.
Vrain nuclear generating stations Plant design Reactor design Safety characteristics Operating experience high temperature gas cooled reactor Plant description Design of major components for the MHTGR nuclear island 5.2.2.1. Reactor core and reactor internals 5.2.2.2. Steel vessels 5.2.2.3. Steam generator 5.2.2.4. Main helium circulator 5.2.2.5. Neutron control devices 5.2.2.6. Fuel handling equipment 5.2.2.7. Shutdown cooling system Safety of modular high temperature gas cooled reactors 5.2.3.1. Safety concept 5.2.3.2. Design approach to satisfy the MHTGR safety concept 5.2.3.3. Safety assessment 5.2.3.4. Safety assessment summary
5.2.3.
CHAPTER 6. 6.1.
6.2.
GAS COOLED REACTOR DESIGN AND SAFETY IN JAPAN
HTTR design 6.1.1. Plant design 6.1.2. Reactor system design 6.1.3. Cooling systems design 6.1.4. Irradiation equipment HTTR safety 6.2.1. Safety principles 6.2.2. Safety analysis
CHAPTER 7.
GAS COOLED REACTOR DESIGN AND SAFETY IN THE UNION OF SOVIET SOCIALIST REPUBLICS
7.1. VG-400 plant 7.1.1. Plant 7.1.2. Plant 7.1.3. Plant 7.2. VGM plant 7.2.1. Plant 7.2.2. Plant CHAPTER 8. 8.1.
layout design safety layout safety
GAS COOLED REACTOR DESIGN AND SAFETY IN SWITZERLAND
Gas cooled heating reactor 8.1.1. Plant design 8.1.2. Safety concept
- 3 CHAPTER 9.
GAS COOLED REACTOR DESIGN AND SAFETY IN GERMANY
9.1. AVR 9.1.1. Reactor plant description 9.1.2. Safety characteristics 9.2. THTR-300 nuclear power plant 9.2.1. Plant design 9.2.2. Primary system design 9.2.2.1. Reactor core 9.2.2.2. Fuel circulating system 9.2.3. Secondary system design 9.2.4. Safety concept of the THTR 9.2.4.1. Decay heat removal 9.2.4.2. Reactor core shutdown system 9.2.4.3. Effects from internal and external impacts 9.2.4.4. Safety assessment 9.3. HTR-500 9.3.1. Plant design 9.3.2. Primary system components 9.3.3. Safety Characteristics 9.3.3.1. Safety assessment 9.4. HTR module 9.4.1. Plant design 9.4.2. Primary system design 9.4.3. Safety concept 9.4.3.1. Safety assessment CHAPTER 10. ANNEX:
SUMMARY
INFORMATION EXCHANGE ON GAS COOLED REACTORS
LIST OF PARTICIPANTS
*********
8974p/8
Appendix III Special Issue of Energy, The International Journal High Temperature Helium Gas-Cooled Nuclear Reactors: Past Experience, Current Status and Future Prospects TABLE PREFACE:
OF
CONTENTS
Guest Editors M. Simnad, A.J. Goodjohn, J. Kupitz
FOREWARD:
R.A. Dean and L.S. Blue
INTRODUCTION:
R. Schulten
SECTION 1. HTGR AND NUCLEAR POWER PERSPECTIVES 1.1. Perspectives on the Overall Energy Situation and the Role of the Modular HTGR, Gordon L. Heins (GCRA, USA) 1.2. Construction and Operational Experience with Small and Medium Power Reactors and Expected Improvements, P. Dastidar and V.K. Mahadeva Rao (IAEA) 1.3. Trends in Nuclear Power Reactor Design and Technology, Jurgen Kupitz (IAEA) and A.J. Goodjohn (GCRA/IAEA) SECTION 2. HTGR DEVELOPMENT AND OPERATIONAL EXPERIENCE 2.1 2.2. 2.3. 2.4. 2.5.
SECTION
3.
The Early History of HTGRs, M.T. Simnad (UCSD, USA) Evolution of HTGR Coated Particle Fuel Design, O.M. Stansfield (GA, USA) Fort St. Vrain Operations and Future, H.L Brey (PSCC, USA) THTR Commissioning and Operating Experience, R. Baumer and I. Kalinowski (HTK, Germany) AVR Power Plant in its Last Year of Operation, C. Marnet, G. Ivens and E. Zeirman (AVR, Germany)
CURRENT STATUS AND PLANS FOR THE HTGRS 3.1. 3.2. 3.3. 3.4.
3.5. 3.6.
Summary of HTGR Programs, A.J. Goodjohn, (GCRA, USA) MHTGR Project Development Outlook, D.P. Hoffman and L.D. Mears (Consumers Power; GCRA, USA) Status and Prospects of the HTR-500 Based on the THTR-300 Operational Experience and Recent R & D, W. Theymann (HRB, Mannheim, Germany) Development Prospects of the HTGRs and Their Role in Nuclear Power, N.N. Ponomarev-Stepnoy, V.N. Grebennik, E.S. Glushkov, A.A. Krulev, A.I. Kiryushin and V.V. Bulygin (USSR) Present Status of HTGR Development Program in Japan S. Saito (JAERI, Japan) Studies on the Concepts of HTGR Plants Viable in Japan, S. An and T. Hayashi (Tokai University, Japan)
— 2— 3.7.
Survey of the Activities in Switzerland in the Field of HTGR Development, G. Sarlos, R. Brogli, D. Matthews, K.H. Bucher and W. Helbling (Paul Scherrer Institute, Switzerland) 3.8. Chinese HTR Nuclear Reactor and Demonstration Plant, Dazhong Wang, D. Zhong, Y. Xu (Tsinghua University, China) 3.9. Prospects for HTGRs in Developing Countries, J. Kupitz (IAEA)
SECTION 4.
DESIGN AND TECHNOLOGY DEVELOPMENTS 4.1.
4.2. 4.3. 4.4.
4.5. 4.6. 4.7. 4.8.
4.9.
4.10.
4.11.
4.12.
4.13. 4.14.
A Direct-Cycle Gas Turbine Power Plant for Near-Term Application: MGR-GT, L.M. Lidsky, D.D. Lanning, J.E. Staudt, Y.L. Yan (MLT, USA), and H. Kaburaki (JAERI, Japan) and Mr. Hori (IHI, Japan) MHTGR Fuel Performance and Supporting Base, D.T. Goodin (General Atomics, USA) Design Status of the HTR-500 and the HTR-Modul Plants, E. Arndt, R. Fischer (HTR-GmbH, Frankfurt, Germany) The Materials Program for the HTR in Germany: Integrity Concept, Status of the Development of High Temperature Materials and Design Codes, H. Nickel, E. Bodmann, and H.J. Seehafer (KFA, Juelich: HTR, Mannheim; Siemens UB, KWU-Interatom, Germany) Approach to Multi-Module Control of the MHTGR, C M . Woodworth, C. Rodriguez, and T.M. Starr (Stone & Webster; GA; ASEA-BBC; USA) R & D Programs for the HTGRs in Japan, I. Nishiguchi and S. Saito (JAERI, Japan) HTGR Fuel and Fuel Elements, A.S. Chernikov (USSR) Results of Reactor Tests of HTGR Spherical Elements, Y.G. Degal'tsev, A.A. Krulev, I.S. Mosevitskii, N.N. Ponomarev-Stepnoy, N.I. Tikhonov, and V.N. Yakovlev (USSR) Investigations of Radiation Stability of HTGR Coated Particles and Fuel Elements for HTGR, A.S. Chernikov, K.P. Vlasov, A.I. Deryugin, K.N. Koscheeve, and V.I. Tokarev (USSR) Behaviour of HTGR Coated Fuel Particles in High. Temperature Tests, A.S. Chernikov, R.A. Lyutikov, S.D. Kurbakov, V.M. Repnikov, V.V. Khromonozhkin, and G.I. Soloviyou (USSR) Structural Graphite for HTGR, Y. Virgiliev, P.Y. Avramenko, V.N. Grebennik, I.P. Kalyagina, I.G. Lebedev, V.A. Filimonov, and T.N. Shurshakova (USSR) A Stable Alloy for Large Components in HTGRs With Coolant Temperature of 950°C, Y.A. Dushin, N.N. Gribov, V.A. Ignatov, and N.A. Medvedev (USSR) Design Requirements for High-Temperature Metal Component Materials in the MHTGR, A.S. Shenoy and W.S. Betts (GA, USA) Integrated Design of Prestressed Cast Iron Pressure Vessel and Passive Heat Removal System for a 200-MWt MHTGR, B. Beine, V. Kaminski, and W. von Lensa (Siempelkamp; Interatom; KFA, Juelich; Germany)
- 3 SECTION 5.
SAFETY AND LICENSING ISSUES 5.1.
Questions Facing the NRC on the MHTGR, Kenneth C. Rogers (USNRC, USA) 5.2. NRC Review of the U.S. Advanced Reactors, T.L. King (USNRC, USA) 5.3. The Safety of MHTGRs, P.R. Kasten (ORNL, USA) 5.4. Theoretical and Experimental Investigations Into the Safety Behaviour of Small HTRs Under Natural Convection Conditions, H. Barthels, W. Rehm, and W. Jahn (NRC, Juelich, Germany) 5.5. HTR Safety Aspects and Safety Research, W. Kroger and N. Kirch (NRC, Juelich, Germany) 5.6. A Safety Yardstick for Evaluating Nuclear Power Reactors, T.E. Northup and J.M. Bolin (GA, USA) 5.7. The Safety Approach of the MHTGR, F.A. Silady and L.L. Panne (GA, USA) 5.8. The Licensing Experience of the MHTGR,, F.A. Silady, J.C. Cunliffe, and L.P. Walker (GA, USA) 5.9. MHTGR Radionuclide Source Terms for Use in Siting, S.B. Inamati, A.J. Neylan, and F.A. Silady (GA, USA) 5.10. Modeling of Fission Product Release From HTR Fuel for Risk Analysis, J. Bolin, K. Verfondern, T. Dunn, and M. Kania (GA; NRC, Juelich; GA; ORNL; USA) 5.11. Environmental Aspects of MHTGR Operation, A.J. Neylan, D.A. Dilling, and J.M. Cardito (GA; GA; S&WEC; USA) 5.12. Design and Safety Considerations in the HTTR, S. Saito, T. Tanaka, T. Sudo, 0. Baba, S. Shiozawa, and M. Okubo (JAERI, USA) 5.13. Analytical Method and Result of Off-site Exposure During Normal Operation of the HTTR, K. Sawa, H. Mikaml, and S. Saito (JAERI, USA) 5.14. Depressurization Accident Analysis for the HTTR by the TAC-NC Computer Code, K. Kunitomi, I. Nishiguchi, H. Wada, T. Takeda, M. Hishida, Y. Sudo, T. Tanaka, and S. Saito (JAERI, Japan) 5.15. The Influence of the Chemical State of the Fission Products on the Filter Properties of the Graphite Components, H. Barthels, C.-B. Van Der Becken, and N. Iniotakis (NRC, Juelich, USA) 5.16. Aerosol Formation by Graphite Corrosion in Case of Water and Air Ingress to the Core of the HTR, K. Kugeler, C. Epping, P. Schmidtlein, and P. Schneiner (IET, Duisberg, Germany) 5.17. Physical and Chemical Analysis of Interaction Between Oxide Fuels and Pyrocarbon Coating of Coated Particles, R.A. Lyutikov, Y.M. Kromov and A.S. Chernikov (USSR) 5.18. LEU-HTR Critical Experiment Program for the PROTEUS Facility in Switzerland, R. Brogli, K.H. Bucher, R. Chawla, K. Foskolos, H. Luchsinger, D. Matthews, G. Sarlos, and R. Seller (Paul Scherrer Institute, Switzerland) 5.19. Passive Heat Removal from the Core of Small and Medium Sized Pebble Bed Reactors, K. Kugeler and P. Schmidtlein (University of Duisburg, Germany)
- 3 SECTION 6.
MARKET POTENTIAL AND ECONOMICS 6.1.
Economic Assessment of U.S. MHT6R Design, L. Daniel Mears (6CRA, USA) 6.2. Market Prospects of Modular HTR in EEC Countries, F. Albisu, S.F. Garribba, J.C. Lefevre, D. Leuchs, and C. Vivante (EEC, Brussels, Belgium) 6.3. Main Design Aspects of an Advanced Nuclear Plant for the Venezuelan Orinoco Oil Belt Development, H. Carvajal-Osorio (IVIC, Caracas, Venezuela) 6.4. Concepts of Nuclear Heat Supply for Production of Aluminum Oxide and Refineries, M.K. Schad, (Lurgi, Frankfurt, Germany) SECTION 7.
SPECIAL TOPICS 7.1.
Prospects of SMSNR Development in Switzerland and the Influence of the Evolution of District Heating Networks, K. Foskolos, and R. Brogli (Paul Scherrer Institute, Switzerland) 7.2. The Nuclear Battery: A Very Small Reactor Power Supply for Remote Locations, K.W. Kozier (AECL, Canada) 7.3. Water Desalination for S. California with MHTGRs, General Atomics, Bechtel National Corp., and GCRE (GA; BNC; GCRA, USA)
8974p/ll
Safety Concept of Hiyh-Teaperature Reactors Based on the Experience with AVR and THTR
Winfried Wachholz Hochteraperatur-Reaktorbau GmbH Mannheim, FRG Dr. Wolfgang Kroger Institut fur Nukleare Sicherheitsforschung der Kernforschungsanlage Jiilich, FRG 1
Introduction
When can a reactor be considered as "safe"? A reactor is considered as safe and is licensable in the Federal Republic of Germany, if verification has been furnished that the requirements contained in § 7 of the Federal German Atomic Energy Act are met for this reactor:Demonstration of sufficient precautions against damage required according to the actual state of the art, and especially compliance with the dose limits for normal operation and accidents. These requirements result in a deterministic multi-stage safety concept with specified requirements for the engineered safety systems. It is claimed in these requirements that the plant has to be designed so as to withstand specified accidents. A difference is made between plant-internal accidents and accidents resulting from external impacts. In this concept it is aimed at establishing a high tolerance of the plant with regard to technical failure and human error. This is preferably to be obtained by physical inherent characteristics (such as e.g. self-shutdown of the
reactor) and by automation ( 30 rain without operator action). This concept has been further developed by creating a fourth level dealing with improved plant-internal accident management (Table 1 ) . In recent years, however, proposals for enhanced safety of nuclear reactors or a radical change in safety philosophy have been made, e.g./I/. This is characterized by the slogans:"inherently safe", "super safe" and similar expressions. A quantitative definition of these requirements has not yet been established, but it has crystallized as a common objective that even in the event of beyond-design basis accidents evacuation, relocation, and large-scale contamination of ground should not occur. The defition of an"extraordinarily safe" reactor was suggested in /2/ , a reactor which fulfils these criteria can be more acceptable and suitable for erection near urban centers, as serious consequences outside the plant are ruled out even in the event of accident sequences of an extremely low frequency of occurrance. This results in a proposal of quantitative requirements and procedures, respectively: 1. Observance of dose limits which are below the recommended values for plant-external protective actions (sheltering within buildings, evacuation, long-term relocation, areal decontamination) i.e. 1(- 10) rem as a 7-days whole-body dose 10(- 100) rem as a 30-days whole-body dose. The identification of the accident sequences should be performed using PRA/PSA methodology in a range which can be "seriously analysed", i.e. assuming < 10~8/a as cut-off criterion.
X
>M o = A-l
O SB --4 00
2. Verification of low sensitivity to deterministically assumed technical failure and human error (including sabotage) and extreme external impacts by means of hazard analysis. This means an extention of the current multi-stage . concept of safety precautions in nuclear power plants with respect to low-frequency envents ( see Fig. 2 ) . The consequence of this concept is a reactor with enhanced inherent safety characteristics, since exclusive reliance upon active engineered safety systems results in a restriction to lower safety margins. The high-temperature reactor as developed in the Federal Republic of Germany has the potential to fulfil these more stringent safety requirements. 2. Safety of the THTR Although the basic concept of the THTR-300 was designed 20 years ago when these extremely stringent requirements were not yet in existance, this potential has been demonstrated in practice for the THTR. Comprehensive subsequent investigations have shown that the THTR at least meets the short-term requirements of these high safety objectives in practice. This was verified by two official safety evaluations: - Determination of radiation doses in the environment of the THTR-300 resulting from a postulated core heat-up accident (1984) as a basis of emergency protection planning - Safety verification of the THTR by the Reactor Safety Committee (1988).
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2.1 Emergency Protection Planning Upon an order of the responsible State Minister the radiation doses were calculated which would have to be considered in the event of a hypothetical reactor accident in the environment of the THTR-300. In the same calculation the time intervals were determined which would be available for taking countermeasures. These two parameters have been used as a scientific basis for the planning of appropriate emergency measures /3/. A spontaneous irreparable failure of the reactor core cooling at full power operation was assumed as a representative accident sequence. As a result of this accident the reactor core temperature and the helium pressure would slowly rise.In addition it was assumed that the helium purification system which is normally used to control the primary pressure is not available and one of the safety valves in the connecting line fails to close again after having opened eight hours after the beginning of the accident. The overall effect would be a discharge of coolant gas with unfiltered release of radioactivity from the reactor through the stack for about two days. The probability of occurrence of such a scenario is extremely low. After opening of the safety valve mainly noble gases and iodine would be released in the beginning. Release of cesium and strontium would start only more than 30,respectively 50 hours later. The overall amount of hazardous fission products would be relatively low. The radiation doses were pessimistically estimated for an exposure time of 7 days: The whole-body dose remains below 5 rem. The thyroid dose remains generally below 25 rem, if a prohibition of consumption is issued
within 8 hours after the beginning of the accident, (see Fig.3, recommended values of the hazard classes for emergency protection are indicated for comparison.)
- steam generator tube rupture - steam generator depressurization or - activity confinement have been satisfactory.
The doses determined by these calculations have to be considered as a kind of upper limit which - with a high probability - would by far not be reached in the event of a real accident . In any case the accident seguence leaves sufficient time for detecting and analysing the accident situation, measuring the actual radiation and initiating appropriate measures. As the Federal Guidelines have been revised, the emergency planning for the THTR-3000 will be reviewed. For this purpose the cited study will be updated and extended within the next two years.
For a further enhancement of the plant safety some minor hardware changes were recommended in order to permit reali2ation of the following measures: - Filtering of the inlet air for the control room - Emergency feed of the liner cooling system by fire extinguishing water - Possibility of discharge of the primary gas into the helium storage system - Filtered discharge of primary gas through the stack.
2.2 safety Review by
the Reactor Safety CoMissioiL
As a conseguence of the Chernobyl accident the safety of all the nuclear power plants operating in the Federal Republic of Germany has been reviewed including their behaviour in the event of beyond-design basis accidents. This analysis was performed by the Reactor Safety Commission (RSK) and has been completed for the THTR in 1988. In the final evaluation of the RSK the HTR-specific safety characteristics have been fully confirmed/4,5/. A special safety potential is attributed to the prestressed concrete reactor vessel and inherent reactivity control. The verifications furnished for the analysed accident sequences initiated by - depressurization accident with air ingress - reactivity accidents
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In addition, a state-orientated procedure was recommended for the way of handling accidents in the Operating Manual These backfits are , however, not a technical prereguisite for safe plant operation. They do not represent necessary supplements to the safety concept but plant-internal measures to support the emergency protection planning so as to further reduce the anyway low residual risk also in the range of beyond-design basis accidents.
3. Safety Contribution from the Avp The AVR has proven convincingly in more than 20 years of operation the feasibility, functional capability and safety of HTRs with spherical fuel elements.In view of its plant availability of almost 70 % averaged over the years, it has also furnished the most important and comprehensive contribution to the HTR-specific operating experience which can also be used to confirm
the reliability parameters for components exposed to primary gas. Further aspects which have been useful for the safety evaluation are disturbances which entailed unintended plant conditions. By far most of these disturbances were of no safety relevance. The three most important events of this type were: - Unintended feed of air (1971) - Refueling error (1974) - Water ingress (1978). Specific experiments which were performed in the AVR have also been very useful for the safety evaluation and demonstration of relevant safety characteristics. Especially the safety concept of small modular HTRs is mainly based on AVR experience./6/. 4. Hew Concepts The HTR concepts currently under development are based on this experience. Depending on the application and si2e of the plant, different technical solutions are, however, adopted to meet the safety objectives. 4.1 HTR-Modul The HTR-Modul is a reactor unit designed with a rated thermal output of 170 to 200 MW each, depending on its application, a core power density of 3 MW/m3 and a gas outlet temperature of 700 °C. Several units can be combined to form a station of the desired plant size. For the HTR-Modul proven technology has been applied on the basis of the AVR reactor in Julich, the operation of the THTR-300 and the construction and operation of light-water reactors:
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The main design characteristics are as follows: - The design of the reactor core ensures that in the event of any potential accident the fuel element temperature of 1620 °C is not exceeded. - Active core cooling is not required for decay heat removal in the event of accidents; it is sufficient to let the decay heat dissipate by passive heat transfer mechanisms (such as heat conduction, heat radiation, natural convection) to a water-cooled surface cooler of simple design, which is arranged outside the reactor pressure vessel in the reactor cell. - Shutdown of the reactor by absorber elements which are gravity-released upon demand into bore holes in the side reflector outside the reactor core. In the HTR-Modul the core and the steam generator are installed in separate steel pressure vessels in a side-by-side arrangement connected by a coaxial steel duct (Fig. 4 ) . The high standard of quality assurance developed for LWRs according to the principle of "basis safety" has been transferred to HTR conditions and applied to all units. Therefore, spontaneous rupture of the pressure vessel units can be ruled out. The concrete of the reactor cell is lined with a water-cooled surface cooling system (3 x 100 %) designed to absorb heat losses during normal operation and decay heat transported from the inner core to the reflector and further through the vessel wall. This system is of no relevance to safety but only to the protection of components, mainly to keep the maximum temperatures of the pressure vessel at its bearings within the design range of about 400 °C in case of
total loss of forced core cooling. Even if this system should fail, the core temperatures would stabilize inherently (maximum 1620 °C, almost similar to the temperature conditions with the concrete cooling system in operation); a loss of structural integrity is practically not to be expected. Accident control by means of inherent or passive safety has been the guiding design principle for the HTR-MODUL, essentially:
-
No active system is needed to remove the decay heat from the (slim) core or to keep maximum temperatures of fuel elements below failure limits under all conceivable accident conditions. Arrangement of the different pressure vessels of the primary circuit to suppress natural convection and to limit the amount of water potentially ingressing into the reactor core. Core design ensuring reduction of positive reactivity introduced by withdrawal of absorber elements or by water ingress.
Thus it is ensured that no hazard to the environment can arise from the HTR-MODUL power plant, neither during normal operation nor in the event of accidents or beyond-design basis accident sequences.
4.2 MR-BOO The HTR-500 nuclear power plant, a 550 MW steam cycle plant with a core power density of 6.6 MWth/m3 and 700 °C gas outlet temperature, has been developed as an electricity generating plant offering the possibility of process steam and district heat extraction.
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The H T K - 5 0 0 is based to a large extent on the technology licensed and realized in the THTR-300. The reactor with its pebble bed core and the six-loop primary circuit is integrated into a massive prestressed concrete reactor vessel (PCRV) equipped with a leak-tight insulated steel liner which is water-cooled by the Liner Cooling System (LCS). Two independent shutdown systems with simplified drive mechanisms are provided - one with reflector rods for control and hot shutdown, the other with in-core rods for cold shutdown. Nevertheless, the HTR-500 design will feature significant improvements compared to the THTR-300, e.g. it will be equipped with a two-loop independent auxiliary system for decay heat removal making use of natural convection under operating pressure. The allowable failure times for the decay heat removal system and the liner cooling system without impairing the safety are estimated to be 5 and 24 hours, respectively. The HTR-500 plant concept has been modified in the meantime compared to the former design stage (1983) in view of a higher economics as well as an enhanced safety engineering concept. To obtain a further reduction of radiation exposure to the public in case of malfunction of plant components or in case of accidents, the pressure relief system of the prestressed concrete pressure vessel was modified so that the primary coolant exhausted from the safety valve is directed into the helium storage facility instead of the reactor building so that no
radioactivity is released.(see Fig.5). Only if the helium storage facility is not available, an additional safety valve opens into the relief line with metal fiber aerosol filter and the helium content is discharged directly from the primary circuit via the stack to the atmosphere. The ventilation system was backfitted with an accident filter system (metal fiber filter and molecular sieve bed). Thus in the event of a depressurization accident with subsequent core heat-up, the residual primary coolant ascaping from the primary circuit can be discharged through that filter system via the stack. This filter system does not need any active components but makes use of the differential pressure between the reactor building and the atmosphere. For the HTR-500 core heat-up accidents resulting from loss of forced convection in combination with depressurization dominate the source terms. Privided that the primary circuit is depressurized from the beginning, a few percent of the fuel elements reach temperatures (viz. 2500 °C) after about two days, which cause total failure of the essential barriers against release of fission products, i.e. of the coatings of the particles in the fuel elements. But 50 % of the fuel elements are exposed to uncritical temperatures not higher than about 1600 °C. The frequency of this event v.ith escape of radioactivity to the atmosphere is less than 10 reactor year. The effluent radioactive medium is filtered before it is released via the stack.
4.3 Source Terms In case of the advanced HTGR concepts the source terms of severe accidents are up to several orders of magnitude lower than the values specified in /7 and 8/ for new light-water reactors. This can be seen from the table below:
Release in Percent of Core Inventory —8 —7 Probability 10 to 10 per year. Reactor Type Sr
HTR-MODUL HTR 500 LWR /7/8/
io~ 5 % io"4 % 10" 1 %
CS
I,Br
10~ 3 % lO"3 % 1 %
10~ 4 % io" 3 % 1 %
Xe,
10 -4 1 10
+2 %
Sureaary
per
It has been proven by experiments in combination with theoretical investigations that the prestressed
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concrete reactor vessel of the HTR-b00 will not be damaged by loss of forced convection to any degree that could affect the accidental radiological consequences, even if in addition the liner cooling system would not be available. Conservation of the integrity of the PCRV even in the event of conceivable beyond-design basis accidents furnishes the essential basis for activity confinement.
It can thus be summarized: Clearly defined conditions can be stated for the new requirements for an enhancid safety of nuclear power plants. The HTR has not only the potential to meet these
requirements, but these requirements have already been verified and confirmed in practice to a large extent in the AVR and the THTR. The new HTR plant concepts meet these requirements in different ways: .the HTR MODUL mainly by limitation of the maximum fuel element temperatures .the HTR-500 by making use of the high safety potential of the prestressed concrete reactor vessel. Thus the potential activity release for both reactor concepts is about 2 - 3 orders of magnitude lower than the values recently specified for LWRs, and this for accident sequences with probabilities of occurrence of 10~ 7 to io"8/a.
1. A. Weinberg et al. Energy Vol. 10, No.5, pp.661-680, 1985 2. W. Kroger, Nuclear Engineering and Design 109 (1988) 295-298 3. Fassbender et al., Jul - Spez-275,1984 4. Final Report on Results of Safety Review of NPP in the FRG, Recommendation by the RSK, Nov. 1988 5. W.KrSger, H. Nickel, Safety/Licensing Reviews in the Federal Republic of Germany with Emphasis on THTR-300 (accepted for publication in Nuclear Safety, Vol. 4, 1989)
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6. W.Kroger, VDI Berichte 729, pp 255 - 270, ISBN 3-18-090729 - 0,1989. 7. J.E Speelmann, J.A.Heil and J.Hejn, "Severe Reactor Accidents Reconsidered: The Source Term". ECN-PB-88-16. 8. Stuurgroep Project Herbezinning, "Summary of the Nuclear Engergy Review Project, May 1988".
Do Events (plant condition) / Event sequences
Extended Multi-Stage Concept for SafetyPrecautions in Nuclear Power Plants
Method to demonstrate Frequency (mean, per reactor - year) compliance (superior) Description
Design Plant conditions
Safety precautions
Statistics or verifiable PSA
Operation + Incidents 3-10*
10*-
Level 1
Normal Operation
• not tofteexpected for single plant, but practically not to be excluded tor savfrti plants
Quality Assurance
(KTA-UA-SF proposal)
10'
Level 2
Operational disturbances
Design basis accidents
Inherently safe response Accident prevention
Plant Protection System Safety equipment
10 « PSA> entirely to be taxonomy, excluded but rough estimates mechanistically plausible
risk reduction (as supplement to licensing procedure)
Level 4
Beyond - design basis accidents
in-plant accident management (prevention of core damage, protection of containment)
Design basis accidents
Fig. 1
I a
Postulates, «not mechanistlworst-ease/ Cally plausible, hazard anabut Imaginable lyses
10°
30mrem/year (§45)
off- site emergency planning
(KFA-ISF proposal)
Beyond-deslQn basis accidents
10'
i"
Beyond — design basis accidents
'practically, not entirely to be excluded for several plants
PSA-verifiable contributors/ key events
Level 3
(beyond licensing procedure)
• normal or to be expected for single plant
I 101
10*
Srem/event (§28.3)
Consequences (whole - body long-term dose, rem)
Requirement for 'Extraordinary Safe' Nuclear Power Reactors of Next Generation (frequency-oriented dose limits) Fig. 2
S-
L\
Hazard class »
Hazard doss 1 - Sheltering In bufcfag* — EwcuoUon
5_
2.
Whole-body dose
5.
from 2.
inhalation and external
5.
iTQdiation ( rein )
Hazard doss I 2_
10 Distance ( m ) F i g u r e 3 a : Thyroid dose from inhalation of kxiine and tellurium after release of activity in the event of a postulated core heat-upp accident of the THTR 300 as a funptip" of the
(D
ice
(//////• Plant aite)
10
Distance ( m ) F i g u r e 3b: Whole-body dose after activity release in the event of a postulated core heat-up accident of the THTH as a function of the distance. (Main contribution : inhalation of Sr 90 ). ( ////// : plant site)
A
OHRS HCS LCS GPP GS RS VS RB
~7Z2
-
Decay heat removal system Main cooling system Liner cooling system Gas purification plant Gas storage Relief system (PCRV) Ventilation system Reactor building
GS
Fig.
5 :
HTR 500 Modified plant concept
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ko
XA0101479
Depressurization Accident Analysis for the HTTR by the TAC-NC
k'azuhiko KUKITOMI, Isoharu NISHIGUCHI, Tetsuaki TAKEDA, Makoto HISHIDA, Yukio SUDO, Toshiyuki TANAKA and Shinzo SAITO
Department of HTTR Project Japan Atomic Energy Research Institute
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ABSTRACT The two-dimensional thermal a n a l y s i s a n a l y t i c a l code TAC-2D
code TAC-NC i s modified
in order to c a l c u l a t e
from the
temperature t r a n s i e n t s in the
case of l o s s of f o r c e d c o o l i n g a c c i d e n t s of t h e HTTR (High T e m p e r a t u r e engineering Test Reactor) such as a depressurization accident. in the core are affected
In these
conditions,
temperature transients
convection
between hotter and colder regions in the pressure vessel.
TAC-NC code includes a special function n a t u r a l convection
in addition to
by natural The
to calculate heat transfer by
conduction, radiation and forced
convection. Verification of the TAC-NC code was carried out by the comparison of the analytical results with the experimental ones of an air ingress
test.
Analytical results of simulated core temperature were in good agreement with experimental results. Temperature t r a n s i e n t s
during a depressurization accident were
evaluated by the TAC-NC code for the HTTR.
The maximum fuel
temperature
decreases rapidly after
the reactor scram and increases slightly after that
due to decay heat.
maximum fuel temperature, however, remains
The
below
the i n i t i a l maximum fuel temperature since most of the core decay heat is absorbed in the large thermal capacity r e f l e c t o r . The peak vessel temperature
of g r a p h i t e
in the core and
occurs at about 30 hours
beginning of the accident and also remains lower
than the
after the allowable
temperature, even if one of the reactor pressure vessel cooling systems is failed.
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CONTENTS
1. INTRODUCTION 2. ANALYTICAL MODEL OF TAC-NC 3. TAC-NC CODE VERIFICATION 3.1 Test Apparatus and Procedure 3.2 Analytical Condition and Model 3.3 Analytical Results U. TEMPERATURE TRANSIENTS DURING DEPRESSURIZATION ACCIDENT IN HTTR BY TAC-NC 5. CONCLUDING REMARKS
REFERENCES
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1. INTRODUCTION
The JAERI has been developing Che safety analysis codes for the HTTR[1],
For the licensing of the HTTR, the JAERI has been carried out the
necessary R&D for verification of the codes as well as for establishment and upgrading of HTGR technology basis[2,3].
In addition, safety analyses for
postulated accidents were evaluated by these newly developed codes. In the case of the primary pipe rupture accident (depressurization accident), residual heat is removed from the outer surface of the reactor pressure vessel by VCSs (reactor pressure Vessel Cooling Systems) and fuel temperature does not exceed the temperature limit although coolant pressure decreases rapidly according as coolant leaks through, a break area and forced cooling is stopped.
One of the inherent safety features of the HTTR
is that there is no forced cooling system during the depressurization accident. The TAC-NC code was developed to calculate the temperature transients in the core during the depressurization accident and to confirm the inherent safety feature of the HTTR[4]. This paper presents the analytical model and verification results of the TAC-NC by the comparison with the experimental results and analytical results of the depressurization accident of the HTTR.
2. ANALYTICAL MODEL OF TAC-NC
The TAC-NC code was developed to calculate the thermal and hydraulic behavior composed of conduction, radiation, forced and natural convection. Figure 1 shows the thermal and hydraulic behavior during the depressurization accident of the HTTR. Under normal condition, generated heat is mainly removed from the core by primary forced coolant, and 1 to 2 % of generated heat is transfered through the stacked fuel block to the reactor pressure vessel by conduction and radiation and finally
removed
from the outer surface of the reactor pressure vessel by VCSs. In the case of the depressurization accident, the reactor is scrammed immediately by detecting decrease of differential pressure between the
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primary and secondary coolant.
After the reactor scram, forced cooling by
primary coolant is stopped and residual heat is removed from the core by natural convection and radiation by VCSs.
Coolant or air ingressed through
the break area flows upward in the core, turns downward under the top head of the reactor pressure vessel and flows downwards along the inner surface of the reactor pressure vessel by natural convection. The TAC-NC can calculate the steady-state and transient temperatures in the reactor core and pressure vessel under these condition.
It was based on
the TAC-2D code which was developed by GA company to calculate transient temperatures in one dimensional problem[5]. Natural convection of the core is simulated by one-dimensional flow network model. The flow network which consists of flow passages and plenums is assumed for the
calculation of the natural circulation (see Figure 4 ) .
The basic flow network equations to be solved are the steady-state, onedimensional momentum equation for each passage and continuity equation, equation of state and energy equation for each plenum. The momentum equation for each passage has the following form:
Du.
l
•DT
i
-o
1
a?
i
*•
,
(
- ' -J[ ~aT " a ~ —
i
,
+ c
i
)
x
T
1
1
K
I
1 u
i
where c :inlet resistance coefficient in the flow passage d :hydraulic diameter of the flow passage P :fluid pressure u :average flow velocity over the cross section p :fluid density X :friction factor i :flow passage number z :coordinate of flow direction
Dt ~ 3t
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U
3. TAC-NC CODE VERIFICATION
Verification of the TAC-NC code was conducted by the comparison of the analytical results with the experimental ones of the air ingress test. The air ingress test was carried out to evaluate transient behavior of the air in the case of pipe rupture accidents.
3.1. Test Apparatus and Procedure[6]
Figure 2 shows a vertical and horizontal cross sectional view of the test apparatus.
The test apparatus is a about 1/10 scale simulation model
of the HTTR which consists of a reactor core simulator, a plenum simulating simply the high temperature plenum, simulated top and bottom plenums corresponding to the top and bottom covers of the reactor pressure vessel, inlet and outlet pipes corresponding to a primary pipe of the HTTR and measuring system. The reactor core simulator consists of 43 heater pipes and insulations between
heater pipes.
fiber
The length, inner diameter and power of
a heater pipe are 1200 mm, 12.7 mm and 36 W respectively.
The heater pipes
raised the temperature of air up to 450°C. Thermocouples on the heater pipes and other structures are installed to measure temperature transient after the pipe rupture.
Flow rate of natural
convection between hotter and colder region of the test vessel was measured by the ultrasonic flow meter installed at the inlet pipe. The
test procedure is as follows:
(A) Inlet and outlet nozzles are closed with blankflanges. (B) Simulated core temperature is
raised up to the maximum 450°C by the
heater pipes. (C) After the temperature distribution in the test vessel is attained
in a
steady-state condition, the blankflanges are removed from the nozzle and ambient air is ingressed into the test vessel by diffusion and natural convection.
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3.2. Analytical Condition and Model The experimental conditions compared with the analysis are as follows. (A) Total heater power is 200 W (B) Maximum temperature of the heater pipe is 300°C. (C) Inlet temperature of the cooling water is 25°C. Figures 3 and 4 show the analytical model and natural convection flow network.
Natural circulation flow channels consist of simulated core flow
channels, flow passages along the test vessel, pipes and plenums. 3.3. Analytical Results (1) Steady-state temperature in the test vessel Figure 5 shows the steady-state temperature in the test vessel before the blankflanges of the inlet and outlet pipes are opened. The analytical results agree well with the experimental ones. (2) Transient temperature in the test vessel Figures 6 and 7 show the experimental and analytical temperature at the surface of the heater pipes. Analytical temperature at the bottom and middle position of the heater pipes are in good agreement with the experimental one. It is somewhat difficult to predict the temperature at the top position of the heater pipe. Both experimental and analytical temperature at the top of the heater pipes increased about 30°C after 2 hours. This temperature increase was attributed to heat transfer from the bottom and middle position to top by natural convection. Analytical temperature however is about 15°C as low as experimental one. The difference in temperature is due primary to the following reason. The equivalent thermal conductivity of the core simulator is calculated from the volume average of the core simulator such as the heater pipes and insulations. The value depends on both temperature of the core simulator and permeability in the insulations. Permeability is the dominant factor of the local natural convection in the insulation which affects the equivalent
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49-
thermal conductivity of the core simulator.
The present analysis considers
dependence of thermal conductivity on temperature, however, does not consider the permeability change in the insulation. Hence, for the safety analyses of the HTTR, properties such as thermal conductivity, heat capacity and emissivity are considered to be conservative so that
fuel and vessel temperature might be evaluated to be higher than
real temperature.
(3) Flow rate of natural circulation
Figure 8 shows the flow rate of natural convection in the test vessel. Analytical result of flow rate due to natural convection is a little higher than the experimental result. This is because analytical pressure loss of the simulated core is considered to be lower than the experimental one. The results are conservative for the calculation of the graphite oxidation of the core.
4. TEMPERATURE TRANSIENTS DURING DEPRESSURIZATION ACCIDENT IN HTTR BY TAC-NC
The major design feature of the HTTR is that there is no forced cooling for residual heat removal after faults due to inherent safety characteristics such as large heat capacity and low power density of the core. The HTTR cooling system is, therefore, designed simply considering the inherent safety characteristics.
It consists of a main cooling system
(MCS), an auxiliary cooling system (ACS), and two reactor vessel cooling systems (VCSs).
The ACS is in the stand-by condition during normal reactor
operation and operated to remove residual heat from the core when there is a trouble in the MCS.
Both VCSs are operated at each 100% flow rate during
the normal operation in order to cool the biological shield around the reactor vessel, and they serve to cool the reactor vessel and the,core in the case of the depressurization accident.
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Temperature transients during the depressurization accident are evaluated with the TAC-NC in order to confirm the inherent safety characteristics of the HTTR. Initial outlet gas temperature and outlet power before the accident for the analysis are 950 C and 30 MW respectively and the one of the VCSs are assumed to be failed as a single failure. Their behaviors
after the accident depend on the magnitude of the fuel
burn up since the properties of the graphite such as the thermal conductivity changes in high temperature and irradiation exposure.
The
analysis uses the relevant properties so that the highest temperature was obtained in each evaluated structure.
Figure 9 shows the maximum fuel
temperature in the accident at the end of core life, and the maximum pressure vessel temperature
at the beginning of core life.
The maximum
fuel temperature decreases rapidly after the reactor scram and then increases slightly due to decay heat. The peak fuel temperature is 1380 C at about 30 hours after the accident was started. It, however, remains below the initial fuel temperature 1495°C. The vessel temperature increases up to 530°C at about 30 hours and decreases gradually.
It also remains under the
allowable temperature of 550°C.
5. CONCLUDING REMARKS (1) Verification of the TAC-NC code was carried out through the comparison with the results of the air ingress test.
It was confirmed that
the analytical results of temperature were in good agreement with the experimental results when equivalent thermal conductivity was properly used for the calculation. Analytical result of flow rate due to
natural
convection was a little higher than the experimental ones. The results were
conservative for the evaluation of the graphite oxidation in the
core.
(2) The maximum fuel temperature of the HTTR during the depressurization accident does not exceed the intial temperature in this analysis. This is because most of the core decay heat is absorbed by the large thermal capacity of the core.
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k<\
REFERENCES
[1] S . S a i t o ,
" P r e s e n t S t a t u s of HTGR Development Program in Japan", 11th
International Conference on HTGR, Dimitrovgrad USSR,June 19-20, 1989 [2] S . S a i t o ,
et a l . ,
" D e s i g n and S a f e t y C o n s i d e r a t i o n
in the High
Temperature Engineering Test Reactor (HTTR)", IAEA Technical Committee Meeting on Gas-cooled Reactor technology Safety and s i t i n g , Dimitrovgrad USSR, June 21-23, 1989 [3] S.Maruyama, e t a l . , " V e r i f i c a t i o n of I n - c o r e Thermal and Hydraulic Analysis Code FLOWNET/TRUMP for the High Temperature Engineering T e s t Reactor at JAERI", to be published in NURETH-4,October 1989.. [4] K.Kunitomi, et a l . , "Two-dimensional Thermal A n a l y s i s Code TAC-NC for High Temperature Engineering Test Reactor and i t s Verification", JAERI-M 89-001, 1989 [5] S . S . C l a r k and J . F . P e t e r s e n , "TAC-2D A General Purpose Two-Dimensional Heat Transfer Computer Code", GA-9292, September 1969 [6] M.Hishida e t a l . , " S t u d i e s on the Primary Pipe Rupture Accident of a High Temperature Gas Cooled R e a c t o r " , to be published in NURETH-4, October 1989.
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VESSEL COOLING SYSTEM
WATER TOP COVER
UPPER PLENUM HEATER PIPE ( 3 7 * 6 :
NATURAL CONVECTION
REACTOR CORE SIMULATOR .•PRESSURE
RADIATION
.HIGH TEMPERATURE PLENUM
CONDUCTION
UGH TEMPERATURE OUTLET PIPE LOWER PLENUM BOTTOM •'' COVER
BOTTOM COVER
ULTRASONIC FLOW METER INLET PIPE
Fig.1
THERMAL AND HYDRAULIC BEHABIOR DURING DEPRESSURIZATION ACCIDENT OF HTTR
Fig.2 VERTICAL AND HORIZONTAL CROSS SECTIONAL VIEW OF TEST APPARATUS
-UPPER PLENUM
—
_ jrrnnurnr-Uuci j -—
—
——
VESSEL
1
—
—
X TEST
~\
—
— -— —
' - / C O R E SIMULATOR 1
-
-1
UPPER
PLENUM
— —
- ' INSULATION -CORE SIMULATOR FLOW PASSAGES
— — —
A
•yr-
— —
.LOWER PLENUM -FLOW PASSAGE ALONG TEST VESSEL
—
— -OUTLET PIPE
FLOW DIRECTION
--
— — -—— — —
—— — —
——
—
- ——
—
LOWER PLENUM
—
-BOTTOM COVER --
1! 1 1 1 1 1M
AXIAL DIRECTION
-— —
—— — — —
—
-
-INLET PIPE
A —
-INLET RADIAL OIRECTION
Fig.3 A-2
BOTTOM COVER
ANALYTICAL MODEL OF TEST APPARATUS
PIPE
•OUTLET PIPE
Fig. 4 FLOW NETWORK MODEL OF TEST APPARATUS
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CORE SIMULATOR
INSULATION
500 o o400
^ANALHICAL RESULT
§300 <
^PERlMENTALRESUtr5^^
§200 u>
\
400
>
100
EXPERIMENT (MIDDLE) EXPERIMENT ( T O P ) EXPERIMENT (BOTTOM) ANALYTICAL RESULTS
1
0
0.05 RADIAL
0.1 DISTANCE
300
0.2
0.15 { m)
o o
CORE SIMULATOR
sus
SUS
500
cc
UJ
a:
ANALYTICAL RESULT
UJ
300 Q:
P I00
EXPERIMENTAL RESULT
CROSS-SECTION
200 LU
100 0 BOTTOM Fig. 5
0.2
i
1
0.4
0.6
AXIAL
STEADY-STATE
t
0.8
DISTANCE
( m)
5
1.0
TOP
TEMPERATURE IN TEST APPARATUS
10 15 20 ELAPSED TIME ( h )
Fig.6 RELATIONSHIP BETWEEN ANALYTICAL AND EXPERIMENTAL RESULTS OF HEATER PIPE TEMPERATURE (CENTER PIPE)
400 ~ o - EXPERIMENT (MIDOLE' — a — EXPERIMENT (TOP) —*— EXPERIMENT (BOTTOM1 ANALYTICAL RESULTS UJ
cc t
en UJ
5
10 15 20 ELAPSED TIME ( h )
25
Fig. 7 RELATIONSHIP BETWEEN ANALYTICAL AND EXPERIMENTAL RESULTS OF HEATER PIPE TEMPERATURE (PERIPHERAL PIPE)
-EXPERIMENTAL RESULT "ANALYTICAL RESULT
5 10 ELAPSED TIME
15
20
(h)
Fig. 8 RELATIONSHIP BETWEEN ANALYTICAL AND EXPERIMENTAL RESULTS OF FLOW RATE OF NATURAL C0VECT10N A-2
1600
i
I
i
r
1495t ([NITIAL TEMPERATURE!
1400 1200
MAXIMUM FUEL TEMPERATURE ( END OF LIFE)
-1000 UJ
800 ^TEMPERATURE LIMIT OF PRESSURE VESSEL
600 550
'MAXIMUM PRESSURE VESSEL TEMPERATURE BIGINNING OF LIFE )
400 200
i
0
20
i
I
I
l
40 60 80 ELAPSED T I M E ( h )
100
Fig.9 ANALYTICAL RESULTS OF FUEL AND PRESSURE VESSEL TENPERATURE DURING DEPRESSURIZATION ACCIDENT
A-2
XA0101480 CALCULATION STUDIES OF BEHAVIOR OF HTGR SPHERICAL FUEL ELEMENTS DURING AN ACCIDENT INDUCED BY A HIGHPOSITIVE REACTIVITY FAST INTRODUCTION
Gol'tsev A.O.
INTRODUCTION The present paper describes the results of calculated studies of the behavior of a high-temperature gas-cooled reactor during an accident related to the steam generator tube ruptures and the steamwater mixture ingress together with coolant into the reactor core, as well as during a hypotehtical accident associated with a spontaneous extraction of all EPS shim rods at once from the pebble bed of a cooled, unpoisoned reactor, at which one can introduce a maximem positive reactivity determined by physical reactor core features. One of the main objectives of the present paper was to study the reactor capability to self-regulate a heating capacity due to, as it is now customary to say "a built-in (immanent, inherent etc.) safety", at maximum design and beyond-design ("hypothetical" accidents. Therefore, in these calculations it was expected that the emergency safety system does not response, a speeding-up is damped by the feedback (mainly, by Doppler effect), and in this case the core construction elements (refluctor blocks, CSS rods, fuel elements and microfuel elements) at a fast heating-up are not exposed
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to thermal expansion or failure* The dynamics of energy release and temperature in maximum stressed reactor fuel elements was studied in parallel in the present paper* The calculated investigations were carried out, as applied to HTGR with spherical fuel elements, operating on the OTTO principle, with a heating capacity - 1000 MW, and reactor core volume
150 m .
!• CALCULATED RESEARCH PROCEDURE All results in the present paper were obtained from the program; CTAPT3* This program is designed to solve problems of the HTGR dynamics eith spherical fuel elements in one-dimension Z-geometry. A neutron transfer is described in a two-group diffusion approximation by following equations (designations are standard):
or- v It is assumed that: Cf^ , Z-^>
r and ^L./^ - depend on the fuel tempera
ture (T f ), burnup (wt) and water concentration (in an emergency situation); b)
Z/^ j ^ £
^<£»y^ 9 ^*Xe 5 ^SMare
depending on the
graphite temperature C^r)» burnup (wt) and water concentration (in an emergency situation); c) D^, Dg and ^. depend only on the water concentration (f „ Q ) . d) J$j depends on burnup (wt) Por the given burnup (wt) constants are described by a polynomial of the form:
for the group (a) constants, where T^ is a fuel temperature, or by a polynomial of the form
for the group (b) constants, where T M is a temperature of graphite. By a fuel temperature is here meant the temperature of UOg microfuel element kernel. — is a concentration of water molecules in helium; a2..«ag - constant, for the given burnup,coefficients. For group (c) constants the a 2 ,a^ and a^ coefficients equal to zero. Besides,
£\
depends on a concentration of Xe ^
and Sm
,
the concentrations of which are determined from the solution of equations:
= r* (Z A
'/» _ The presence of OSS rods is taken into account by effective additions to absorption cross-sections of the first and second groups. A preparation of neutron-physical constants for the program CTART 3 was accomplished according to the program DRAKOH f1J which is intended to calculate two-group neutron-physical constants for thermal reactors with allowance for thermalization of neutrons, fuel burnup and double heterogeneity of HTGR fuel elements Neutrons-graphi te scattering is calculated in a non-coherent approximation. A water (steam) scattering is taken into account from the free gas moA-3
del. The fuel burnup calculation in the spherical cell (the fuel element with spherical helium layer) is carried out bearing in mind an accumulation of plutonium isotopes, Xe, Sm and slags. Heat from fuel elements to gas is described transfer by equations:
1
at t
s
S* J
dpw
S - passing cross-section for"
P
7^ 6?) - gas temperature
R
~ temperature of wall of a
^zpwTQi
{
fuel element or reflector block
const
- helium density - helium pressure.
A heat transfer ober a solid phase (fuel element or reflector blook being in the reactor zone with coordinate Z) from a center to a surface is described by a set of equations:
i, *) - RK« £ (TK (z, i) - Tfft (z, with boundary conditions:
VlQz(o^i)»O;
-Ae^
where q(r,z) is an energy release in kernel (q,(r,z)=O ±n the reflector block); 7K(ZJ2JJ
'6rz(Zj2-J~ are temperatures for a kernel and
graphite; £ - is a coordinate of a point over the fuel element radius; or the reflector block; - is a thermal resistance of microfuel element coatings.
\4"is a kernel volume. A - i s a graphite volume per one kernel e:pf-
is an effective heat conductive heat conductivity coeffici-
ent. For £
- is a graphite heat conductivity coef-
ficient. The heat transfer by radiation as well as residual heat in the given version for CTARTSprogram is not taken into account.
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Go
2. BEHAVIOR OP REACTOR DURING ACCIDENT INDUCED BY A WATER INGRESS IN THE GORE An accident of such a type can occur during ruptures of steam generator popes in one of reactor plant loops. A rate and a character of ingress as well as a total steam-water mixture mass depend on a wide variety of reasons the analysis of which is an independent problem. It was assumed in the present paper that: 1. The maximum water amount in the reactor core can come up to 300 kg. 2. The emergency protection system does not operate, 3. The coolant flow rate and its temperature at the reactor inlet do not change and equal, respectively, to - 340 kg/sec and 623°K. 4. Water concentration in every point of the reactor core changes in time according to the linear low,
jo (2, t) = miH-ffsoo > mattO; ^
{t - -jfcjjj
where: P-IQQ is a water concentration in helium, at which the total water mass in the reactor core equals to 300 kg. t.,QQ is a time during which the water concentration in helium at the reactor input amounts to j^OO* V..- is a mean mass helium flow rate in the reactor core. For t~>vv A n vaHi
lues equalling to 4 sec, 8 sec, 16 sec and 40 sec,
•
figures 2.1 and
2,2 show the plots illustrating changes in the reactor power and maximum temperature of microfuel element kernels. It is seen that in all cases the maximum overheat of microfuel elements after power flash amounts to -~ 25O°K. This is explained by the fact that the maximum value for excessive positive reactivity is the same in all cases, but the compensation of this effect is mainly accomplished due to the Doppler effect. Therefore, as far as the reactor initial state in all variants is the same and the excessive positive reactivity
value is the same, then the fuel elements overheat should also be the same. Here an emphasis should be placed upon two effects which cannot be taken into account in a standard point model for the reactor dynamics. The first one is that during the reactor speeding-up a main energy is released at the core upper part, which results in an overheating not all fuel elements but only a part of them (see Pig.2.3)• A neglect of this factor in the reactor point model can lead to a mardked error in determining a value of heat amlunt released at a "flash" of power. The second is the effect of slow increase of a reactor power. After the initial "flash", caused by redistribution
of heat between upper and lower fuel element la-
yers (see Pig.2.4)• The result of such a redistribution is the appearance of low positive reactivity. Prom fig. 2.4 is seen that by 6-th minute the field of temperatures in the reactor core was stabilized and its heithg profile repeats precisely the initial temperatures profile (tsO), The calculations show that the 300 kg water ingress into the reactor core at a not-operated neutron capacity automatic regulator and emergency protection system will result in the increase of: the reactor power by 1.5 times and the coolant outlet temperature by " 300°K. In this case, if such an amount of water enters quickly the reactor core (during several minutes), then the time of approaching to a new stationary state is determined by the time of reactor core thermal delay, which ad hoc comes up to
5-10 min.
If the water mass increase in the reactor core up to 300 kg proceeds for a long time (over 5-10 min), then the time of come out to a new stationary state will most likely be determined by the time of positive reactivity insertion (by the period for which the water
amount in the core increases up to 300 kg)#
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3. THE EFFJECT OF KERNEL SIZES AND MICROFUEL ELEMENT CLADDINGS THERMAL RESISTANCE ON THE PROCESS OF SPEEDING-UP INDUCED BY A WATER INGRESS Any process of a change in a nuclear reactor power is due to a change in a balance of rates of absorption and neutron generation, or, as it is customary to say, is dictated by an insertion of reactivity (positive or negative) into a reactor. In our case (the water ingress accident) the reactivity changes in time. Firstly, the water concentration increases, which gives a positive contribution in the system reactivity; secondly, the fuel (UOp) temperature grows and, consequently, an absorption of moderated neutrons on resonances U ^
(Doppler effect) increases. This process makes
a negative contribution to a total reactivity. The rate of positive reactivity insertion is stipulated by the rate of water concentration increase in the reactor core or its separate regions and, in its turn, is due to the sizes leaks in the steam generators, the mean-mass helium rate in the reactor core (flow rate), and by other reasons. The rate of negative reactivity insertion at the *) expense of the Soppier effect ', depends on the rate of fuel temperature increase, more precisely, the temperature of microfuel element kernels. The rate of increase in kernels temperature depends on the density and heat capacity of uranium dioxide that made up the kernel and on the rate of heat dissipation from kernels to the matrix graphite. The heat removal rate, in its turn, is determined by kernel sizes and thermal resistance of microfuel element 'As it was mentioned above, all calculated studies in this paper were carried out assuming that during the accident process the emergency protection system does not operate A-3
(oh
claddings. In this case, as it was previously shown in a number of peprs U *
2,3
, the absorption of moderated neutrons on resonances
depends markedly on the size of microfuel element kernel. And
this means that the fuel reactivity temperature coefficient (Doppler effect) and a reactivity "vapor" effect depend also on the microfuel element kernel size. In paper [ 3] it was shown that the lesser the kernel diameter is, the higher the vapor effect of reac?tivity. On the other hand, it is evident that the lesser the kernel diameter, the quicker the heat generated in the kernel, will be removed to graphite and, hence, the slower the uranium temperature will be increased and negative reactivity - will be inserted, under otherwise equal conditions. That is to say, the both effects (the "vapor" reactivity effect and the rate of heat removal from uranium kernel) when decreased the kernel diameter, lead to th increase of a total positive reactivity inserted during an accident, and as a consequence of this, to the increase in power splash. Pigs. 3.1 and 3*2 show how the reactor power and the fuel maximum temperature change when 300 kg of water enters the reactor core for 4 sec. It is seen that the change in reactor characteristics during an accident (during its first seconds) depends essen*) tially on the size of microfuel element kernels
. The main con-
tribution to the change of emergency process parameters was made by the effect of change in neutron-physical properties (the "vapor" reactivity effect) due to the change in kernel diameter. The ef*) In the previous section calculations were run for fuel elements with kernels of 500 mu in diameter. In general, the fuel element with parameters -* c/fCL =600; enrichment - x= 6.5%, the kernel diameter d,=500mu, the temperature conductivity coefficient for microfuel element claddings - R^M= 0.02, was accepted as a base variant. A-3
feet of accelerated heat removal from the kernel to the matrix graphite from small size microfuel elements changes slightly the emergency process characteristics (see Table 3*1) Table 3.1. MAIN CHARACTERISTICS OP AN EMERGENCY PROCESS (300 KG-WATER INGRESS FOR 4 SEC) FOR A REACTOR WITH MICROPUEL ELEMENTS OP VARIOUS DIAMETERS* Characteristics
Kernel diameter - d,., mu 250
500
Reactor maximum power during a "neutron blash", MW
10104,
Reactor power on 80-th second of emergency process, MW
1§29_ 1500
1496
Puel maximum temperature before to a start of accidents, °K
1242 1241
1245
Puel maximum temperature on 1015-th second
16046 1466
Puel maximum temperature on 90-th
1605
second
—
Energy release in reactor core for the 54600 first 15 sec. of accident, MJoule 39100
U 6 ?
750
1
?88 1493 1280 1255
1447 1482 1480
1497
37900
— 28100 35800
*) The lower figure in the first and third column was obtained on the assumption that neutron-physical constants do not depend on the kernel diameter. For these cases use was made of constants of the base variant.
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66"
For the same reason the parameters and character of emergency process do not practically depend on the value of effective heat transfer coefficient R ^ (in the recommended region of its variation - 0.01-0.03 W/K), Only in the case when its value will be less almost by an order of magnitude (R K M = 0.002 W / K ) ; the character and parameters of the emergency process change by a mared" value (see figs. 3.3
and 3.4).
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4. REACTOR SPEEDING-UP DURING ITS START-UP, BEING DUE TO SPONTANEOUS EXTRACTION OF EPS-SHIM RODS One of the most severe nuclear accidents relating to the hypothetical class, is an accidence induced by a spontaneous withdrawal of EPS shim rods from a pebble bed core. The most extreme even of such a hypotehtical accident when all EPS shim rods begin to be withdrawn spontaneously, is considered in the present section. Here we will discuss a situation when the rods are withdrawn from a cold (293°K) unpoisoned reactor (we assume that the initial neutron flux gives a heat capacity of 0.1 MW and the coolant
flow rate at the
reactor core inlet is nominal - 340 kg/sec), when the inserted posi~tive reactivity value is maximum possible for the given reactor. The calculations show that this value is the sum of the temperature ef-v feet of reactivity^ 5% A K / K and the effect of X e 1 3 5 stationary poisoning " yfo AK/Kf that all in all comes up to ^ 1 6 B
££
(here
r Q±f =* °»5% A K/K). According to estimations, to compensate such a positive reactivity the EPS shim rods should be deepened in the pebble bed at a depth of 70-100 cm (the calculation from the CTAPT 3 code stationary version gives this value s 74 cm). The question - with which rate can this reactivity be inserted - is significant for a final fomulation of the problem. However, in this paper this question was not studied in detail. The calculations were carried out for two variants: rods are withdrawn with a constant acceleration and rods are removed with a constant speed. The acceleration and speed values were present by parameters within the limits of 2 *^ 1000-10 cm/sec and 100-1 cm/sec; respectively ' *) It was shown below that the withdrawal of rods with acceleration 1000 cm/sec
corresponds to an "instant" extraction of positi-
ve reactivity /-v 16
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TABLE 4.1. MAIN CHARACTBRISTIGS OF THE EMERGENCY PROCESS INDUCED BY THE WITHDRAWAL OP EPS SHIM RODS WITH A CONSTANT ACCELERATION
Acceleration, era/sec
Characteristics
Reactor power at the pulse peak, GWt
1OOO
100
50
•25
10
156
180
144
105
67
13. Some characteristics of the emergency process with the withdrawal of rods with a constant acceleration, obtained from calculations, are given in Pigs. 4,1, 4,2 and in Table 4.1. Prom the figures is seen p
that for the acceleration values a = 1000 cm/sec
the development of
the speeding-up process with consequent self-quenching are similar in quality. At the beginning of speeding-up, at the due to a high rate of positive reactivity insertion the heat generation rate in kernels exceeds its removal rate to the matrix graphite. As a results, the fuel temperature (kernels) reaches very quickly (for tenth parts of a second) the values at which the negative temperature Doppler effect compensates the positive reactivity. But in this case the reactor power attains the value ^ 1000 000 MW (100 N nom).
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Since the delayed neutrons had no time to be accumulated yet in the reactor during this process, the drop of power at the beginning proceeds very rapidly. Therefore, the flash-up of power has the shape of very narrow (half-width
0.4-0.5 sec), high, symmetrical peak
(see Table 4»1»)» Then the kernels' temperature drops due to the heat removal to the matrix graphite. The rate of power decreases and it on the 14-15-th second attains the value ~ 2OO0MW. After this a slow growth of power begins which is caused by the appearance of low positive reactivity due to cooling-down the upper heated fuelelements (in which a main heat has been liberated) and heating-up the lower fuel elements. In this case the coordinate of fuel element location point with a maximum fuel temperature moves step by step downwards at its own value does not exceed 25OO°K. It should be borne in mind that the process of the rods withdrawal from the pebble bed with the acceleration ~" 1000 cm/sec
simulates in fact the
process of "instantaneous" reactivity extraction Q% A k/k. Indeed, the rods being at the 74 cm depth in the pebble bed, are withdrawn
2
ff
I HZ
from it completely for the time f a » "y 1 0 0 0 ~ ' the significant growth of power and temperature begin after 0.1 sec (see Pig.4.1). Let us now consider the reactor behaviour at an accident when SPS-shim rods are withdrawn from the reactor with a constant speed v = const. Pig. 4.3 illustrates the character of changes in the reactor power while withdrawal the rods with the speed 1 cm/sec (^0.1
|3 eff/sec). Tke power behavior is seen to be of an oscilla-
ting character. In this case, the fuel maximum temperature during A/ 50 sec increased monotonously approaching the plateau; during this entire process it does not exceed the value 1900°K. With the in crease in the speed of rods withdrawal (see Pig.4.4) the period between pulses in reduced and
they themselves are elongated becoming
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narrower and higher. Over the limit at very great speeds of rods withdrawal7^ 50-100 cm/sec all the pulses are mixed together into one, and the process becomes completely analogous to the process caused by the withdrawal of rods with a constant acceleration a = 10-100 cm/sec
(see above). The possibility of existence of
such a process (a power oscillation at the reactivity insertion according to a linear low) is pointed out in papers / 4 , 5 , 6 / . The reason owing to which the reactor comes to such a regime is the superposition of two processes of reactivity insertion: the positive one - through the linear low and the negative (because of the heat liberation and temperature growth) - according to the e xp one nti al one. A high degree of HTGR inherent safety is ensured not only by the negative temperature coefficient of reactivity, but also by the high heat capacity of the core. Thus, in the accident with the control rods rejected with a constant acceleration of 980 cm /s from the cold dispoisoned reactor the rate of growth of the outlet helium temperature is aboutHO K/s, though the maximum helium temperature in the core at the initial stage of the process grows much more quickly (see Pig.4,5)• This results from that the heat is released mainly in the upper part of the core, the core bottom remaining practically cold (T«*35OK). Therefore, helium flowing through the core from top to bottom and heated up by the upper hot fuel elements transmits the heat to the cold bottom fuel elements and the bottom reflector blocks which act as a heat accumulator. As a result, the rate of growth of the outlet temperature remains relatively low, ~" 10 K/s. In the case of the prompt core heating it is necessary to take into account the decreasing helium density which leads to the pressure growth and
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eventually to a decrease in the inlet flow rate and an increase in the outlet flow rate (see Fig.4«5)« Indeed, the temperature dependence of the mass of gas in the core is described sufficiently well by the ideal gas law: according to which the mass of gas m in the volume is inversely proportional to its temperature T. However, we have postulated the invariability of the gas volume V and pressure P (see the input equations) as well as the constancy of the inlet weight flow of gas. Therefore, in heating the excess gas mass causes the temporal increase of the outlet flow. CONCLUSIONS 1. The calculations carried out have shown the necessity of employment of the reactor model with distributed parameters to calcu: late dynamical processes in HTGR with spherical fuel elements. 2. In all considered in the present paper variants of accident processes the calculations have shown that owing to the high reactor core heat capacity, microfuel elements placed homogeneously in the fuel element graphit matrix and the negative temperature coefficient of fuel reactivity, the rate of heat release and removal processes in the simulated reactor fuel elements in such that the probability of fuel melting during the accident
caused by the
high positive reactivity fast insertion is lack. 3. In a number of emergency processes (a rapid insertion of positive reactivity /v 16 B Q^^ for 1-J-2 sec) the rate of growth of temperature gradients is rather high (on a micro fuel element 2000K for 0.2 sec, and aspherical fuel element
1000°K for 0.5-0.7 sec).This,
firstly requires the experimental designing studyings to determine a maximum possible reactivity and a rate of its insertion determined
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by design features of the specific reactor type;
and, secondly,
asks for the study of proceeding the mechanical and physical-chemical processes both in a single micro fuel element (an interaction of U0 2 with cladding materials) and in a spherucal fuel element as a whole. 4. While withdrawn the rods from a cool,
poisoned reactor with
the constant speed -VI-10 cm/sec the change of power has the oscillating character. The nature o£ oscillations are stipulated by the fact that the positive reactivity is inserted according to the linear low and the negative (the growth of power and temperature) according to the exponential one. 5. The results of calculated investigations on the emergency process induced by the water ingress into the reactor core have indicated that the rate of water ingress gives rise to the character of time behavior of the reactor power and fuel temperature. The reactor, when the same amount of water enters it, in a certain time (/v 10-15 min) comes out to a new stationary state. The increase of the reactor power
W and fuel temperature at the new state depends
only on the quantity of water entering the reactor core. For 300 kg water A W = 50% lSLrtni, and A T 1 " 8 1 ~ nom fuel
250-300°K (A fuel overheating).
An error in determining the vapor effect value does not change qualitatively the process character. Only A W andaT1118'35" are varied. fuel Thus, for example, the ~2 times increase of"the vapor" effect will result in the ^ 2-times growth of T^el* 6. The influence of micro fuel elements on characteristics of the emergency process induced by the water ingress in the reactor core has been investigated. It was found that this effect is mainly stipulated by the increase in the reactivity "vapor" effect at the decrease in the kernel diameter. The process of accelerated heat removal from small diameter kernels affects insignificantly the
92-
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emergency process characteristics* The change of temperature conductivity coefficient value R ^
* 0.02 WK
by 50% does
practically not affect the character of emergency process proceeding. REFERENCES 1.
Gol'tsev A.0« Abstract for the DrcA<\UN program.: Voprosy atomnoi nauki i tekhniki. Ser. Phisika i tekhnika yadernykh reaktorov 1985, vyp. 5, p« 71•
2.
Gol'tsev A.O. The selection of optimum fuel cycle for HTGR operating on OTTO principle with uranium oxide-base fuel, Voprosy atomnoi nauki i tekhniki. Ser.: Atomno-vodorodnaya energetika, 1978, vyp. 1(4), p. 95.
3.
Evseev V.P., Karpov V.A., SobilevA.M., Sukharev Yu.P., Udyanskii Yu#N. The feasibility study for selection of physical
characteristics of the VG-400 reactor. Voprosy atomnoi
nauki i tekhniki. Ser.: Atomno-vodorodnaya energetika i tekkhnologiya, 1982, vyp. 1(11), p.38. 4.
Reactor Handbook. Second edition revised and anlarged. Vol. Ill, part A, Physics. Edited by H.Soodak. Interscience publish ers. A division of John Wiley and Sons. JTew-York-London, 1962.
5«
Kramerov A.Ya., Shevelev Ya.V. Engineering calculations for nuclear reactors. M., Atomizdat, 1964*
6.
Khetrik D. The dynamics of nuclear reactors M., Atomizdat,
1975.
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W.GWt 5 -
The change in reactor power when 300 kg water enters the reactor
8 sec,
4 -
40 sec,,
3 -
2 ~
1 -
10
20
30
40
'$• 50
60
70
80
Pig. 2.1. max The change in fuel maximum temperature when 300 kg water enters the reactor. Designations are the same as in Pig.2.1. 1500-
uoa 1300
1200-
1100
Pig. 2.2.
sec
100 0
10
20
"eV
70 A-3
80
iThe fuel temperature distribution over the reactor core fuel height at various moments of the accident process (300 kg 1500 —| ^-o^^ water ingress for 4 sec.)
125OH 10
'sec
ZOtSec
1000_|
750
500 Z,m
4.8 Pig. 2.3.
The fuel temperature distribution over the reactor core height at various moments of the accident process (300 kg water ingress for 8 sec) max fuel
pL
600 sec
1500"
1250.J
sec 1000 _J
750
^
Pig. 2.4. 500
T
4.'8
Z,m A-3
Change in the reactor power when 300 kg water enters for 4 sec; for fuel elements with kernels of the diameter:
GWt
5 H
250^
500
V ,sec
10
li K
i
i
20
30
i
40
50
60
70
O'hange in the fuel maximum temperature in the reactor at the 300 kg water ingress for 4 sec.
1600
1500-
A-3
80
GWt Change in the reactor power at the 300 kg water ingress for 4 >sec for fuel elements ; with various temperature conductivity coefficients of micro fuel element claddings (/K^
4-
0.02 0.002
0.03 o
3_
2_
1_
, .sec I 10
20
30
40
50
60
I 70
i
80
K
1
1600H
Fig.3.3. Change in the fuel maximum temperature at the 300 kg water ingress for 4 sec. Designations are the same as in Fig.3.3.
1500-
I400-
I300-
200-
1100
60
0
80
Fig.3.4. A-3
Change in the fuel maximum temperature during the accident process induced by withdrawal of EPS-shim rods with a constant acceleration - a
max
a=1000cm /sec* 3000 -
2500 H
2000
1500
1000-
500"
»sec Pig. 4.1
- A-3
T km
Change in maximum temperature drops between the kernel and matrix T j ^ and between the fuel element center and its surface :J TC£J — o - o — during the accident process induced by EPS-shim rods with drawal with a constant acceleration - a, CM/SCCZ
3
V. sec
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GWt max Teff
2000"
1500_
1000-
50Q_
Qj
0 0
Change in time of the reactor power and fuel maximum temperature at thft withdrawal of J3PS-shim rods with speed 1 cm/sec from the cool, poisoned reactor*
Change in time of the reactor power at the withdrawal of EPS-shim rods r.ith 'tK'-J .•"•••--a6d
With the speed 6 cm/seo
-20
5 cm/sec
4 cm/.aeo
.15
10
7? A-3
u
Change in the outlet helium flow and temperature,
out kg/s L
He,K
G
+ and T
. and maximum helium temperature in the
core, T m ^ x . The inlet helium temperature and flow are constant and equal to 293K and 340 kg/s, respectively. During the accident the control rods are withdrawn at a velocity of V = 980-ZT, cm/s. max THe
.
.— •—
T out
Gout
i
i
15
20
25
30
Z ,s Pig.4.5.
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XA0101481
GA-A19674
MHTGR RADIONUCLIDE SOURCE TERMS FOR USE IN SITING by S.B. INAMATI, A.J. NEYLAN, and F.A. SILAOY AND L . L . WALKER (SWEC)
This is a preprint of a paper to be presented at the IAEA Technical Committee Meeting on "Gas-Cooled Reactor Technology, Safety and Siting" June 21-23, 1989, Dimitrovgrad, USSR and to be published in the Proceedings.
Work supported by Department of Energy Contract DE-AC03-89SF17885
GENERAL ATOMICS PROJECT 7600 JUNE 1989
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DISCLAIMER This report was prepared as an account of work sponsored by the United States Government. Neither the United States nor the United States Department of Energy, 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, mark, 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.
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MHTGR RADIONUCLIDE SOURCE TERMS FOR USE IN SITING ABSTRACT This paper presents the site suitability source terms for the Modular High Temperature Gas-Cooled Reactor (MHTGR). The MHTGR program has identified a spectrum of accident release scenarios that form a basis for determining the MHTGR site suitability source terms. This basis, consistent with 10CFR100 guidelines, includes an evaluation of a spectrum of accidents to determine a set of conservative source terms. The spectrum of accidents is represented by a set of Licensing Basis Events (LBEs). The subset of LBEs that involve radionuclide release from the plant is identified. The MHTGR approach evaluates site suitability with respect to the regulatory criteria applicable not only for off-normal events, but also for normal operation and emergency planning. The radionuclide inventories available for potential release from the MHTGR and the release behavior of these inventories are characterized. A summary of the source terms for selected key radionuclides are given. The time-dependent MHTGR source terms account for a mix of accident phenomena, failure states of barriers, number of reactor modules, and chemical attack conditions that are consistent with the unique characteristics of the MHTGR.
SELECTION OF RADIONUCLIDE RELEASE SCENARIOS The criteria to evaluate suitability of proposed sites for stationary nuclear power reactors under various radionuclide release scenarios are described in 10CFR100 (Ref. 1). To facilitate the evaluation process, 10CFR100 (Ref. 1) provides generic guidance to develop source terms by considering a spectrum of postulated radionuclide release scenarios that would result in potential hazards "not exceeded by those from any accident considered credible." Thus, the identification of credible, postulated radionuclide release scenarios for the MHTGR (Ref. 2) is the first step in the evaluation of siting suitability.
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A spectrum of possible radionuclide release scenarios was identified for the MHTGR program, consisting of a set of LBEs, to form a basis for determining the site suitability source terms. The selected LBEs are consistent with 10CFR100 guidelines and yield a set of conservative source terms. These MHTGR LBEs, shown in Table 1, are further characterized by the following: 1.
They are logically chosen, following a structured method to consider a wide spectrum of possible events.
Events were selected by a safety risk
assessment [Level 3 per Nuclear Regulatory Commission (NRC) NUREG/ CR 2300; Ref. 5]. 2.
They are bounding, by considering all conceivably credible accidents including those not expected to occur in the lifetime of several hundred MHTGR plants (down to a mean frequency of 5 x 10-7/plant year).
3.
They are comprehensive, since they account for a mix of accident scenarios yielding varying release rates, release mixes, failure states of barriers, and release from multiple reactor modules.
4.
The radionuclide release estimates are conservative, since they include uncertainties combined satistically in calculations.
5.
The radionuclide release estimates include enhanced safety margins, consistent with NRC's Advanced Reactor Policy (Ref. 6).
Table 1 shows specific release scenarios that contribute to radionuclide release from the plant, since not all scenarios result in release of radionuclides from the plant. The mean frequency and the applicable criteria against which compliance is demonstrated are identified in Table 1. The MHTGR approach for site suitability evaluation includes compliance with the regulatory criteria that are applicable not only for offnormal events, but also for normal operation and emergency planning.
Approach to Radionuclide Control The MHTGR configuration was developed to ensure the integrity of the standard (no defect and within specification) fuel particles such that the radionuclide inventory is retained within the fuel particles under all credibly conceivable events. Thus, the only available significant source for potential radionuclide release is outside the standard particles.
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TABLE 1 MHTGR RELEASE SCENARIOS SELECTED FOR SITE SUITABILITY SOURCE TERMS Release Scenarios
Frequency/ Plant Year
Applicable Criteria
Small releases associated with RCCS air and other service system sources
Normal plant operation
10CFR50 Appendix I (Ref. 3)
Small primary coolant leak with forced core cooling (AOO-5)
0.3
10CFR50 Appendix I (Ref. 3)
Moisture inleakage without forced core cooling (DBE-7)
5x10- 5
10CFR100(Ref. 1)
Primary coolant leak with forced core cooling (DBE-10)
0.01
10CFR100(Ref. 1)
Primary coolant leak without forced core cooling (DBE-11)
3x10'4
10CFR100(Ref. 1)
Moisture inleakage with delayed steam generator isolation and without forced core cooling (EPBE-1)
3x10" 7
Lower PAG (sheltering; Ref. 4)
Moisture inleakage with delayed steam generator isolation and with forced core cooling (EPBE-2)
4x10" 6
Lower PAG (sheltering; Ref. 4)
Primary coolant leak in four modules without forced core cooling (EPBE-3)
7x10~ 7
Lower PAG (sheltering; Ref. 4)
Legend AOO = Anticipated operational occurrences DBE = Design basis events EPBE = Emergency planning basis events
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Table 2 illustrates, for example, sources of 1-131 (which is the dominant contributor to thyroid dose) available for release in one module and the relative timing characteristics of associated release mechanisms.
As shown, the smallest sources, the
circulating and plateout activities within the primary circuit, have the potential for the fastest release to the reactor building. Because this release is linked to the accidental leakage of the gaseous helium coolant from the vessel system, this release could characteristically occur within minutes. The remaining sources of 1-131 radionuclide are within the core graphite, but outside of standard, intact particles, and take longer to be released. Since the mechanisms for release from these defective fuel particles depend on core temperature, which increases very slowly due to the large heat capacity of the massive graphite moderator and low power density of the core, these releases characteristically occur over hours to many days.
However, it is possible for a small fraction of the inventory from fuel
particles with as-manufactured defective coatings to be released rapidly. The potential for such a rapid release (minutes to hours) from these defective particles is postulated with events in which high core temperatures occur coincident with a large moisture ingress providing reactants that can hydrolyze the carbide portion (approximately 7%) of the UCO fuel. In summary, the potential activity releases, as shown in Table 2, can be grouped into two broad categories, a small early release and a larger delayed release.
Normal Operation Releases Insignificantly small quantities of radionuclide effluents to the environment in the form of gases and liquids occur during the normal MHTGR plant operation. These releases result from the radioactive gas waste system and liquid waste system, respectively. Table 3 provides a summary of expected annual gaseous and liquid release from a four module MHTGR plant. The concomitant doses for gaseous effluent releases are well below 10CFR50 Appendix I limits (Ref. 3), with margins of an order of magnitude or higher. In comparison, the expected radioactive liquid effluent releases have margins of five to eight orders of magnitude against the maximum concentration limits of 10CFR20 (Ref. 7).
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TABLE 2 1-131 (a> INVENTORY AVAILABLE FROM A SINGLE MODULE MHTGR FOR POTENTIAL RELEASE Release Mechanisms Source Characterization
Inventory (Ci)
Timing of Release
From Core
From Primary Circuit
A. Circulating
0.02
Minutes
—
Leakage flow (He depressurization)
B. Plateout
20.0
Minutes
—
Leakage flow (He depressurization) and moisture (water ingress)
C. Outside standard particles 1. Nonintact [failed SiC
]
Hours(c)-days Temperature (loss of forced Leakage flow (He depressurization) cooling) and moisture (water ingress)
2. Contamination
93.0
Hours-days
Temperature (loss of forced Leakage flow (He depressurization) cooling)
9.3 X10 6
> days
Temperature (no event identified)
D. Standard particles
——
: (^Contributes to thyroid dose. b ( )Silicon carbide (SiC); outer pyrolytic carbon (OPyC). (^Approximately 33 curies of the inventory (representing the UC 2 in nonintact fuel particles) is subject to release within minutes under hydrolyzing conditions that may be encountered in rare MHTGR accidents.
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TABLE 3 RADIONUCLIDE EFFLUENTS FROM A FOUR MODULE MHTGR DURING NORMAL PLANT OPERATION Nuclide
MHTGR Releases (Ci/Year)
A. Expected Annual Gaseous Release H-3 Kr-85 Xe-133 Ar-41
10 40 10 20
B. Expected Annual Liquid Release 1-131 Cs-137 Ba-140
4.9 x10" 6 2.2 x10" 4 3.1 x10" 7
Off-Normal Event Releases A mechanistic assessment of the off-normal events was performed and a summary of the equilibrium source terms for a few key radionuclides are given in Table 4 for release scenarios that could result in potential radionuclide release from the plant. The source terms shown in Table 4 are for a single, specific, and independent event and therefore the source terms are not additive. Further, the MHTGR plant is located primarily below grade with no stack and hence any potential radionuclide releases are contained at or below ground level. The equilibrium source terms shown in Table 4 are insignificantly small. Thus, the resulting offsite does are negligibly small and are well within the Protective Action Guides (Ref. 4) and hence do not require drills for offsite public sheltering or evacuation. The time dependence of the release of the radionuclide 1-131 from the core due to a loss of core cooling and a loss of coolant in all four modules for one specific event (EPBE-3) is shown in Fig. 1 for illustration. As seen, the release occurs slowly spanning several days. The release from the reactor vessel to the reactor building is limited due to radioactive decay and the lack of a driving force from the reactor vessel. The slow release characteristics allow further retention in the reactor building by naturally occurring physical phenomena such as plateout and settling. Since there is no driving force to transport radionuclides out of the reactor building, the radionuclide release to the environment is further limited. In summary, the cumulative source terms developed for the MHTGR plant siting suitability include a mix of release scenarios, failure states of barriers, and multiple reactor modules. A-4
TABLE 4 MHTGR EQUILIBRIUM SOURCE TERMS OF KEY NUCLIDES FOR LICENSING BASIS EVENTS Cumulative Fractional Release to the Environment^) LBE
No. of Modules
Release Duration (h)
Kr-88
Sr-90
1-131
Cs-137
AOO-5
1
0 to 1.25
1.4x10-7
1.1 X10"11
3.1 x10" 10
1.5 x10" 10
DBE-7
1
0 to 0.02
7.5 x10- 9
5.4x10~y
3.1 x10" 8
9.2x10-7
DBE-10
1
0to1
2.2x10-7
1.1 x10" 10
8.0 X10"10 3.4 x10" 9
DBE-11
1
0to8 0 to 100
2.4 x10" 8 3.3 x10" 8
2.2X10- 1 * 9.9 x10" 12
4.5 x10" 9 2.8x10-7
2.6 xiO" 1 1 4.4 X10"11
EPBE-1
1
0to8 0 to 100
9.7x10-7 9.9X10-7
4x10" 8 4x 10"8
3.6x10- 7 4.9x10-7
3.6x10' b ' 3.8 x10" 6
EPBE-2
1
0to8
2.6x10-7
1.1 x10"7
3.6x10- 7
1.9 x10" 5
EPBE-3
4
0to8 0 to 100
8.2 x10" 8 1.6X10-7
8.2 x10" 12 3.8 xiO" 1 1
1.5 x10" 8 9.5x10-7
1.0 x 10-™ 1.7x10"10
^'Fractions of initial radionuclide inventory (single module) of 9.9 x ; 10 B Ci of Kr-88, 7.4 x 105 Ci of Sr-90, 9.3 x 106 Ci of 1-131, and 8.6 x 105 Ci of Cs-137.
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TIME PAST INITIATING EVENT (DAYS) 4
6
TO REACTOR BUILDING
cc
«x
20
40
60
.——I——+——^—~^— 80 100 120 140 160
200
TIME PAST INITIATING EVENT (HOURS)
Fig. 1. MHTGR cumulative retention of 1-131 during EPBE-3
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CONCLUSION The inherent characteristics and passive safety features of the MHTGR provide a solid basis for developing a mechanistic basis for siting. The MHTGR site suitability source terms can be characterized and evaluated by a representative set of potential release scenarios and resultant radionuclide releases to the environment. An evaluation shows that the source terms developed for the siting suitability of the MHTGR are benign, requiring no sheltering or evacuation of the public beyond the site boundary.
ACKNOWLEDGMENT The authors would like to thank the U.S. DOE for approval to publish this work, which was supported by the San Francisco Operations Office, Contract DE-AC03-89SF17885.
REFERENCES 1.
"Reactor Siting Criteria," Title 10, U.S. Code of Federal Regulations, Part 100 (10CFR100), Revised January 1, 1985.
2.
"The Design Status of the United States Department of Energy Modular High Temperature Gas-Cooled Reactor," Mills, R. R., June 1989, Modular HTGR Plant Design Control Office, San Diego, California 92138.
3.
"Domestic Licensing of Production and Utilization Facilities," Title 10, U.S. Code of Federal Regulations, Part 50 (10CFR50), Revised January 1988.
4.
"Manual of Protective Action Guides and Protective Actions for Nuclear Incidents," Environmental Protection Agency Report EPA-520/1 -75-001, September 1975 (Revised June 1980).
5.
"PRA Procedures Guide," U.S. Nuclear Regulatory Commission Report NUREG/CR-2300, January 1983.
6.
U.S. Nuclear Regulatory Commission, "Policy for the Regulation of Advanced Nuclear Power Plants," Federal Register, Vol. 51, July 8, 1986.
7.
"Standard for Protection Against Radiation," Title 10, U.S. Code of Federal Regulations, Part 20 (10CFR20), Revised January 1988.
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XA0101482 TRANSFER OP FISSILE MATERIAL THROUGH SHIELDING COATINGS IN EMERGENCY HEATING OF HTGR COATED PARTICLES AoN.Gudkov, S.G.Zhuravkov, M.A.Koptev, A.D.Karepin
ABSTRACT The measurement results of leakage dynamics of fissile material from the coated particles within a temperature range of I2OO + 20CO°C are given. The methods of carrying out the experiments are briefly described. The relation of the leakage rate of uraniun-235 from CP with the pyrocarbonic coatings has been obtained.
The measurement of the fissile material (FM) leakage from the HTGR CP at temperatures of exceeding- IOOC 0 is interesting both for predicting the aftereffects of the fuel emergency overheating and the evaluation of the matrix graphite contamination in the production of ball CPEL. The results of the investigations (1,2)
show that at temperatures of exceeding I500°C a conside-
rable transfer of uranium is observed in the carbonic CP coatings. At the same time no data on the uranium yield from CP have been published. The aim of the present work is to investigate the
FM
leakage from BiSo CP within a temperature range of
I2CC + 2CCC°C. A-5
- 2 -
The experiments have beeen carried out at the plant including the vacuum chamber with a device for heating CP and sampling the fissile material. The FH settling from tr>e molecular flow leaving the sample takes place on the cold changeable collectors during the pre-determined time of exposure. The uranium amount on the collectors is measured by the radiographic method of the fission fragment tracks after which the PM number is calculated which has left the CP for the time of exposure. The sample temperature is measured by a pyrometer. The pressure during the experiment does not exceed IC"" torr. Pig.I gives the graphs of the relations of the uranium-235 leakage rates and time at a temperature of 1900°C from the micro CP of Type BiSC from HTI and LTI layers of dense pyrocarbons. The characteristic peculiarities of the presented results: - fast ( <1 10 minutes)reach of the stationary leakage rate of the both CP types; - a considerable difference in the equilibrium rates of , leakage for the two CP types. Pig.2 gives the graphs of the relations of the FM leakage from CP and annealing time. It also shows the time relations of the uranium-235 in the coating and core of the both CP types. A rather surprising fact can be considered that tlie PM amount having come out of CP with the HTI coating iu 3C minutes exceeds one order
the PH content in the coating and bv I0fe©ursof annealing
makes up already procents of the initial PM amount in CP. The FM leakage 10
from CP with the LTI pyrocarbon by 10 , hours makes up
of the initial content in CP that is approximately equal to
the FM content in the coating to the same time. A-5
O\5
- 3 -
For cheching up a direct measurement
of ?M loss from CP for
the annealing time taking into consideration a change in the intensiti of gamma radiation with an energy of 185.7 keV has been carried out. The results of the gamma spectrometric measurements match the results of measurements of the F
flow from the sample.
Fig.3 gives the results of the investigation (3)
in the
simultaneous measurement of the cesium flow through the graphite memebrane and Cs amount accumulated by the membrane. The construction of the curves in Figure 3 qualitatively coincides with the analogous curves for uranium in Figures I and 2 , namely: - quick establishment of the stationary flow in the membrane outlet for Cs and outlet of thePyC coating for uranium; - continuing of the Cs accumulation in the memebrane and uranium in the coating after establishment of the statio~ nary flow in the outlet. It can be supposed that the CS transfer mechanism as well as uranium in the carbonic materials is just the same. In this case, on the one hand, the uranium transfer can be described within the model limits suggested in (4) for describing the results of the investigation (3), on the other hand, the investigation of the uranium transfer regularities can be considered as the modelling of the cesium
transfer in the CP coatings. In favour of the
assumption of the similar mechanism of the cesium and uranium transfer in the CP coatings it is possible to also give the other experimental results: - characteristic peculiarities of the Cs and U propagation in BiSC particles are equal, Fig.4; - for Cs and U a comparatively high mobility is characterisA-5
- 4-
tic in the pyrocarbonic layers of the coatings and for Cs and U layer SiC serves as the main barrier. Fig.b presents the time relations of the uranium-235 from CP with the dense PyC coating of type HTI in the temperature range of I20C * I9OC°C. Fig.6 shows the relation of the meam leakage rate from the same samples and the reverse temperature which is well described by Equation
Ru = RQexp (-Q/RT) with an activation
energy of Q = 76-6 kCal/mol. By the the results of the work it is possible to formulate the following conclusions; - there exists a considerable leakage of the fissile material from BiSC CP at a temperature I20C°C and above. At a temperature of I9CO°C BiSC CP with the HTI coating loses for several hours some percents of the initial amoumt of the fissile material;- the characteristic peculiarities of the leakage dynamics and accumulation in diffusion of uranium and cesium atoms through the CP coatings rathsr resemble which permits to suppose
the similar transfer mechanism of these nuclides
and a possibility of using the results of investigation of the uranium behavour for describing the cesium migration; - mean rate of the uranium-235 leakage from BiSC CP can be described by relation Ru = R exp(-Q/RT) with an activation energy of Q - 76-6 kCal/mol in the temperature range of I2CC + I9CO°C. A-5
R E F E R E N C E .
1. R.W.Daton, CT.N.Cxley, C.W.Townley. Ceramic Coated Particle Fuels. Journ.of Nucl.Mater., 1964, V.II, No.I, pp.1-34. 2. AoN.Gudkov, A.A.Kozap, A.D.Kurepin. Peculiarities of Profile Formation of Fissile Material Concentration in HTGR Coated Particles. Voprocy atomtoy nauki n tekhnilci. ser.Atomto-vodorodnaya energetika i tekhnologiya. Vyp.I, 1338, str.70-71.(Russian) 3. A.B.Riedinger, C.B.Hilstead, L.R.Zuiawalt. Proc.Fifth Carbon Conf. 1963, V.II, p.405. 4. D.Chandra, H.H.Norman. Diffusion of Cesium through Graphite. Journ.of Nucl.Mater., 1976, V.2, No.2,pp.293-310. p
5. Zollrr P. Das Transortverhalten der Spaltproducte Cesium und Strcsncium in beschichteten Brennstoffteilchen fur Hochtemperaturreaktoren unter Bestrahiungen. Jul-1324, Julich, 1976,
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- 6 -
CAPTIONS TC FIGURES
Fig.I. Uranium leakage rate from BiSC CP, temperature I9OC°C: I - LTI fyC, 2 - HTI PyC. Fig.2. Uranium leakage dynamics from BiSC GP and uranium content in PyC, temperature I9CO°C: I - LTI, leakage from CP; HTI, leakage from CP;
2 - LTI, concent inPyC;
4 - HTI, content in RjC;
3 -.
5 - LTI,
HTI - content in kernel Fig.3. Cs leakage"and accumulation dynamics in its diffusion through graphite memebrane, (3): I - leakage rate;
2 - leakage;
3 - content in memebrane.
Fig.4. Uranium and Cs concentration profiles in BISC CP: I - uranium;
2 - Cs.(5)
Fig.5. Uranium-235 leakage from BiSO CP with dense PyC layer of HTI type. Fig.6« Relation of uranium-235 mean leakage rate and reverse temperature.
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XA0101483
Analytical Method and Result of Off-site Exposure during Normal
Operation of High Temperature Engineering Test Reactor (HTTR)
Kazuhiro SAWA, Hisashi MIKAMI and Sinzo SAITO
Department of HTTR Project Japan Atomic Energy Research Institute
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Aos
ABSTRACT
This paper describes analytical method and result of off-site radiation exposure
during
normal operation of the High Temperature
Engineering Test Reactor (HTTR).
Even in normal operation, some amounts of fission products are assumed to be released from the fuel particles with defect of coating layers. In the evaluation of off-site radiation exposure, 1% of coated fuel particles is conservatively assumed to be failed in the HTTR. The release paths of fission products from the core to the stack are divided into two patterns, namely continuous release and periodical release. Continuous releases of fission products are leaked from the reactor containment vessel which contains fission products leaked from the primary coolant system. On the other hand, periodical release of fission products are brought by the purge of radioactive gas
from the
decay tank which accumulates the gaseous wastes from the primary helium purification
system
and the primary coolant sampling system and by the
fuel handling system. Off-gas from the sweep gas sampling system for the fuel failure test is also considered.
The annual off-site radiation exposure from the HTTR is 0.77 jiSv/y in effective dose equivalent. This value is sufficiently lower than the reference dose of 50 jjSv/y in normal operation.
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/106
CONTENTS 1. Introduction 2. Analytical Method 2.1 Release paths and patterns of fission products 2.2 Calculation method 2.2.1 Amount of fission products released from the stack 2.2.2 Off-site radiation exposure 2.3 Sources of fission products 3. Effective dose equivalent in the off-site
4. Conclusion
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1.
Introduction
The Japan Atomic Energy Research Institute (JAERI) is going to build the High Temperature Engineering Test Reactor (HTTR) at Oarai Research Establishment of JAERI^^.
The Oarai Research Establishment site is
1,500 m wide and 1,700 m long, and about 5 km far from the center of Oarai town of about 20,000 inhabitants. The nearest site boundary from the HTTR facility is about 280 m. The off-site radiation exposure during normal operation of a nuclear facility is one of the essential matters in the licensing for its installation. The dose limit for the public is 1 mSv/y including the exposure from natural radiation. The Japanese Nuclear Safety Commission set up separately the reference dose of 50 jiSv/y for normal operation of nuclear power plants which are installed
on the same site. This
reference dose has been applied in practice also for research reactors.
The HTTR is a high temperature helium gas cooled reactor using coated particle fuels. The number of coated fuel particles in the core amounts to one billion. Therefore, it is almost impossible to fabricate all of the fuel particles without any defects. On the other hand, some amount of helium gas contained in the primary cooling circuit leaks inevitably to the atmosphere through the reactor containment vessel or the reactor building during normal operation, and primary helium gas possibly contains fission products. The gaseous fission products might be also released to the atmosphere from the decay tank and the fuel handling system periodically or irregularly. These radioactive sources and leak paths for normal operation of the HTTR are evaluated and the off-site radiation exposure is assessed.
2, 2.1
Analytical Method Release paths and patterns of fission products
The release paths of fission products from the core to the stack which should be considered in analysis are divided'into seven paths as shown in Fig. 1. i
A os
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(1) Gaseous fission products together with helium gas leak from the primary coolant system to the reactor containment vessel and are transported to the stack through the ventilation system when the pressure in the containment is regulated during normal operation or the air in the containment is replaced with fresh one during the period of reactor shut-down.
(2) The coolant purification system in the primary coolant circuit is outside the containment vessel. Therefore,
gaseous fission products
together with helium gas leak from the primary coolant purification system to the reactor building and are transported to the stack, although iodines are removed by the charcoal trap which is located upper stream and inside the containment vessel.
(3) Gaseous fission products leak from the primary coolant sampling system to the reactor building and are transported to the stack.
(4) Gaseous fission products contained in the primary coolant sampling gas are once stored in the decay tank and purged from it to the stack.
(5) Off-gas is generated by reclamation of the cold charcoal trap and molecular sieve trap which removes noble gases and tritium from the primary coolant,
respectively.
Gaseous fission products and tritium
contained in off-gas are purged from the decay tank to the stack.
(6) Gaseous fission products contained in off-gas from the fuel handling system are purged to the stack through filtering system during refueling.
(7) Fuel compact irradiation experiment with release of fission products is planned. In the case, the released gaseous fission products are purged from the sweep gas sampling system to the stack through the charcoal trap.
The release types of radioactive materials from the stack are summarized corresponding to the release paths as follows.
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@
Continuous release by the regulation of the containment internal pressure during normal operation as described in (1).
(g) Periodical or intermittent release when the air in the containment is replaced with the fresh one during reactor shut-down as described in (1). (§) Continuous release by the ventilation of the reactor building as described in (2) and (3).
^y Periodical
release from
the gaseous waste treatment system as
described in (4), (5) and (6). (§) Periodical release from the sweep gas sampling system as described in (7).
2.2
Calculation method
2.2.1 Amount of fission products released from the stack
In order to calculate the amount of fission products released from the stack by the release types described above, following equations and parameters are employed.
^ ) Fission products released by the regulation of the vessel internal pressure during normal operation, Q J Q (Bq/y). V
O* 1 Cc«
— s
yV - -LCV*A.-(l-g.)-(t,i . i l cv
where,
t
(1
~\ A .* ° )•
A*i-tl
(1)
Vp:volume of the air released by the regulation of vessel internal pressure per year,
including compressed air
used in vessel, 5,600 (Nm / y ) , V(-jy:volume of reactor containment vessel, 2,800 (Nm ),, L^y:leakage rate of helium from primary coolant system to containment vessel, 2.3x10
(s ) ,
A^iactivity of nuclide i in primary coolant, (Bq),
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tf-rremoval efficiency of ventilation system, 0.9 for I, tjtstorage time in containment vessel, 3.8x10° (s), A.;:decay constant of nuclide i, (s~ ) .
@
Fission products released by the replacement of the air in the
containment vessel during period of reactor shut-down, Qjy (Bq/y).
Q
iv= YLcv-V('-V- ' ~ V '
where,
ng:number of reactor shut-down, 5 (y ) .
(§) Fission products leak from the primary coolant purification system, 0-2 (Bq/y)an d the primary coolant sampling system, Q3 (Bq/y), and are released by ventilation of the reactor building. Q
where,
2=
L
p'Ai-(1-3?Ci)*t2
Lp:leakage rate of helium from primary coolant purification system, 7.0xl0~9 (s"1) V Q^'removal efficiency of charcoal trap in primary coolant system, 0.99 for I, t2=reactor operation time per year, 1.9x10
(s/y),
Lg:leakage rate of helium from primary coolant sampling system, 2.3xlO~9 (s" 1 ), Vpi:removal factor by plate-out, 0.1 for I. (2) Fission products are contained in off-gas from the primary coolant sampling system, Q^ (Bq/y), the cold charcoal trap and the molecular sieve trap, Q^ (Bq/y) and the fuel handling system, Qg (Bq/y), and are released by purge from the gaseous waste treatment system.
Q
4=
-T- ' A j . ^ . e - V ^
(5)
A-6
Y
where,
I-*
Vg:volume of sampling gas, 200 (Nm / y ) , V:volume of primary coolant gas, 3,000 (Nm ) , to .'storage time in decay tank, 2.6x10 (s), n^:number of gas purge per year, 6 (y ) , W^rflow rate of cold charcoal trap, 1.4 xlO
(kg/s) for
noble gases and of molecular sieve trap,5.6x10
(kg/s)
for tritium, W:inventory of primary coolant, 670 (kg), t^:storage time in a trap, 5.2x10 (s), ^{Ij!fraction of gaseous tritium to reclaimed tritium, 0.1 for tritium, t 5 :cooling time in a trap, 4.3x10° ( s ) , Vprvolume of
off-gas
from
fuel
handling
system,
340
3
(Nm /y), t^:cooling time from reactor shut-down until beginning of refueling, 8.64xlO5 ( s ) . (§) Fission
products released
by the fuel
compacts
irradiation
experiment, Qy (Bq/y).
Q
where,
7
=
n^rnumber of experiments per year, 12 (y
•)»
Ag^ractivity of nuclide i released by the experiment, (Bq), 7} g^:removal efficiency of charcoal trap in the sweep gas sampling system, 0.9 for I, t^:storage time in charcoal trap, 5.5x10 (s).
A-6
2.2.2
Off-site radiation exposure The effective dose equivalent, H (jiSv/y), is calculated as the sum
of the dose equivalents by external gamma-ray exposure by noble gas, internal thyroid exposure by iodine and internal body exposure by tritium. H = H y +Hj + HT where, H :dose equivalent by external gamma-ray exposure by noble f gases, (uSv/y), H.:dose equivalent by internal thyroid exposure by iodines, QiSv/y), H-:dose equivalent by internal body exposure by tritium, (^Sv/y). In the HTTR, the fission products are released from the stack of 80 m in height to the atmosphere. The released fission products are transported by wind and diffuse. The relative concentration of fission products in the air, X /Q (h/m^), ±s calculated by following diffusion equation. 2
1 (X/Q) =
• exp (2^-3600-cr -a -U y z
-*-r ) 2cr y
2 X Cexp { -
where,
( Z
"^
2 > + exp { -
(z+
.h)
} ]
(10)
Qrrelease rate of fission products from the stack, (Bq/h), Uraverage wind speed at representative height of emission, (m/s), h:height of emission, (m), n — n. T O *
n
v. 1 1 )
hg:height of the stack,
v:the upward speed of the plume,
DQ:diameter of the stack, 0* „ , 0" „: standard deviation of the cloud width in horizontal ydirection and vertical z-direction, (m).
6
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The concentration of fission products in the given direction is calculated superposing the contribution of the concentration of its neighboring directions. The sum of the inverse of wind speed for each wind direction and for each atmospheric stability is used to calculate the concentration of fission products. In the case of discontinuous release, the binomial distribution is applied to evaluate the frequency of fission products release for the given and both neighboring wind directions. On the reliability of binomial distribution, a value of 67% is adopted.
The
dose
equivalent
by
the external gamma-ray
exposure is
calculated based on the air absorbed dose. The relative air absorbed dose, D/Q (uGy/Bq), is calculated by integrating the concentration of fission products.
(D/Q) = K1-I'lla• 0ST -°° X°° 0/ ! °.—
in v
where,
2
5- -B(,ur)-U/Q)-dx-dvdz
(12)
K-^:conversion factor from activity to absorbed dose, (dis«m3«;iGy/MeV-Bq'h), E:effective gamma energy of noble gases, (MeV/dis), # Q :true linear absorption coefficient of gamma-rays for air,
(nT 1 ),
H'.total linear absorption coefficient of gamma-rays for air,
(m" 1 ),
r:distance from a point in the radioactive cloud to the receptor, (m), B(#r):dose build-up factor.
The internal thyroid exposure by iodine is evaluated by considering the three pathways, inhalation, ingestion of leafy vegetables and ingestion of cow's milk, and the internal body exposure by tritium is evaluated by considering inhalation and skin permeation. The dose equivalents by iodine and tritium are calculated by concentration in the air at ground in accordance with the method which is recommended by ICRP. Publication 30. The maximum X /Q and D/Q for each release type are shown in Table 1.
.
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2.3
Sources of fission products During normal operation of the HTTR, coatings of fuel kernel can
retain noble gas and iodine. Therefore, it can be considered that these fission products are mainly released from the fuel particles with defect of coating layers by diffusion. The calculational model has been developed to evaluate the ratio of release to birth of fission product, (R/B), for fuel compact containing coated fuel particles with defect. no
(R/B) of nuclide i is calculated based on (R/B) of
Kr in accordance
with following equations.
= 1.56X10 3 -exp ( - ±?M-
) +1.29X10"2
(R/B) = K X (-ip^i- )1/2X(R/B) i
i
A
Kr-88
i
where, (R/B)^:(R/B) value of nuclide i from coating failure fuel particle, Ttfuel temperature, (K), K^rparameter
of the precursor effect of nuclide i
determined by the experiment.
The calculational model is verified by comparing the calculated (R/B) with the measured in the OGL-1 fuel element irradiation tests at JMTR^ '. An example of verification results is shown in Fig. 2. On the other hand, iodine plates-out on the inner surfaces of the primary coolant system. The removal efficiency of iodine is calculated by PLAIN code^' by considering the conditions of plate-out in the HTTR, such as concentration of iodine, coolant flow condition, temperature of inner surfaces, etc. The removal efficiency is estimated to be 10% per pass. PLAIN code is verified by the OGL-1 experiment. An example of verification results is shown in Fig. 3.
Then, the activities of fission products in the primary coolant, A(Bq),
are calculated by considering removal of the primary coolant
purification system and removal by plateout on the inner surface of the primary coolant system for iodine.
8
Ate
3.2X10I0-/l A.
where,
=
- 7pi
(15)
P:reactor thermal power, (MW), 0:fuel failure fraction, Y-rcumulative fission yield of nuclide i, (fission ) ,
W'rp.;flow rate in traps of primary coolant purification system,
1.4x10
(kg/s) for
noble gas and 5.6x10
(kg/s) for iodine and tritium, 77V-:removal efficiency of charcoal trap and cold charcoal trap, 0.99 for I and 0.90 for noble gas, respectively, W:flow rate in primary circuit, 10.2 (kg/s). The amount of tritium in the primary coolant is calculated by considering the generation from impurities in primary coolant and graphite blocks, boron absorber and tritium recovery test samples which are loaded in the core. The calculated activities in the primary coolant are summarized in Table 2. In this place, 1% of fuel failure fraction is assumed very conservatively to hold operational flexibility although the fuel failure fraction at the fabrication is expected to be less than 0.02%. Total activities in the primary coolant are 9.4x10 4.7X10
11
Bq for iodine in
131
MeV-Bq for noble gas,
I equivalent and 4.1xlO10 Bq for tritium.
The amount of fission products released by the fuel compacts irradiation experiment is 2.0xl08 MeV'Bq for noble gas per test.
The amount of fission products released summarized
in Table 3 for
each
release
type.
from the stack is In Table
3,
the
contribution of the fuel failure test is also included. In the test, activity in the primary coolant is permitted to raise up to 4.4x10 • MeV' Bq-h for noble gas and five tests are allowed per year in maximum.
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/M6
3.
Effective dose equivalent in the off-site The dose equivalent in the off-site is evaluated for external
gamma-ray exposure, internal thyroid exposure and internal body exposure based on the calculational model for various leakage and transport paths described above. The dose equivalent in the 16 directions by external gamma ray exposure on the site boundary is shown in Fig. 4. The maximum value is 0.69 ^uSv/y in the south west direction. The dose equivalents by internal thyroid exposure and internal body exposure are 0.058 )iSv/y and 0.019 jaSv/y, respectively in the west south west direction. Therefore, the" effective dose equivalent during normal operation of the HTTR is 0.77 )iSv/y. Considering the amount of radioactive materials released from the JMTR which has been installed on the same site, the effective dose equivalent becomes 13 jiSv/y.
4.
Conclusion
Analytical method and result of off-site•radiation exposure from the HTTR during normal operation are evaluated and assessed. The annual effective dose equivalent is 0.77 ^iSv/y. The result is sufficiently lower than the reference dose of 50 jiSv/y, although very conservative assumption and parameters are taken such as 1% fuel failure.
References (1) S.Saito, "Present Status of HTGR Development Program in Japan", 11th International Conference on HTGR, Dimitrovgrad, USSR, June 19-20, 1989 (2) K.Sawa, et. al., "Analysis method of the coated fuel particle failure and fractional release from fuel elements of HTTR" (in Japanese), JAERI-M 88-258, (1988). (3) O.Baba, et. al., "Fission products plate-out analysis code in the HTGR -PLAIN-" (in Japanese), JAERI-M 88-266, (1988).
A-6 10
Aft
Table 1 Maximum ( X / Q ) and ( D / Q ) for various release type
(*/Q) (h/m3)
(D/Q) (^Gy/MeV-Bq)
By regulation of containment internal pressure
5.3X10" 11
5.1X10" 14
By replacement of air in containment
6.3X10" 11
5.2X10" 14
By ventilation of reactor building
5.3X10" 11
5.1X10" 14
By purge from gaseous waste treatment system
5.2X10" 11
6.5X10- 14
By fuel compacts irradiation experiment
6.6X10- 11
5.4X10" 14
Release types
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11
Table 2
Radioactive nuclides and activities in the priiary coolant
Nuclides
A c t i v i t i e s (Bq)
Kr-83a
2.6X1013
Kr-85«
9 . 4 X 10 1 2
Kr-85
3.0X109
Kr-87
1.3X1013
Kr-88
1.5X1013
Kr-89
6.7X1012
Kr-90
4.6X1012
Xe-131a
1.0X10n
Xe-133i
1.1X1012
Xe-133
2.4X1013
Xe-135«
1.3X1013
Xe-135
4.1X1013
Xe-137
6.1X1O12
Xe-138
1.1X1O 1 3
Xe-139
2.1X1012
1-131
1.1X1011
1-132
1.3X1012
1-133
6 . 5 X 1O11
1-134
3 . I X 10 1 2
1-135
1.1X1O 1 2
H-3
4.1X1O10
A-6 12
M°\
Table 3 Amount of fission products released from the stack
Noble gas (MeV-Bq/y)
Iodine (Bq/y in 131 I eq.)
By regulation of containment pressure
7.4X1O 1 0
4.0X10 8
By replacement of air in containment
2.1X10 1 1
1.3X10 9
By v e n t i l a t i o n of building
1.7X10 1 3
1.4X10 9
By purge from gaseous
1.2X1O 10
5.8X10 8
2.4X10 9
4.9X10 8
Release types
Tritium (Bq/y)
1.1X10 1 3
waste treatment system By fuel compacts i r r a d i a t i o n experiment
A-6
13
Alo
N>
Stack Reactor Building Filtering System
Reactor Containment Vessel >->•
Fuel Handling System
(6)
Filtering System
(1
Decay Tank leakage
leakage
Molecular. Sieve Trap Core
\
Filtering System
(5) v
Cold Charcoal Trap
Charcoal Trap
(4)
Gaseous Waste Treatment System
(/leakage) (3)
Primary Coolant Sampling Pipe
Capsule Primary Coolant System
Charcoal Trap (7) Sweep Gas Sampling Facility for Fuel Compacts Irradiation Experiment
Fig. 1 Release Paths of Fission Products A-6
163
!
1
i
y
1-131
O
Q™
Xe-135m o \c
or-,,,,.
-80m -
cPKr-87 Kr-88/
-
Xe-138
o
•I to
5
°o Kr-89 /( / > e-137
-
i
10
10n
10"' Measured
Fig.2
^y,
Iff
(R/B)
Comparison of the Calculated ( R / B ) with the Measured ( R / B )
A-6
15
10" CM
Coolant Flow
f 108
o o
10'
o
a> o
§
o
o
-
Measured
— Calculated (PLAIN) I0 4 1000 High Temperature Duct
Duct
800
In-pile Tube
Heat Exchanger
Cooler
a>
200 -
0
0
20
40
60
80
Distance from Fuel Top ( m )
Fig. 3
Plateout Distribution of
t31
I in OGL- 1 A-6
16
CO. 13J NNE
[0. 33] NW Meteorological Observation
CO. 30] WNW
Pacific Ocean
Tower-dSK
CO. 6 9 3 SW
CO. 183 SE
SSE
•
. ' . "
C/iSv/y]
Fig. 4
Dose Equivalent in the 16 Directions by External Gamma-ray Exposure on the HTTR Site Boundary A-6 17
,
.
A
Exclusion Area Boundary \ JAERI-PNG Boundary
XA0101484 ADVANCED GAS COOLED REACTORS - DESIGNING FOR SAFETY by Barry A Keen Head of Engineering Development Unit NNC Limited Booths Hall Chelford Road Knutsford Cheshire UK
Introduction The Advanced Gas-Cooled Reactor Power Stations recently completed at Heysham in Lancashire, England, and Torness in East Lothian, Scotland represent the current stage of development of the commercial AGR. Each power station has two reactor turbo-generator units designed for a total station output of 2 x 660 MW(e) gross although powers in excess of this have been achieved and it is currently intended to uprate this as far as possible. The design of both stations has been based on the successful operating AGRs at Hinkley Point and Hunterston which have now been in-service for almost 15 years, although minor changes were made to meet new safety requirements and to make improvements suggested by operating experience. The construction of these new AGRs has been to programme and within budget. Full commercial load for the first reactor at Torness was achieved in August 1988 with the other three reactors following over the subsequent 15 months. This paper summarises the safety principles and guidelines for the design of the reactors and discusses how some of the main features of the safety case meet these safety requirements. The paper also summarises the design problems which arose during the construction period and explains how these problems were solved with the minimum delay to programme. Safety Principles and Guidelines The overall probabilistic design safety guidelines applying in the UK require that the total frequency of all accidents that could lead to an uncontrolled release of radioactivity should be no greater than about 10~?/year i n order to avoid individual faults causing an excessive contribution to the overall station risk of 10~^/year. Higher frequency faults are acceptable at commensurately lower releases. The practical interpretation of these guidelines in terms of the design development has been to provide effective protective features and systems to ensure that the reactor pressure vessel, internal structures, and fuel are maintained within safe limits for all fault sequences more frequent
A-7
than 10~7/year. Thus, a design basis is determined within which the total envelope of initiating faults and fault sequences is considered. The aim is to show that even for the most limiting sequences, the possibility of any accidental release of a significant quantity of radioactivity can be discounted. The frequency of those sequences falling outside the design basis is calculated and shown to be acceptably low. Good design, with the objective of reducing the probability of faults occurring, and the provision of reliable protection systems form the basis of the design approach. Reliability is achieved through the adoption of appropriate design standards and the use of redundancy. Following the more frequent faults for which very high reliabilities of protection are needed, it is not considered that redundancy alone within a single system is sufficient. The safety guidelines require that diverse means of protection are provided as a defence against common mode failure. The guidelines require specific attention to be given to the potential consequences of internal hazards, i.e. those arising from failures within the power station, and external hazards, i.e. those which are a feature of the site in terms of both natural and man-made phenomena. These guidelines are responsible, in particular, for the layout, segregation and qualification of plant and systems as necessary for each of the possible hazards, which can themselves be considered as potential common cause failures. Finally, an important principle having an influence on the design is the requirement that in the event of an initiating fault occurring, the benefit of operator actions to improve or ensure safety should not be claimed within at least 30 minutes post-fault. This has led to the provision of automatically initiated and controlled systems and the avoidance of dependence on operators to identify and control fault conditions in the short term with the risk of taking precipitate actions. 3
Development of the Safety Case The following sub-sections discuss various aspects of providing the final safety submission to the UK licensing authorities.
3.1
The use of probabilistic safety assessment The adoption of a probabilistic approach has had a major influence on design and safety assessment. Figure 1 shows the relationships which indicate that the PSA has not been a separate activity but that it has had an integrated role in ensuring that the design and safety case develop in sympathy. This integrated role has developed as the design proceeded from initial concept through Preliminary Safety Report, Pre-construction Safety Report and station safety report stages and will continue through operation. The methodology comprises three major parts:
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i)
Fault schedule The fault schedule is a comprehensive set of initiating events broken down into distinct categories. Within each category individual events are identified which lead to a similar reactor response and are significant in respect of both frequency and consequence.
ii)
Fault sequence analysis In satisfying the probabilistic guidelines the objective for each initiating event has been to provide and justify adequate redundancy and diversity of protection to trip, shutdown and cool the reactor. The development of the safety case involves determining the family of possible fault sequences which could follow each initiating event depending upon the operation of the protection systems. Thus the role of PSA is twofold.
iii)
i)
Firstly, all sequences where the frequency is sufficiently high must be identified and confirmed as having acceptable consequences.
ii)
Secondly, the frequencies of all sequences outside this design basis are assessed and integrated for comparison with the probabilistic targets.
Event and fault tree analysis The execution of the PSA has used a combination of event tree and fault tree methodology. Event trees which consider the development of faults in time have been used in quantifying the frequency of fault sequences involved in the operation of the reactor trip and shutdown systems. Fault trees have been used to quantify the integrated arrangement of gas circulators, main boilers, decay heat boilers and the associated auxiliary service systems including electrical supplies. Different fault trees are used to reflect the different cooling requirements and plant availability following each initiating event. This approach permits the functional dependencies arising from the division of the primary circuit into four quadrants and the associated arrangement of essential supplies to be correctly represented.
In carrying out the analysis, as far as possible the aim has been to adopt generic reliability data and thereby avoid the need for specific justification. This has been possible in most cases but for some specialised components alternative approaches have been followed. In some cases, reliability data has not been readily available and certain judgements have been made on the basis of engineering and operational experience. This is not inappropriate. The PSA has proved a ready means of establishing the importance of the individual component reliability data to prove acceptable while focusing on more critical items for which special attention can be given. A particular application of sensitivity analysis has been the numerical consideration given to common mode
A-7
failure. Avoidance of such dependent failures has been achieved by careful attention to design specifications, quality assurance, construction, installation, commissioning and operating procedures. The main objectives of the application of probabilistic safety guidelines in the design approach to safety for the UK AGR are: i)
To assist in the development of the design to achieve a very low probability of a radioactive release.
ii)
To achieve the first objective primarily by the provision of a high standard of protection system design such that the probability of events leading to possible core damage is itself sufficiently low irrespective of the further probability of consequential events leading to a significant release.
iii)
To provide a framework for a comprehensive and systematic assessment using both qualitative and quantitative measures to compare the importance of one aspect of the design with another.
iv)
To ensure a balanced design is developed such that no single events or fault sequences make significant contributions to the overall risk.
In examining these objectives, it should be noted that none make any attempt to establish an acceptable level of safety of the reactor simply in terms of an absolute index of risk. The guidelines are specified and used as a powerful aid in ensuring an adequately safe design is achieved and as a means of providing a very detailed insight into the design and operation of the protection systems. However, the numerical results of the analysis does indicate a very low probability of a radioactive release. The total frequency of fault sequences which could lead to exceeding plant and fuel safety margins, i.e. potentially leading to a damaged core, is calculated to be close to 10~6 per reactor year. Notwithstanding this result, the very long timescale for response of the primary circuit following most faults means that many of the sequences considered to be unacceptable in the probability analysis will result in adequate heat removal achieved by operator action some time (hours) after the initial incident and before any radiological hazard occurs. The more practical value of the analysis is the ability to systematically identify strengths and weaknesses in the design and implement protective measures in the most effective manner resulting in a balanced design. The analysis also provides an effective basis for the practical translation of probabilistic safety guidelines into the operational stage. In the UK, operating instructions are applied to control the state of the reactor and associated systems during operation to ensure that the power station is always operated within safe limits determined by the outcome of the design safety analysis. In respect of requirements to take essential plant out of service for maintenance while the reactor remains at power, the probability analysis provides an appropriate means of identifying the importance of individual and combinations of plant outages. The design safety guidelines require specific protection system reliabilities to be met during maintenance and the analysis permits both
A-7
AU
an assessment to be made against such guidelines and the development of effective operating instructions allowing the greatest operational flexibility consistent with satisfying safety requirements. While the role of PSA has developed rapidly over the last decade and has been a major influence on the design and safety assessment for Heysham 2 and Torness it must be remembered that it is only one part of the overall design and safety assessment. It is to be noted that the CEGB Design Safety Guidelines comprise 20 separate annexes which cover all aspects of design and safety, and define acceptable standards developed from the experience gained from previous generations of reactor design and operation. The effect of these and other measures is to complement the use of PSA in assuring overall adequacy. The use of PSA in design has assisted in making sure that design decisions have been arrived at systematically and justifiably and has avoided major design modifications as the project has proceeded.Despite the value of its contribution to the development of the design and safety assessment for Heysham 2 and Torness, there are important limitations to the use of PSA. In some respects, the methodology still awaits development. However, the introduction of explicit safety guidelines and the"use of a probabilistic approach to safety from the earliest stages of design have meant that it has been possible to incorporate all the major safety features before construction started. 3.2
Design approach to hazards Protection against hazards is a requirement of the safety guidelines and their possibility and consequences are specifically taken into account in the design approach. Internal hazards are those whose source is attributable to failures within the power station while external hazards are those whose source is outside of the station and include both natural and man-made phenomena. These are discussed in turn. 3.2.1 Internal hazards The design approach followed is to recognise hazards by their consequence and systematically examine the plant and systems within the power station to identify potential causes. Internal hazards considered include fire, flooding, dropped loads, hot gas or steam release, pipe whip, missiles, failure of rotating machinery, release of toxic substances, and failure of pressurised systems (e.g. gas storage tanks). Defences adopted in the protection against internal hazards depend on the nature and potential consequences but may include i)
Avoidance or minimisation of hazard potential, e.g. use of non-combustible materials
ii)
Layout, e.g. remote location and careful orientation of high pressure storage tanks in respect of vital plant or systems
iii) Separation, e.g. provision of sufficient space between diverse systems or redundant parts of one system such that the consequences of a hazard are limited
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iv)
Segregation, e.g. provision of rated fire barriers between groups of components to limit the extent of a hazard
The recognition and treatment of hazards has a fundamental effect on the overall arrangement of systems. This is carried out through to the detailed segregation of electrical power and control cables which is arranged to satisfy principles which limit the impairment of redundancy within systems in the event of a hazard at any location on the station. As an example of the application of these principles, a depressurisation fault arising from failure of a sidewall penetration leads to a hot gas release hazard which because of barriers within the reactor building can affect the operability of plant serving only one of the four quadrants. Similarly a major failure of the main feed and condensate system, e.g. catastrophic deaerator failure, can only affect the operation of the emergency feed system (which is in the same area) leaving the decay heat boiler feed system on the opposite side of the reactor building unaffected. 3.2.2 External hazards The design approach is initially a site survey to quantify the frequency and severity of potential external hazards. Clearly, the subsequent treatment of external hazards is site-specific. A comprehensive range of possible hazards is studied and the possibility of adverse effects on the power station is examined. In addition to site location and defences, those requiring specific design solutions for plant and systems to satisfy the safety guidelines are considered further. For Heysham II and Torness these are earthquake and high wind for which suitable qualification of plant and systems to withstand the consequences of these hazards is necessary. Although layout and separation or segregation of plant and systems may be important in terms of limiting the consequences of failures induced by the specific hazards, these are not the prime defence adopted. The approach followed is to specify for the site an appropriate intensity for each hazard and then to qualify sufficient plant to withstand the effects of the hazard, including possible consequential effects of the failure of non-qualified plant, to ensure safe reactor shutdown and decay heat removal with adequate reliability. 3.3
Seismic design The rarity of destructive earthquakes in the UK means that seismic design needs to be incorporated only in equipment necessary to provide safe shutdown and cooling of reactors. Two main protection levels were chosen: i)
The "Safe Shutdown Earthquake" (SSE) is only likely to happen once in 10,000 years with a peak ground acceleration of 0.25g. The safety components designed against this include those associated with safely tripping and cooling the reactors.
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ii)
The "Operator Shutdown Earthquake" works at a much lower level of ground movement. Not all safety equipment is given a seismic designation, but some safety-related equipment is required for protection against other reactor faults that could occur at relatively frequent intervals. Because of this, the station operator will trip the reactors in case the non-seismic equipment has been damaged by an earthquake. An alarm operates at the chosen horizontal ground acceleration of 0.05g - l/20th of the force of gravity.
Seismic classification of components is required because of the large number of them in a power station and their wide range of functions. The classification provides a systematic basis for the design of an individual component so that the final overall power station would possess a coherent level of seismic capability. i)
Class A equipment has to withstand an SSE and still function properly. This classification applies also to buildings and structures which house or support Class A components.
ii)
Class B relates to equipment whose failure in an earthquake could affect safety-related seismic plant. Where necessary, appropriate strengthening or upgrading to aseismic standards was undertaken.
iii)
Outside these categories, other components are allowed to fail as there were no safety implications for the plant.
In all, 72 basic components or systems have been designed to survive a safe shutdown earthquake. Once the ground motions were chosen, it was necessary to translate these into loading conditions for incorporation into design specifications. The dynamic response of components and structures are not solely dependent on their mass, stiffness and damping characteristics but depend also upon the interaction between the different buildings and the ground itself. The analysis carried out was broken down into two parts: soil-structure interaction studies and component studies, where the former provided the input to the latter. The soil-structure interaction studies investigated the behaviour of the overall system where the modelling of the structures and components was simplified but still retained the important features that govern their dynamic response. Actual seismic qualification of equipment is done by: -predicting the equipment's performance under analysis -qualifying by combined testing and analysis -shaker table testing under simulated seismic conditions -comparison with similar equipment already qualified.
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A 3/1
Most components could he qualified by analysis, using mathematical and computer techniques. A combination of analysis and testing was used where a basic dynamic characteristic was unknown. In addition to analysis of dynamic responses, the parameters of stress, strain and displacement had limits set on them. The need to have aseismic components has an effect on the total design of a power station. In relation to aseismic components there is an increased likelihood of design changes because of aseismic requirements and those design changes have then to be checked for aseismicity. Another consideration is the need to avoid excessively conservative designs because of the particular characteristics of analytical and test methods used. The overall impact of seismic criteria was kept to an acceptable minimum by the four stages of careful selection of the SSE ground motion to represent UK seismic activity, component classification, phasing the overall station analysis with the concept of the Design Basis Earthquake being used for early work where design/layout changes were most likely. Selecting the most appropriate method to qualify equipment. Owing to the non-linear response or low natural frequency, additional, more detailed, analysis had to be used on the boilers, fuel assembly, charge machine, charge hall crane, reactor core and pressure feed water tank. 3.4
The design of AGR thermal shield details for high temperature applications The AGR pre-stressed concrete vessel (PCPV) liner, the gas baffle and the guide tubes are thermally shielded from the circulating coolant gas by a ceramic fibre insulant. The insulation is retained by a series of thin stainless steel plates and cylinders which are attached to the shielded structures by various means. In the plane areas of the PCPV liner and gas baffle the primary retention consists of a threaded stud welded to the shielded structure, and the cover plate is locked at the required position by means of a threaded clamping arrangement. The shield details around upstands and penetrations are frequently more complex and utilise cylindrical components. A significant proportion of the thermal shield components are exposed to coolant gas temperatures of 425-650°C when the reactors operate at full power. At these temperatures creep effects become significant and the designer must utilise design rules which recognise these effects. There are two basic routes for design by analysis contained within design codes. The elastic route requires the use of linear elastic analysis methods and contains a high degree of conservatism in many of the design rules. This is intended to cover the uncertainties which exist when dealing with components that are subject to plastic and creep deformation. The alternative inelastic route requires a better estimate of the plastic and creep strains that a component receives during service and provides less conservative design rules in recognition of this improvement. Ideally, the designer needs to use simple methods of analysis with design rules which guarantee a minimum of conservatism. The present rules are particularly severe when designing thin components subjected to thermally
A-7
induced loads, such as are seen in the AGR thermal shield. A need was identified therefore to develop simplified methods of design by analysis specific to the shield components. It has been demonstrated that simplified methods of analysis can be applied which bound the inelastic stress-strain behaviour. These methods allow for an improved assessment against incremental behaviour and creep-fatigue damage criteria using the results of linear elastic analysis. If elastically calculated strain ranges are less than 0.1% then elastic shakedown will occur and creep-fatigue damage will be estimated with reasonable accuracy using the creep-fatigue design curves of the design code with elastically calculated quantities. For strain ranges in excess of 0.1% the use of design factors on best estimate cyclic frequencies gives an acceptable assessment of thermal shield components, since the additional damage due to elastic follow-up is effectively included in the creep-fatigue assessment. Utilising these guide-lines there is no requirement to compute additional creep damage separately since it is adequately catered for in the design creep-fatigue curves. A demonstration that components are not subjected to excessive incremental behaviour is dependent on the mechanical properties assumed and provides the major limiting design feature when utilising elastic analysis. Particular attention has been given to the design of Alloy 800 cylindrical bellows units operating at temperatures up to 650°C. An extensive test programme has been carried out to consider the particular effect of cold work on the cyclic performance of a unit. From the mechanical properties it was found that significant incremental behaviour was not likely and that the resulting creep-fatigue endurance was a slight improvement over the anticipated annealed material behaviour. The creep-fatigue design line constructed from a consideration of material properties and inelastic behaviour at convolution crests is intended for use with simple elastic calculations and is an example of the application of simplified methods to the design of thermal shield components. 4
Resolving Design Problems During Construction All Project Management Teams which are set up to drive complex projects like the design, construction and commissioning of a nuclear power station must realise from the beginning that unexpected problems will arise at various stages of the project. They must therefore have an organisation available which can swing into operation quickly to determine the cause of the problem and then to propose, agree and implement the preferred remedy. This section explains how two major problems which did arise were resolved with minimum project disruption.
4.1
Standpipe Cracking The roof of the PCPV is penetrated by a series of carbon steel fuel standpipes which form a continuation of the PCPV liner. The standpipes are water cooled and thermally shielded by means of a stainless steel liner containing ceramic fibre insulant. The temperature at the .top of the standpipe is monitored during operation to ensure that the closure arrangement is not at risk of losing effective sealing. Temperatures normally are of the order of 60°C, but during 1985 the temperature on one standpipe at Hunterston 'B' began to approach the operator action limit of 150°C.
A-7
2>3
4.2
Control Rod Vibration Programmed dates were achieved throughout the first 6 years of the construction programme for each of the two reactors at Heysham 2 and Torness. Following the unfuelled engineering runs at both Heysham 2 and Torness, however, inspection of the control rod assemblies revealed wear marks on the control rods and immediate checks were then made on all the control rods and their channels. These revealed that there were spiral and circumferential wear marks on the outside of a number of rods and corresponding marks in the guide tubes and graphite core channels. Wear was also found in the joints between the control rod sections and on some of the chains which suspend them. The worst worn rod joints and worn chains were replaced but it was realised that serious wear problems could develop on the others if they were used for the reactors' thirty year life span. The control rods are the reactor's regulating and primary shutdown system and their vital safety role means that they must be of the highest possible integrity. In order to review all possible causes of the problem, NNC constructed a full scale pressurised CO2 rig and several water rigs and used a half length scale atmosphere air rig, followed by a full length full scale air rig. Other pressurised rigs in the UK were modified to assist in testing. In parallel with investigations on these rigs, tribological studies were carried out and it quickly became clear that the configuration of the cooling gas inlet ports to the control rod channels imparted a swirl to the gas (Fig. 5 ) , which in turn generated a precessional movement of the control rods within the guide tubes and channels and so gave rise to the wear marks. These port arrangements differed from previous designs, because the control rod and fuel standpipe nozzles in the gas baffle were inverted at Heysham 2 and Torness relative to Hinkley and Hunterston in order to provide adequate dome cooling. The solution adopted after much testing involved blanking off the inlet holes with collars clamped and welded around the original inlet ports, and drilling 32 new holes in each guide tube lower down. The size, number and location of the new holes was only decided after studying results from tests on numerous different hole arrays. Putting these changes into effect required an unusual combination of thin men and remotely operated machinery. One contractor supplied the thin men who managed to squeeze between the guide tubes to position and weld the blanking collars and to drill some of the new holes while another produced machines, controlled by operators ten metres above on the pile cap, capable of accurately positioning and drilling new holes in the guide tubes from inside them, in a fraction of the time which was taken to do it manually from outside. Access conditions in the reactor were very difficult, and the remote machine helped to shorten considerably the time taken to drill the new holes. The modifications have been successfully carried out on all four reactors and new tests on the control rods show that the activity has been reduced to a very low level, and that they should last the planned thirty year life of the stations.
A-7
The examples illustrated above have led to an overall delay to the first reactors at Heysham 2 and Torness of about 9 months but the second two reactors were not significantly delayed. Acknowledgements The author is indebted tt> his colleagues at NNG who provided the source material for this paper
DESIGN DEVELOPMENT
Reliability Requirements
Fault sequences SAFETY ANALYSIS
PSA Fault consequences
ROLE OF PSA IN DESIGN DEVELOPMENT AND SAFETY ANALYSIS
FUEL STANDPIPE CLOSURE
GAS FLOWFUEL ASSEMBLY PLUG UNIT
•FUEL STANDPIPE
-STANDPIPE INSULATION CONCRETE COOLING WATER PIPES
HEAT SHIELD •
LINER ROOF
NOTE:- DIMENSIONS EXAGGERATED TO EMPHASIZE GAPS 6 GAS FLOW PATHS
ROOF INSULATION
HOT REACTOR GAS' ENTERS STANDPIPE
DRUM COOLED GAS EXITS FROM CRACKS
STANDPIPE THERMAL SYPHON FLOWS F?G A-7
STANDPIPE INNER TUBE
STANDPIPE INNER TUBE T
i
BIOSHIELD Hill
BIOSHIELD
,;.
BIOSHIELD COLLAR-
BIOSHIELD COLLAR
HEAT SHIELD -
HEAT SHIELD
RESTRICTOR-
RESTRICTOR
UPPER CIMBALJOINT
NOTE> RELATIVE HORIZONTAL 6 VERTICAL SCALES DISTORTED
CROSSFLOW •
GAS FLOWS-OLD HEAT SHIELD
UPPER JOINT
NOTE- RELATIVE HORIZONTAL 6 VERTICAL SCALES DISTORTED
CROSSFLOW
CAS FLOWS-NEW HEAT SHIELD
GAS BAFFLE DOME CONTROL NOZZLE CONTROL ROD CONTROL GUIDE TUBE
CASTELLATED NUT
rr& 5
PLAN ON X-X
HEYSHAM 2 /TORNESS CONTROL ROD INLET PORT CONFIGURATION A-7
FIG 4
The mechanism causing the temperature rise was considered to be associated with weld cracking at the bottom of the standpipe liner, which would lead to thermal syphon flows up the bore and down the outer of the standpipe liner. (Fig. 2) The cracking was confirmed when, during a statutory shutdown of the reactor in 1985, the faulty liner was replaced by cutting through the bottom, removing the old and rewelding a new liner. During this outage a survey of all the peripheral standpipes accessible by man entry was carried out and revealed a significant proportion of the liner welds to be defective. A safety case was made on the basis of forwarning by temperature monitoring of any situation that could lead to a safety hazard and the reactor was returned to power. Economically there were strong incentives to determine the cause of the cracking and to remedy it quickly. The technical investigation including the introduction of specially instrumented standpipe liners and fuel plug units to monitor operating conditions showed that the loading was thermally induced, with moderately high cycle fatigue associated with thermal syphon flows occurring around the plug unit heat shield.(Fig. 3) These flow conditions led to large temperature asymmetries (200°C diametral difference) which were unstable and gave rise to the fatigue cracking of some welds in the standpipe liner. Since the new station design was very similar to the operating reactors it was to be expected that a similar phenomenon would occur, and it was shown that an unacceptable design life would result unless some late modification was introduced. To solve the problem, thermal shielding is retained around the central structural component but a series of baffle plates are introduced to promote better mixing of the convective flows and hence reduce the diametral gas temperature differences imposed on the standpipe liner. (Fig. 4) The principle was demonstrated initially utilising half scale water rigs which gave reasonable representation of flow characteristics and a guide to changes in the temperature field. Thermal hydraulic material codes were also developed to explore sensitivity to such aspects as baffle size and position, and cross flow conditions. The final demonstrations were achieved by installing a number of heavily instrumented plug units with the proposed heat shield modification into the operating reactors. Data obtained from these units confirmed that the magnitude of thermal asymmetry was halved, that the level of thermal instability was reduced, and that the structural integrity of the plug unit and its modified components was adequately retained. This confirmation was obtained approximately 12 months after the problem was first revealed at which point manufacture of the modification was well advanced and backfitting on the new plug units at sites had begun. As a consequence the full station complement of heat shields was modified prior to power raising without influencing the overall programme to completion of Heysham 2 and Torness beyond that brought about by the control rod instability problem discussed in the next section.
A-7
XA0101485 Y.A.Dushin, If .If .Gribov, V.A.Ignatov, U.A.Medvedev Structure stable alloy for large-sized equipment of kigk temperature gas cooled reactor with coolant temperature of 95O°C Ike material selection for kigk temperature gas cooled reactor is determined by tke requirements of processing, structure stability, keat resistance. For tke production of kigk temperature keat exckanger components for reactor nuclear system Bf-4.00 of 1060 Mw keat power, 1100 semifinisked items are called for, ranging from 14x2mm tubes to 2200x150mm skells (Pig.1). fke material must serve for 5.10
- 2,5.10
kours atthe temperature up to 95O°C.
Tke ckaracteristic properties of XH55MB
alloy
Under mentioned above circumstances we should orientate ourselves on strained Ifi-base alloys witk strengthening elements Mo and W. In these alloys a possible elose-packed phases precipitation under operating conditions causes alarm - solid solution impoverishes in refractory elements, loses strengtk,
and brittle
particles of (3" (\L~~) £ - phases initiate premature material fracture. In XH55MBI4 alloy tke main alloying elements (Cr, Mo, W) are in such proportions as to prevent it /\/', After long exposure at 900°C only fine dispersed carbides particles are found (Pig.2). "Cup" fracture confirms a ductile intragranuiar fracture mode; fracture toughness even after exposure at unfavourable temperature of 750 C exceeds 120 j/cm . for comparison Hastelloy XR data are given - an intermetallide /fy-phase precipitation and fracture in the zone of these brittle particles and cracks are observed ; A-8
/I
- 2p
the fracture tougkness is reduced to 30;j/cm
after high tempera-
ture exposure,. A moderate Mo and W concentration in XH55MB
alloy permits
to deform ingots of 6t weigkt and manufacture all necessary semifinished items, but limits kigk temperature strengtk. Tke problem is solved by Zr alloying. In tkis case it should be remembered tkat operating stresses at maximum temperature are one ordme lower tkan tke yield strengtk. Tkus, in creep a skear along grain boundaries plays an important role, it is accompanied by pores formation and metal density reduction (plastic loosening). Zirconium concentrates on grai boundaries, inkibits intergranular slip (tke displacement along grain boudaries is reduced by 5 - 10 times), pores and cracks formation; provided tkat Zr concentration increases from 0,04 to 0,09%, tke loosening
AP/fo
at 95O°C
will decrease by tke factor of 10, and creep life will increase by tkree times. Tkus. it is possible to gain a karmonious combination of processing and service ckaracteristics in tkis alloy. fke material processing properties are illustrated in Fig.4, wkere billets for large-sized reactor and kigk temperature intermediate keat exckanger B -400 components are skown. fke welding of tke alloy in tkicknesses up to 172mm is mastered, also by automatic metkods: electron-beam, submerged arc and electroslag welding, fke meckanical properties of weld metal permit to forge roll welded skeets and plates. Tke alloy is adaptable to forming, cutting out, expantion, reduction, coiling, flattening. As to cutting tke alloy refers to tke tkird group and is among tke austenitic steels. In strengtk, creep, life, radiation properties, kydrogen permeability tke material is quite similar t© tke alloys ©f Hastelloy X type (?ig.5), A-8
- 3fke advantage of XH55MBU, alloy is its kigk ductility (as well as skort-term and long-term) and kence good fatigue resistance tkis ckaracteristics is especially important under conditions of reactor cooling. Plastic loosening and creep life fke kigk alloy ductility and structure stability permit to simplify tke pkysical creep model and to obtain illustrative solutions for estimation of tke material creep rupture bekaviour. Ifovozkilov 's concept is used as tke basis - plastic deformation causes proportional material density reduction A ? . fhe meckanisms of suck plastic loosening are valid for kigk temperature creepAf/fc
tke current deformation £ distinctly correlates with ,
fracture takes place
by tke critical value
. I?
76 1.4 % (?ig»6). It permits to identify tke ratio
/A/*-
with a well-known damage parameter CO and t® use Robot-
nov's model of delayed ductile-brittle fracture /4/. In tkis respect it migkt be well to point out - tke kinetic equation of damage
CO— BS^I-to)
wkere K>-0 ,Q<£ i
is not speculative, but follows from tke direct density estimation at various creep stages (performed by prof. Xuraanin's ckair) - tke density reduction as a result of creep accounts for 1-2$, but for ductile materials such as XH55MBU, alloy
tke long-term
ductility (tke reduction of area) remains equal to more tkan 20 %* Tkus, tke geometric reduction of area is almost in order kigker tkan its pkysical reduction by tke action of pores, and tke
material
may be considered uncompressible. - tke suggestion tkat damages do not influence on tke meckanism A-8
0
and rate of steady-state creep £
is confirmed for ductile
structure stable materials by parallel creep tests and tensile tests performed at 800-950°C
at different stress levels, from
O.I6V2.a 1 1 experimental points fall on
"
witk the same values of In summary determinrd
ft .
?
the creep life T ^ and long-term ductility T^
i-(l-a-o~rK)^
<"£ - ^/n- A'&n
•5-s a
cree
P life by ideal fracture,
If
(X- (n-K)A/(4~ty)&.
(\-==---K
, the relationships
will not differ from those obtained by L.M.Kachanov. is well correlated with factor again four; ft7 K > A curves
curve
as /<5~/
^ where
A
£ •= A • O
^v ((5C)
}
&
B
Index
ft
, and the constants number is
* They are determined from experimental
£ (ft) of limitted duration on the basis of the
program "Kvartet". The results of subsequent creep-rupture tests (up to 2,5.10
hours) confirm the reliability of solutions for
the materials being considered. Within the equipment service life, the creep life of XH55MBU, alloy is slightly reduced as compared with an ideal lifetime
/ by 5.10 hours
&c (/^*)
i s sl
/
T^. /X±
= 0,84, the bent of the curve
igilt/, long-term ductility
relatively high level
remains at a
/ 30 %/.
transient creep regularities It is customary to assume that at high temperature the creep process rapidly becomes stable and a material operates in a stable regime.
However, in helium circuits (due to low stresses)
transient creep is delayed and encompass the permissible deformation values. A-8
— 5— Within the mechanical plastic loosening mechanism it is possible to combine the rate £ and deformation <£ of transient creep with a well-studied steady process rate by way of two experimental constants -
]f~ /stabilisation rate/ and
/maximum unsteady defor-
mation/ /6/:
fhe
T
£
parameters at a given temperature T
are the
function of initial stress Oo . Only by moderate values of / and relatively high values of (50 % these parameters do not respond to the temperature and stress changes. Then solutions appear as particular relations with }f= ConS"ty^r^^4garlier
^.^ s a m e relations
with constant ratios were obtained and examined on the basis of dislocation mechanism of creep /7/. therefore the relationships are universal and may be used to generalise the experimental data in the wide range of temperatures and stresses, and
If
£
unsensitivity to {} @G values is an indication of dislocation creep without intergranular shears and metal loosening. Experimental and calculated results for XH55MBU, alloy are shown in Fig.7. Plotted on the abscissa
&/Z
is a deformation
portion responsible for a steady-state creep, plotted on the ordinate is a total deformation By relatively high
Q > 0,15^
£. , including
a transient creep.
experimental points
for all
temperatures are grouped together on a dashed curve, described by the relation with constants
Q = 85, £ 4 - 6.10
. As the stress is
reduced, the deflection to greater values of £ is observed, a cleavage of the curve takes place, from the £/£u (£) plot it is evident that in the permissible range of £ (up to 2 %) ;, may be in some extent greater than
steady-state. A-8
creep.
-6The corresponding difference in deformations reaches nearly 2 %. It is evident that isochronic and relaxation curves essential for strength calculations must be computed from transient creep characteristics. Jig. 3 shows a heat exchanger component of the intermediate primary circuit of BP4OO power system
and a structure of stresses
in typical tube sections. Compensating and thermal stresses play an.important role. After 10 hours these stresses are reduced by 1 , 5 - 2 times(thanks to relaxation): in the upper section - due to high initial value, in the bottom - due to high temperature. The actual effect may be less because the supposed tubes lifeservice is shorter in nominal operating conditions. In this case, however,because of the high initial creep rate the relaxation procgSi is of fundamental importance, and a component service life is greatly influenced by a transient creep rate. The role of gas circuit medium In gas cooled reactor circuits the medium influence on the material is not confined to a surface - a metal structure may be effected to a considerable depth, and in thin-walled components over the entire section. The medium constitution (?ig.9) is dictated by various factors: in the primary circuit
- the reaction
of leaking steam with graphite of the active section; in the intermediate circuit - impurity atmosphere identification
to form
an oxide film at the surface of heat exchangers tubes and thus to reduce hydrogen permiability. In the processing circuit metan conversion takes place. The principle difference between coolant and gas compositions is not in the active elements partial pressure, but in their correlations, which determine the type of corrosion damage A-8
-7For the primary circuit it is carburization, for processing hydrogenat ion. The medium role is clearly evident by XH55MB
alloy thin
specimens tests (thickness - 1 mm). As a processing circuit a reaction tube of the primary reforming "Azoty
was used tyanovskiy plant
as the primary and intermediate circuits - helium stands
CHD-1, CHD-2. After 12.10 fa exposure in a processing medium the carbon content remained at the initial level, and hydrogen concentration was increased by 4 - 40 times, which is in good agreement with calculated data obtained by using typical solubility constants of hydrogen and its partial pressure in gas medium. The hydrogenation effect shows itself most vividly in metal embrittlement, the elongation after rupture at 20°C is reduced approximately in •3 times. In the coolant of the intermediate circuit the essential changes in the alloy composition and properties were not found thanks to a low carbon activity /
after 3.103h
the carburization
a protective film formation was observed,
with
in the alloy the opposite
processes are being developed - carburization and carbides dissolution.
As a result an interesting chromium, carbon and a carbide
phase distribution in specimen thickness is seen, which influence on the material mechanical properties. A special program "Oka" is developed to calculate the local composition / see a paper Dushin Ju.A., Ivanov A.V., Syshchikov
A.V.
Long-term carburization
of thin-walled components in the primary circuit of HTGR/.
A-8
8
Литература i;- Артемова E.H., Ананьева M;AV, Вергазов А.Н. и др. Жаропрочный оплав для высокотемператрных энергетических установок с гелиевым теплоносителем // Вопросы атомной науки и техники. Сер. Атомиоводородная энергетика. - 1982. - Вып. 2 О С.87-93. ! 21 Душин Ю.А., Медведев Щ А., Грибов H.H. и др. Технологические свойства сплава ХН55МВЦ применительно к высокотемпературному теплообмеиному оборудованию // Вопросы атомной науки и техники. Сер. Атомно-водородная энергетика. - 1988. - Вып. I. С в 94-95. 3. Новожилов в;в. О пластическом разрыхлении // Прикладная механика, - 1965. - Т. 29, С.681-689. 4. Работнов К Ш . Ползучесть элементов конструкций. - М.: Наука, 1966. - 752 с; 5. Душин KJJAi, Медведев Щ А . , Артемова E.H. Пластичные материалы в свете 'Моделей вязко-хрупкого разрушения // Известия АН СССР. Металлы. - 1989, tè 2. C.I70-I74. 6. Вакуленко A^Ai, Душин 10. А., Медведев H.A. Неустановившаяся ползучесть металлов при повышенных температурах и пониженных напряжениях // Проблемы прочности. - '1988. - !•& 10. С. 54-57. 7. Amin К.T., MuMie'^ee А.К. and Dorn J.E. A universal Law for High-temperature diffusion controlled transient Creep // J. Meoh. Solids. - 1970. - V.18. P.413-426.
8. Володин 0.-И..-,' Душин Ю.А,, Звездин ЮоИ 0 и др. Наводораживание металлических материалов ВГ-400 в среде технологического контура // Вопросы атомной науки и техники. Сер. Атомно-водородная энергетика. - 1988; - Вып. Г. С.96-97. А-8
- 9 Рис. I АТОМНАЯ
ЭН ЕР ГО-ТЕХНОЛОГИЧЕСКАЯ
УСТАНОВКА
ВГ-400
D
ГЕХНОЛОГИЧЕСКМЙ КОНТУР
В ПАРОГЕНЕРАТОР I ЭНЕРГЕТИЧЕСКОГО | КОНТУРА 'ü
Для теллообмешшкоБ требуется
тонн
IIOO
полуфабрикатов из никелевого сплава ХН55ШЦ , в том числе 2^0
км ТРУБ
ГОРЯЧЕПРЕССОВАННЫЕ ОБЕЧАЙКИ КОВАНЫЙ
2200*150
Ш * 2 мм, ТРУБЫ
203*9
ММ,
мм,
СОРТ Ф 350 ММ,
ПЛИТЫ ТОЛЩИНОЙ
&0 "ММ .
Л-8
-
40-
СТРУКТУРНАЯ УСТОЙЧИВОСТЬ СПЛАВА Основные элементы, мэс.%
Марка
ХП55ВД
0,08 1,5
19
XII50MB ( типо
0,03 1,5
20
Mo
W
Fe
6,0
2,5
16
0,9
19
55
СТРУКТУРА ПОСЛЕ 3000ч ВЫДЕРЖКИ ПРИ 900°С ХН55МВЦ ХН50МВ
x 3000
M - фаза (Nî-20,Mo-7/,W-8
Карбиды УДАРНАЯ ВЯЗКОСТЬ ПРИ 20°С после
адо 300 о
23 О
f8?
г-
• 1
выдержки
3000ч
.. J 1
200
...
1 #
1 •
#
100
é
О V"
750 ХН55МВЦ
у
800 850 900 Температура выдержки, и С О ХН50МВ.
950
Л-8
.
_. .Рис. 3 . _.. ЭФФЕКТ МИКРОЛЕГИРОВАНИЯ ЦИРКОНИЕМ Без циркония 0,10 иис.% циркония
Смещение по границам зерен и распределение цирконии в зерне ( 950°С, общая деформация 2 % )
Относительное изменение плотности олловэ по длине образца после исиытония но длительную прочность ( 950°С, 13 МПэ ) • О
0,09 мпс. %1г 0,04 нас. %Zr
,Тр = .6ЦО0 ч = 2100 ч tVp Л-8
-
\2-
Рис.4 ТЕХНОЛОГИЧЕСКИЕ СВОЙСТВА
02200мм
Сварно-кованое кольцо теплообменника ВГ-400
0900мм
Цельнокованое кольцо разгрузочной трубы ВГ-400
Сварное соединение (толщина 144мм, электрошлаковая сварка )
Сплющивание трубы
ЛЦ°\
ф 108x8 мм
Л-8
- 43Рис. 5
СЛУЖЕБНЫЕ СВОЙСТВА СПЛАВА ПРИ 950° С ( экспериментальные данные ) ПРЕДЕЛ
ПРОЧНОСТИ
СУЖЕНИЕ
ПРИ РАЗРЫВЕ
СКОРОСТЬ
ПОЛЗУЧЕСТИ
ПРЕДЕЛ
ДЛИТЕЛЬНОЙ
— 155 МПа
(5g
ПРИ НАПРЯЖЕНИИ
ПЛАСТИЧНОСТЬ
ПРЕДЕЛЬНАЯ
ДЕФОРМАЦИЯ
ПРЕДЕЛЬНЫЙ ПОРОГ
10 МПа
£у-
*.Ш-ш0-1
6 - 12 МПа ^#, —
ПРОЧНОСТИ НА БАЗЕ Ю * ч
ДЛИТЕЛЬНАЯ
ПРЕДЕЛ ВЫНОСЛИВОСТИ
- 96 %
У
НА БАЗЕ
НА БАЗЕ
НА БАЗЕ
ФЛЮЕНС [(fij
САМОСВАРИВАНИЯ
Ю4ч
2 С Of
*J J
Ю 3 ЦИКЛОВ
Ю 7 ЦИКЛОВ
у
Sa-
[ O j
ВОДОРОДОПРОНИЦАЕМОСГЬ ®
0,7 % НО МПа
Ç
- 8 • Ю18нейтр/см 2 (
Е-> 0,1 Мэв )
5 МПа ГО"5
см3 ( н . у . ) см*-с
(g) - в условиях первого контура
Л-8
A so
JO
Рис.6. РАЗРЫХЛЕНИЕ Lflß0
, ДОЛГОВЕЧНОСТЬТ* И
ДЛИТЕЛЬНАЯ ПЛАСТИЧНОСТЬ^
20
10 О -0,5 -1,0
Деформация £ и разрыхлениеkßlßo в процессе ползучести при 950°С, П МПэ
п
60
ос
20
—&
x
-Ciось
/г
-с 0
I0
в
too
Ъ
f
50
—"iüj;
15
с
о
О
ф - опорные точки , О - контрольные точки — - расчетные кривые .
/15/1
А-8
Рис.
7.
СКОРОСТЬ £ И ДЕФОРМАЦИЯ £ В НЕУСТАНОВИВШЕЙСЯ СТАДИИ ПОЛЗУЧЕСТИ
г— О
2
4
t
$
Зависимость деформации ползучести £ от линейной составляющей 6 t / f при 850 - 950°С и по пряже ни их,
h —^——j
£
г
•*
. •• •
#
•• •
. . •* *
• •"
0 2-Ю
Относительная скорость стадии при 950°С, 10 МИа • - опытные точки, расчетные кривые
А52-
8
8«Ю
8
гг.ч
деформация £ в неустоновившейся
. . . уст оно вившаяся стадия ползучести. д 8
-16 _
Рис.8.
СТРУКТУРА
И РЕЛАКСАЦИЯ
НАПРЯЖЕНИЙ В ТШООБМЕШЮЙ ТРУБЕ1
Iff МПэ ; 60
I
40 •
20 66О°С
<э кт
6.6
б, МПа
932°С
- компенсационные - мембранные - термические напряжения
уровень напряжения через 10
ч,
уровень нопряжени-я через-5 • 10
ч Л-8
- nРОЛЬ СРЕДЫ ГАЗОВЫХ КОНТУРОВ Парциальные давления г а з о в , Па Контур
250
Первый 15,5.10
Промежуточный
со.
GO
Н.
Не
10
250
О,
СН,
25
25 25
150
550
г,5-ю- |2,5'Ю- 2,5-10 5
Технологический
Концентрация водорода [\\] , углерода fc] и удлинение при р а з рыве <%- после .выдержки в газовых контурах ( толщина образца 1мм) Технологический Первый ( • ) и промежуточный ( о ) контур 12 000 часов
[Hl
*
0,002
() Г о
0,001
(У& О/Р(
и о
о
)
о
3 000 часов
[с],
пр О, мас.%
контур 1 •
0,10
£
о
0,05
0
о—
о
5, г ОО
О
40 ) )
20
о
О 650
700
750
800
'
650
750
850
950 0
Температура выдержки в газовом контуре, G Л-8
XA0101486 MAIN PRINCIPLES OF LOW-POWER HTGR RADIATION SAFETY ENSURANCE Bylkin, B.K., Grebennik, V.N., Kir;jushin, A.I., Kuzovkov, N.G., Ponomorjev-Stepnoi, N.N., Hruljev,A.A., Yanushevich, I.V. The report to be presented at the IAEA TCM on GCR Technology, Safety and Siting, Dimitrovgrad, USSR, 21-23 June 1989.
A low-power reactor plant VGM is being developed as a pilotcommercial plant and has to be a prototype for the future nuclear power plants for different energy-process purposes and must show HTGR properties ensuring their absolute safety under all operating conditions, including emergency ones. HTGR is characterized by the passive inherent safety means availability, which are not dependent on nuclear power plant personnel actions. Realization of these eliminates HTGR harmful radiation and ecological influence on population and nature. First of all to such means and properties we refer: - a great range between operating temperatures and material and structure damage temperatures; the qualities of high-temperature fuel used, the damage of which rises under the temperatures above 1600°C (Table I) are of particular importance; - high thermal response time due to the high material and structure heat capacity, and low core energy intensity; - a possibility of ensuring a reactivity negative temperature coefficient all over the temperature range;
A-9
Ass
2.
- one-phase coolant and its chemical inertia. Wow 7GM design development has not been completed. Therefore an analysis of designing experience and passive safety means realization in the low-power plant VGR-5O and its modifications, many parameters of which are close to CGM (Table II), are useful. Solving the safety problem has been carried out according to the regulatory documentation requirements being in force in the country and HTGR special features have been taken into consideration, as well. During using the plant VGR-5O (VGM) the main sources of the primary equipment radioactive contamination are: - fission products releasing from fuel elements into a helium coolant; - primary material erosion-corrosion products; - graphite dust contaminated by radioactive products; - disintegration products of fuel elements at manoenvring
them
in a circulation loop. During VGR-5O operation under nominal conditions safety elements were ensured by the requirements imposed for the fuel elements on leakproofness which followed from the restriction of the primary equipment serviceability after the short-lived fission product decay (Fig.1). These requirements were formulated as follows: - a fuel element disintegration limit for normal operation, defining a primary coolant activity level is characterized by a relative balanced leakage the radionuclide Xe-133 equal to (R/B) =£ 10" (the first design limit). Safety under emergency conditions was also ensured by the requirements imposed for the fuel element on leakproofness, which
J&6
A-9
3.
arised out of ensuring the population safety in the vicinity of the plant; at all the deviations from the normal plant operation primary fission product activity doesn't have to exceed the activity value typical of the gaseous non-leakproofness of fuel elements, the relative balanced Xe-133 leakage of which is (R/B) ^lo"-3 (the second design limit). An additional requirement imposed for the fuel elements was a restriction residing in that Cs-137 and Sr-90 leakage under all operating conditions has to be less than 10 c/yr. Table III shows the main independent on each other barriers localizing the radioactive fission products. They are chosen on the basis of the analysis of the fission product behaviour in the HTGR primary circuit and the efficiency of the systems localizing them. 1. Spherical graphite fuel elements of 6 cm in diameter in the cores of which microfuel elements with P C-SiC-P C coatings are dispersed to the number of 10 , provide the fission product activity fragmentation and localization at the monitored level of the fission products leakage out of fuel elements when the plant operates under the nominal and some emergency conditions. 2. A chemical cleaning system provides helium coolant purity for chemically active impurities at the level less than one mill"" , C0 2 < 5, CO -c 50, H 2 0 <; 2, N £ <: 25, CH 4 < 2
5, H 2 <. 50, Ar < 1
2
at the chemically active impurities rate equivalent to ca 100g H 2 0/h, the requirements pointed restrict the circuit and core material corrosion and argon activity and C-14 formation, aa
well.
The cleaning system is the second fission product localization system (Pig.2). A-9
/IF?-
3. Helium self-purification by iodine, cesium and other radionuclides deposition on "cold" surfaces (at T <^ 450°C for iodine, at T <; 75O°C for cesium) provides the low activity level of helium itself. 4. A mechanical cleaning and fuel element burn-ups monitoring system provides dust, defective and burned-up fuel element withdrawal out of the circulation loop. 5. Sealing of equipment, rooms and the plant as a whole limits radionuclide ejection out of the plant boundaries under all the emergencies without plant destruction. Table 4 presents data characterizing the primary coolant activity level of the plant VGR-50 under nominal and some emergency conditions which are obtained by using the parameters given in Table 3. The following main emergency conditions have been considered: - the core cooling failure conditions resulting in the fuel element temperature increase; - helium chemical cleaning system failure; - steam generator tubes break-resulting in the interaction of fuel element graphite
shells and reactor reflector with steam;
- lateral breakage of duel pipeline (MPA) at which increase in fuel temperature and fuel element shell oxidation by air take place. Temperature conditions of fuel use for different operating conditions are presented in Fig.3. Under all the conditions pointed the requirements imposed for safety are met. The fission product and coolant activity calculations are made by using the relationship /1/:
.fafa-Tj
5.
where R( T.) is the fission product rate at instant of time £ 1 ; T., T. .. is the fuel temperature at instant of time 3* and preceeding j
«"* •
it j-1;
Av(Tn'""Tn*»i) i 0
a
fraction of fuel elements the temperature
of which has changed from T. - to T.'jn is the spectrum characteristics of defective, contaminated and microfuel elements; ^f is a fraction of defective microfuel elements in fuel; £6? is a fraction of fissile
material contaminating the fuel element structures. After
the events at the Chernobyl nuclear power plant the design of the plant VGR-50 has been modified with the purpose of providing the higher safety of it and approaching its characteristics to VGM parameters. It's provided: - in all emergency situations a reactor shutdown must be carried out by freely falling rods and the reactivity compensation system based on the absorbing ball insertion into a side reflector channel; - in all emergency situations core heat release removal is performed by the surface cooling system consisting of three independent channels based on a passive principle; - using an additional circulation loop capacity of which is upto 3% of nominal, to cool the core - filling the reactor pit rooms with nitrogen to restrict the core graphite oxidation. Applying a helium moisture content sensor and quick-acting isolation valves for water and steam and steam generator quick draining facilities and also the main plant equipment state diagnostics systems and safety systems contributes to the increase in the plant safety. A-9
6. To ensure reliability of functioning, required redundancy and sectioning of separate equipment, systems and protective safety systems, providing desired efficiency at independent on an initial even failure of one of some independent system channels, are provided. Emergency situations due to sequential failure practically of all the active and localizing safety systems the occurance pro—7 bability of which is less than 10 ' and an accident with entire destruction of all localizing systems, including reactor building destruction have been analyzed (Table 5 ) . Calculation of activity release out of the destructed reactor boundaries has been conducted according to relationship:
where A. is i-th radionuclide activity deposited on the circuit surface during the plant operating time (for cesium A«
was assumed to
be 50c); Kj is an emergency desorption coefficient (K^ = 0.02 J?or iodine, ICj s 0.01 for cesium); K 2 is a coefficient of depositing in the plant destructed under depressurization (K2 = 0 . 7 for iodine, K 2 = 0»9 for cesium); JC, is a coefficient of depositing in the plant destructed at heating up (Ky*o.7 for iodine, K-s-Q.S for cesium). To evaluate the fission product release out of the reactor building boundaries in the case of reactor destruction, the knowledge of the fission product retention by the reflector, reactor structures and destructed reactor building structures is extremely important. At present these data are not practically available, thus their values were assumed in an expert way at a conservative approximati on. A-9
7. Table 5 sums up the results of radionuclide release assessment under severe emergency situations temperature conditions of which deviate from nominal values by 500°C in the worst case. The analysis has shown that at such accidents volatile fission product leakage is determined by fuel the temperature of which is in the range 1500-1800°C and a fraction of which is very low. The latter fact confirms the importance of delailed core temperature conditions knowledge (Table 7)« Now there are not enough data on Cs, Sr and Ag behavior in graphite and reactor plant structures to essess these radionuclides leakage out of the core and reactor boundaries correctly. In this direction it is necessary to perform an approximate research program. The results presented in Table 6, are obtained without taking into consideration active oxidation of graphite. Consideration of this may result in substantial increase in fission product leakage (Sr and Cs in particular). Thus there are some topical goals: - to develop effective means for graphite oxidation protection, including creation of graphite characterized by limited oxidation rate at a direct contact with air; - to use fuel compositions resistant to carbonization in graphite environment at temperatures upto 20000°C. Thus, the radiation safety analysis of 250 MW small-sized plants has demonstrated the possibility of their safety ensured at all the accidents including improbable severe ones. Solving a problem of fuel and graphite chemical resistance at the accidents is a supplementary conditions of these plants.
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8. REFERENCES 1.
Bykin, B.K., Glebov, V.P., Hruljjov, A.A. et al. "VGR-50 plant radiation safety", AYE T book, issue No. 8, 1988, M., Energoatomizdat, p. 260.
2. Goltsov, A.O., Hruljov, A.A. "Calculational investigation of different
factors influence on ETGR primary circuit activity",
VANT book, series AVE T, 1986, issue 2, p. 57.
TABLE I Operating temperatures and separate structural component failure temperatures
Material
Graphite
Operating temperature
Failure temperature,
°C
°C
750-1100
Fission Product retertion temperature, °C volatile non-volatile
3600 (sublimation)
1100-1200
1600-1800
partially up to melting
partially up to melting
uo 2
1100-1300
2800
U 0 2 in fuel element composition
1100-1300
1600-2100
1600-2100
1600-2100
300-350
800-900
partiallyafter breakdown
partially after break-down
50-100
950-1100
Cs, Sr, 3 partially after break-down
partially after breakdown
Thermal insolation (upper)
Metal shell
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TABLE 2. The principle parameter of the plant YGR-50 and the HTGR plant of module type Value
Parameter VGR-5O
HTGR-module
1. Heat power, MW
136
200-250
2. Primary coolant
Helium
Helium
3. Helium inlet/outlet temperature, °C
280/810
300/750 (950)
4. Helium pressure, MPa
4,0
5.0-6.0
5. Core energy intensity, MW/m3
5.8
6. Fuel element type, diameter, cm
7. Primary equipment arrangement
3.0
spheretype loop-type
A-9
TABLE 3 The main radioactivity retention barriers (nominal condition)
Barrier
Barrier type
Fragmentation and localization
passive
fuel element 2.1O5-1O6 microfuel element 1.1O9-1O10
passive
^-%^IP. for 133Xe at...
Microfuel element and fuel element coating Helium purification system Primary coolant selfpurification from J, Cs, Sr
Mechanical cleaning system of helium from dust Defective fuel element with drawal system Primary gas-tight equipment
Parameter
active
passive
radionuclides are deposited on "cold" surfaces of equipment are sorbed by graphite dust.
active
active passive
helium leakage rate
0.3 V %/& inzr/a^/?^
Special ventilation system of the primary rooms
active
Productive shell
passive
Aerosol and iodine filters Plow rate 1.0.104 m 3 /h, cleaning coefficient Leakage rate
i no v'%/ iSanitary-protective zone
/16 5"
passive
Determined by Geajplan A-9
TABLE 4 Primary activity under different emergency conditions, TBq (calculation)
Radionuclides
Krypton Xenon Gaseous fission products
5,1 3,2
12
8,3
56
44
Water ingress into core
10 8,1 18
The seaccident
5,6
26
—
3,7
23
—
9,3
48 5,6.I0~2
37 I ,6J0 2 3,7.I02 44 0 0,55 I .5J0 2 0:,18 3 3 2 3 0,31 9,6.I0~2 9,6..I0~ 9.6.I0" 9,6 .I0~ I ,2 JO" 37
Iodine Tritium ' Cesium-137
37
81
7,4.JO""3 7.4.I0"3 7,4.ICT3 I ,2.I0~2
Strontium-90 51
Purifi- Fuel heating\t~™<~. T 3Nominal a y a tcation em up of failure 1OO°C
Cr, 5 9 Pe, 6 0 Co,
5,6.J0~ 2 5,6.I0~2 5,6 .I0~2
2,4
0,31 7.4.I0"2 —
30
58
Co, C - H graphite dust Total
*)
46
I.4.I02 1,0 ,I02
48
3 ,8.I02 9,3.I02
activity of tritium in helium activity at the second design limit of fuel element failure.
A-9
/\GG
TABLE 5Hypothetical emergency condition characteristics
No
Steins
Maximum Graphite Final fuel tem- Oxidastate perature, tion characte°K ristics
1.
Blowers disconnection + core failure not more than for 20 hs
nominal
2.
Core failure,belowers disconnection and failure of gadolinium system, of reactivity control after 20 hs
500 periodically after 20 hs
3.
Duel pipeline break, core failure not more than for 20 hs, failure of one out of three blowers and nitrogen system
100
1.2 for 20 hs
4«
The same as in item 3, including all systems of reactivity control
500 periodically
2 for 20 hs
5.
The severest design accident + failure 2 blowers out of 3 ones
6.
The severest design accident + all blowers failure + all reactivity control system failure + nitrogen system failure
Oxidation of 1.8t graphite for 40 hs
500
+500 periodically
0.05 for 1 h
A-9
Oxidation of 10t graphite for 1300ns
TABLE 5
7.
The severest design accident + 1 blower failure+ + steam generator by failure + nitrogen system failure
500
6 for 20 hs
8.
The severest design accident + 1 blower failure + + steam generator box failure with nitrogen accumulation under a protective shell
nominal
3.6 for 18 hs
9.
The severest design accident + 1 blower failure + + steam generator box failure + protective shell and nitrogen system failure
500
9.5 for 20 hs
Oxidation of lot graphite for 60 hs
Oxidation to microfuel ele ments for 40
hs
10. The severest design accident blower failure + steam generator box failure (with nitrogen "washing" of the primary circuit.
nominal
11. The severest design accident + 1 blower failure + all protective and localizing systems
500 periodically
1 for 20 hs
Oxidation to microfuel elements for 30 hs
A-9
TABLE 6 Radionuclide leakage absolute value assessment in hypothetical emergency situations (with building destruction) (Hm=25Q MW, fuel elements period T«=2 yr, exposure t=20 hs, 00 = 2.1O"5) Radionuc lide
&
-85
Xe~ 133 3 - 131 3 - 133
a
- 90 - 137 - 110
Relative leakage from fuel elements
5-10-2 1,5 10-3 2 10-3 6 10-4 5 10-5 1,1 10-3 1,5 10-3
from fuel elements
Yieldt from core
out of thel out of the reactor bo- reactor undaries building boundaries
4,5 IO 3
4,5 10
4.5 10°
4,5
2,3 10
2.3 I0 4
1.3 10
T 3 TO4 5.4 IO3 1.6 I01 3,2 IO2 3,0 IO3
2,3 I0 4 . q 9,2
5,4 5.4 10 6,4 3,0
IO4 4,6 IO 3
Tr\«J O
1,9 10"
3,8 3,2
0,3
1,610'
3,2 I0
1,5 I0:
3,0 I0
A-9
b°\
TABLE 7 Gaseous fission product and J leakage parameters at hypothetical accident in a module reactor
Temperature range
Relative leakage
Fuel fraction
800-900
0,156
8.HT2
900-1000
0,136
0,11
I000-II00
0,193
0,15
II00-I200
0,110
0,3
1200-1300
0,121
0,35
1300-1400
0,178
4,5
1400-1500
0,098
0,7
1500-1600
0,052
1600-1700
0,035
1700-1800
0,0104
1800-1900
6.I0" 4
1900-2000
io- 4
I I I I I
0,3 0,43 0,7 I I I I I I I I I
±8 a relative leakage from a fuel element, including microfuel elements failed is a relative leakage governed by fuel element and microfuel element graphite.
A-9
8
1d~> PHC. i
W
3iEKCHMOCTk MousiocTH acnu
Hi ctfopynoDSHKH ncpDoro K o i n y p i O T
1 — KUitoH oCny
7 CJT;
- 30 cyr, —
4
- 7 cja nocnc
fir
A-9
9 IIpiiHUMnnajibHafl cxei/a pacnpeflejieHHH aKTHBHOCTH B BTTP
3 a riDej;ejr: 10
PMC.
A-9
-3
W
-A
10
-5
10 \~
-6
300
SOD
700
900
£{00
1300
1S00
7~e /*w ep a m y p a.
TewnepaTypHbie ycjiosHH pa<5oTLi I-HOMMHajibHK," pe^Hw; 2-neperpeB TonjiHBa Ha 100°C; 3-MfIA AN/AT-RORR TonjiMBa B TenriepaTypHOM A-9
Figure 1. Primary equipment dose rate dependence on degree of fuel element leakproofness 1 - canyon of irradiation 2 - sphere-type fuel element control means 3 - steam generator boxes 4 - rooms for the distribution, irradiator discharge, rejection and pumping systems - - -
- exposure for 7 days - exposure for 30 days - exposure for 7 days
after steam generator desactiva-
tion. Figure 3* Temperature conditions of fuel use: 1 - a nominal condition; 2 - overheating of fuel by 100°C; 3 - the severest design accident ,jN/ A T - fuel fraction in temperature range. Figure 2. Schematic diagram of activity distribution in HTGR (nominal condition) 1 - activity; 2 - reactor; 3 - fuel element 4 - microfuel element 5 - coolant 6 - purification system 7 - release out of the plant.
A-9
XA0101487 £*%W £5S£.3 **'t?l • ^ ^
HTR-GmbH
Gesellschaft fur Hochtemperaturreaktoren
Gemeinscnafrsunternenmen der Asea Brown Boveri AG und der Siemens AG
"-^#
n 11\
IAEA TECHNICAL COMMITTEE MEETING ON GCR TECHNOLOGY, SAFETY AND SITING Dimitrovgrad, USSR, June 21-23, 1989
Design Status of the HTR 500 Power Plant and the HTR Module Power Plant
by Erhard Arndt, HTR-GmbH, Frankfurt, FRG Reimer Fischer, HTR-GmbH, Frankfurt, FRG
Presented at the IAEA Technical Committee Meeting on Gas-Cooled Reactor Technology, Safety and Siting
Ate
*&M
B-l
1.
Conceptual Layout of the HTR 500 The HTR 500 nuclear power station is a dual-cycle plant with high-temperature reactor and steam power plant for electrical power generation, with the additional possibility of process steam and district heat extraction. The design of the HTR 500 makes considerable use of the technology embodied in the THTR 300. The simplifications and optimizations reflected in the HTR 500 design are based on practical experience with the THTR 300. Thus the transition to commercial high-temperature nuclear power plants means little risk for utilities and manufacturers. Its main features are: Integrated design of the primary system components in a single-cavity concrete pressure vessel. Use of standardized components and proven materials from the THTR 300 wherever possible. Separation of operating and safety systems, leading to simple design. Accident control making use of the slow transient characteristics of the high temperature reactor. Forecasts show a trend towards power units in the size range of 400 to 600 MW. The HTR 500 is a well-suited answer, as it can compete with other nuclear and conventional power plants of comparable size. The arrangement of the HTR 500 power plant is shown in Fig. 1. In the center the reactor confinement building is situated housing the prestressed concrete reactor vessel including the primary system, the shutdown facilities, parts of the decay heat removal system and other safety-relevant components. In addition the reactor confinement building acts as a protection against external impacts. • Fig. 2 shows the prestressed concrete reactor vessel with its internals: The pebble bed is completely enclosed by the gra-
B-l
- 2 -
phite reflector, resulting in a cylindrical core volume of about 200 m 3 . The fuel element spheres are added continuously from above. Spent fuel elements are discharged from the core through three discharge pipes without interruption of power operation. The helium coolant flows downwards through the reactor core, being heated from 265°C to 720°C at a pressure of 55 bar. The energy absorbed by the helium is transferred to six steam generators through which the primary helium flows in an upward direction. The circulators assigned to each steam generator transfer the cooled helium to the cold gas plenum enclosing all the peripheral areas of the prestressed concrete reactor vessel. Thus the liner, the liner cooling system, the control and shutdown rods, and the metal reactor internals are only exposed to the cold gas. Control and shutdown of the reactor are accomplished by means of absorber rods which either drop into boreholes in the side reflector or are directly inserted into the pebble bed. Two auxiliary heat-exchangers with their own circulators are arranged between the steam generators. In the event of failure of the main heat sink, the decay heat is removed by these two heat-exchangers via a separate two-loop decay heat removal system. The feedwater in the steam generators flows downwards, in counter-current to the flow of helium. On entering the steam generators, the feedwater has a temperature of 190°C. The main steam enters the turbine at 530°C and 180 bars, passing through the high-pressure and intermediate-pressure sections before being directed through a water separator for draining and, finally, expansion to the condenser pressure of 60 mbars in the low-pressure section.
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- 3 -
Due to the primary coolant's high temperature level, steam qualities can be achieved with the HTR which correspond to those usually obtained in fossil-fuelled power plants. Due to the strict separation of the operating and safety systems, a conventional steam/feedwater circuit can be used which differs in no way from that of a conventional power plant. For a group of German and Swiss Utilities a Preliminary Safety Report has been recently completed together with design specifications for the major systems and components. A preliminary risk analysis has been performed, an updating in accordance with the Safety Report is underway.
2.
Conceptual Layout of the HTR-Module The HTR Module has a thermal power of 200 MW per unit equivalent to 80 Mwe for electricity production in smaller grids. Moreover it can be utilized for the cogeneration of process steam or district heat. Electricity and nuclear process heat can be used directly for the upgrading of coal and steam reforming of methane. One to eight modules can be connected to one or more parallel steam/power conversions systems. This leads to a wide power range between 80 and 640 MWe with flexible adaptation to the client's requirements. The HTR Module employs the high-temeprature reactor technology realized and proven at AVR. The application of steel pressure vessels makes use of the proven pressure vessel technology of light water reactors. Pig. 3 shows a longitudinal section through the HTR Module with side-by-side arrangement of reactor and steam generator.
B-l
- 4 -
The main features are: Pebble bed core with TRISO-coated fuel particles Coolant flow through the core from top to bottom Maximum fuel temperature even in case of beyond-design accidents not exceeding 1600°C Arrangement of the gravity-driven control rods and absorber balls in reflector bore-holes Integrated arrangement of all primary components in 2 ferritic steel pressure vessels, arranged in parallel Passive residual heat removal via threefold redundant cavity coolers (with fire hose connections) Conventional steam-water cycle. Fig. 4 shows a section of the reactor building. Engineering work on the detailed design has reached such an advanced status that detailed drawings, schemes and stress analyses etc. have already been elaborated for the main components and systems. As far as possible only proven components will be used. A licensing procedure for a site independent license permit has been initiated in 1987. It is expected that the experts will submit their draft expertises in the middle of 1989 to the Reactor Safety Committee.
Engineering Details The purpose of this paper is to provide recent working results regarding the HTR 500 and HTR-Module in the field of research, development and engineering.
B-l
- 5 -
3.1
HTR Design Criteria In the context of the licensing procedure for the THTR 300, the lack of nuclear rules and guidelines especially applicable to HTRs was significant. There is, therefore, a need for technical rules for the wide range of typical HTR-components. In the research and development project "Design Criteria for High-Temperature Metallic and Ceramic Components and the Prestressed Concrete Vessel of future HTR Plants"- carried out under the sponsorship of the Federal Ministry for Research and Technology- fundamental principles and basic data were worked out to establish German Nuclear Safety Standards for the design of HTR-components, Fig. 5. The project began in 1984 and the research work was divided between several working groups and task forces, under participation of several institutions and companies. The coordination has been carried out by the Nuclear Research Centre Juelich. The project has been classified into four sections: A: B: C: D:
Technical safety boundary conditions Metallic structural components Prestressed concrete pressure vessel Graphite structural components
Design criteria are advanced far enough to be the basis for setting up rules for design and construction of HTR-components. Construction and operation experience from gas-cooled reactors and high-temperature technology proven in conventional plants have furnished a most significant contribution to these design criteria and will ensure that the standardization of rules, which are currently being established, has a solid practical background.
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A go
- 6 3.2
Structure of a Pebble Bed Core Fig. 6 shows the structure of the HTR 500 core. The core structure consists of the side reflector, the bottom reflector, the top reflector and the pebble bed of spherical fuel elements. For design optimization and verification a series of tests have been performed with models of various sizes. The purpose of these experiments is to determine the forces upon the reflector side wall and the core bottom structure in horizontal and vertical direction due to dead load, pressure of pebble bed and insertion forces of incore-rods. The test results have verified the chosen design: The forces are comparable for both initial and equi-
-
librium core conditons. 10 insertion procedures without pebble circulation are sufficient to find out the max. forces. The total vertical rod forces amount to 6.480 kN for max. insertion depth. The maximum load upon one side reflector block does not exceed 10 kN. Core bottom structure loads are low. The maximum forces upon one block of the first layer do not exceed 25 kN.
In addition to the above experiments various models were tested to verify the seismic safety of the whole structure. Dynamically the core structure is a many-body structure with non-linear force-deformation couplings. The integrity of the side and bottom reflector under seismic loads is kept by radially orientated spring packs which transfer the loads to the thermal shield. These spring packs must be stiff against earthquake, but must allow radial thermal movements of the core structure under normal operating conditions. Large-scale models of the side reflector with pebble bed and top reflector were tested on the "HRB Vibration Test Facility Juelich", Fig. 7. The results of these tests were compared with those of analytical methods which were developed in parallel.
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AK/\
- 7 The good-natured behaviour of the pebble bed under dynamic excitation has been confirmed. Due to the granular structure of the statistical pebble bed, high damping occurs during seismic excitation, which is, however, reduced with increasing depth of the pebble bed because of restriction of movement. It is possible to describe the seismic behaviour of the pebble bed core analytically. The one- and two-dimensional test configurations of the top reflector were used to analyse resonance and lumping effects. The experimental results were verified by the computer codes. The experimental investigations of the side reflector are underway. The results show a very good-natured behaviour under the impact of horizontal vibrations. The system is stable under seismic load. Small, rigid-body motions of single side reflector blocks were detected, but they did not cause any global ovalization of the complete ring structure. First Finite Element calculations with contact and friction between the blocks show a close agreement with the experimental results. The results of both test series shall be used in the design of the HTR-Module as well.
3.3
Design and Test of High Temperature Components In addition to the steam generator experience at the AVR and THTR tests have been performed at the Interatom KVK Test Facility. These tests have been extended up to temperatures of 950°C for chemical processes: D-Tube He/He Heat Exchanger 10 MW Helix He/He Heat Exchanger 10 MW
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12.
- 8-
Hot Header for He/He Heat Exchanger with original dimensions Auxiliary cooler for decay heat removal Primary Hot Gas Duct, Compensator and Bend Secondary Hot Gas Duct, Compensator and Bend Secondary Hot Gas Valves The 10 MW U-tube heat exchanger and hot gas duct results will be discussed in more detail in the following. 10 MW U-Tube Heat Exchanger Pig, 8 shows the test unit. The heat exchanger tubes are bent to a U-shape. Primary helium flows along the tubes outside in a countercurrent arrangement, secondary helium flows inside the tubes. The hot header is positioned in the upper part of the heat exchanger, resulting in a rather short central tube. The cold gas header is separated from the support plate and is suspended by springs. Pour phases of tests have been performed from 1985 to 1986 with a total testing time of about 4700 h, 1560 hours at temperatures > 950°C. The test results are as follows: Steady-state operation at 40 - 100 % with 950°C on the primary side and 900°C on the secondary side (Fig. 9 ) . There is a good agreement between the design and the test data, particularly in the hot region of the U-tube bundle. Due to a helium bypass flow the design and test values differ in the cold region. • • Non-steady state tests were successfully performed/ such as: Tenfold startup and shutdown of the heat exchanger at a transient level of + ,1 K/min over a temperature range of 750°C.
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- 9 -
.
Cutoff of flow through the IHX at 25 % load over 1,5 h. No measurable convection resulted. Fifty thermal cycles at 7 K/min and a temperature rise of 300°C.
Simulation of disturbed primary inlet temperature between + 10 K/min and - 50 K/min for a max. temperature difference of 200°C. The inspection of the U-tube heat exchanger showed the essential components, such as tubes, hot header and insulation to be in good condition. Only the sheet metal envelope of the tube sections exhibited signs of leakage close to the bend region. But this envelope is only installed for testing and will not exist in the fullscale heat exchanger. The bearing forces for the load transfer system of the cold header and the bundle are within the range of the calculated values. The measured vibrations did not result in any noteworthy loading of the heat exchanger tubes. Successful tests showed leak tightness between the primary and secondary side of the IHX. Test of the Helicoil He/He Heat Exchanger and of the Hot Header in the KVK Test Facility showed favourable resultsJas w e l l .
. .
*.-,•..;-.•
.
>
.
.
.•
;•
•• : •
-;
Primary Hot Gas Duct Fig. 10 shows; the test section of the coaxial primary hot gas duct with fiber, insulation. The gas liner- consists, of graphite and Carbon Fiber Composite. Adjustable supporting elements -fix the gas liner radially and axially. Tests -on?the.primary hot gas duct under normal operating conditionsifeave;been completed. Pig.-. 11 gives a comparison of
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/i
- 10 -
test results and design data. As a significant result, the heat loss from the hot to the cold side was substantially lower than calculated. Typical temperature profiles of the cold support tube, of the pressure vessel tube and of the hot gas liner are shown in the upper part of the figure. The bandwidth of the temperature distribution is included. While the maximum bandwidth is 16°C at the gas liner, values of not more than 5°C were measured at the pressure vessel and support tube. Tests with coaxial bends as well as with coaxial compensators led to similar results. The mechanical behaviour of the components proved to be excellent, including results of more than 1500 mechanical load cycles on the compensator.
3.4
Helium Circulators with Active Magnetic Bearings In modern plant concepts for high-temperature reactors the helium circulators are arranged in vertical position. The magnetic bearings for these circulators will be contribute to the desired simplification and the high availability of the HTR. By the introduction of this technology it is possible to abandon the complicated and expensive facilities for lubricant supply of conventional circulators. Fig. 12 shows an cross section of the prototype circulator with active magnetic bearings. It is representative for all types of circulators envisaged for the different HTR concepts. In designing and testing the prototype circulator, the overall scope of requirements is met by fulfilling the most stringent individual requirements. Construction of the prototype circulator will be started in mid-1989.
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- 11 -
A precondition for the use of active magnetic bearings is the installation of catcher bearings to avoid unintended contact between rotor and static machine parts with the magnets deenergized. In view of the size and weight of the circulator shafts to be controlled, ABB has analysed and experimentally verified the design and construction of the catcher bearings in an anticipated R & D program. Fig. 13 shows the catcher bearing test stand which is simultaneously used to demonstrate the suitability of the magnetic bearings. The catcher bearings are roller bearings of special design, which are capable of withstanding more than 60 drops with grease film lubrication. The dry-lubricated variant, which can withstand more than 10 drops from max. circulator speed of 6000 rpm, also meets the requirements of the reactor. According to the time-schedule the principal experiments are to be completed in 1993.
3.5
Steel Fiber Concrete in the HTR 500 PCRV Steel fiber concrete is a type of concrete to which steel fibers are added to improve its characteristics. Normally steel fibers of 0.2 to 1.0 mm dia and 10 to 80 mm length are added. Addition of steel fibers does not replace the prestressing tendons but permits to reduce or even completely abandon the mild reinforcement. For economic reasons it should be tried to obtain a short construction period of the PCRV. Since the duration of the construction period is essentially determined by the installation of the mild reinforcement, the use of steel fibers allows to reduce the construction period. It is therefore envisaged to use steel fiber concrete without or with mild reinforcement only in selected areas for the approx. 60 cm thick shell around the liner. In addition it is intended to
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- 12 -
use steel fiber concrete for the full height of the stand pipe region of the vessel head. Thus it will be possible to do without reinforcement rods whose installation is timeconsuming because of the lack of space, Fig. 2. A further incentive for using steel fiber concrete is the increase of the vessel safety. The guaranteed tensile strength and especially the toughness of this material result in an increase of the safety of the structural elements. Within a research project a suitable steel fiber concrete has being developed on the basis of the type of concrete used for HTRs. The program includes the material development, testing and calculation.
3.6
Dissimilar Weld Joints for Steam Pipes
The high temperature section of the steam generator including the header uses the material Incoloy 800. For the live steam pipes a ferritic material is used. Steam generators and life steam pipes are connected via dissimilar weld joints. Fig.14. To verify the integrity concept for such joints the r&d program MINERVA has been initiated. MINERVA is part of the project for the verification of the integrity of the water/steam circuit of high-temperature reactor plants for the designed lifetime. MINERVA is designed for a testing time of 20.000 h. The experiment started in January 1987. The impact on the dissimilar weld in the intersection between life steam header and life steam piping is characterized by peak stresses due to different coefficients of thermal expansion of the materials. These peak stresses are reduced during holding times at ser-
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- 13 -
vice temperature by relaxation. The main deformation is shifted to the heat-affected zone of the ferritic material, Fig. 15. The analytical determination of accumulated creep, damage and crack growth in dissimilar metal weld joints during plant operation is verified by experiments with specimens and component tests under test conditions simulating plant operation. In addition, the non-destructive experimental methods are evaluated with respect to their accuracy in service application, 3.7
Computer Codes for Core Layout For the layout and analysis of pebble bed cores concerning neutron physics and thermal hydraulics a family of computer codes is available. Those codes have been validated with the results obtained from KAHTER, AVR and THTR. As an example the code family MOCCA for neutron physics used at HRB and the Paul Scherrer Institute will be presented below. Similar codes are available at Interatom (Code ZIRKUS) and KFA (Code VSOP). The nuclear design calculations for HTRs are performed using the Modulear Core CAlculation System MOCCA in combination with existing code systems or stand-alone programs. Examples for the last two categories are the 3-dimensional Finite Element Code DIFGEN and the 1- and 2-dimensional Transport Codes ANISN and DOT IV.3. The code systems are applicable for HTR's using the thorium/uranium cycle with high enriched uranium or the uranium cycle with low enriched uranium. Fig. 16 presents the generation of the nuclear working libraries for use in the MOCCA System. At present the transition is made from the ENDF/B4 (Evaluated Nuclear Data File B Version 4) based working libraries to libraries based on JEF-1
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- 14 -
(Joint European File Version 1 ) . These data libraries are validated by comparing calculational results with experimental and operational experience gained with the experimental facilities KAHTER and the HTRs AVR and THTR. Additional experiments will be performed with the critical facility PROTEUS. Fig. 17 gives a simplified overview of the MOCCA system. The modules of the MOCCA system may be combined in a highly flexible way by the DRIVER routine SPIDER. The modules are communicating with each other only via standard interface files. These standard interface files are produced by an input processor or the modules themselves. The most important modules are: - SPAX:
Cross section generating code using zero-dimensional diffusion theory. In the resolved resonance region mixtures of different fuel and absorber elements with grain structure may be homogenized by use of an equivalence principle.
- QDEIS:
Two-dimensional diffusion program with SchererGerwin treatment of the large void above the pebble bed. - MOSTE: Program for calculating the temperature distribution using two-dimensional gas kinetics, an empirical pressure loss equation and one dimensional heat conduction. - MALAGA: A code performing the depletion calculation. - COLA: A code which models the fuel circulation with and without: recirculation of spherical elements. For special three dimensional problems the COMMIX/STINT program codes are available at Interatom. For thermal hydraulic calculations all HTR partners use codes on the basis of the THERMIX program.
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- 15 3.8
Status of Fuel Element Testing The qualification of fuel elements for future HTR-plants comprises the tasks: -
Establishing of advanced manufacturing processes and quality control methods, Irradiation testing of representative fuel elements, Accident simulation testing of irradiated fuel elements.
Spherical fuel elements with D02~TRIS0-partides have been irradiated in several test reactors and in the AVR. The AVR was loaded with large quantities of production fuel elements (approx. 50 000), which is a sound statistical basis for the evaluations. The present status of the irradiation experiments is shown in Fig. 18. It is demonstrated that the max. fast fluence/burnup of discharged HTR 500 and HTR-Module elements are well covered by the irradiation tests. In all these experiments no inservice failure of coated particles has been observed. Accident simulation tests at elevated temperatures with irradiated fuel elements provided important data for safety and risk analyses. Isothermal heating at temperatures up to 2100°C and ramp tests up to 2500°C (50 K/h) verified the retention capability of TRISO fuel under extreme thermal conditions. Fig. 19 shows the measured release of the noble gas krypton-85 during thermal heating, which is also representative for the radiologically more important nuclide iodine-131. At 1600°C the heating time can be extended to 500 h without a measurable increase of noble gas or iodine release. The measured cesium-137 release is given in Fig. 20. Cesium release exceeded in all cases the krypton rate because of the diffusion through the coating.
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- 16 -
The conclusions of the accident simulation tests are: During all design basis accidents (T < 1600°C) the integrity of the TRISO coated particles is maintained. In hypothetical events (frequency < 10~ 7 /a) the majority of the fuel retains the fission products to a large extent. * * *
Summing up, we can say that the HTR 500 and the HTR Module are good candidates for nuclear energy production and for chemical processes. The development of the HTR reactor allows to start the construction in the 90s. It is, therefore, a special challenge to convince the decision-makers to use the advantages which the high-temperature reactor system can offer. The R & D work has been sponsored by the BMFT, MWMT and the Swiss Government. Mayor contributions are made by KFA and the Paul Scherrer Institute.
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HTR 500 - Nuclear Power Station Longitudinal Section
Spent fuel building
nm
Reactor building
Turbine hall
Reactor auxiliary building
+ 33,5
+ 12,5
30,0
,
48,0
. I-
42 5
'
12,5
66,0
199,0 • • - • • •
—
..
.
.
^
89.14-1
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HTR 500 Reactor Pressure Vessel with Internals
Steel fibre concrete
Prestressed concrete reactor vessel Incore rod Reflector rodAuxiliary heat exchanger
Main circulator Main steam line "^Feedwater line Steam generator Thermal shield
Reactor core Graphite reflector Auxiliarycirculator -Fuel element discharge pipe
89.14-2
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Primary Circuit of an HTR - Module 0 9400 mm
1 2 3 4 5 6 7 8 9 10 11 12 13 14
nm
Pebble Bed Pressure Vessel Fuel Discharge Small Absorber Balls Reflector Rod Fuel Loading Steam Generator: Pipe Assembly Outer Shroud Feed Line Live Steam Line Blower Hot Gas Duct Surface Cooler Insulation
89.14-3
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HTR - Module Section through Reactor Building
nm
Reactor
Steam Generator
89.14-4
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1\
Research Program " HTR Design Criteria "
SECTION A Technical saftey boundary conditions! Protection goals
SECTION B Metallic components Material assessment
SECTION C Prestressed Concrete pressure vessel Prestressed
concrete
SECTION D Graphite structural components Material properties non-irradiated
structure Safety principles
Manufacture and inspection
Liner system
Material properties irradiated
Integrity concept
Physical and mechanical data
Vessel closures
Corrosion
Life fraction rules
Heat insulation system
Loads and stress levels
Stress categories, failure modes
Calculation analysis
Dimensioning, stress analysis
Construction 89.14-5
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Core Structure of a High-Temperature Reactor
Ait n i K
89.14-6 B-l
2-D-Ring Model of the Side Reflector with Pebble Bed
n • i\
89.14-7
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°\K
10 MW He/He U-Tube heat Exchanger
Secondary Outlet
ii
Secondary Inlet
Operational Data Primary Row Rate
Support Plate Central Duct Cold Gas Header
11\
3,0 kg/s Temperature 950/293 °C Pressure 39,9 bar Diff. Pressure 0,5 bar Power
Secondary 2,9 kg/s 900/220 °C 43,5 bar 1,0 bar
10,1 I
Outer Annulus Pressure Vessel Hot Header
24980
Number of tubes
180
Tube Dimension
0 20 x 20
Tube Material
2.4663 (Nicrofer 5520 Co)
Vessel Material
1.6368 (WB 36)
Structure Material
1.7380 (10CrMo910)
1.6311 (20 MnMoNi 55) 1.5415 (15 Mo 3)
Insulation Hot Gas Central Duct Mixing Device
89.14-8 B-l
Comparison of Design and Test Data of the U-Tube Helium Heat Exchanger
t
1000
s
600
a a>
nm
800
est
(Calculation 400
200 10
15
20
25 30 Tube Length [m]
TEMPERATURE PROFILE OF THE U-TUBE HEAT EXCHANGER Primary Side
Load 1 0 0 %
Temperature • •
Secondary Side
Design
Test
Design
Test
950 293
950 306
220 900
217 894
39,9
40,0
43,6
43,3
[°C]
Inlet Outlet
Pressure
[bar]
Pressure Drop
[bar]
0,50
0,40
1,0
0,86
Mass Flow Rate
[kg/s]
3,0
2,97
2,9
2,71
Helium Velocity
[m/s]
17,6
Overall Heat Transfer Coefficient [W/m 2 K] Heat Transfer [MW]
17,5 Design 446 10,24
42,6
45,6 Test 442 9,53
89.14-9 B-l
lOo
Primary Hot Gas Duct with Graphite Liner and Rbre Insulation
if
8000
Cross-section A - A
16
Pressure vessel Support structure Insulation Liner Displacement body
Operational Data
Detail B Axial support element —-~ —~—
^ AI2O3
Hot gas side 3kg/s Row rate: Temperature: 950 °C Pressure: 40 bar 20m/s Velocity:
Cold gas side 3kg/s 293 °C 39 bar 8,5 m/s
Detail C Radial support element Dimensions and Material Displacement body: Liner: Support structure: Pressure vessel:
a 660 x 4 a 880 x 50/30 u. 10 a 1020 x 25 a 1130 x 25
1.4876 (incoloy 800 H)
Graphite/CFC 1.5415 (15 Mo 3) 1.5415 (15 Mo 3) 89.14-10
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r°
n I
Primary Hot Gas Duct with Graphit Liner and Rbre Insulation 970 960 950 Hot Gas Liner 940
L_—.
Temperature
—<
-Graphite
CFC-
TEST
C°C]
•
Hot Gas
950
961
•
Cold Gas
293
327
Mass Flow Rate
[kg/s]
Pressure
[bar]
Helium Velocity
[m/s]
3,0
2,9
40,0
38,0
•
Hot Gas
19,1
19,6
•
Cold Gas
9,1
9,5
36,0
12,0
3,0
1,0
Heat Exchange per Meter
320
DESIGN
[Kw/m]
Temperature Drop per Meter [°C/m]
A B C 89.14-11
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Cross Section of the Prototype Circulator MALVE I
1111\
Top
Bottom
89.14-12
ZQ3
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Retainer bearing test facility with active magnetic bearings
89.14-13
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Location of the Dissimilar Weld
n
11\
Service data Primary circuit 'Secondary circuit medium : helium temperature: 700 °C pressure : 5 MPa
water steam 535 °C 19 MPa
Dissimilar weid
Life steam pipe - * - Turbine
Feed water pipe —* Preheater
89.14-14
Zos
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Structure of the Program
niil
Dissimilar weld X 20 CrMoV 121/X 10 NiCrAUi 32 20
I
1
1B Service loading constitutive laws
Fundamental tests
c
t
1°
Tests with complex loading
Calculation complex loading tests
h*
Calculation for component tests
Component test MINERVA G Verification of life time integrity J Surveillance concept
Code case
89.14-15
2o&
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1 1
Code Systems for HTR Nuclear Core Design
n
in
Data Source File
Data Processing System
Working Libraries for Various Application Areas
Modular Code System for Core Design
89.14-16
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MOCCA Modular Core Calculation System INPUT
MODULES
OUTPUT
CROSS SECTIONS Module: SPAX
DIFFUSION Module: QUEIS
TEMPERATURE Module: MOSTE
DEPLETION Module: MALAGA
FUEL CIRCULATION Module: COLA
89.14-1?
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Irradiation Experience with TRISO - Particles for HTR - Fuel Elements
mil
10
9-
o •o
8 o
oo
o
7-
o o A
8
5-
I
"(9
0
LL
to
3-
£
HTR-Modul
= o :AVR 1-
oooo o o o 50
100
150
Burnup (GWd/t) 89.14-18
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Kr - 85 Release from Fuel Elements with UO2 TRISO Particles
10 10
20
30
Heating time [h] -
89.14-19
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ZAO
Cs -137 Release from Fuel Elements with UO2 TRISO Particles
o JO) 0)
cc "(5 o
S LL CO
8
100
200 300 Heating time [h]
400
500 89.14-20
lAA
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XA0101488
HIGH TEMPERATURE TECHNOLOGICAL HEAT EXCHANGERS AND STEAM GENERATORS WITH HELICAL COIL ASSEMBLY TUBE BUNDLE
Korotaev O . J . ,
Mizonov N . V . , N i k o l a e v s k y V . B . , Nszarov
E.K.
1). Centralnii Kotlo-TurMnnyi Institut, 194021, Leningrad, Politechnicheskaja ul. h. 24. 2 ) . Leningradskii Politechnicheskyi Institut, 195251, Leningrad, Politechnicheskaja ul., h.29. 3). Gosudarstvennyi Institut Azotnoj Promishlennosti, 109815, Moscow, Chkalova ul., h.50.
B-2
INTRODUCTION Analysis of thermal hydraulis characteristics of nuclear steam generators with different tube bundle arrangements and waste heat boilers for ammonia production units was performed on the base of operating experience results and research and development data. The present report involves the obtained information. The estimations of steam generator performances and repair-ability are given© The significant temperature profile of the primary and secondary coolant flows are attributed to all steam generator designs. The intermediate mixing is found to be an effective means of temperature profile overcoming. At present the only means to provide an effective mixing in heat exchangers of the followng types: straight tubes, Field tubes, platen tubes and multibank helical coil tubes (with complicated bend distribution along their length) are section arrangements in series in conduction with forced and natural mixing in connecting , lines* The heat exchangers with tube bundles in a from of helical coils (with small radius of coiling) are characterised by the unique possibility to from several effective mixing zones inside tube bundle along the primary coolant flow. The tube bundle consists of undressed assemblies manufactured as one-to-three bank multipass helical coils with several strongly twisted crossbars placed between the helical coil sections along assembly length. The crossbars act
as mixing
intensifier and separating device simultaneously.
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The experimental and theoretical computations are carried out to justify the steam generator design feasibility. The base advantages achieved by an effective intermediate mixing coupled with the design feature (tube bundle assembly arrangement) allow to consider the proposed steam generator * and heat exchanger designs for NEPPHG and NEP as competitive ones with other designs. I.
Main features of processing heat exchangers for ammonia production units
The pilot-commercial nuclear energy technological plant for combined ammonia and electric power generation is one
of
variants of usage of high temperature gas cooled reactors
de-
signing in the USSR. High potential thermal energy, carried by helium coolant of intermediate circuit from nuclear reactor to ammonia production unit, is used in hydrocarbon vapour conversion and steam generation processes for hydrogen containing gas producing. The generated steam is needed for nitrogen-hydrogen mixture compressor driving (1, 2 ) . The layout of the steam generation system in the ammonia production
unit is shown in Fig. 1.
Thee feed-water flow of 350 t/h is delivered by pump 1 through low pressure heaters 2 (heated by gas, steam condensate or nitrogen-hydrogen mixture) to deacrator 3. Then the
water
NEPPHG - nuclear energe plant for power and heat generation. NEP - nuclear energy plant. B-2
Fig 1 The steam generating schematic diagram 1 - pump; 2,5—7» - heaters; 3 - deaerator head; 4- - feed pump; 8 - steam header; 9-11 - waste-boilers (1 stage after mine methan conversion, 11 stage after carbon oxide conversion); 12-13 - helium evaporator and superheaters; 14 - secondary boiling tank; 15 - turbine for driving of gas synthesis compressor; 16-17 - condensing turbines for air, natural gas and ammonia compressor driving; 18 - the thrott .ling facility; 19 - Condeser; 20 - Condensate injector.
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heated to 1O2°C is pumped by high pressure pump 4 to high pressure heaters 5-7 (heated by converted gas, nitrogen-hydrogen and nitrogen-hydrogen-ammonia mixture). After that water heated to 250 C flows to steam header 8, from it
the water, at
the
saturation temperature 317°C, is distributed to waste boilers 9-11. Evaporators 9-10 are located along the primary coolant ftcrf behind the methane-and carbon oxide conversion units where conversion process is followed by heat release. Evaporator 12 is heated by helium
intermediate circuit. The saturated steam
from steam header is supplied to superheaters 13 with helium heating. Then, the steam, superheated to 500 C, at pressure 11 MPa is delivered to turbines and is also used for technological needs. Thus, the steam generation system in ammonia production units is equipped by the water heaters and evaporators, heated by converted gas, and evaporator and superheaters, heated
by
helium. The selection of the above described equipment may be based on the available designs, used in steam generating system with fire heating or on the alternative designs, developed for electric power needs for example. Moreover it is considered at present the scope of the steam generating system modification by using once-through helium steam generators and waste heat boilers in ammonia production unit. The operating experience of the existing ammonia production units revealed the waste boilers to be the least reliable component of steam generating system. The waste heat boile'r tube bundle consists of Field tubes, heated by the converted gas with inlet/outlet temperatures 1000/6000C at pressure 3,2 MPa,Pig.2.
B-2
45
3 Pig 2. Waste-boiler layout 1 - Feed water; 2 - Gas? 3 - Cooling water; 4 - Steam-water emulsion; 5 - Gas,
B-2
Swelling
and disrupture of the outer Field tubes under the
tube wall overheating action was observed during the operation. Ip may be attributed to the nonuniform distribution of water flow through tubes and gas flow through intertube space, resulted in the separate tube overheating at high heating
gas
temperature. Taking into account the shown and other disadvantages of available heat exchanger types for steam generating system, a problem is arised to develop an alternative design of heat exchangers, including the development of the new equipment
with
helium heating for nitrogen industry. 2.
Large temperature differences through tube bundles
In foreign steam generator engineering for nuclear energy plants with sodium and gas coolants it is used preferable helical coil tube bundle (in a form of multibank multipass helical coil) instead of platen and straight ones on the base of engineering-economic considerations. By contrast with it the tube bundles with small radius of coiling are used in the USSR. It is a forced decision and the above mentioned design is inferior in multibank helical coil one. In this connection the necessity is arised to evaluate the promising concepts of these two steam generators designed for electric power generation and technological needs on the base of
designing and research experience. The most significant operational data and information
on
structural materials published in scientific works on AGR Hey-
B-2
sham I and Hartlpool I steam generators with multibank helical coil heat exchangers (3, 4 ) . These steam generator heated by COp (65O°C and 5 MPa) are characterized by high unit power - 250 Mwt (th) and standard steam-water parameters - 157/543 C and 16,9 MPa. The tube bundle consists of 285 helical tubes forming 19 ring banks around the central displacer. To compensate the hydraulic irregularity effect on heat exchange rate along the ring banks, a hydraulic control of feedwater flow distribution by an orifice was carried out. Steam generating tubes of economizer section are manufactured from carbon steel, the tolerable tube wall temperature is 350 C. The evaporator section is manufactured from ferritic steel 9Cr - 1 Mo, the tolerable wall temperature is 500°C, the superheater section - from austenitic steel, the-tolerable wall temperature is 65O°C.
An additional restriction is imposed
steam temperature of the last section. Its temperature be more by 30°C of the saturation one for preventing
on should
water
drop carry - through superheater and austenitic steel corrosive cracking. The steam generator operating experience revealed the following: The significant temperature difference along the ring banks were observed, that was not predicted at the design stage, Fig. 3. The temperature differences... followed by the available tolerances on structural steels allow to bring the plant to output not more than 64% from rated one (Pig. 3, compare curves 1 and 2)?when using the initial flow diagram of feed water dis-
8
B-2
480 460
*r
V—X'
a
440
A
r 38(?{ 56O 500 X
460
X
IT
X
;
*
\3
y
42a
X
*
420 400
f
3
f
7 9
//
*?
?
Fig. 3 Tube temperature distribution along the ring banks 1 - 70% load\ 2-64% s.w., 100% gas j 3 - 78,5% load, new profiles Heysham 1, 4 - 73* l o a d § Hartlepool 1. 5
ZZo
B-2
tribution, and to output not more than 78,5% from rated one, when using new throttle devices. Temperature differences for any of 16 steam generators were appeared to have an own nature and value (compare curves 1 and 4 ) . The results of suitable excecuted computations showed, these temperature differencies not to be connected with deviations of tolerances during tube bundle manufacturing process. Thus, it should be noted, that there are random geometry deviations of multibank helical coil tube bundles from designed one owing to the specified manufacture tolerances and quite possible development of residual deformations during this
tube
bundle to be under service. The tube bundle deformation
re-
sults in. such intertube space flow distribution through the ring banks, that being relatively stable for each steam generator, the flow distribution is changed randomly from one steam generator to another. The above described phenomena may be the main reason of unsatisfactory prediction of the new throttle devices on the deterministic mathematical model approach
to
heat mass transfer in tube bundle. The mathematical model described in the reference (3) was not verified by the experimental data of the work (4). Moreover the steam temperature decrease caused by nonunifrom heat load on tubes may be turn out an important factor even in view of a promising use of more resistant structural materials. The mechanism of steam temperature difference development and mean temperature decrease at the steam generator outlet are given in Pig. 4. The tube bundle geometry deviation from design one causes a disturbance of ratio of coolant
10 B-2
224
Fig.4 Mechanism of large temperature difference development and steam mean temperature decrease due to the hydraulic irregularities.
II
212-
B-2
flow on aide of intertube space, falling at one tube, to coolant flow inside the tube. Spatial response of a multitube steam generator on this input disturbance may be plotted analogous with diode detecting by using mono tube steam generator characteristic, that is the steam outlet temperature dependence of isolated steam generating circuit on the coolant flow ratio. The selected work point A corresponds to the power steam general.orn of gas-cooled nuclear power plants, and point B, approaching to the primary maximum coolant temperature corresponds to the nuclear energy plant for power and heat generation and steam generators for nuclear energy plant with fast reactor. outlet Tn the first case the significant temperature difference is arised at some decreasing of mean steam temperature. In the second case steam temperature decreasing may be large enough owing to the strong nonlinearity of steam generating
circuit characteristics in point B. Shutting off the damaged tubes and deposit development in-
side tubes may give rise to negative effects. The only method to equalize heat load on tubes of the multibank helical coil steam generators is practically the hydraulic control of feed water flow distribution. It should be realized on the base of direct temperature measurements separately for each steam generator. The operating experience of the AGR reactor steam generators with platen heat exchangers (5) showed that the tube bundle temperature measurements for revealing the temperature processes involved are quite complex and expensive ones.
12
B_2
It seems to be even more undesirable to replace periodically the throttling
devices on the existing AGR steam generators
or to equip the steam generators by a control system of feed water flow distribution
in the tube bundles as in the case of
THTR-300 reactor plant. If the above mentioned problem will be neglected, the rated service life and safety operation of steam generator, design outlet parameters also can not be assured.
and
The above
mentioned factors should not be considered as arguments against the multibank helical concept in favour of platen and oncethrough designs. The real approach to this problem allows to consider thermal hydraulic irregularities of random or deterministic natures (Pig. 5) as inherent features of shell-tube heat exchangers (including the steam generators) of any known tube bundle
ar-
rangement. It is confirmed by the world wide development ' of deterministic circuits or hompgenious models for two- and three-dimensional thermal hydraulic analysis of heat exchangers with different coolants, various tube bundle arrangements and mutual coolant flow circuits. The first attempts of heat exchanger probabilistic modeling are given in references (6, 7 ) . They were aimed to evaluate such a static tube bundle characteristic as an equivalent tube deflection on the base of experimental date for straight tube sodium - sodium heat exchangers.
In the intermediate sodium - sodium heat ex-
changers of Phenix nuclear power plant, the sodium temperature difference at the tube outlet of the second circuit sometimes reached 32°C, at the design value of 15°C (8).
13
B-2
7
8 Q
10
Fig, 5. Hydraulic irregularity development 1.- Determinated irregularities; 2 - The side inlet and outlet; 3 - Single side supply; 4- - Sealing leakage} 5 - Leakage along the shell; 6 - The distributing header; 7 - Random irregularities} 8 - tube bundle manufacturing inaccurasies; 9 - tube dimension tolerances; 9 - tube deflection.
14 B-2
IIS
This temperature difference was computed on the results of tube bundle hydrodynamic experimental investigations
on
full-scale model. The intertube space temperature
differences.., insufficient
rate of heat exchange and tube deflections in straight
tube
bundle of high temperature intermediate helium - helium heat exchangers are noted in references (8). The large temperature differences
between the tubes of
45 MW experimental straight tube steam generator with sodium heating
are discovered in the 162 tube economizer-evaporator,
sections (10). The tube bundle geometry deviation from design one
is
found in a platen heat exchanger steam generators (5) and in a helical coil (with small radius of coiling) steam generator (11, 12). the
Experimental data analysis showed mean steam temperature decrease by 15-2O°C at helium parameters: 700°C and 4 MPa (work point A, Pig. 4)t°&e attributed to the tube bundle geometry deviation from design one. 3.
The stabilizing
effect of coolant mixing
The coolant flow mixing along the tube is an effective means for decreasing the sensibility of heat exchanger
and
steam generator outlet characteristics to the thermal hydraulic irregularities. However, the natural mixing
efficiency appeared to be in-
sufficient to equalize the tube thermal load at the qui'te large tube number (platen tube bank, ring banks in package). The sections arranged in series coupled with the and natural mixing in connecting
forced
circuits is an effective
15 B-2
coolant mixing means for all known tube bundle arrangements (straight tube Field's tube, platen and multibank helical coil). The tube bundle, consisting of helical coils (with small radius of coiling) is the exclusion from this rule. cently,
:
Until re-
: of such tube arrangement was inferior
to the multibank helical coil one and was considered as a forced decision. The received limited experimental data may be as it seemed to enhance this estimations (11, 12). Indeed, the experimental data static processing executed according to ZMEI PG programme on three-dimensional thermal hydraulic analysis of these steam generators (13), showed the significant equivalent deflection of 3mm value for helical coil modules of experimental section of st-eam generator. The characteristics of this experimental section are given in references (11, 12). The temperature differencies of experimental section, predicted by programms (with deflection modules being included) do not exceed the temperature differencies of multibank helical coil .AGR steam generators (taking into account deflection modules too). But if deflection modules are shut off, the received results would became unsuitable (Fig. 6, the left part). On the other side, this arrangement is a
unique means for
providing primary coolant forced mixing in the existing or specially formed zones between the helical parts of coil module. To provide an effective intermediate mixing, the strong twisted crossbars were installed in the technological gaps. Thus, the heat transfer assembly (module) arranged in helical
16 B-2
St&e \
o
9MMo
x
\
Jar
\
•dieom -Jiweiheai
S3-D-D-D
700
-0—0-
H
31
2 21 4
//
/ /
Sf
Fig. 6 Cell temperature distribution at coil module shut off X - metall; o - outlet steam; A - steam; D - outlet helium
17 •
r i
B-2
coil tube section of interchanging length ensures the heat transfer between the primary and secondary coolants and twisted crossbar sections (Pig. 7 ) . At the hexagonal packing geometries of heat transfer assemblies in tube bundle( Fig. 8) the twisted crossbars
form
several chambers of the effective intermediate mixing of primary coolant along the tube bundle and can act simultaneously as a separating device with fin-to-fin contact. The mixing efficiency in these chambers is 10-20 times more than the natural one in the intertube space of helical coils (14). The mean mixing efficiency increases by 5-10 times at twisted crossbar overall length, equal to 25-50% of the tube bundle one. The crossbars installed at the tube bundle inlet and outlet smoothes the inlet irregularities and the primary coolant temperature field at the tube bundle outlet, when the
damaged
assemblies are shut off. Inspite of the worse initial parameters (without forced mixing) the helical coil (with small radius of coiling) steam generators with several chambers (5-7) of intermediate forced mixing along coolant flow appear to have basic advantages over the multibank helical coil tube bundle steam generators for example. Three dimensional thermal hydraulic analysis executed according to ZMEI .PG programme showed intermediate mixing to be effective enough to reduce to acceptable level such negative disturbance effects as deviation of tube geometry from design one and shutting off coil modules. 18
VL°\
A-A
B-B
Fig. 7 Arrangement of undressed mini helical coil heat transfer assembly. 19
B-2
Srff
oo < o o OOI'OO
o o <> o o O O ' > OO 0 0 ) 0 0
Fig. 8 Tube bundle assembly arrangement.
20
23/1
B-2
To illustrate the above phenomena it is given in Fig. 6. The profiles of steam temperature (tubes and helium in cells along the main diagonal of hexahedral shell when the angular feed water assembly (61 assemblies in total) is shut off. In the left part of the figure it is given the corresponding temperature profiles of natural mixing in the tube bundle. The profiles of the forced mixing for the six high levels are given in the right part of the figure. As it is seen from the present example of computation, the development of intermediate mixing
allows to assure the
constraints imposed on the AGR steam generator structural material, even in the case of such strong disturbance, as an angular chamber shutting off in the multimodular tube bundle. Moreover, the effective mixing has a stabilizing effect on steam generating tube static stability (taking into account deposits)
(15) and on reliability of prediction by mathemati-
cal modeling (taking into account the uncertainty of reference data and the static deviation of tube bundle geometry from the design one). The reliable prediction (and reproducibility) of the steam generator parameters in operating long life conditions allows to refuse in principle from a detailed temperature control and a control of feed water flow distribution in tubes. 4.
Arrangement of multibank helical coil tube bundle or assembly
Helical coils (v/ith small radius of coiling) are often considered as singlebank (one or multipass) helical coils, which are opposed to the multibank (about 20 banks) multipass
21
helical coils formed, multibank helical coil tube bundle. One- , mini- and multibank helical coil arrangements may be examined from the common positions. The requirement of identical heat loads on tubes of any, i-th ring
bank may be easily complied with the following con-
ditions: 1) The tube length should be uniform
L. - 1 , V CJT ) D . ) 2 -r (n. S ) 2 - idem, X
XX
X
(1)
Zx
where Iji
-
the number of coiling in one tube;
D.
-
mean diameter of coiling
n.
-
the number of tubes in bank;
S
-
axial pitch of tube spacing.
- The number of cross stream-lined tube banks along z axis should be identical Z± - 1 1 . n i - idem -
(2)
ratio of intertube space coolant flow falling at i-th ring bank) to coolant flow in tubes of this bank should be identical (or the corresponding pitch to tube diameter ratio).
_s
a-
=
idem,
where Sr> = rr (D. - - D.) - radial pi d._ - inner diameter of tube.
(3)
These conditions are complied with the following arrangement formulae: -
for mean diameter of coiling D. = 2S R . i;
i = M l n _ , M l n . 1..., M Q u t
3
(4)
where M. ^
1
ML,+^> M.^ -
-
first (inner) ring bank number;
-
last (outer) bank number;
for the number of tube in bank n± = A . i
:
where
(5)
for the whole numbers A = 1, 2, 3 ... ratio of symmetry
is satisfied exactly and for the fractional numbers - approximately. Then, the other geometric characteristics can be defined by the formulae: - each tube length
L -z. y
4 2S
/
+ Sz2
(6)
- the number of ring banks N
bank "
M
out - M in +
1
- the number of tubes in multibank helical "tube " 2 \ ""in
+ M
out' • "bank
— multibank helical coil length H = Z . S9
'
(9)
23 B-2
central displacer diameter D
displ.
== (2M
in - 1) . S R
(10)
-outer shell diameter D
sh =
(2M
out
+ 1)
• SR
Table 1 shows the formula
(11)
validity on the example of
multibank helical coil steam, generator with different coolants. The fractional coefficient A corresponds to the tube elongation owing to their number reduction and it is used for steam generators with downward steamjwater mixture in tubes. Rounding of the tube;. number' ring
bank
to the whole one
forms some asymmetry that may be compensated by the hydraulic control of feed water flow distribution. The given arrangement formulae are true both for singlepass helical coils with small radius of coiling (A = I; M^n = M
. = 1 ) , and for several- and multibank helical coils, that
makes their comparison easier. 5.
Heat transfer assemblies in a form of helical coils
An attempt was made to realize the constructive advantages of multibank helical coil tube bundle assembly and module assemblies in design, consisting of several minibank helical coils (with small radius of coiling). It is appeared, as it will be shown later, that 2-3 bank helical coils from 5 or 9 tubes are the optimal steam generating assemblies with central displacer. Inside of this displacer, the downcomer throttling pipes are placed as in the case of multibank helical coils.
24 B-2
23>s
Such module is essentially a minihelical coil, in which there are no obstacles for strong twisted crossbars to be installed between the helical tube sections. Mini helical coil steam generators compares favourable with the multibank helical coils in heat transfer surface compactness. As to repairability, production engineering,
flow
detection and installation, the possibilities to develop
a
wide range of the minihelical coil heat exchangers on the base of standartization within the system may gain even an advantage over the multibank helical coil owing to tube bundle assembly arrangement concept. Tube bundle assembly arrangement allow to realize in principle the flexible three'staged fail-safe concept in the case of the steam generator tube leakages. 1.
Shut off the wholly faulted assemblies quickly without
steam generator operating state loss. 2.
Inspection and plugging of separate damaged tubes dur-
ing the preventive repair, remotely or at the direct access with the first loop to be covered (medium steam generator
re-
pair) . 3.
Dismounting and replacement of separate nonrepairable
assemblies with the first loop to be uncovered (capital repair of steam generator). There are three alternative tube assembly arrangements in a form of helical coil. A.
Several (for instance, 19) singlebank coils (with small radius of coiling) are placed in parallel in the assembly shell, Pig. 9. If even one coil in such assembly is shut off, it would result in nonadmissable temperature shifts. 25
A-A
8-B
C-C
Fig. 9 Arrangement of heat transfer assembly from coil modulus
26 B-2
Therefore the assembly with damaged tube should be shut off quickly on steam-water and primary coolant sides only followed by replacement during repair. B.
In the assembly shell it is placed one severalbank
mini helical coil (from 19 tubes for example) allowing plug a tube without assembly operating state loss. In such
a case
it is advisable to connect each tube with upper part of distributing and collecting headers of each assembly with downcomer tubes to be placed inside of the central displacer, Pig. 10. The assembly should be shut off on the water-steam and primary the coolant sides, but the faulted assembly with limited number of damaged tubes may be repaired without replacement. C.
The undressed mini helical coil (Fig. 7) with a forced
mixing of primary coolant should not be shut off on the primary coolant side and faulted assemblies may be repaired without their replacement independent of the number of damaged tubes. Unlike the above mentioned dressed assemblies, the steam generator operating state loss at assemblies, being shut off quickly on water-steam side, is attributed to their mutual lay-out. Namely, according to the temperature shift conditions, it is allowed to shut off not more than 3 inner adjacent assemblies. If the mini helical coils are placed in a tube bundle of hexagonal geometry in addition to the inner ring cells inside mini helical coil, the triangular cells between the ajacent mini helical coils are formed. The circumferencial and angular cells are formed near the shell. Let us consider the conditions, when the triangular cells will be equivalent to ring ones without application of any symmetry providing units (for instance,
27
Fig. 10 Arrangement of dressed mini helical coil assembly.
28 B-2
displacers in triangular cells). Then, the arrangement concept shall be expressed as follows:
4P., d
ri
<
*
for any form of cell, provided that the hydraulic resistance coefficient does not depend on the cell form. In the case of the hydraulic diameter estimations, the perimeter Hi
can be regarded as a circle, corresponding to
the mean diameter of coiling and cross— section P. - as pitch to tube diameter ratio along the free-stream flow, that is,between such circles. If the gap between the outer coiU/KS of tubes of the adjacent assemblies should be equal to A coiling SR &
, the radial pitch
of
provides condition of equivalence of triangular 10.74 - d * A
and
(13)
ring cells. Proceeding from the relative radial pitch for multibank helical coil tube bundle taken to be S^/d = 1.5 *• 2.0, R o£ the increasing of the number coil banks over 4 is considered as unadvisable, the value of M
.
= 1 * 4 is preferred, but for
the mini helical coil with a displacer the value of M should be equal 2 * 3 . The displacer may be installed instead of the first coil bank (M , = 1 ) , then its diameter is equal: (14) R oux er = 3S
(
)
it worth of note, that changes of coil steepness from bank to bank results in variation of heat transfer ratio and hydraulic resistance coefficient inside the tubes. For threebank mini helical coil with a central displacer heat transfer ratio
is
29 B-2
1,2 - 1T4 times more (with variation of - 1-8%), and hydraulic resistance coefficient is 1,23 - 1,36 times more (with variation of - 4-6%) then for straight tubes. The central displacer part in two-three-bank mini helical ooil is equal to 10-20% of the whole tube bundle cross-section. The table 2 illustrates three possible assembly arrangements. These arrangements provide approximately uniform tube bundle dimensions and helium steam - generator (reactor VG-400) output. For A-arrangement, the assemblies correspond to the design, described in reference (19)• The comparative metal requirement for three assembly arrangements is related to the helical tube section vireight, which is identical for all cases, with
allowance for surface.
Table 3 shows the comparative initial metal requirement with
axiowance
for
the damaged tubes shutting off.
The table also gives the summary comparative characteristics of the different assembly arrangement steam generator with the account of the repairability during the service life. These characteristic
analyses based on the assumptions, that
10%
of tubes would be out of service during 30 year operation, and the preventive repair to plug some tubes and replacement
of
faulted assemblies should be performed once in a year. It should be noted that the equipment to shut off quickly the faulted assembly on the steam-water side provides the significant decrease of the unscheduled outages induced by the steam generator damage. It concerns all variants of assembly arrangements. Only for the undressed mini helical coil assemblies
30 B-2
(variant V ) , the assembly replacement would not required, plug:
of some damaged tube is performed during the pre-
ventive repair. In the rest cases (variant B and especially A) one assembly set is not enough for overall service life; that increases the summary metal requirement and results in the periodical tube bundle capital repair.
31
B 2
"
Conclusion Thug, the proposed high temperature heat exchangers and steam generators of new designed may be ilsed in chemical and petroleum chemical industries and for high temperature gas cooled nuclear energy-technological plants, where technological processes of heat transfer and steam generating are proceeding at 900
coolant temperature.
The heat exchange surface of mini assemblies from helical coils allo to clean the waste heat boiler heating surface by vibration without tube bundle strength decrease. The commercial inducing vibrating cleaners can be used for the periodic tube oscillatory motion,
by electromechanical vibrator (20). It seems
reasonable to consider feasibility of continuous vibratory cleaning with the help of tube vibration induced by a coolant flow. The twisted crossbars, located between the helical tube sections,can intensify the tube vibration induced by coolant flow in an intertube space. In this case this'phenomena may be considered as a positive factor but it requires, however, the development of cial distance facilities to prevent the tube damage due
speto
the vibration effect. The high repairability of heat exchangers and steam generators with heat transfer surfaces in the form of mini helical coil assemblies allows to decrease outage duration to minimum to when some tubes have been failed and reduce
the repair
ex-
penses.
32 B-2
The advantages of such tube bundle arrangement in comparison with multibank helical coil, straight tube, platen ones and so on, assure their thermal engineering safety during long life operation already at the development stage. That is mean the indentification assurance to the specified tolerance deviations of real operational characteristics arising due to the tube bundle geometry deformation, deposit development and damaged tube shutting off. These advantages are realized, by the unique possibility to provide the stabilization
of coolant mixing, that is
achieved by simple design features of the mini helical coil tube bundle, namely. In this case, the detailed temperature and hydrolic control of flow distribution during the operation may be excluded, in principle, what is necessary for the other designs, as a rule. Assembly arrangement concept of the tube bundle for heat exchangers, designed for the different purposes, allows to standardize and unificate heat transfer assemblies in a form of the limited bank of optimal geometries of undressed mini helical coil assemblies. It creates the base for different heat transfer ecmipment development on the base of unification and commercial standartization within the system. In turn this approach provides the base for automatization of production engineering, flaw detection, reliable reproducibility and prediction of main characteristics of standard components, the modification of installation and dismantling processes and repairability of tube bundles.
33 B-2
Development of the unificated system from mini helical coil assemblies allows to design and manufacture heat exchangers the and steam generators within wide range of operating conditions without additional expenses on the research and development work.
34 B-2
Table 1 Multibank helical coil steam generator parameters THTR 300 Unit power, MW Primary coolant: pressure/resist., atm temperature, C HpO: pressure/ resist, atm. temperature, C Tubes: number diameter, mm length, m full surface, m
2
Multibank helical coil A Coil bank number of ring bank Pinch: S R , mm Sn-, mm relative pitch o T> ft
D
KWU
AGR
Super Phenix
250
co 2
750 Na
39.5/0.39 250/750
200 He . 60/1.5 245/700
186/54
190/20
180/550
200/530
125 He
80
314
285
357
25 196
22x2.8
31.5
106
96
12 30
2200
2710
25x2.6 91.5 2570
1/3
1/2
9.11...23
17.18...39
1
1
6.7...24
13.14...29
15
23
19
17
38.33
31
60.8
45
30.8
35
40
34
1.93x1.27
1.80x1.36
1.53x1.23
1.41x1.59
7 O
mean relative pitch Diameter: displacer, m shell, m displacer part,% helical coils Number of bank along axis p 3 Compactness,m /m
1.37
1.50
1.57
1.56
0.65
1.04
0.67
1.125
1.80
2.45
2.98 t
13
17.4
8.28
9.4
4.9 10.0
2.65 18 10.9
268
57.5 16,17
270 50 18
250
321
38.7 3,4
42.7 19
Note. The present data were obtained for the design, modified on the base" of limited information and given formulae, therefore they can be differ a little from the rated one.
35 B-2
Table 2 Helical coil assembly arrangements Comparative configurations of heat transfer assemblies
A PG 94BO
B mini helical coil (dressed)
Number of:assemblies tubes in assemblies tubes in steam generator in whole Tube dimensions: length, rated, m Diameter, mm Assembly pitch: axial S ,mm radial mm Between coils, mm Diameter: displacer, mm coil/pass
19 19 361
37 9 333
37 9 333
40
43.4
43.4
16x2.5 20 36
16x2.5 20 36 308. 108x4 144/2 216/3 288/4
16x2.5 20 32.7 92
(wrapper) dress, mm Bank number along axis Z
500x4 194
108x4 144/2 216/3 288/4 316/4 191
flow Length: inlet section,m coils, m techno!. clearance twisted crossbars, m outlet section, m the whole tube bundle,m
2,64 0.70 3.88 . 0.3x6=1.8
2.70 0,70 3.82 1.8
65.4/1
Hydraulic resistance,kg/sm coils twisted crossbar the whole bundle : Metal requirement, % coils dress (wrapper) displacer twisted crossbars the whole bundle
1.1 7.48
1.1 7.42
0.262
0.210
0.262
0.210
100% 58%
100% 71% 23.6%
158%
195%
V mini heli cal coil (undressed
191 2.70 0.4 3.82 2.82 0.4
7.44 0.210 0.048 2.258 100%
23.6% 17.1% 141%
36 B-2
Table 3 Helical coil tube bundle assembly characteristics with allowance for repairability
Comparative configuration of heat transfer assemblies
A PG 90B0
B mini, helical coil (dressed)
V mini helical coil (undressed)
- Surface margin for faulted tubes shut off
1.21
1.1
1.1
Tabulated value of metal requirement at the steam generator start up conditions
1.23
1.38
1.0
Unscheduled repair possibility
0.4%
0.45%
0.4 - 0.5%
Necessary set of assembly with allowance for replacement of. damaged one during 30 years
2.68
1.22
1.0
Tabulated value of metal requirement for 30 years
3.3
1.7
1.0
..
37
B-2
REFERENCES 1.
Sjavrikov A.Ya., Gohgut A.P., Enkov V.I., Krugljansky V.Ya., Veranjan R.S., Nasarov E.K., Nikolaevsky V.B., Lctkina O.S., Krotov L.I., Nikolaeva L.Z. "The apparatus-technological equipment of intermediate helium loop of the nuclean power technological plant VTRG". Questions of Huclear Science and Technology, ser. AVET, M., 1984, issue I, p. 22-26.
2.
Stoljarsky A.Ya. "The nuclear-technological systems on the base of high-temperature reactors" M., Energoizdat, 1988, p. 152.
3.
Green G.H., Lis J., Hitchoock J.A. "Modelling of AGR boiler thermal performance". Nuclear Energy, 1985, vol. 24, N 6, 367-380.
4.
Kettle D.B.
The referruling of the pod boilers at Hartle-
pool and Heysham I advanced gas-cooled reactors". Nucl. Energy, 1986, vol. 25, N 6, 371-376. 5.
Morris A.1/7. "A technique for estimating steady state temperature variation tube-to-tube within an AGR boiler from measurement external to the pressure vessel". Boiler dynamics and control in nuclear power stations, BNES, London, 1979, 379-389.
6.
Gotovsky M.A., Efanov A.D., Misonov H.V., Firsova E.W., ¥uriev Yu.S. "The hydraulic nonuniformity effect on the liquid-metal heat exchanger intensity", Preprint, FEY-1131, Obminsk, 1981, p. 15.
7.
Blagoveschensky A.J., Mizonov N.W., Gotovsky M.A., Fedorovich E.D.,
Firsova E.W. Thermal efficiency of heat
ex-
changers: manufacturing inaccuracies and tube bundle defor-
38
B-2
mation effects. Heat Transfer Engineering, 1985, v. 6, IT 1, 52-57. 8.
Carbonnier F.L., Lapicore A., Solly J.A.
Ten years operat-
ing experience with large components of the Phenix plant. Past breeder reactors and trend, 1985, 2, pros. cer. IAEA-SM, Vienna: IAEA, 1986, 3-15. 9.
Bjurgsmjuller P. "Steam generators and heat exchangers of gas-cooled reactors. The engineering
development state in
Switzerland". Izvestija, the Academy of Science, BSSR, ser. physical-energy sciences, N 1, special issue, Minsk, 1983, p. 16. 10. Lannou L., Llory M., Quinet J.L., Vambenepe G.
Generateur
de vaper experimental 45 MW Stein Industrie, Rapport EDF-DER, HT/32/77.01.1977. 11. Bezlepkin V.V., Kalachev D.M., Korotaev O.I., Kuznetzov N.M., Lomaev S.Y., Mizonov N.V., Pedorovich E.D. "The problems of helium steam generator development for coal steam gasification units and. methan conversion". Questions
of
Nuclear Science and Engineering, ser. AVET, M., 1982, issue 2, p. 18-23. 12. Bezlepkin V.V., Korotaev O.I., Mizonov N.V., Simkin B.P., Fedorovich E.D. "The thermal hydraulic experiments of helium steam generator test section. In the book: Two phase flows. Heat transfer and nonsteady state processes in
the
power generating equipment components". L., Science, 1987, p. 215-221. 13. Korotaev O.I.
Mizanov N.V., Bezlepkin V.V. "ZMEY-PG pro-
gramms for three dimensional thermal hydraulic steam generator computations". Information List N 88-145, LCMTI, 1988.
39 B-2
14.
Bezlepkin V.V., Korotaev O.I., Koschevaja L.I., Mizonov N.V, "The intermediant mixing intensification in coil steam generators and heat exchangers", Scientific works ZKTI, issue 247, L., 1988, p. 11-21.
15»
Brown M., Layland M.W. Tube to tube excursive instability sensitivities and transients. "Boiler Dynamics and control in nuclear power stations". BNES, London, 1979, p.151-158.
16.
Henry Ch., Elter C.
"Thermohydraulic verification during
THTR steam generators commisioning, IAEA, Specialist meeting on technology of steam generators for gas cooled reactors, 9-12 march, 1987, Winterthur, 14 p. 17.
Elter C , Jurgens B., Schar G.
First operating experience
with the steam generators of the THTR nuclear power plant "
18.
•
, 24
p.
The power station as a complex of high temperature reactors - modulus. The concept of installation and radiation safety". Craftwerk Union AG, Interat GmBX, 1984, - p. 79-
19.
Mitenkov F.M., Goiovko V.F. , Uschakov .F.A., Juriev Ju.S. "The nuclear plant heat exchanger designing". M. "Energoatomizdat", 1988. - p. 296.
20.
Tchelokov Ja.M., Awakumov A.M., Sazikin Ju.K. "The cleaning of waste-boiler heating
surfaces". Energoatomizdat,
1984, p. 160.
40
B-2
XA0101489
A.L. Malevski, A.Ya. Stoliarevski. V.T. Vladimirov, E.A. Larin, V.V. Lesnykh, Yu.V. Naumov, I.L. Fedotov. THE CHOICE OF EQUIPMENT MIX AND PARAMETERS FOR HTGR-BASED NUCLEAR COGENERATION PLANTS INTRODUCTION Improvement of boat and electricity supply systems based on cogeneration is one of the high-priority problems in energy development of the USSR. Fossil fuel consumption for heat supply exceeds now its use for electricity production and amounts to about 30% of the total demands. District heating provides about 80 million t.c.e. of energy resources coserved annually and meets about 50% of heat consumption of the country, including about 30% due to cogeneration. The share of natural gas and liquid fuel in the fuel consumption for district heating is about 70%. The analysis of heat consumption dynamics in individual regions and industrial-urban agglomerations shows the necessity of constructing cogeneration plants with the total capacity of about 60 million kW till the year 2000. However, their construction causes some serious problems. ,The most important of them are provision of environmentally clean fuels for cogeneration plants and provision of clear air. The limited reserves of oil and natural gas and the growing expenditures on their production require more intensive introduction of nuclear energy in the national energy balance. Possible use of nuclear energy based on light-water reactors for substitution of deficient hldrocarbon fuels is limited by the physical, technical and economic factors and requirements of safety. Further development of nuclear energy in the USSR can be realized on a new technological base with construction of domestic reactors of increased and ultimate safety. The most promising reactors under design are high-temperature gas-cooled reactors (HTGR) of low and medium capacity with the
B-3
intrinsic property of safety. HTGR of low (about 200-250 MW(th) in a steel vessel), medium (about 500 MW(th) in a steel-concrete vessel) and high (about 1000-2500 MW(th)in a prestressed concrete vessel) are now designed and studied in the country. At outlet helium temperatures of 920-1020 K it is possible to create steamturbine installations producing both electricity and steam and hot water. If the helium temperature at the core outlet reaches 1120-1220 K, it will be possible to create a single-loop HTGRbased gas-turbine installation using waste heat for heat supply. The economic feasibility of creating industrial and heating plants with HTGR, rational fieldes of their application in cogeneration systems can be determined after complex optimization analysis of schemes and their main parameters considering the whole complex of really influencing factors in their operation. INDUSTRIAL AND HEATING PLANTS WTTH MODULAR-TYPE HTGR. The main requirement to nuclear sources for district heating is their safety resulting from the necessity of their more close siting to industrial and civil consumers. The basic host schemes of industrial and boating plants wi tli modular-typo HTGR are realized on the following principles: - process third loop;
steam
for
consumers
should
be supplied from the
- hot water for public services should be supplied from the third 1oop; - the three-loop schemes foresee a trapping loop. Fig. 1 "'presents basic heat schemes of industrial and heating plants based on the 200-250 MW(th) reactors. The considered level of initial steam pressure can be achieved by steam reheating or separation. The use of extracted steam for heating simplifies the steam generator design,- reduces the heating surface, and, as a result, increases its safety. The use of a once-through-type steam generator with high steam pressure increases requirements to feed-water quality. The costs on chemical water treatment can be decreased by the process steam production in a steamgenerating installation (SGI). Water of the heat network is B-3
heated in the two-stage preheater. If a set of the optimized thermodynamic and discharge parameters of a working medium is characterized by vector X, a set of schemes and variants of structural, load and functional redundancy by vector Y, and the share of capacity reserve in the district heating system accounts for R, the optimization problem of parameters and characteristics of HTGR-based industrial and heating plants is stated as follows: opt
(X.Y.R) P Qr = cons);
: min 3y
(X.Y.R)
Tt; Tp = idom; x""'n < X < X™"* ; Y e Y
R = min R [ H(X.Y.R) > H
]; H
F (X,Y) < F(X.Y) < F where Q
;
(X.Y.R) > H H (X,Y)
- reactor thermal capacity, T , T
at the reactor inlet and outlet; X . , X min
- helium temperature - a
feasible region
max
for changes in the independent parameters X; Y - a feasible set of schemes and variants of structural. load and functional redundancy; H ' (X,Y,R) -the index of reliable supply of * ** consumers with thermal and electric energy; F (X,Y), F (X,Y)a feasible region of dependent variables and their functions. Here the minimized functional are the total expenditures 3sum for the whole service life, that are discounted to the operation start, including: . - cost on uranium production, conversion, enrichment, fuel element production, processing and burial of waste fuel; - capital investment on plant construction; - cost on main and auxiliary equipment; - cost of installation works; - operating costs; - cost of all natural and labour resources used; - additional costs on provision
of
equal
energy
and social B-3
The reliability indices of cogenoration systems are calculated by the complex of hierarchically constructed models.
The
reliability indices of cogenerntion for each type of
product are
determined at
the upper
level.
The calculation method is based
on the probabilistic calculations and system coefficient
of
coefficient
of
supply
with
system
reliability indices.
product
operation
The latter
of
state analysis. The each
type
efficiency
are
and
the
taken
as
index is the probability of the
system output being not loss than that required by the load curve at any time. The middle level of the complex includes calculation of the structural regarding
the
reliability
structural,
individual elements. indices
is
based
on
of
combined
load
and
energy
installations
functional
redundancy of
The
calculation
method
for reliability
the
application
of the Markov model and
determines reliability indices that are interdependent electric and
heat energy production.
The reliability indices of
energy equipment elements are determined at the third probabilistic
model
constructed
as to the
for
calculating
level. The failure-free
operation is based on the? probability that the current loads will not exceed
their critical values. The required reliability
of oogeneration features
of
is
achieved
equipment
by
the
proper
choice
level
of design
elements, application of the structural,
load and functional redundancy of the plant and the use of system reserve. The program
of optimization
The problem of optimizing solved
at
individual
stage
1.
studies consisted of two stages.
the main parameters of heat schemes was
The
results
showed
steam pa ratno t. n rn accounted
that
the
for- 1.7. 5-18.5
level
of
MPa and 018-
838 K. The pressure of steam reheating was within the ranges 1.01.5 MPa at a temperature of 583-592 K. The extracted steam with a pressure of
8.0-9.0 MPa
was rationally used as a heating steam.
In the scheme with steam separation the pressure amounted to 0.10.15 MPa. At the pressure of the supplied steam above 1.8-1.9 MPa a
single-stage
installation extraction
is is
steam
extraction
realized, reasonable.
at Table
in
lower 1
the
steam-generating
pressures
presents
a
the
comparative analysis of heat schemes of HTGR-based
two-stage results
of
industrial and
B-3
heating
plnnts.
Table 1. Comparative charnct.eristic of industrial
and heating
plant r. Indices
Three-loop
Two-loop with
Two-loop with
with SGI
SGI and reheating
SGI and steam separation
Heat
s o u r c e s , MW
Electric in
capacity,
2x2 50"
2x250
102. 9
110 . 7
1.17 . 0
1J 0 . 0
117.2
12 3.5
136.4
143.4
150.0
956.7
1013.8
1065.1
2x250
(th) MW
denign
conid i t. ions in
nutumn-spring
cond i t aons in s u m m e r Yearly
conditions
e l e c t r i c i ty
production, Yearly
MWh
process
steam
produc t ion, thousand Yearly
t/year
heat
for p u b l i c
of
588. 7
588.7
676 . 7
676.7
676.7
production services,
MWh/year. Saving
588. 7
discounted
costs, million
rubles/year
The table shows
that
the
1 .92
0. 95
0.0
scheme
of
the
plant
with steam
separation is most- efficient . Stage
2
of
optimization
structural scheme of capacity.
The
studies was intended for choosing the
nuclear
studies
cogeneration
resulted
in
plant
and
the unit
the economically
expedient
structural, scheme of nuclear cogeneration plants (NCP) . • A double-
B-3
block scheme
with two
reactor units
with a capacity of 200-250
MW(th) and a turboset has economic advantages. A block-type NCP wit.h GTU, based on large HTGR. The prospects of using Power Installation
(NPI) are due
-reduction of capital -elimination of
the gas-turbine
unit
(GTU)
in Nuclear
to the following factors:
investment;
constraints on
possible uso of "dry" cooling
water consumption owing
to the
lower!;;
-achiovenient of high electric
efficiency nnd
heat
utilization
coefficient; -possibility
for
wide-range
produced without reducing Due
to
all
plants with
these
Complex
the
nuclear
cogeneration
the installation
type of NPI of a similar
ones.
optimization was
generation.
GTM-based
studies on
of
schemes,
equipment arrangement at NCP, Fig.l,
in the amount of heat
be considered as a perspective
and lechnico-economic type-as up-to-date
the electricity
facts
HTGR can
variation
carried
on
parameters and ways for the
whose
during
discrete programming problem
basic
scheme
decomposition
with
the
is
of
given in
continuously
application
of computer
modelling of all the main processes occurring in NPI / I / . The minimized whole lifetime comissioning, -costs
on
processing, secondary
functional, i.e. (3ncp) that
total costs
are discounted
on NCP during
to the
includes
the following components:
fuel.
uranium
(
manufactor ing
fuel
product! on
the
fuel
and
its
year of plant
enrichment, fuel
elements,
selling
the
bred);
-cost of the plant
construction;
-cost of the main and auxiliary equipment of NCP; -cost of installation works; -operating etc . ) .
costs
(salary,depreciation
Reduction of the variants based
on
the
use
of
to
the
marginal
electricity. Solution of the
and maintenance costs,
equal costs
similar
economic on
effect was
nuclear
optimization
fuel
and
problem
with
B-3
account
of
different
conditions
following peculiarities: costs
on
NCP
for
for
the total
each
NCP
operation
costs on
condition
of
NCP is
plant
has
the
the sum of
operation.
possibilities for NCP equipment operation were tested
in
The
all the
corid i t ions . For
manoeuvring
the
heat
load,
produced
for consumers, the
mcthode of by-passing gas-water coolers 4 and 6 '.(Fig . 2 ) ^ was used. Consideration was given to NCP with HTGR of a monoblock type of 1060 MW(th) with a pebble-bed
core,
that
operates
in uranium-
plutonium fuel cycle. The installation will be put into operation in 2010 and will operate in three modes
with combined
of electricity and heat in the form of hot water and Calculations were
production
steam.
made on the BESM-6 computer with the help of
software package "PEGAS" that uses detailed models of reactor and heat-transformation parts of NPI / 2 - 3 / . During technico-economic optimization of energy installation of any type it is advisable to its designs.
Within the
cover all
framework of
difficult and hardly advisable to parameter
change,
type
of
a
values of steam and hot water region
is
narrowed
the possible the problem
study
the
variants of solved it is
possible
range of
scheme, variants of arrangement, loads in
by preliminary
detail. The optimization
technico-economic studies on
NCP that persue the goals: - estimation of the possible range of cogeneration loads; - substantiation of the arrangement
type for
the equipment of
tho first loop; - determination
of
the
impact
of
cogeneration
load on the
parameters and type of the NCP heat scheme. The advisability of such that the
is caused
by the fact,
applied mathematical model of NCP has a large number of
independent optimized search
an approach
for
all
parameters the
factors
(up would
to
40) require
and
the overall
time-consuming
computations. At the stage of preliminary studies NCP was considered steam
generation.
(Scheme
in
(Fig.2
pressure and temperature at the reactor
had
no
steam
outlet were
without
loop. Gas taken equal
to f> MPa and 1223 K respectively). Height and radius of the core,
B-3
radius of micro-fuel elements, enrichment, of the loaded fuel, gas temperature
at
the
reactor
turbine and before the
inlet,
helium
second compressor,
pressure after the
coolant rate
in heat
exchangers and pipelines, height of "dry" cooling tower, diameter of its mouth and number of columns were considered as independent optimi zed pa r am e tni-n . The results
of studies
allowed a number of conclusions
to be
made (for greater detail see / 2 - 3 / ) : 1) Economic efficiency of NCP depends
on the
different
cogeneration load value Q
is supplied to the consumers dependence is
technological
O K follows:
in the form with
schemes of
(Fig.3). If heat
of hot
water only, the
Q < 300 MW.heat load factor
(heat load - to - reactor
thermal capacity ratio) s < 0.283, the
most complex
the minimum discounted costs; with
MW < Q c
scheme has
< 500 MW ( 0.283 < e < 0.472 ) the
300
preference is given
to the scheme with a regenerator and compressor, and with Q > 500 MW ( e > 0.472) the well
(Fig.3).
simplest scheme
becomes competitive as
If the scheme has a loop with a steam generator and
gas cooler, conventionally called second compression providing the
stage should
steam
"steam"
of helium
side), since
in the loop with a
a
one,
then the
be given up due to necessity of
high temperature
the low-pressure
as
after regenerator
(on
the required temperature of water
genera tor,that
is
determined
by the
parameters of generated steam, is sufficiently high. 2)
By
the
criterion
of
minimum discounted costs on NCP the
integral arrangement of all the equipment reactor in
the multi-cavity
strong vessel
than the separate arrangement of a (difference
in
the
amount
in the
of
first loop and
is more advantageous
turboset and
heat exchangers
discounted costs is 1-3 million
rubles/year). When a turboset is arranged in the same vessel with heat exchanger1,
IITGR and
the following two variants are equally
economic for 3ncp: a) with a two-flow
turbocompressor, located
in one horizontal
cavity under the core and heat exchangers; b) with
a four-flow
turbocompressor
located in two horizontal
cavities under the core and heat-exchangers. Heat
exchangers,consisting
of
four
sections,
are
located
B-3
nymmetrioally
around
the
core
preference should be given to the since in
in eight vertical cavities. But first variant
of arrangement,
the second
variant difficulties with the location of a
device for unloading
the spent fuel elements from the reactor may
B
ri se . 3) With
fcho
given hoot
s c h e m e of N C P t h e o p t i m u m
p a r a m e t e r s of
HTGR and equipment of a heat-trans format ion part depend on the value of Q (Fig.4). The behaviour of NCP electric capacity, coefficient of heat utilization and capital investment versus the growth of cogeneration load allows the following values of Q to be recommended: for NCP operation without considerable reduction of electricity production Q c = 200-300 MW. On the base of preliminary studies of NCP with combined production of electricity,steam and hot water, the heat scheme was selected with a regenerator, one stage of compression and integral arrongoment of the hent-transformation equipment of the first loop in the multi-cavity prestressed concrete vessel. According to this scheme, all the heat exchangers consisting of four sections, are located in pairs in accord with the results of preliminary studies: section of a regenerator - section of "steam" gas cooler; section of a gas cooler for heat productionRootion of a " termini* £ " cooler; a two-flow single-shaft turbocompressor is in the horizontal cavity under HTGR and heat exchangers. The program stages:
of
optimization
studies
consisted
of several
1) At the first stage, in the range of pressures {P ) and heatings (AT ) of helium in the reactor (P = 5-9 MPa, AT = 300600 K ) , joint optimization of HTGR parameters and heattransformation equipment was held with the use of detailed models of reactor and heat-transforming parts. At the given stage the most typical parameters of .steam and water in heat networks, cogeneration capacity, ratio between steam and hot water loads, duration of NCP operation in different conditions (Table 2) were chosen on the base of the analysis of a number of large energy consumers in the European part of the USSR by the values of loads and steam and hot water parameters, characteristics of yearly
B-3
load curves, etc (Table 2 ) . Table 2. Characteristic of NCP operation conditions (SG-steam generator, NWH-network water)
N
Operation condition
Duration of operation condition (hours)
1. 2. 3.
Design Autumn-spring Summer
110 4954 2936
Thermal capacity (MW) SG
NWH
215 215 150
185 170 130
The parameters of generated heat were given as follows: steam pressure - 1 MPa, temperature of feed water and steam - 423 K and 453 K respectively, pressure of water in heat networks - 1.5 MPa, water temperature at the reactor inlet and outlet, respectively343 and 423 K. Temperature of helium at the HTGR outlet was taken equal to 1223 K, and some constraints were imposed on the following reactor parameters: average energy intensity of the core Q < 6 MW/m ; temperature of fuel element centre T < 1523 K; temperature gradient of fuel elements grad (Tf ^) < 25000 grades/m; energy intensity of a fuel element q v = 3 kW/ball; fuel burnout B( = 150 MW day/kg. Besides the values enumerated in the previous sections (excluding the parameters, related to the second compression stage), gas pressure in the reactor was also optimized. According to the applied decomposition of the problem, the value of 3ncp was found as a sum of the corresponding values for reactor (RP) and heat-transformation (HTP) parts: 3ncp=3rp + 3htp. Studies have shown, that the value of 3rp (that is subject to the influence of pressure in the reactor, gas consumption, core size, indices of a fuel cycle) monotonously increases with the growth of P and AT in the studied range of their change. Cost of heat-transformation part 3htp have a distinct minimum in the region of gas heating values of 500-600 K in the reactor,
B-3
that is determined by electrical
the behaviour
efficiency
(Fig:.
of capital
5a).
Fig.
dependence of discounted cost of NCP on the rector. The
investments and
5b
illustrates
gas pressure
in the
minimum is observed at all the values of gas heating
in the reactor and is shifted from P r = 6.6 MPa at P
the
= 7.2 MPa at
AT
r
AT f = 600 K to
= 300 K.
I"
In
the
course
electricity,
of
steam
minimum
discounted
heating
AT
studies and
at
hot
costs
the
water
3ncp
given
typical ratios of
production
are
(Table
observed
2 ) , the
at the values of
= 600 K and gas pressure in the rector P p = 6.6 MPa.
2) The goal of the second stage was to study
the dependence of
the optimum technico-economic characteristics of NCP on the value of total
thermal load
and marginal
costs on
electric and heat
energy. Dependences of the optimum parameters of schemes and technicoeconomic
characteristics
load were obtained on
of
NCP
the base
on the value of total thermal
of optimization
studies carried
on . Fig. 6a illustrates the dependences of useful electric capacity of
NCP
and
dependences
capital
invest mnnts
(parameters
of
a
on
the
coolant
value
in
the
of
Qc . Other
nodes,
design
characteristics of equipment, rate of a coolant in
the pipelines
and
in
equipment
components,
temperature
heads
the
heat
exchangers, cost characteristics) are presented in a similar way. Analysis of the
impact
of
marginal
costs
was
performed by
maximum, medium and minimum values of marginal costs, typical of the economic regions in the European part of the USSR: South, Center and
Urals respectively. Marginal costs on steam load were
taken 20% higher, than marginal costs of calculations
on hot
water. The results
of the marginal costs on heat and total costs on
NCP for different values of Qc Table 3 and Fig. 6b.
and marginal
costs are
given in
B-3
Table 3. Economic efficiency of NCP at different coRonoration loads. 3disc Qc, MW.
Location of NCP
3ncp
3ncp+ 3disc
million rubles/year
250
Urals Centre South
20.5 30.87 33.42
400
Urals Centre "South
15.03 16.1 17.63
550
Urals Centre South
0.0 1.08 2.21
0.0
20.5 30.87 33.42
15.92
30.55 32.02 33.55
38.73
38.73 39.81 . 40.94
Tt is seen from the figure, that the discounted costs on NCP increase monotonously with the growth of heat load. Increment in the marginal component of costs leads to the distinct minima in the total costs. Values of Qc in the studied range of changes in the marginal costs on heat, at which the total costs on NCP have the minima, vary from 250 to 320 MW. Hence, the optimum value of Qc for the given type of GTU-based NCP with HTGR in the regions with minimum level of marginal costs on heat is equal to 250 MW ( the heat in this case the heat load factor is 0.234). In the regions with maximum level of marginal costs on heat the optimum value of Q accounts for 320 MW (heat load factor is 0.3). The dependences of NCP parameters on Qc similar to those given in Fig. 6a being available, the optimum load and any parameters and performances of NCP can be obtaned using Fig. 6b for the required area of NCP allocation.
B-3
CONCLUTIONS The optimization studios on Nuclear Cogeneration Plants with HTGR allow the following conclusions to be made. 1) NCP, based on the modular-type HTGR and steam turbine installat ions: - The results showed that the level of individual steam parameters accounted for 17.5-18.5 MPa and 818-838 K. The pressure of steam reheating was within the ranges 1.0-1.5 MPa at n temperature of 583-592 K. The extracted steam with a pressure of 8.0-9.0 MPa was rationally used as a heating steam. In the scheme with steam separation the pressure amounted to 0.1-0.15 MPa. At the pressure of the supplied steam above 1.8-1.9 MPa a single-stage steam extraction in the steam-generating installation is realized, at lower pressures a two-stage extraction is reasonable. - A double-block scheme with two reactor units with a capacity of 200-250 MW(th) and a turboset has economic advantages. 2)The block-type NCP with GTU, based on large HTGR: -With the integral arrangement of equipment preference is given to the variant with location of the one-shaft, two-flow turbocompressor in one horizontal cavity, located in the same vessel with HTGR and heat exchanger below the core. - If heat is supplied to the consumers in the form of hot water only, the dependence is as follows: with Q < 300 MW ( heat load factor £ < 0.283 ) the most, complex scheme has the minimum discounted costs; with 300 MW < Q c < 500 MW (0.283 < « < 0.472) the preference is given to the scheme with a regenerator and compressor, and with Q > 500 MW ( « > 0.472) the simplest scheme becomes competitive as well. - If the scheme has a loop with a steam generator and gas cooler, the second compression stage should be given up and for NCP operation without considerable reduction of electricity production Qcis recommended to be less than 300 MW. The following parameters of HTGR coolant are rocommended: P = 6.6 MPa, AT =
B-3
600 K ( the outlet temperature of heleum in reactor T p = 1223 K ) . - The optimum value of Qc for the given type of GTU-based NCP with HTGR in the regions with minimum level of marginal costs on heat is equal to 250 MW ( heat load factor is 0.234). In the regions with maximum level of marginal costs on heat the optimum value of Qc accounts for 320 MW (heat load factor is 0.3). -The relation between area of location, optimum cogeneration capacity and optimum NCP parameters are shown; the algorithm for obtaining the set of NCP parameters for any area of location is given.
REFERENCES. 1. Mathematical modelling an optimization of thermal energy units/ Popyrin L.S./ -M., Energy. 1978, 416 p../in Russian/. 2. Study of the impact of cogeneration load on the parameters and type of a scheme of NCP with HTGR / Malevski A. L., Vladimirov V.T., Naumov Yu., V.// Voprosy atomnoi nauk 1 i tehniki. Ser.: Atomno-vodorodnaya energetika i tehnologiya. 1988, vyp. 3. -p. 6-9,/in Russian/. 3. Impact of the equipment arrangement of the first loop on the parameters of GTU-based NCP with HTGR /Malevski A.L., Vladimirov V.T.. Stolyarevski A. Ya.//Voprosy atomnoi naiki i tehniki. Ser.: Atomno-vodorodnaya energetika i tehnologiya. 1988, vyp. 3. P. 1012/in Russian/.
B-3
I?
-w ^H5—I
T >7
Fig, 1, Basic thermal schemes of industrial and heating plants with HTGR 1 - reactor; 2 - steam generator, 3 - high-pressure cylinder, 4a - steam reheater, 4"b - separator; 5 - low-pressure cylinder; 6 - water heaters of cogeneration system; 7 - steam generating installation.
B-3
6s*
Pig, 2. Basic flow-diagram of GTU-based NCP with HTGR, 1 - HTGRj 2 - turbine; 3 - regenerator; 4 - "steam" gas cooler; 5 - steam generator; 6 - gas cooler for heat production; 7 "terminal" gas cooler; 8 - water heater of cogeneration system;
B-3
a) o
10 -
03 0)
H
I
S-
g
•H H H •H
B
-10 -
3. Competitiveness of NCP schemes versus the most complete scheme (Fig. 2) with different thermal load# 1 - scheme without regenerator, with one compression stage; 2 - scheme with regenerator and one compression stage; 3 - scheme with regenerator and two compression stages.
03
H
1 § •H r-t 1-1
•H S
« 350
O
Pig. 4. Electric capacity and capital investments on NCP with different thermal load for the scheme with regenerator and two compression stages. B-3
160
m
o a>
H
•H
o
•H
o •H H H •H
0 £5
eoo
S00
-a
0)
g
•H H H •H
28
E O
i7
15 Pig.
s 5
8
MPa
Dependence of a) capital investments and electric efficiency of NCP (I -P =5.10 MPa, 2 - P =7.15 MPa, 3 - P =9.20 MPa) and b) discounted costs on NCP on the pressure (P ) and heating (/iT ) of gae. =300K, 2 -
B-3
02
a> r-i
e o
•H r-i H •H
s
g
0) 03
§
•H H
P O !
200
300
400
500
Pig, 6, Dependence of a) capital investments and electric capacity of NCP and b) discounted costs on NCP with and without account of the marginal costs on heat (1 - Urals, 2 - Centre, 3 - South of the European part of the USSR) on the value of thermal load. B_3
XA0101490 Prospects for Application of High—temperature Holium Reactor ( HTHR ) to Provide for Power Nesds in Refineries and Petrochemical Plants
E.A.Feigin, E.A.Raud, E.G.Rom&nova, P.A.Panasenka, V.M.Nikitin
Crude oil refining and patrochem.'cal industry is among the most energy—intensive industries in the USSR. It holds the first place among branches of the heavy . ir.dustry in steam consumption; second place, fbl lowing ferrous metallurgy, in fuel consumption; third place, following ferrous and nonferrous metallurgy in electric power consumption. Crtcit? oil refining e.nd petrochemical processes themselves require energy equivalent to 6 - 14 X of the crude nil processed. The structure of power consumption in various plants in oil refining and petrochemical industry in the USSR cer* be •~'*eprt'v»en'ted as follows: 40 - 60 '.'. - as fuel j 30 - 40 X - an s'..u?am; 10 - 15 '/. " as electric power ( Bee Fig. 1 ) . Thus, about half of the energy consumed i«s used as fuel oil and fuel gas. Over the last years conridoruible? efforts have been made to try and reduce; the power consumption in the oil refining and petrochemical industries plants. Among the main anergysaving Measures t;re the implementation of new technologies, allowing t.o increase the heart products yields and reduce process temperaturesj higher efficiency in fuel i;se .in furnaefcs; maximum use of heat avi-.ilable in products streams, becirpr- insulation of vessels and piping, etc. Despite that and due to ths advent of heavier types of feedstocks power consumption in the oil refining and petrochemical industry may grow up to 14 — 20 '/. on cruds oil procsiis»exi. Refineries and petrochemical complexes arc? the p—ime consumers of energy in the industry. Energy requirements of B-4
refineries and petrochemical plants vary from 1OO to 8000 riwt ( heat ) with the most common requirements ranging from 700 to 1300 Mwt ( Fig.2 ) . Steam to plants is supplied from thermal power stations ( TPS ) whereas heat required for the process is generated in tubu' heaters, whera* n fuel oil or gas is burnt for the purposeIt should be notej that characteristic for the oil refining and petrochemical industries plants is a high rate of fuel oil use as compared to plants in chemical and fertilizer industries. Fuel oil consumption amour.tr to about 1/3 of the total fuel burnt in the refinery ( G-ss Fig.3 ) . Specific capacity of tubs haaters where crude oil and hydrocarbon products ^xrs exposed to h«?
Fig.5, which also shows that most of the energy consumed is within the 300 to 400°C temperature range. Of the high-temperature processes run at 800 - 900r»C the most important are pyrolysis and conversion. Products obtained in pyrolysis are basic feedstocks for the petrochemical industry. Conversion of hydrocarbon feed is dedicated to the production of hydrogen. Characteristic Q — T diagram of the petrochemical complex including low-, medium— and high-temperature? oroceeses is shown in Fig.6. Hidh-temperature processes in the petrochemical complex structure amount to at least 10 V.. In the near future due to the advent of sulfur and heavy crude oils the importance of high—temperature procR-jnes and especially of conversion will grow high. Hydrogsn produced in conversion is used to upgrade hydrocarbon products in hydrotreating and hydrods— . sul f urisation processes. In doihc so the higher ""is the? upgrading rate, the higher is the hydrogen consumption ( Fig.7 ) . x
h(? oil rt?f.lning and pertorrhemical industry of the U3SR ii. facing now the following main problems: 1. Upgrading of heavy residual fractions, mainly atmospheric residue, upgrading of sulfur and bituminous crude oil, and expansion of hydrogen production capacities directly p.«.iscciate»d with the upgrading. 2. Iropro\-ements in the environment protection in industrial areas. 3. Reduction in utilisation of hydrocarbon feedstocks as fuel. One of the promising ways o* solving the above problems is the use of HTHR to meet ths energy requirements in the refinery and petrochemical plant-"-. A number of problems nesd to be resolved in order to implement HTHR in oil refineries and petrochemical plants. One of the major problems is making nuclear power source more reliable and safe. The problem is still more complicated because of fire and explosion hazards associated with the operation of refineries and petrochemical plants. B-4
Another major problem is creating systems ensuring proper heat transfer from HTHR to process consumers. Studies on the subject have shown that there are two ways of solving the problem: heat transfer can be arranged by means of short-range and long-range heat-transfer systems [ 1 ] . In the short-range heat-transfer mode heat from the power—source is transferred to process units by means of a circulating heat—carrier. Fig.8 shows an example of scheme for the short-range heat—supply featuring the use cf heat generated by HTHR for the pyrolysis of hydrocarbon feed, heating and processing of hydrocarbon products, and raising of steam. High-temperature heat is transferred by l.?lium to pyrnlysis plant located adjacent to the nuclear reactor, whereas the medium-temperature heat is sent to the units processing crude heat-carrier.
oil through
the use
of
an
intermediate
In the scheme shown on Fig.8 the helium outlet temperature from the HTHR is 950°C. In as much as low- and medium— •*:<=mperature processes in the crude oil ^r-ocfrssing account for about 90 7. of the total, it seems promising to use HTHR with helium temperature at the outlet from the reactor being 650 to 750t"C. In this case heat from the source can be transferred by means of an intermediate heatcarrier ( Fig.9 ) . The short-range heat supply can be employed when the reactor is located within several miles from the cp'isuiner. With HTHR being located farther off ( tens or hundreds of milf?s ) from the acsosiated consumers, long—range heat supply systems may be employed, based on chemothermic cycle, such as convers.\on-methanatirn cy.-le C 2 3. In this case heat distribution in the plants can be arranged from centralized power source? ( mechariators ) by means of heatcarrier loops. For the heat supply systrm to be properly designed 3n appropriate heat-carrier should be selected. In principle, the industrial heat supply systems may use both gaseous fluids: helium, steam, C 0 3 and liquid fluids: metdi and salt melt«= or organic heat-transfer fluids. B-4
Gaseous heat-transfer fluids characterized by high thermal stability are however low heat-transfer agents. This necessitates the application of large heat transfer areas and results in high transportation costs. To reduce the volume of the heat-carrier used in Lhe system it seems practical to increase the prtssure in the system. This however will r e m i t in significantly higher cost of the loop, more stringent requirements to ensure safe operation of the system. Liquid—mr'tal hc?at~carriers have inherently the highest h«at-transfe*- capabilities. One of the merits of liquid metals is their low viscosity. The most widely used metal melts are potc.ss.ium, sodium, their mixture and also outec^ic melts of lead and bismuth. However alkali metals are explosive, whereas lead and bismuth are rather scarce. These shortcomings make all the advantages of liquid .T.etals null snd void and precludes at least at tht? beginning thej r une in the refinery heat supply systems. Liquid—salt heat carrieris have poorer heat—transfer nr-opo'-ties than liquid metals. Their advantage rests with their low-cost and wide u«=;e in the industry. Thus, in crude oil refining end petrochemical industry there is a long-term experience in application of nitrate—nitrite mixtures at catalytic cracking units. The advantage of organic heat-carrier appliration is in that they are very common in -Lhe industry's plants. The use of these heat-carriers until resently was limited to temperatures of 360 — 380^0. Over the last years new organic heat—carriers have been developed, capable of operating at temperatures above 400°C. Adding regeneration units to the heat-carrier circulation system and application of thermal stabilir.iny additives will permit to raise the maxin.am allowable operating temperature in the circulation cytle» still higher. A cost comparison study has been made for circulation cycle filled with different heat-carriers in order to try and select a sutable heat-carrier, tt feasibility study made for the case of heat furnished from the HTHR to a crude unit B-4
with the capacity of 150 Mwt ( heat ) has shown that the cost of systems using liquid heat-carriers vs. gas filled systems is about twice lower. It was found that in the gaseous systems the most expensive equipment are compressors. Their share in the total cost of the circulation cycle accounts for about 40 %. In the case of liquid heating systems the most costly is the hsat— exchanging equipment, which accounts for about 70 7. of the total cost ( Table 1 ) . From the above analysis it follows that heat supply from HTHR located several miles from crude units, operating at 360 - ^ O O ^ C , can be arranged with the use of organic heat-carr.iers. At higher operating temperatures n.ost acceptable are saline heat carriers. However tht? final selection of a heat-carrier requires extra research work to be carried out, including that on a large-scale basis in order to improve the" technology and equipment"'of "the heat"— supply systems. In conclusion we would, like1 once again to point out the problems which can be solved by implementing HTMR at the crude oil processing and petrochemical plants: improving pollution control, making more hydrocarbon fuel available for other use, intensification of operation of process units, and making them less fire hazardous, increasing the power efficiency of process plants. REFERENCES 1. E.A.Feigin, E.A.Raud, E.B.Ramanova, etc. "Prospects for application of nuclear reactors to prcvine energy to refineries and petrochemical plants". M.'TSNIITEK'EFTECHTM* , 1785, 1O8 pp. 2. A.Ya.Stolyarevsky. "Nuclear technology on the basis of high-temperature reactors". M. 'ENERGOATOMIZDAT', 1TS8, 152 pp.
B-4
Ensrgg refuiuif
in snare in o> petraenei industn
Fuel oil and Fuel gas 40-60%
-p
Plants
9}istriktion of oil Rifining anct Petrochemical Plants according to hmt. consumption
Heat ISO
TOO COO TO 2500 " SiBp; 3T0O ^3^0 4303
' K9B 6T0C BDGO tlOH N/lW (therm)
'-t ^
CO SS
o Hydrogen production
/ractionation unit
f1
Gas .A
Crude tradz \ oil
oil / — — / •^ radio nation
Hydrogen
+
Besulfurisation
f Diesel oil Desuliarisation cut Gasoline cut
Catalytic
Gasoline
Steam
cut
crackm9
Dewaxinj
Diesel
oil
Gasoline
Reforming Polyethylene Qlefi nes
Petrochemical processes
Detergents Ethyl Qicoboi Synthetic Ftber _
Rubber Basic
crude
distillation
scheme
B-4
Reflnerj B/T diagram <500
40Q
?B0 2D0
100
Q.Mw 100
ZBO
?00
400
WO
600
700
Oil product Heating capacity [istrlbution aesordino te teeiperatiire a \ Mw %
200-?OO ?80-400 20
280
?70
4-^,7
IEVI
^Bfl'600 90
?o
6,2 B-4
800
PetreGhamical eotn 8/T dlajr
"100
ZOO
proouot Ksaxtng u llstrlbirtion aossrEilng ii tomperature \m 20Q-?0G 700-400 408-W ?00^BO 500-700
°G
MW /c
22,4
?24
44,?
42,2 •?,?
It B-4
Hydrogen eons urnption/crucU ? ° / 0 4,25 •
I
s
S
P.
0,70 0,00
p
-
-f
HE consumption Hrkiirotnatmtj
of
distillation proctucts 80
6§
70
75
80
85
motCffUEi
pfoiu.ctioi}/crui5;
y consumption for refining of sulfuric oils, B-4
... n
i
i
1
£ V
I
i
1
bI
I v>
THH
SHARE
OF DIFFERENT;
COST
OF SHORT
DISTANCE
EQUIPMENT
TVPE
HEATING
SYSTEMS
Cost, NETWORK
CC- k
ELEMENTS
550 °c
550 °C 700 °C 550 °c 700°C
0
6 ,0
25,2
PUMPS, COMPRESSORS,etc
3
0,3
35,
HEATEXCHANGBRS
7
i ,5 67 ,0
80,0
30 : 0
—
2
/
)^i
CO 2
Hi i
PIPELINES
HEATCARRIARS •
TNE
/o
OHC /i2O°C
IN
35
5 20,
28
31,6
38, 0
^2,0 :
0
3, 1
B-4
Short distance heating systems cost of primary distillation units with different heat carnzrs
55O°C
• Cost, thousand rb/s OHC CC-4 CO 2 He 42O°C 550°C 700°C 550°C 700°C
1177
582
1056
3090
3090
2923
Insulation
88
79
(54
255
255
240
Pfperacks
Z«2
212
2i2
212
212
2(2
212
•118 19 498!
58
(06
309
309
292
292
69 4113
19
4350
4350
4875
4875
4485
3685
2161
4115
3058
142
462
3?0
370
5
5
19747 12662
11605
network
zlemenis
Pipes
Steel
structures
Pumps ^ compressors, etc. H ea/ex changers
0,04
He aj carriers
Total :
6595 5255
6492 12271
2923
B-
Energy Consumption in Wain Oil Refining and Petrochemical Process Type
Process
feed
Final
Products
Processes Energy Temperature inheater consumption,
processes , °C
Crude oil
Gasoline (xtmosphzrie and vacuum aasoils} vacuum residue Wydrocracking Vacuum oasoil Gaso/ine, jetand diesel oil kerosine dizsel Mydrotreated kerosine U T t i oil and diesei oil Low octane Reforming tfledium High octane gasoline acisoline temperature HyoLrotreoled Catalytic \acuum gas oil High octane gasoline cracking \li\cuum residue Cracking bottoms, dies el VisSreakino. oil component7 yacuurn oasoi! Steam cracking [thane 1 gaso'/f, Lower olefines, aromatic High gasoline hydro car bones temperature Hudroge/7 Conversion
Low temoerature
Distillation
refinery gazes
40-60
300-400
25-35
if Q 0'550
/0-20
800-goo B-4
XA0101491 NEUTRON-PHYSICAL ASPECTS OP THE HTR CONCEPT WITH SPHERICAL FUEL ELEMENTS N.N.Ponomarev-Stepnoy, V.N.Grebennik,E.S.Glushkov, N.E.Kucharkin A report to be presented at the Technical Committee Meeting IAEA on HTGRs June 21-23, 1989, Dimitrovgrad, USSR The development of a high-temperature reactor with He coolant and Carbon moderator is held currently. This reactor has a number of advantages over traditional reactors. The use of multilayer coated particles dispersed in a carbon matrix of a carbon moderator and inert
coolant -
He allows to achieve high-temperature level in such reactors along with high-reliability and safety of their operation. Great safety of these reactors is determined by: 1) the negative temperature reactivity coefficient in all the temperature ranges; 2) high termal heat capacity of the core; 3) use of particles with a multilayer sheath that provides the retention of fission products inside fuel particles. Because of all this HTR can be employed not only in electric power
generation, but also in high-potential heat for technological goals production. HTR is of interest because of its versatility with regard for use of different fuel cycles with U238
a n d Til
232
a s fer-tile
materials.
There are 2 fuel element concepts for such reactors - prismatic elements (in the USA) and spherical elements (in FRG and in the USSR). In the second case the HTR core is formed
as free loaded spherical
elements, cooled by He, pumped through the core. We think, the use of optimum-size spherical elements leads to better conditions of their operation in the core regading for
tensily-de-
formed state of fuel elements; it also makes their work off in the reactor loop tests more rapid. Besides, the use of spherical elements makes the refuelling problems easier due to the ability of spherical elements to roll under the gravity.
B-5
2. As a result of optimisation of spherical elements and fuel particles parameters it was accepted that: 1) spherical element diameter is 60 mm; 2) fuel compacts(containing particles) diameter is 50 mm; 3) kernel particles diameter is 500/M ; 4) fuel particles coating is 3-layer of pyrocarbon and silieone carbide. In pebble-bed HTR reactors there exist
several possibilities of
refuelling: - only once - in the so-called OTTO cycle and many-times refuelling (MTR). In MTR it is necessary to have a special device at the
reactor outlet that is going to control the level of fuel burn-up in eaci: fuel element and to grade them according to it. Fuel elements that have achieved high burn-up must be extracted from the reactor and those that have not, must be returned to the core. In this case fuel elements with different burn-up levels are mixed more or less equally in the core and that leads to the traditional distribution of neutron density throughout the reactor volume that has relatively small unevennesses. In the OTTO-cycle fuel elements pass over the core from above down only once and that leads to the sharp change of the fuel isotope structure along the core height (from the fresh fuel in the upper part of the core to the burned to the required level fuel in the lower part of the core). This results in a specific high-altitude distribution of a neutron flow and of heat excretion
with its maximum in the upper part
of the core and with a great fall to the lower part of the core. Thus, this method of refuelling influences greatly
the neutron-physical
HTR performances including heat excretion distribution, reactivity effects and the control organs efficiency. The development of two HTR reactors is held currently in the USSR: small power producing reactor "VGM" and great power producing reactor YG-400 f1
, 6 3 . Great power producing reactor "VG"-400 uses low en-
riched uranium (LEU) in contrast to TtfTR operating in FRG using Th
rw
fuel and MTR. Main performances are listed in Table I. B-5
O
3.
Physical calculations of a "VG"-400 reactor were held in 2 stages. At the first stage the calculations of a stationary (balanced) regime of operation of a rated power reactor with OTTO-cycle with the help of til a " Gavrosh'1 program in diffusion approximation with regard for the fuel burn-up and for reactor temperature fields calculations (assuming the absence of poison rods) were made. According to these calculations isotope structure distribution throughout the core with regard for fuel burn-up; average fuel compaign neutron and temperature fields; heat excretion distribution etc was determined. Such reactivity effects as the effect of reactor poisoning due to Xe-joj-; temperature effect; effect of the water (steam) entering the core etc. were determined in these calculations also. At the second staye of calculations diffusion approximation for determining neutron-
physical performances of VG-400 at different positions of control organs can not be used. That's why MCV £43 three-dimentional program was used to make the calculations. This program is based on Monte-Carlo method and accounts for structural features of the main reactor components; moderation features and neutrons thermalisation; the presence of technological void. Calculations based on MCU program require much time and that»s why they must be done for the limited number of variants and for the simplified reactor model. The continuous distribution of isotope structure along the core volume that is determined by the fuel burn-up in the OTTO-cycle was replaced in these calculations by the division of the core along the height in the form of steps into 6 zones, in each of which there was assumed to be average isotope structure.
Scheme of such reactor model, used in the Monte-Carlo method calculations is given in Figures 1 and 2. Some results of neutron-physical calculations of VG-400 reactor are given belowt'J. B-5
4. Table 2 lists the reactivity effects that must be compensated with the help of control organs to provide the reactor subcriticality with the maximum value of K *- - that is in cold nonpoisoned condition.
In Figure 3 you can see the character of heat excretion distribution along the height of the reactor core for 2 variants, differing in
fresh fuel enrichment (6.5% and 10%). It is seen that maximum of the heat excretion distributions i$. removed to the upper boundary of the core which is connected with the loading of fresh fuel from the top of the core and with the fuel burn-up as it moves downwards in OTTO-cycle. Dependence of inserted rods systems efficiency from the depth of their insertion in reactor (carbon top reflector, void, core) for 2 variants differing in fresh fuel enrichment (6.5% and 10%) is given in Figure 4« One can see that control organs efficiency at the insertion of control organs in pebble ped for 240 cm is a greater value (
4K/k s 26%) for the variant with 10% fuel enrichment. Inserted rods system efficiency increases for 20%/when using 10%
enriched fuel is determined by the remove of the maximum of the thermal neutrons flow (heat excretion) position to the upper boundary of the core loading in comparison with the maximum position for 6.5% enriched fuel. Comparing the values of reactivity effects and of the inserted compensative rods system
efficiency one can state that the con-
trol organs efficiency is enough to support reactor subcriticality in a cold non-poisoned state and to provide reactor operation at power. One must note great dependence of the heat excretion distribution from the depth of the ICR insertion (Fig.5). In Fig.5 heat excretion distributions along the height of the core at different ICR system positions (for 6.5% enrichment of fresh fuel) are given. VQ-400 reactor reactivity changes during spherical elements extraction from the core were calculated. These calculations were made with the help of MCU program for the reactor in a cold state with B-5
5. TABLE I VG-400 REACTOR PERFORMANCES FOR THE RATED REGIME OF OPERATION (OTTO)
Performances
Unit
Value
Reactor power (heat)
MWH
1060
Coolant temperature (He): at the reactor inlet at the reactor outlet
°C °C
300 950
Average fuel temperature
^C
870
Average moderator temperature in the core
°C
810
Core diameter
m
6.4
Core height (equivalent)
m
4.8
mm
60
U Content in one fuel element
g
6.15
Enrichment for Ugoc
%
6.5(10)
Particle kernel diameter
my*
500
Equivalent thickness of technological void Carbon top (butt-end) reflector equivalent thickness (CTR)
m
0.5
m
1.2
Carbon bottom (butt-end) reflector equivalent thickness (CBR)
m
1.4
Carbon side reflector thickness (SR)
m
1.0
Fuel element diameter
Number of incore inserted compensative rods (ICR) Number of absorber rods in side reflector Absorber rods diameter for B.C poison (outer/inner)
55 24 mm
110/90
B-5
6.
TABLE 2 REACTIVITY EFFECTS FOR VG-4OO
4K/K, % No.
Reactivity effects
6.5.% enriched fuel 1.
Operative reactivity supply
2.
Reactivity change due to the
10% enriched fuel
1.5
1.5
cooling of the core
4.8
5.3
3.
Stationary poisoning
3.0
3.0
4.
Reactivity effect when water enters the core
0.80
1.85
Supply for subcriticallty
1.0
1.0
11.1
12.6
5.
TOTAL
B-5
7. the ICR inserted in a pebble-bed (the depth of insertion in the core 240 cm). At the decrease of the height of the pebble-bed due to the extraction of the fuel elements it was considered that the control organs position against the CTR is fixed. That means that
during
extracti-
on of spherical elements the core is removed downwards leaving the control organs. What does it lead to regarding reactivity? Decrease of the core height leads to technological void increase and, therefore, to the neutron leakage increase. Decrease of the depth of ICR insertion into the core in this situation leads to decrease of their efficiency, and, therefore, to reactivity increase. Besides, the extraction of the most bumed-up fuel elements and drawing at the same time less burned-up fuel near to CBR also leads to reactivity increase. The calculations confirmed that all these effects can lead to reactivity increase at the extraction of fuel elements from the reactor. The results of these calculations are given in Figure 6. One can see (curve I) the change of VG-400 reactor reactivity in a cold nonpoisoned state due to decrease of the core height
H in a fixed position
of ICR (in extreme low position agains CTR). Curve I shows the reactor with 6.5% enrichment of fresh fuel and it is rationed so that reactor in a hot state at a rated core height with stationary poisoning should have reactivity supply - 1.5% (operative supply at the withdrawn control organs).
In Figure 6 one can see that during extraction of spherical elements positive reactivity can appear (run
out). It s maximum value is achie-
ved at core height decrease for 240 cm (that is for the depth of IRC insertion) - 4K/K a (4.5 - 2 ) % .
B-5
8. It is important to note that run out of reactivity is displayed during a long period of time, that is determined by the productivity of unloading - loading device; le ;>1.o day.
time sca-
Increase of reactivity can be eliminated if
passive transfer of ICR together with the pebble-bed were provided. In Figure 6 you can also see the results of a reactivity change for the reactor in an operation state with the control organs withdrawn (curve 2 ) . It can be seen that in this case the extraction of the fuel elements with burned up fuel doesn't lead to reactivity increase. In Figure 7 one ~can see the dependence of a VG-400 reactivity change from the steam mater mixture content in the core. Module concept of atomic plants design is being widely discussed in different countries nowadays. For the first time in the USSR the module concept of atomic plants (AP) design was suggested by S.M.Feynberg as an atomic power plant (APP) with the fast breeder using He as a coolant [,57 • The idea of module concept is that a small power reactor and therefore of a email size (module) is contained in a metal vessel that is fabricated at a plant and then transported to the AP location by ground. Increasing of power of AP is realized by the simple combining of reactor modules of the same type and their joint placing at the AP. Fabrication of a module reactor at a plant raises its operation reliability; allows to control the quality of reactor and equipment fabrication; shortens the time of fabrication and by making licencing easier it shortens the time of putting AP into operation. That's why AP with module reactors can be compared and even more economic in comparison with big, much power producing reactors, that are fabricated at the AP construction site. This concept was reported in 1972 in Minsk at the
IAEA
experts
conference. B-5
9. But the idea that it was necessary to increase power of a single reactor in order to make AP more economic, that prevailed at that time, didn't allow the specialists to appraise the concept of a module reactor And only several years later module concept of AP based on small power producing reactors-modules was admitted in our country - VGR-5O project and abroad. Calculated research is made to develop safe reactor with physical safety characteristics inherent in it and with passive devices of cooling of the reactor after the stop at norlal operation and control organs in the core and to place them in a radial reflector if possible. That leads to the structural simplification of the control system and provides more safety of VGM. In VGM design it is very important to choose divisible circulation of fuel elements through the core and that's why much attention is paid to it a calculation analysis. MTR and OTTO - cycle were discussed. U-loadjLn§ in fuel elements, fuel enrichment and sizes of the core were
varied
• Main calculations
were made with the help of "Gavrosh" complex program, allowing to make neutron-physical and ifcermohydraulic calculations of the reactor with regard for burn-up and refuelling [3] . Calculated estimations of reactivity effects and of efficiency of control organs, placed in the core and in the radial reflector were made. The results of this research, and VGM characteristics chosen with its help and adopted for the following checking are given below. Reactivity effects and control organs efficiency is very important when choosing VGM-reactor main characteristics. As calculated research shows efficiency of the control organs, placed in a radial reflection depends greatly on the core diameter When the core diameter becomes less such control organs efficiency increases. That's why there exists a trend to choose a variant with a small core diameter. It is adopted nowadays that the core diameter is 3 m. This is the maximum value.
10. When the core diameter increases, efficiency of the control organs, placed in SR will not be enough to provide reactivity balance. Main characteristics of VGM-reactor for the
rated operation regime
are given in Table 3. The specific feature of VGM-reactor is the prolongation of the core H/D^3 - it is connected with a desire to decrease the average-specific power of the core in order to provide passive cooling of the core from the residual thermal excretion even if the coolant system is damaged. (Fuel temperature in such a case doesn't exceed 1600°C and fuel particles are hermetically sealed regarding for the fission products). It is characteristic for reactors with H/D>3 that control organs position greatly influence
axial thermal excretion distribution. It is
very important for the calculation of temperature fields. The character of heat excretion dependence along the core height with the withdrawn an, incerted (to the midium height against the core/rods is given in Figure 8. It is seen that the position of control organs in a SR influences greatly the axial distribution of heat excretion along the core. Dependence of rod efficiency from the number of rods their distance from the core centre was studies for VGM-reactor. In Figure 9 you can see dependences of rod groups in the SR from the number of rods in a group (^corea^50 cm, R r Q j center*^-' cm). It is seen that the efficiency of a rod system increases with repletion when the number of rods increases. In Figure 10 the dependence of efficiency of the absorbing rod system placed in a SR from the radius of their position is given. In the same figure one can see the data concerning the efficiency of the circular absorbing coating ( 1 cm thick; of B / ) as the upper estimation of the absorbing rod system efficiency. Main reactivity effects from the balanced operating regime of a B-5
11. VGM-reactor, that must be compensated by the control organs are listed in Table 4. Comparing the obtained values of reactivity effects and control organs efficiency, One can see, that it is possible to provide reactivity balance if placing control and compensation organs in a SR of a VGM-reactor with a core diameter not more than 3 meters. But this problem must be studied more detailed at the further stages of this reactor development. Special attention should be paid to critical bench tests.
B-5
12.
TABLE 3 MAIN VGM-REACTOR PERFORMANCES FOR THE RATED REGIME OP OPERATION (12-THOSS REFUELLING)
Performances
Unit
Reactor power (heat)
Value
Mwtt
200
°c °c
300
Coolant temperature (He) at the reactor inlet at the reactor outlet
750(950)
Core diameter
m
3
Core height (equivalent)
m
9.4
Fuel element diameter
mm
Initial U content in one fuel element
g
7
Initial enrichment for Upoc
%
8
Partical kernel diameter
m/n
500
Equivalent thickness of a technological void
m
0.5
m
1.0
m
1.4
m
1.0
60
Carbon top (butt-end) reflector equivalent thickness (CTR) Carbon bottom (butt-end) reflector
equiva-
lent thickness (CER) Carbon side reflector thickness (SR) Number of
KLAK
cemals in a SR
22
Number of absorbing rods canals in a SR (at the maximum rods incertion - for 1/2
24
of the core height) Divisibility of the fuel circulation
12
through the core
B-5
13. TABLE 4 MAIN VGM-REACTOR REACTIVITY EFFECTS
AK/K; %
Reactivity effect 1.
Change of reactivity at the turning of rated power into partial (50$) power
2.
Reactivity change due to the cooling 7.0
of the core and XB^^C decay 3»
1.4
Reactivity effect when water enteres the core 200 kg
1.2
4.
Change of pebble-bed density
0.5
5.
Regulation supply
0.5
6.
Supply for subcriticality
1.0
TOTAL
3CM
11.6
B-5
u. Fig,1.
Calculation diagram (scheme) of a VG-400-reactor for the height.
Pig.2.
Calculation diagram of a VG-400-reactor in plan.
Pig.3.
Character of thermal excretion dependence for the core height 1 - initial fuel enrichment for U-235 - 6.5%* 2 - initial fuel enrichment for U-235 - 10%.
Pig.4.
Dependence of the inserted rods system efficiency from the depth of insertion. 1 - initial fuel enrichment for U-235 - 6.5% 2. initial fuel enrichment for U-235 - 10%.
Fig.5«
Axial distribution of thermal excretion at different positions of ICR system. Curves 1,2,3,4,5,6 correspond to the different depths of ICR insertion and are marked
with
Figures 1,2,3,4,5,6, Pig.6.
Change of a reactor reactivity due to the core height
decrea-
se at the fixed extreme low position of ICR rods. 1. ICR rods are inserted into the core (reactor is in a cold non-poisoned state). 2. - ICR rods are withdrawn from the reactor (reactor is in operation state). Pig.7.
Dependence of a reactor reactivity change due to the steam water mixture in the core. 1. initial fuel enrichment for U-235 - 6.5% 2. initial fuel enrichment for U-235 - 10%.
Pig.8.
Character of the axial distribution of heat excretion in VGM core 1. SR absorbing rods are withdrawn 2. SR absorbing rods are inserted (1/2 of the core height).
15.
Fig.9.
Dependence of rods group efficiency in a radial reflector ^R
center
=
1
^5 cm) from the number of rods in a group
( 1 - N c /N v = 600; 2 - N c /N v = 550). Pig*10.
Dependence of the absorbing rods (1-3) and absorbing coating (4) efficiency from their distance from the core center in a radial reflector ( 1 - H r o d » 6; 2 - HTQd = 20; 3 = 30).
B-5
16 LITERATURE 1. Комаров Е.В., Митенков М.Ф., Лаптев ф.В. и др. Атомная энерготехнологическая установка ВГ-400. Атомная энергия, т. 47, вып. 2, с. 79, 1979 г. 2. Бедениг Д. Газоохлаждаемые высокотемпературные реакторы. (Перевод с немецкого). М., Атомиздат, с. 224, 1975 г. 3. Савандер В.И., Сарычев В.А., Цибульский В.Ф. Методика расчёта стационарного режима работы ВТГР с засыпной активной зоной Вопросы атомной науки и техники, серия: Физика и техника ядерных реакторов, вып. 8 (45), М., 1984 г. 4. Лиман Г.Ф., Майоров Л.В., Юдкевич М.С. Пакет программ МС[/ для решения методом Монте-Карло задач переноса излучения в реакторах. Вопросы атомной науки и техники, серия: Физика и техника ядерных реакторов, вып. 7, с. 27, 1985 г. 5. Фейнберг С М . Бридер с газовым теплоносителем. Идеология жёсткого и мягкого спектра. Доклад на совещании экспертов МАГАТЭ, г. Минск, 1972 г. 6. Глушков Е.С., Гребенник В.Н., Евсеев В.И. и др. Выбор и обоснование физических характеристик и геометрии активной зоны ВТГР малой мощности. Доклад на Советско-западногерманском семинаре , 17-23 июля 1988 г., г. Москва. 7. Глушков Е.С., Демин В.Е., Майкова Л.К., Сарафанов Н.И. Эффекты реактивности в ВТГР с однократным прохождением шаровых твэлов через активную зону. Доклад на Советско-западногерманском семинаре "Безопасность ВТГР" 22-29 апреля 1989г., Москва В-5
Cxei/a БГ-400 по гысоте
PEC.
В-5
LOHC7;isH cxewa -реш.тсра БТ-400 в п/^ле.
- ячейка бокового отра^лте.'ш БО-I - ячейка бокового отрз^гелл БО-2 - ячейка A3 без стергля
О
- ячейка A3 со стержнем СУЗ
Рис. 2 В-5
, кВт/шар 5,0 _
1,0
Рио. 3 . Характор зависимости тошюввдвлвния по высоте активной зоны. 1 - начальное обогащение топлива по урану-235 - 6,5 % , 2 - начальное обогащение топлива по урану-235 - 10 % • В-5
о Зависимость эффективности системы погружных стержней от глубины погружо1шя (холодное состояние).
к 25
20
15 Активная зона 10
Н, см 400 Рис.
4
1 - начальное обогащение топлива по ураяу-235 2 - начальное обогащение топлива-по урану-235
х 5 = 6,5 % Хс = 10 *./
%
В-5
отн.ед
Зависимость удельного энерговцдоления от высоты активной зоны по оси реактора при различных глубинах погружения поглощающих стержней (х5=6,5 %)
активная
зона
УУУУУУУУУУУ7 '<У/7/,\ 3 V7777777/Y777 /7/7'
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400 Рис.
5
600
650 В-5
Изменение реактивности реактора от уменьшения высоты активной зоны при фиксированном крайнем нижнем положении стержней в холодном разотравленном состоянии (кривая I) и. для реактора без стержней в рабочем состоянии (кривая 2 ) .
о
к о
^.
дН,
-10
-15
-20 РИС.
6
В-5
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I\
2,0 J
1,0 -J
0,5
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о
500 I00Ü 1500 2000 Рис. 7 Зависимость изменения реактивности реактора от количества пароводяной смеси, находящейся в активной зоне, 1 - начальное обогащение топлива по U -235 - 6,5 % , 2 - начальное обогащение тошшва по 17 -235 - 10 % • В-5
2500
(p
io Ц/Н Рис.8 Характер осевого распределения тепловыделения по активной зоне ВГМ 1 - поглощащие стержни БО извлечены 2 - поглощающие стержни БО погружены (на 1/2 высоты активной зоны)
В-5
N
zs
10
30
40
Л/ штук
Рис. В. Зависимость эффективности группы стержней в радиальном отражателе (Нц^СТв=165см) от числа стержней в группе t î - ^ = 6 0 0 ; 2- ^ = 4 4 9 )
В-5
3/1 ъ
4
-
2
-
I
150
160
180
190
£,сы
200
Рис. £ 0 Зависимость эффективности поглощающих стержней (кривые 1+3) Е поглощающей оболочки (кривая 4) от радиуса расположения их в радиальном отражателе (1-/?ст=6шт; 2- Пст=20шт; 3 - Ист=30шт) В-5
XA0101492 Improved Safety Nuclear Power - and - Heating Plant with HTGR of Modular Type. Bogoyavlenskii, E.G., Vinogradov, V.P., GlebovV.?., Grebennik, V.N., Ponomarev-Stepnoi, /£ J$., Hruljov, A.A.
Technical Oommitte meeting on Gas-Cooled Reactor Technology, Safety and Siting. Dimitrovgrad, USSR, 21-23 June 1989The recent accidents at the Three-laile-Island Nuclear power plant (USA), at the Chernobyl Nuclear Power Plant (USSR) and emergency situation in HTGR plants in FRG have radically changed an attitude to safety problems in nuclear engineering and caused an increased interest to developments of the higher safety nuclear plants of new generation - high-temperature gas-cooled reactors (HTGR)/I, 2/. The HTGR-module nuclear plant / 4 / is a universal power source for combined generation of electric power and process steam, for direct heating or for use in commercial process units generating methanol and gasifying coal, etc. HTGR-module concept is characterised by the standartized reactor units of the equal power, which may be integrated into the plants of the higher power on request. An improved safety nuclear power-and-heating plant concept of HTRG-module is based on ensuring the module safety as a whole /3/« The HTGR nuclear power-and-heating plant safety improved is ensured by maximal employment of the internal (i.e. inherent reactor safety features) passive safety means, the following criteria form the basis of the concept: - the passive means of safety ensurance, e.g., natural coolant circulation in.the primary loop and heat removal systems, which provide residual heat release removal out of the core during all the time period required for accident elimination, repair and dismounting;
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- a reactor trip by itself if "heeman factor" fails; - a reactor plant enables to localize any reactor subsystem failure, i.e. the principle of accident non-propagation about the system is realized. Using the equipment fabrication technologies available in combination with the passive safety means permits to reduce the amount of safety systems and to exclude operator errors, which improves the system reliability on the whole, simplifies the possible accident risk assessment and increases a confidence in nuclear engineering. Additionally, radiation safety of nuclear-power-and-heating plant HTGB-module units are ensured by 5 barriers, which prevent gaseous and solid fissin product release into the environment. I-st barrier - a multi-layer ceramic coating of fuel particles (microfuel elements); 2-d barrier - a spherical fuel element graphite matrix with fuel particles dispersed in it; 3-d barrier - a fuel element graphite shell; 4-th barrier - strong-firm metal reactor vessel operating at moderate temperatures (300-350 G) and pressure 6 r.'IPa and enseering the service life not less than 40 yrs due to the lower neutron flux which falls onto the vessel, in comparison with a pressurized water reactor vessel; 5-th barrier - a reactor pit having safety valves and characterized by sufficient degree of leakproofness. Materials applied in "safety" barriers provide considerable temperature margin, i.e. the material operation temperature is 2-3 times lower than the temperature of change in its state. The improved safety HTG-R-module power-and-he at ing plant is based on two basic arrangements of reactor heat removal and utilization; in the former HTGR heat power supply to the steam-turbine cycle working medium is performed throughjtwo helium loops, excluding water ingress from a steam-generator to a reactor core under any operating conditions and in the latter the heat power supply
3/6
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from a reactor to the back-pressure turbine working medium is fulfilled through an intermediate helium lopp and an outside steam generator and to heat-extraction steam turbine plant working medium through a steam generator built in HTGR. Using the above mentioned arrangements allows to unify designarrangement approaches to the nuclear power-and-heating plant of different power and of HTGR-module different applications. Integral arrangement of the reactor equipment which gives the opportunity not to use the primary pipelines of large diameter, ensures the plant compactness and reduction in specific metal content from 5.03 t/MW to 3.19 t/Ktf in comparison with the unit arrangement of the reactor equipment. Further, the paper describes 200 IvM-j^ HTGR-module in the strongfiz-m vessel of which e, helium-to-helium heat exchanger is placed to increase the radiation safety and to prevent the water ingress into the reactor core. The basic technical data of the module are given in Table I. Table I Item
Value
1. The plant thermal power, T.'ivv
2 x 200 = 400
2. The plant electric power, Mtf
110
3. Power produced in the form of lower-grade heat, MW
230
4. The reactor thermal power, MVV
200
5. A type of a reactor
heterogeneous thermal-neutron, high-temperature, gas-cooled
6. The primary parameters of the plant
\
6.1. Coolant
helium
6.2. Pressure, MPa
6 c-i
6.3. The coolant temperature,
C
at the core inlet
320
at the core outlet
750
7. The secondary (intermediate) parameters of the plant 7.1. Coolant
helium
7.2. Pressure, MPa
5,5
7.3. The coolant temperature, G at the heat exchanger inlet
280
at the heat exchanger outlet
600
8. The third parameters of the plant: 8.1. Coolant
Water, superheated steam
8.2. The superheated steam pressure, MPa
13
8.3. The superheated steam temperature, C
555
8.4. The feed water temperature,
230
C
9. A number of the module principle equipment: 9.1. Reactor
2
9.2. Steam generators
2
9.3. The primary blowers
I
9.4. Steam-turbine plants
I
The principle plant-module arrangement is shown in Figure I The module consists of the following basic systems: 1. The primary module circuit. 2. The intermediate (secondary) circuit. 3. The steam-generating (third) circuit.
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4. The reactor active cooling system (RAGS). 5. The reactor passive cooling system (RPGS). 6. The fuel element circulation charge and discharge system (F3CCD3). 7. The system of filling the rooms with nitrogen (SFRN). The primary module circuit is placed in the reactor vessel and consists of the spherical fuel element bed core, 4 sections of the helium-to-helium heat exchanger and the main blower. The primary circuit arrangement provides the natural coolant circu-
lation under normal operating conditions, under start-up, heating-up, cooling, spinning reserve conditions and in emergency situations related with de-energizing the main blower electric motor. The primary equipment mix and sectionalization ensures the module operation on power less than nominal at the main blower and any heat-exchanger section disconnection. The intermediate (secondary) module circuit is generally located outside of the reactor vessel and consists of the helium-tohelium heat exchanger, two blower pipelines with check valves and the steam generator. Among the intermediate citcuit equipment the only equipment placed in the strong-firm reactor vessel is the helium-to-helium heat-exchanger sections. The module steam-generating (the third) circuit is located outside the reactor vessel and consists of two steam generators, feed pumps and other unified equipment. The reactor active cooling system is included into the module intermediate circuit and consists of pipelines, two check valves and two helium-to-air heat exchanger. RACS with the help of the helium-to-air heat exchanger provides the core heat removal by natural circulation under normal operating conditions: under the start-up, heating up, spinning reserve, shut-down, cooling conditions and in all the design emergency situations, as well. RAGS
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provides an essential saving of electric power consumption on own demands because under mentioned conditions the blowers don't operate. Power tapped by RAGS is 3-5% of the core nominal power. RACS is included into the forth independent circuit of the module and consists of a heat transfer surface located round the periphery of the reactor pit, headers, pipelines and two air heat exchangers. RPCS is filled with water and operated in the natural circulation mode. RPGS is meant for maintaining the required reactor pit concrete temperature under normal operating conditions and for reactor vessel heat removal and maintaining the vessel metal temperature within the allowable limits in the emergency situations. The RPGS power is 1-3% of the core power. Thus, RAGS, RPCS and the coolant natural circulation in the reactor ensures the reliable core heat removal under any emergency and hipothetical conditions. FdCCDS has two independent fuel element circulation loops, each of which consists of pneumatic unloading facilities, partitioning mechanisms, tanks for the fuel elements damaged, fuel element dump tanks, an accelerating and retarding devices shut off and sphere lines. The fuel element charge, discharge and circulation system is included in the plant primary circuit and is meant for the initial and subsequent fuel element charges into the core, for the fuel element discharge out of the reactor and for conducting the fuel element circulation during the operating period. SFRN is an emergency system providing the plant safety. SFRN eliminates the substantial fuel element graphite coating and graphite laying oxidation under the primary depressurization SERN consists of the headers with atomizers located in the reactor vessel and pit and in FECCDS rooms, and pipelines and a cylinder manifold with the nitrogen store. SFRN operates at sharp descrease in the primary helium pressure which is associated with the circuit emergency depressurization. SFRN has 100% redundancy in pipelines of nitrogen supply into the rooms. c-i
The plant-module consists of two equel power reactors located in separate rooms. Sucli a building arrangement enables to commission the plant step-by-step. j-Jvery reactor-module is placed in a concrete pit. Round the concrete reactor pit periphery the heat transfer surface of R?C3 having top and bottom headers, is located. To ensure the reliability under operating and emergency conditions the RPC3 heat transfer surface is devided into three sections. At the top of the pit there are the intermediate coolant headers to which the helium-to-helium heat exchanger section pipelines are connected. The top pit room, where the upper reactor unit is located, is unserviceable at the nominal reactor power. At the bottom building area the rooms are located in which there are
rejected and
discharged fuel element tanks. The tanks mentioned are emptied by the special handling machine. The FECCDS rooms are restrictedly serviceable ones. The primary equipment of the module is maximally brought nearer to the reactor pit which enables the pipelines to be minimum short and the primary equipment-• "T..rooms space to be minimal. The reactor VGR-2OO comprises the vessel with a head, the helium- to -helium heat exchanger placed in the reactor head, the control and safety system (CSS) drives, in-pile metalworks, the upper and annular radiation protection, the top reflector, the graphite laying and the core formed by fuel element filling. The VGR-2OO reactor structure is shown in Figure 2. A VGR-2OO reactor coolant flow pattern is chosen in such a way that the metalwork and vessel metal temperature doesn't exeed 300°C during the operating process. The reactor vessel dimentions allow to transport it by water, road and rail. In the reactor head there are located 22 reactor control and protection system absorbing rod dribes, the helium-to-helium heat exchanger section chambers and 2 in-pile monitoring gauges. The special reactor feature is the absence of the fuel elec-i
8 ment discharge channel and as the consequence the absence of the fuel element branch pipe of large diameter at the bottom of the reactor, which eliminates the possibility of breaking down the latter and pouring the fuel elements out underneath the reactor. Additionally, the most safety area for storing the fuel elements is used. It is the reactor core out of which the heat removal is performed by all the module active and passive cooling systems. The radiation protection being in the reactor provides the decrease in radiation to the blower, the main vessel joint, the heat exchanger heat transfer surface and the control and protection system drive to the required level. The radiation protection mix and dimentions ensure the specific activity of helium gained in the helium-to-helium heat exchanger at the permissible level as in the water-cooled and water-moderated reactors. The VGR-200 reactor core is formed by free charging the spherical fuel elements filling the space formed by the inner graphite laying surface. The fuel elements move by gravity into the core downwards. The core coolant flow is fulfilled by the blower or natural circulation upv/ards. Under all the reactor and plant operating conditions the core gives the negative reactivity coefficient and together witn the control and protection system, reactor active and passive cooling systems ensures the module safety under any emergency conditions, including non-contemplated ones. In each of the reactors there is an in-pile monitoring system with the helpjof which during operation the coolant and graphite laying block temperatures, neutron fluxes in core height and radius, blower heads, etc. are measured. Additionally, in the reactor thermometric monitoring of the main joint studs state, torus sealing, reactor vessel and head flanges, control and protection system drives are carried out, the helium-to-helium heat exchanger leakproofness is monitored, metal and welds radiation monitoring of the reactor vessel and head is perc-i
formed during operation with the specimens located in the reactor. The main HTGR-module safety feature is defined by the selected type of a reactor with the sphere-type fuel element bed core, the temperature of which is limited independently within the values where decay product release out of the fuel elements doesn't occur even when all the active cooling systems have failed and the fuel loss of coolant have taken place. Consequently, HTG-Ii-module is nonhazadous for environment both under normal operating conditions and emergency ones and may be situated in close vicinity to populated localities. Thus, the improved safety HTGR-module power-and-heat plant is a new safety generation of nuclear -olants characterized by a wide scope of use in the world power industry.
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3 2 2>
REFERENCES
1. Pilippov et.al "HTGH gas-cooled nuclear plants", Energornashinostrojenje, 1983, N I, p.p. 25-30. 2. Vinogradov, V.P. et.al. "High-temperature gas-cooled reactor VGR-50", Voprosy atornnoi nauki i techniki. Ser. Atomno-vodorodnaya energetika i technologiya, 1982, issue 2 (12), p.p. 2-7. 3. Ponomorjov-Stepnoi, N.N., Slesarev, I.N. "Nuclear industry safety and efficiency is the basis for activities on nuclear reactors of new generation". Atomnaya Snergiya, 1988, V.64, issue I, p.p. 40-46. 4. Meers, L., Brown, M. "Studying HTGR of modular version in USA", Atomnaya technika za rubezhom, 1986, N 04, p.T). 32-35.
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Fig. I Schematic diagram of VGR-module for power-and-heat nuclear plant: 1 - the reactor core; 2 - the helium-to-helium heat exchanger section; 3 - the primary blower; 4 - the emergency cooling heat exchanger; 5 - the steam generator; 6 - the secondary "blower; 7 - the rejected and pumping facility; 8 - the fuel element discharge vessel; 9 - the separating device; IC - the failed fuel element tank; 11 - the handling device; 12 - BPOiS heat exchanger; 13 - the air heat exchanger; 14 - the accelerating device; 15 - the retarding device Pig. 2 Reactor VGR-2OC 1 - the reactor head; 2 - the helium-to-helium heat exchanger; 3 - annular radiation protection; 4 - the upper radiation protection; 5 - the gas collector; 6 - the top reflecter; 7 - the core; 8 - the control and protection system absorbing rod; 9 - the reactor vessel; 10 - the handling device; 11 - the graphite laying.
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( LURG1)
XA0101493
UTILIZATION OF PROCESS HEAT FROM THE HTRM IN THE CHEMICAL AND RELATED INDUSTRIES
M. Schad H. Barnert R. Candeli
(Lurgi) (KFA Jiilich) (Interatom)
IAEA Technical Committee Meeting on Gas-Cooled Reactor Technology, Safety and Siting Dimitrovgrad, USSR, 21 - 23 June 1989
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328
( LURGI )
- 1 -
UTILIZATION OF PROCESS HEAT FROM THE HTRM IN THE CHEMICAL AND RELATED INDUSTRIES M. K. Schad (Lurgi GmbH, Frankfurt/M, West Germany) Abstract: Lurgi investigated the feasibility of supplying industrial processes with heat and energy from a Module High Temperature Reactor in an extensive study. This study shows that there are several processes suitable for coupling with the HTRM almost immediately and only require the layouts are tested. The most interesting process in this respect with high market potential are aluminium oxide production and crude oil refining. Nulcear Process Heat, HTRM, Aluminium Oxide, Crude Oil Refinery Introduction About 3 years ago Lurgi, as plant contractor, supported by BMFT began to search for suitable processes to supply process heat and energy from the HTRM and assess them, with the scope of a study. The purpose of these endeavours was to extended the application range of the HTRM and to establish potential uses so that the work still necessary to introduce the HTRM as process heat supplier could be approached with purpose and effectively. The result of the study is: There are processes for which process heat and energy supplied by the HTRM can be used immediately. Only layout trials are practically necessary for this, i.e. no major long-term development work will be necessary. Known technologies only require to be optimally combined. Economic operation is possible and a potential market on hand. In addition arguments such as polution control, saving of resources and non-existant infrastructure can favour the competitiveness of the HTRM. The range of application of the HTRM for supplying process heat and energy is there where large deposits of raw materials are exploited in the long-term and where considerable quantities of process heat and energy are required for their convertion into intermediate and final products. However, the HTRM was not designed market-specific for the process plant operator consciously with the wide and large potential in mind. 'In order to gain this potential - it is equivalent in the order of size to the process steam and power supply - for the HTRM, the HTRM must be offered to meet the requirements of the customer with flexible operating parameters smaller reactor sizes variable helium inlet and outlet temperatures and as reasonably priced as possible.
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32M
(LURGI)
- 2-
The processes to be combined with the HTRM must, of course, be adjusted to the HRTM in regard to their operating parameters. In the second study supported by BMFT, begun in autumn 1988, the optimum concept both for combining the refinery and combining aluminium oxide production with the HTRM is being prepared by Lurgi in cooperation with Interatom. This work shall have progressed so far by the beginning of 1990 at the latest that serious discussion can be held with potential customers. Processes Investigated Table 1 shows processes Lurgi considers suitable for combining with the HTRM and their assessment.
•
Heavy Oil Production
•
Refinery
•
Oil Sand -
•
Oil Shale-
•
Seawater Desalination
•
Aluminium Oxide Production
Retorting
•
Hydrogen
•
Methanol Production
•
Ammonia Production
•
Petrochemical
•
Cement
•
Iron Ore Sintering
•
Iron Ore Pelletizing
•
Coal Gasification
•
Coarse Ceramics
Tab. 1 Processes Examined for Combination with HTRM
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( LURGI )
- 3 -
The potential processes are summarized in the topmost group in this table. After the concept is laid down they only demand layout checks and in regard to economics some have already a market today. The middle group require either considerable development or currently there is no market for them. Both reasons apply to the bottom group. The topmost four processes are mineral oil technological processes. In mineral oil technology temperatures in the range of 600 - 700 °C are required, i.e. only conventional engineering technology is necessary, if hydrogen production is neglected. At the most 10 - 15 % of the energy requirement of such a plant is required for this purpose and byproducts not suitable for use elsewhere can be applied as energy source here. In the case of sea-water desalination the temperatures are still lower. Aluminium oxide production does indeed require process temperatures of 950°C - 1000 °C but the process heat requirement exceeding 800 °C is only about 2% of the energy demand of the overall plant so that it has little influence on the economics of the plant if this residual process heat is supplied by the HTRM, e.g. through power generation. The combination of the HTRM with an aluminium oxide plant is, however, especially interesting as the world's aluminium industry is currently extending its capacities. The reformer and the steam generator are developed for supplying heat to the hydrogen generation facility in methanol and ammonia plants. Some of the plants installed in the Western world have, however, been shutdown. Therefore, application for the HTRM will be possibly only in countries with greatly expanding economies when new plants are built. There are a large number of plants in the petrochemical industry and, therefore, revamping is to be expected but here considerably development work must be done to be able to combine the HTRM with petrochemical plants. The bottommost group involves processes whose highest process temperatures are about 1000 °C. Only 60 - 80 % of the required energy can be supply by the HTRM. In their case some of the heat is transferred in a temperature range which cannot be utilized without trouble with the materials available on the market. Extensive development work is, therefore, required. Combining the HTRM with Process Plants The problems still to be solved in regard to combining the HTRM with process plants will now be illustrated with the refinery and the aluminium oxide production, two processes viewed as being close to the market:
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2>2>A
( LURGI )
200 MW
300'
750°
-
4 -
200 MW
300*
750°
A V 250°
_Nuclear
700°
250'
700°
Conventionell
Crude Oil* Distillation 154MW
Vacuum-' flasher 52 MW
Propane * Deasphalt. 60 MW
vacuum Residue * Desulfuriz. 22 MW
Vacuum Distillation Desulfuriz 20 MW
Inter-Dist.* Desulfuriz. 17 MW
Benzene * Desulfuriz. 13 MW
Reformer * 62 MW
* With Steam Generator
Fig. 1 Process Heat Supply of a Refinery with HTRM The refinery (see Fig. 1) comprises a large number of individual plants combined with one another. The majority of these plants require heat generated today by combusting byproducts, some of which are suitable for further processing. The trend is towards residue-free refineries. Therefore, in future these byproducts will be further processed and must be replaced by other sources of energy. The throughput of a refinery is about 6 - 7 million tons of crude oil per year requiring a thermal rating of about 400 MW. These quantities of process heat must be transmitted using an intermediate heat transfer loop. It is more economical to built, maintain and repair the large number of heat exchangers conventionally and not to nuclear specifications. The HTRM and the necessary He/He intermediate heat exchangers and valves required to effect this combination - the requisite HTRM helium outlet temperature is 750 °C •-. were developed during the Project Nuclear Process Heat. In addition only heat exchangers to transfer the heat from the secondary helium to the process medium are required. Fig. 2 shows the first draft of a heliumheated crude oil heater. It is a steel vessel insulated on the inside. The hot helium enters at the bottom, flows through a tube bundle filled with crude oil and leaves the C-2
( LURGI )
Feed Water Steam
- 5 -
Crude Oil 220 "C 2.5bar
17 250
Crude Oil + Crude Oil Vapor 370 °C 1.5 bar i
Helium 250 °C ~ 40 bar
Fig. 2 Crude Oil Heater steel vessel at the bottom in the center. It is to be observed that the media to be heated are heated extremely gently. A simple component which must only be laid out and tested. It is considerably smaller than conventional fired heaters and permits more uniform heating of the process medium than they do. : Aluminium oxide production from bauxite requires three stages {see. Fig. 3). In the first the aluminium hydroxide is leached from the bauxite by caustic soda solution. For this purpose hot, liquid salt heats the caustic soda solution to 250 °Crequiring a helium liquid salt heater. In the second stage the aluminium hydroxide is separated from the caustic soda solution. Process steam is required here, which is produced in a steam generator and at the same time used to cover the plant power demand. The third stage, calcinating the aluminium hydroxide to aluminium oxide, is carried out in a fluid bed heater. All the heat exchangers required for this HTRM combination are operated conventionally and must only be converted for heating with secondary helium.;•; :
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C LURGI )
- 6 -
Liquid Salt
250 °C
260 °C
Fig. 3 Aluminium Oxide Production The concept of the fluid bed heater is shown in Fig. 4. Aluminium hydroxide enters at 200 °C on the left and leaves on the right as aluminium oxide at 950 - 1000 °C. The aluminium oxide is brought up to 950 - 1000 °C with electric heating in the last fluid bed heater chamber. The helium enters the fluid bed at 850 °C and leaves it at 680 °C.
Air + Steam 460 °C
Al
4 500
A Air 700 °C Helium heated
Electrically heated
Fig. 4 Fluid Bed Heater
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( LURGI )
- 7-
In the main liquid or gaseous hydrocarbons are to be heated gently in the refinery. That also applies to the overall mineral oil technology so. that the development of the combination of refinery with HTRM in principle covers the requirements for mineral oil technology. Similar applies to aluminium oxide production in regard to transferring heat to fine-particle solids. Economics The question in regard to the possible applications of the HTRM is: Can the combination of HTRM and process plant be operated economically?
(2 x 200 MW, Refinery) (2x170MW,AI 2 O 3 )
16
18 /
20
Years of Operation
/
(2x 170 + SL, AI2O3 -10%) 17CIMW + SI., AI,03)
Fig. 5 Accumulated Cost Difference: 2 HTRM - Heavy Heating Oil (240 DM/t) Inflation Rate: 3%/a, Heavy Oil 5%/a Fig. 5 shows the estimate of the accumulated cost differences for process heat and energy supplied by two HTRMs and by the combustion of fuel oil. At today's fuel oil prices and an inflation rate of 2 % higher for the fuel oil as the general rate the HTRM would be more favourable in'regard to cost than the fuel oil after approximately 14 years of operation. The bottom curve of this cost difference estimate shows the combination of two 170 MW HTRMs with the aluminium oxide plant. The 950 °C helium outlet temperature generated by the HTRM are not necessary for this purpose. A higher HTRM output and, thus, at the same time the cost reduction due to higher quality material not being required, should permit a considerable cost reduction. C-2
( LURGI )
- 8 -
The second curve from the bottom shows the cost difference at 10 % lower capital cost for the HTRM. A higher reduction ought certainly to be possible for the above example but does, however, pressume that the HTRM industry offers the HTRM with more flexible operating parameters. The third curve from the bottom compares fuel oil with two HTRMs without intermediate circulation system. The curve does, however, not take into account possible reactor-building-like capsuling of the process plant components, which would be the primary loop components in this plant concept. For processes having their greatest heat requirement in the upper temperature range of the HTRM the intermediate heat transfer system must, however, be dropped which is only possible if the helium is extremely pure in every operating condition. The prerequisites for this is improved retention of fission products in the fuel element and sufficient capacity in the helium cleaning system. The topmost curve shows the cost comparison for HTRMs with 200 MW output as required for refineries. This combination is better in regard to cost due to the specific higher reactor power density despite a secondary heat transfer system. Further considerable savings ought certainly to be possible due to process adaption and more flexibility in the HTRM operating data. A lack of infrastructure or a stronger oil price increase can shift these curves in favour of the HTRM. The question above is thus answered in the affirmative. Justification of the Development The development of combining the HTRM with process plants is only worthwhile, if the possible demand promises a larger market. A very rough assessment of the HTRM market potential was carried out for this reasons (see Tab. 2 ) . For this purpose the energy requirement of the installed capacities of the processes examined by Lurgi were devided in each case by the output of the suitable HTRM (170 or 200 MW). This shows that the installed heat requirement is equivalent to a capacity of about 2000 HTRMs. Even only 1 or 2 % of this - due to replacement - is a good HTRM market which justifies pushing the necessary development work ahead. This even applies when one disregards the currently less attractive processes in the bottom third of this table, since particular large numbers are attainable for retorting oil shale, sea-water desaltination and coal gasification.
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( LURGI )
- 9-
WorldProduction (Mio-tons/a)
HTRMPotential
25
40 (1514)
3000
850
42
40
41 [14]
104 [36]
Petrochemicals
46
266
Methanol Production
25
50
Ammonia Production
100
270
Iron Ore Sintering
400
80
Iron Ore Pelletizing
350
47
1000
450
Heavy Oil Production Refinery Oil Sand - Retorting Oil Shale - Retorting Sea-Water Desalination Aluminium Oxide Production
Coal Gasification Cement
2 2233 1 % = 22 () with respect to known reserves I] planed
Tab. 2 Estimated HTRM Potential Requirements on the HTRM As shown, the potential market in process heat for the HTRM is of considerable size. To ensure that the HTRM obtains a percentage of this appropriate to its technical possibilities, its operating and design parameters must be more flexible, i.e. meet the requirements of a customer spectrum, as large as possible. The average plant complex examined by Lurgi has a process heat and energy requirements of about 400 WM-th. In general it will not be shutdown, partial or completely, if at all, for a general overhaul or due to the failure of components. - The operators of some refineries informed us that their plants have been in operation for 30 years without interruption. - These are conditions presuming a constant but flexible process heat and energy supply. 400 MW-th output can be generated with two HTRMs. If, however, one of these HTRM must be shutdown, half the process plant, when not all, must be taken out of operation. An undesirable situation and the operator must be protected from the financial losses caused by production loss. This is only to be counteracted by several HTRMs with lower output. Therefore, the HTRM ought to be offered with appropriate capacities below 100 MW-th in addition to the usual one of 170 and 200 MW-th. G-2
( LURGI )
- 10 -
In regard to the average plant this means: The plant would be equipped with 5 HTRM and an additional HTRM or fossil-fired heater as standby. Each HTRM could thus be maintained or repaired indepently without influencing process plant operation. The HTRMs could be selected in such a way that they more effectively cover the necessary energy demand over the temperature range of the process. Rough cost estimates have shown that smaller HTRMs can be up to about 20% specifically more expensive at constantly ensured as well as effective process heat and energy supply. Heat is f.i. only required in the range 900 °C to 700 °C for generating hydrogen for pure hydrogen or for manufacturing ammonia and methanol. The remaining heat can only be used for generating power when the HTRM cannot also supply another plant. This is an economic disadvantage when power cannot be sold due to the lack of an infrastructure. At a higher HTRM helium inlet temperature more heat could be supplied to the reformer. Similar also applies to coal gasification, sintering and pelletizing iron ore. In the case of aluminium oxide production a low helium outlet temperature, e.g. 850 °C would lead to lower material costs and higher reactor output and, thus, compensate the extra costs for the intermediate heat transfer loop. Summary To summarize it can be said: The HTRM can technically be combined immediately with some process plants. Combination with these plants requires no development but only testing layout data. The HTRM can supply process plant with process heat and energy with high probability at better cost than heavy fuel oil. For example, in the case of aluminium oxide production as in the case of the majority of other processes the discharge of flue gases is eliminated completely and in refineries too insofar as the residues can be further processed. There is a market for various HTRM process plant combinations. Other factors such as polution control, saving of resources and lack of infrastructure favour the introduction. To utilize these HTRM advantages the more easily realizable HTRM combinations ought to be prepared, more complicate ones can follow and profit from the simpler ones. However, for this purpose it is necessary that the HTRM industry offer the HTRM with flexible operating parameters. The HTR can contribute considerably to the reduction of CO2 problem.
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( LURGI)
- 11 -
As opposed to these there are acceptance by the public and in part hesitation of process plant operators in regard to other technologies. Despite this the prospects for the application of this type of nuclear energy are favourable enough due to the high potential as well as the possible competitiveness to push ahead, especially in regard to the more easily introduced combinations. All that is missing are prepared concepts to discuss with potential customers.
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XA0101494
IAEA TECHNICAL COMMITTEE MEETING ON GAS-COOLED REACTOR TECHNOLOGY, SAFETY AND SITING
DIMITROVGRAD, USSR, JUNE 21 - 23, 1989.
SAFETY ASSESSMENT PRINCIPLES FOR REACTOR PROTECTION SYSTEMS IN THE UNITED KINGDOM
W PHILP
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CONTENTS
1.
Introduction.
2.
HM Nuclear Installations Inspectorate (Nil). Assessment Principles.
3.
Nil Assessment Principles and Societal Risks.
4.
Principles and Guidance for the Assessment of Reactor Protection Systems.
5.
Assessment of Reactor Shutdown Systems. 5.1
Introduction
5.2
Deterministic Requirements.
5.3
Reliability of Shutdown Systems.
6.
References.
7.
Figures.
8.
Annex 1 - HM Nuclear Installations Inspectorate Safety Assessment Principles for Reactor Protection Systems.
ACKNOWLEDGEMENTS The author is grateful to his Nil colleagues for helpful comments on the manuscript.
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1.
INTRODUCTION
v: T'-•/"•••..•
In the United Kingdom the legislation for the safety of nuclear installations is the Health and Safety at Work etc Act 1974 and the associated statutory provisions in the Nuclear Installations Act 1965.
Under these Acts a nuclear site licence has to be granted by
the Health and Safety Executive (HSE) for all commercial nuclear installations.
HM Nuclear Installations Inspectorate (Nil) is that
part of the Health and Safety Executive responsible for administering this licensing function.
The general requirements for the safety of
nuclear power plants are contained in the Nuclear Installations Act 1.965 and the specific requirements are the responsibility of the Nuclear Installations Inspectorate.
These allow different forms of
requirements, for example, as conditions attached to the site licence. The duty of the Nuclear Installations Inspectorate is to see that the appropriate standards are developed, achieved and maintained by the plant operators, and to monitor and regulate the safety of the plant by means of its powers under the licence.
The Nil carry out this
duty by assessment of the proposed sites and nuclear plant designs, by the establishment of safety requirements and by inspection for compliance with these requirements. 2.
HM NUCLEAR INSTALLATIONS INSPECTORATE ASSESSMENT PRINCIPLES
The Inspectorate does not issue standards or codes of practice for nuclear plant, but it requires each plant operator to develop its own safety criteria and requirements. These criteria are not formally approved, though they normally make reference to national and international standards, such as the British Standards Institution. The Inspectorate has developed safety assessment principles for nuclear power reactors 1 which have been developed primarily as a guide to its own staff, but also with a view to assisting designers and operators. These principles form a statement of the Inspectorate's views and requirements on particular aspects of design safety assessment of reactors and their ancillary plant.
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The principles are divided into three basic categories. The first is a set of fundamental principles on radiological protection, the second sets out basic principles on the limitation of the radiological consequences of the operation of a nuclear plant in both normal and accident conditions and the third category is concerned with the engineering features of the plant. Many of the safety assessment principles are concerned with the engineered safeguards and.the protection system used to preserve the integrity of the physical containments in fault situations.
The
basic requirements being that the reactor can be shut down and cooled after shutdown with high reliability.
The nature and extent of
assessment of an installation design depends on whether the plant is of an established or a new type.
In the case of the Magnox and AGR
reactors the designs have followed a pattern of steady development in the UK and the plant operators have developed design safety criteria 7 to provide guidelines for the design and performance of the reactor protection equipment.
As part of the engineering assessment
of the design the Inspectorate's assessor checks the results of fault analysis with the reliability and probability estimates for systems and components of the equipment.
Such a numerical approach has been
a feature of the Inspectorate's safety assessment for many years. More recently, plant operators have been expected to include a risk assessment for proposed nuclear reactors.
This is proving to be of
value as a means of assessing the performance of safety-related components and systems and revealing possible weaknesses.
However
there are uncertainties associated with the data available and the Inspectorate does not consider that a safety case for a reactor protection system can be based solely on this approach.
The
following sections discuss other components of the safety case. 3.
Nil ASSESSMENT PRINCIPLES AND SOCIETAL RISKS
There has been considerable public debate in the United Kingdom about the risk from nuclear installations and in response to a recommendation in the report of the public inquiry into the Sizewell B station The Health and Safety Executive published the document "The Tolerability of Risk from Nuclear Power Stations2" in February 1988. The document's purpose was to provide an account, for the interested 3
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public, of the way the nuclear risk is assessed and regulated, and a comparison to other risks. It suggests that a dividing line can be drawn between the level of risk which can be tolerated by the public and that which is intolerable and it discusses the relationship between the proposed tolerable risk levels and the Nil's safety assessment principles. An indication is given of the doses and hence risk which in practice result from the application of these principles and of the operating regime applied by the Generating Boards. The document makes it clear that there is no such thing as zero risk in human activities.
Instead of aiming at the unachievable the main
tests that are applied in regulating risk involve determining: a.
"whether a given risk is so great or the outcome so unacceptable that it must be refused altogether; or
b.
whether the risk is, or has been made, so small that no further precaution is necessary; or
c.
if a risk falls between these two states, that it has been reduced to the lowest level practicable, bearing in mind the benefits flowing from its acceptance and taking into account the costs of any further reduction. The injection laid down in safety law is that any risk must be reduced so far as reasonably practicable, or to a level which is "as low as reasonably practicable" (ALARP principle)".
(By risk is meant the likelihood of a specified undesired event/s occurring within a specified period or in specified circumstances). These concepts are shown in Figure 2, taken from the HSE document. Above a certain level a risk is regarded as intolerable and is forbidden whatever the benefit might be. For example, for adult workers, continued exposure above the dose limits set out in the A
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Ionising Radiations Regulations 1985 9 corresponds to this level of risk (bearing in mind the other risks to which the worker might be exposed from his occupation). Below such levels an activity can take place but the closer the risk from it is to the intolerable level the greater the pressure to pursue further safety improvement: this is the ALARP region. In the region of very low risk the regulators need not press for more improvements but the operator is still required to make the risk as low as reasonably practicable. In the case of nuclear power stations, it is the risk from radiation that has to be considered. The main dose limits at present in force in the UK are: Whole body dose limits currently in force in the UK Dose limit for adult employees
50 mSv
In any calendar year
15 mSv
In any calendar year
Dose limit for members of the public
5 mSv
In any calendar year
Annual average dose limit for members of the public exposed over long periods
1 mSv
Investigation level for adult employees
There are other, equivalent, limits eg for doses to organs and other parts of the body and for females of reproductive capacity. In converting these dose limits into risk (of death by contracting cancer) it is currently reckoned that the factor to use is 1 in 100 for every 1,000 mSv received by the whole body.
However, these risk
factors are at present under world-wide review and in the next year or two it may become accepted that the risk is higher than current assumptions by a factor of 2 or 3. C-3
The Tolerability of Risk paper proposes that the maximum tolerable risk for any individual member of the public should be not less than ten times lower than that for a worker, ie 1 in 10,000 per annum. Such a level equates to the average risk of dying in a traffic accident, and is less than everyone's general chance of about 1 in 3 00 per annum of contracting fatal cancer.
The HSE document
therefore suggests that a "broadly acceptable" risk to an individual "below which, so long as precautions are maintained, it would not be reasonable to insist on expensive further improvements to standards" would be one of 1 in a million per annum." In regulating individual and societal risks in accident situations the probabilistic techniques used are a very useful tool but they can only complement good engineering and not supplant it. They can ensure a systematic approach and help with the achievement of a balanced design but there can be considerable uncertainties in the figures that are derived and the estimation of probabilities in areas such as hazards and human factors present difficulties. At best, therefore, probabilistic techniques can only provide an indication, depending on the assumptions made, as to what the risk from a nuclear power station is likely to be. The Nil assessment principles express the standards applied by the Nuclear Inspectorate assessors, and the tolerability of risk document is a paper for public discussion aimed at clarification and definition of the underlying policy. 4.
PRINCIPLES AND GUIDANCE FOR THE ASSESSMENT OF REACTOR PROTECTION SYSTEMS
4.1
Nil SAFETY ASSESSMENT PRINCIPLES (SAP)
The Nil S.A.P.'s define the protection system and the principles to be used in its assessment (see Annex 1 ) . A protection system is all that equipment provided to act in response to a fault so as to prevent, limit or otherwise control the development of any unsafe state in the plant. Each part of the protection system is assigned a specific function and more than one function may have to be performed
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by more than one system to control certain faults.
The actions of
the protection system are defined as: a.
Protective Action The single specified action performed by a channel or group of identified channels, eg reactor trip from one parameter identified in the schedule of identified faults for the reactor.
b.
Protective Function
The combined objective of one or more
protective actions, eg trip reactor. 4.2
Nil ASSESSMENT GUIDES FOR REACTOR PROTECTION SYSTEMS
Assessment: Guides have been developed from experience gained in the past and provide to the assessor reference and guidance from previous assessments. AG1 Single Failure Criterion This document gives guidance on the objective and the application of single failure criterion defined in Principle 112 - No single failure within the protection system should prevent any protective action achieving its required performance in the presence of any specified fault. AG2 Diversity requirements Principle 250 states the CMF limitations for protection equipment should be in the range corresponding to one failure per 10""-* to 10"^ demands. Diversity is normally necessary where the required reliability of a protective function exceeds approximately 10~ 4 failures per demand. AG3 Computer Based Protection This document gives guidance on the application to digital computer protection systems of
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34}
Principles 112 - Single failure criterion, 122 - Diversity of fault detection, 250 - Limitation on reliability. AG4 Fire Protection This document gives guidance on protection against fire hazards. AG5 Qualification of Equipment Qualification of equipment requirements for use in reactor protection systems. AG6 Safety Related Instrumentation Safety-related instrumentation is defined in the introduction of section 3.5 of the S.A.P.'s and the relevant Principles are 132 to 142. 5.0
Nil ASSESSMENT OF REACTOR SHUTDOWN SYSTEMS
5.1
INTRODUCTION
The basic principles laid out in the following pages relate to protection systems for operating UK gas cooled reactors. The application of the assessment principles to a particular system will depend, to some extent, upon the type of plant and the reliability requirements which have to be met. The assessment requirements stated are intended as a guide to illustrate practices which have been used in the past. They are not intended to be exhaustive. 5.2
DETERMINISTIC REQUIREMENTS
5.2.1
General a.
Safety equipment should be segregated from "non-safety" equipment and it should be kept in a locked cubicle with restricted access under administrative control.
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b.
There should be an adequate combination of equipment, with suitable overlap in the ranges of the equipment, and suitable correlation in their functions, to ensure that the plant is protected when an operational veto is correctly applied.
c.
Where operational vetoes are required as part of the plant operation they should be so arranged that in the event of their being applied incorrectly, on either increasing or decreasing signal, the plant shall be shut-down by the protection system.
5.2.2
Multi-Channel Systems a.
The various channels in a multi-channel system should be kept separate at all times; i.e., they should be designed as completely separate as possible, avoiding common connections or sources, in order to limit the possibility of a common fault inhibiting more than one channel.
In
particular, separate cubicles should be used for different groups of parameter instruments and input modules to provide a sufficient amount of segregation. b.
Fail-safe aspects of design should be applied as far as possible, including failure to safety upon loss of power supplies.
A sensor may have to be provided for this
purpose. c.
Each channel in a protection system should be completely segregated from other channels, preferably using separate cubicles. This reduces the risk of inadvertently putting a fault on another exposed channel when one is being maintained, modified or checked. If the channels are in one cubicle, it becomes difficult to substantiate a low probability of failure. The more compact the channels, then generally the greater the danger of faults occurring by cross connections, etc.
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d.
The two ends of a channel should be convincingly segregated, preferably in separate cubicles. Reset contacts and end-relay cross connections should be subjected to careful consideration in this context.
e.
Only one channel should be accessible at any one time for maintenance, modification or checking.
This is usually
achieved by using an interlocking arrangement between cubicles whereby only one key is available. f.
In addition to redundancy being provided for each parameter, at least 2 diverse methods of measurement should be used to detect each frequent fault condition.
g.
System performance (accuracy, response, stability and reliability) should remain within the required specifications under all expected environmental conditions, including environments due to plant fault conditions, and including conditions where electrical and magnetic interference fields may be present.
5.2.3
Controls and Information Presentation
It is important that control settings associated with protection systems should be correctly set at all times and that they cannot normally be re-adjusted without access to keys involving administrative control. The following principles apply. a.
Operational controls should always be calibrated. Arbitrary calibrations should be accompanied by a calibration plate mounted next to the control.
Calibration
should display meaningful units. b.
Where possible, controls should be kept under lock and key.
10
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c.
Control settings should be visible at all times.
d.
Main controls, such as high flux trip settings, should have a range limited to the maximum safe trip level and, in general, multi-range instruments should not be used in the protection system.
e.
After a trip initiation, the system should not be capable of being reset until the parameters being measured for protection purposes have returned to a safe value.
They
should operate as directly as possible into the final shutdown devices. f.
Manual tripping facilities in the form of an emergency push button or switch should at least be provided at the main control and instrument desk.
They should operate as
directly as possible into the final shutdown devices. 5.2.4
Testing and Maintenance
Testing and maintenance procedures have an important bearing on safety and the following principles apply. a.
Each redundant system must be proof tested at regular intervals determined by the reliability requirements; this proof testing procedure should not lessen the degree of protection required, and be carried out without interfering with reactor operation.
b.
After a safety system or part of a safety system has been maintained, it should be thoroughly proof tested against inadvertent short circuits, open connections and other faults which may have been introduced by the maintenance procedure.
c.
A two-out-of-three system should be put into a one-out-oftwo state upon the removal of equipment for maintenance or testing.
While maintaining two-out-of-three systems,
11
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shorting links and other by-passing facilities should not be used. d.
Proof testing signals should be injected at the earliest possible point in the system while the observed results should be taken from the latest possible point in the system under test.
Testing should be possible at any time
without interfering with the reactor operation at power. e.
5.2.5
Test signals should be of a shape which simulates actual operational or fault conditions. Sensors
The following points are made regarding sensors: a.
A direct measurement of the effect guarded against should be made whenever possible.
b.
Signal amplitudes from sensors should preferably be high enough to preclude the need for head amplifiers to be positioned close to areas of high radiation (particularly neutrons), of refuelling activities or of other areas where equipment may suffer damage.
c.
Cables should run between functional units without intermediate junctions.
5.2.6
Human Factors
The quantification of human reliability factors is difficult. Yet human factors enter into every design, operation and maintenance technique. Even if a system is made as automatic as possible, the human factors will always be present. Periods of non-routine operation are particularly prone to human error. (A non-routine operation might for instance be a particular experiment carried out
12
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under unusual conditions, or it might be the introduction of a new operating method.) About three-quarters of human operator accidents are due to maloperation rather than accidental operation. (Maloperation is when an operator in carrying out a required operation performs an unwanted operation; an accidental operation is one where an operation was made when no operation of any sort was required, eg someone leaning on a pushbutton or switch.) In the United Kingdom a considerable proportion of the safety improvements being made to reactors are aimed at reducing the possibility of human error causing a major nuclear accident. Increasing attention has been focused on the importance of what is termed "the operator/plant interface" and the crucial role of operating staff in both preventing and coping with fault conditions. The aim of the UK reactor licensees has been to protect against operator error by providing automatic control systems for the initiation and control of post trip cooling so that no operator action is required during the first 3 0 minutes after the trip.
Also
the plant operators pay considerable attention to the training of all operating staff, with extensive use of simulators designed to familiarise them with abnormal operating conditions. The following assessment principles relating to human factors are applied by the Nil. a.
The system or equipment design should be such that, as far as reasonably possible all operations necessary are of a routine nature.
b.
Designs should attempt to minimise the possibility of maloperation rather than concentrating too much on preventing accidental operation.
c.
The aim should be to rely as far as possible on automatic systems of a fail safe design to protect against all credible faults and maintain the reactor parameters within acceptable limits.
13
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5. 3
RELIABILITY OF SHUTDOWN SYSTEMS
Shutdown systems with only a single means of inserting negative reactivity have been accepted as having a reliability limit of 10~ 5 failures on demand. Only 25% of the control rods (ie typically 10 out of 40 rods) are required to shut down the reactor, and reliability analysis demonstrates that the rod system will meet the 10~5 failures per demand when one or two of the rods fail to enter the core. Because of the large margins that exist between the calculations and the requirements, therefore, the common mode limitation of 10~ 5 failures per demand for control rods is considered justified. Channels using relay logic have been accepted as limited at 10~ 4 failures per demand by common mode effects. On operational reactors to date only channels of the reactor shutdown system using LADDIC elements have been credited with a 10~ 5 failures on demand. In most UK reactor shutdown systems the system reliability is limited by common mode failure of the end relays or the rod trip contactors. To improve reliability the shutdown system is proof tested at regular intervals when the reactor is at power.
UK Central Electricity Generating Board (CEGB) stations test one channel and its associated instruments every month. guidelines
7
The CEGB design
require pessimistic limits on the reliability that can be
claimed for relays etc used in reactor protection channels.
Because
of the pessimistic reliability allowed for protection components CEGB require staggered 3 monthly testing with the reactor at power to demonstrate their reliability claims on most systems. The diagram on the following page is an attempt to summarise the reliabilities which could be expected from the designs of shutdown systems for operating stations. The range of the bars A to E indicate the reliabilities which might be achieved, using the different systems. The bars indicating the possible ranges of the reliability show how the final performance is dependent upon a number of factors; for example, the proof testing interval.
14
c-3
A^UjDEJPJRELlABILlTY RANGFS FOFTVARIQUS SHUTDOWNSYSTFM TYPE! TYPES OF SYSTEM
1E-1
1E-2
1E-3
1E-4
1E-5
1E-6
1E-7
1E-8 1E-8
A - NON REDUNDANT B - PARTIALLY REDUNDANT C - PARTLY DIVERSE TRIP SYSTEM AND REDUNDANT ABSORBER D - DIVERSE TRIP SYSTEM AND REDUNDANT ABSORBER E- DIVERSE TRIP SYSTEM AND DIVERSE ABSORBER
G-3
ASS"
Several types of system are considered and are defined as follows. The non redundant system "A" consists of a single safety monitoring assembly.
The partially redundant system "B"
contains a number of
single safety monitoring assemblies operating in redundancy logic; an example being two-out-of-three logic.
The partly diverse system "C"
is one where at least two separate and physically different parameters are used to detect or control a particular plant condition.
In this case, physically different types of measuring
instruments and sensors are used.
The partial aspect of the
diversity arises because outputs from the 2 parameter measuring systems are fed into a common safety shutdown system.
The fully
diverse trip circuits "D" is a further extension of system "C" but where the instrument outputs are fed into two separate guardline systems, both physically different and separate.
These then feed
into a redundant shutdown system, for example control rods. In E where diverse trip circuits are specified, these are each similar to system "C", but are quite different from one another. For example, one system may measure temperature and pressure, the other will measure neutron flux and flow rate. The whole arrangement from the sensors through to the shut down devices themselves are designed (and operate) on different principles in the separate systems. It is advantageous if two separate design and construction teams are also used in order to ensure segregation of techniques. For the neutron absorber two means of inserting negative reactor must be used, for example, control rods and the Boron Ball Shutdown Devices (BBSD) in the Magnox reactors.
15
C-3
REFERENCES 1.
HM Nuclear Installations Inspectorate, "Safety Assessment Principles for Nuclear Power Reactors", HMSO, ISBN 0 11 883642 0, 1979.
2.
Health and Safety Executive, "The tolerability of risk from nuclear power stations", HMSO ISBN 0 11 883982 9, 1988.
3.
Kelly G N, Hemming C R, Harbison S A, "Procedures to relate the Nil's safety assessment principles for nuclear reactors to risk", NRPB-R.189, HMSO, 1985.
4.
Harbison S A and Kelly G N, "An interpretation of the Nuclear Inspectorate's safety assessment principles for accidental releases", IAEA Seminar on Implications of Probabilistic Risk Assessment, Blackpool, 18-22 March, 1985, IAEA-SR-111/20.
5.
HM Nuclear Installations Inspectorate, "Safety Assessment Principles for Nuclear Power Reactors: Amendment Sheet No. 1", HSE, December, 1988. Hinkley Point "C" Public Inquiry Document S39.
6.
HSC "Comments received on the Tolerability of Risk from Nuclear Power Stations", 1988, HMSO ISBN 0 11 885481.
7.
Central Electricity Generating Board - Design Safety Criteria for CEGB Nuclear Power Stations HS/R167/81 (Revised March 82).
8.
Central Electricity Generating Board - Advanced Gas Cooled Reactor Design Safety Guidelines.
9.
The Ionising Radiations Regulations, 1985 No 1333, ISBN 0 11 057333 1.
C-3 16
7.
FIGURES
1.
The United Kingdom Nuclear Installations Inspectorate.
2.
Tolerable and Actual Levels of Risk to Workers and the Public.
3.
Reactor Trip System Instrumentation.
4.
Schematic of a Reactor Shutdown System.
5.
Principle of the Double 2-out-of-3 Protection System.
The Author W Philp is a Principal Inspector with HM Nuclear Installations Inspectorate. The views expressed in this paper are thos of the Author and do not necessarily- represent those of the Nuclear Installations Inspectorate.
17
c-3
3RANCM A
Section t
Parliamentary. Ministerial. Heaun ana Saltty Commission Business. Liaison wiin otnef Government Oeoartments. Advisory Committee on
Section 2
P'eiicensmg. Nil orincioies ana standards, suing aoucv
Section 3
international liaison, researcn co-oramation. training, decommissioning coney
Section •»
Radiation grotection oolicv. Aavisory Committee on me Transoott ot Raaioactive Materials
Section i
Saietv analysis, fuel 3enavoui. anysics
Section 2
Waste management, cnemical engineering, c'ftemistry
Section 3
Meaitn onysics ana raaioiogicai 3fotectton. fission oroauct :ransoon. cnticautv
Section
Comouters. ergonomics, control ana instrumentation
Section i
Sleei structures, metallurgy fracture mecnames
Section 2
Containment engineering, concrete pressure vessels, ixiernai nazaras
Section 3
Quality assurance, msoection ana examination oroceaures
Section &
Electrical ana mecnamcai engineering
POLICY
BRANCH 3 ASSESSMENT (PHYSICS)
HM CHlEr INSPECTOR OF NUCLEAR INSTALLATIONS
3RANCH C ASSESSMENT ; ENGINEERING!
3PANCH 0 j I I i \
INSPECTION CHEMICAL PLANT. SMALL INSTALLATIONS ANO NON-LICENSED SITSSt
.
Section i
Seiialieid. Ongg. Winascaie Nuciear Laos I U K A E A )
Section 2
Omer 8NFL sites. Sons Royce. Amersnam international, aesearcn Reactors. MOD. U K A E A sues
Section 3
Protect msaection. sianmng ana naison
Section i
Wyifa. Hunterston A ana 3 HinKiev Point A and 8
Section 2
Oungeness A. Siieweil A ana 3. 3erKeiey. OldDury. 3radweii
Section 3
Oungeness 8. Hartieoooi. Heysnam t ana 2. Torness
Section i
Caider Hall. Chaoeicross. Trawsivnyad. Emergency Arrangements
3RANCH E INSPECTION (POWER REACTORS)
C-3
Fiq. L
Tolerable and actual levels of risk to workers and the public Where appropnate the specified risk ranges assume that the risk factors will be increased as recently suggested by NRPB
in
Suggested maximum toleraole risk to workers in any industry
Suggested maximum tolerable risk to any member of the public from any largescale industrial hazard
'0>
1 m 10*
== Range of risk to Hj average radiation H worker
1 m 10'
li :;«; SI
11
risk Range of to members of the public living near nuclear installations from normal operation
if issSS
I
1 in 10°
Range of risk M to members of H the public living a near nuclear install- • ations from any kind I of nuclear accident ««s IN
1 in 10
Range of risk to the average member of the UK public from normal operation plus possible nuclear accidents
"7C !t is /e". difficult :o assign a c.-ccaomty to the ':SK oorne ov peooie .vno live dose to a slant from its normal ooeranc r ii^cs a^v aoses .vnicn .rav oe receivea sv nciviauais are not onw /erv small but are unascertamaDie: 'or instance :"'•< 3 verv '«•/; oeocie ivi^g ;:cse :o .i 'ew siants are 'eguianv -?xcosec "he -estimate gives ^niy a oroaa ic:ea ot ' " ^ "SKS :iorr-r r.v :re .vro.e 'ince 3' isnc:'? •••nnrl ;:ose -^r^ougr' :o "e itfecled. on SSSSIITIIS;:-; jssumctions.
C-3 Fiq . 2
REACTOR TRIP SYSTEM INSTRUMENTATION POWER SUPPLY INSTRUMENT SUPPLIES
SENSOR
AMPLIFIER ETC. !
TRIP UNIT
FOR SIGNAL CONDITIONING
1. RELAY SYSTEM COIL
SENSOR
AMPLIFIER ETC. FOR SIGNAL CONDITIONING
2. LADDIC SYSTEM
PROTECTION CHANNEL INPUT RELAY
TO REACTOR PROTECTION 3 OUTPUTS REQUIRED FOR 2 OUT OF 3 LOGIC
A SIMPLE REACTOR SHUTDOWN SYSTEM
POWER SUPPLY
POWER SUPPLY END RELAYS
B
8/
TEMPERATURE
LINE RESET
C/
\
B
T1 T2 T3
\ T1 /
T2 /
PI
P2/
T3
/
I
P1 P2
P3/
r
P3 I MAINTENANCE iVETO .
PRESSURE
CONTROL RODS
X\ (M X X /
Y /
Z /
j \J
END
X
1
\
RELAYS o
POWER SUPPLY POWER SUPPLY
(M) HAMOCK METER
FiG 4 C-3
Z
PROTECTION CHANNELS PARAMETER
INSTRUMENT OUTPUT RELAYS
r
TEST TRIP PUSH BUTTON
TO SHUT DOWN DEVICES
PRINCIPLE OF THE DOUBLE 2-OUT-OF-3 PROTECTION CHANNEL •
•
RG 5
C-3
ANNEX 1
HM NUCLEAR INSTALLATIONS INSPECTORATE SAFETY ASSESSMENT PRINCIPLES FOR REACTOR PROTECTION SYSTEMS
The following principles are an extract from HM Nuclear Installations Inspectorate safety assessment principles for nuclear reactors.
Protection system
Introduction The principles in this section are concerned with the equipment and systems which are provided co ensure nuclear safety in che event or" plant faults or possible plant maioperation and with instrumentation whose failure or maloperation has a nuclear safety significance. Such equipment may be divided into two categories: (a) Protection system. All equipment or systems which act directly in the event of fauits :o prevent damage that may lead to the escape of radioactivity, e.g. chat equipment provided to:— interlock against unsafe modes o( operation: prevent, limit or delay the escape of fission products following a fault: crip the reactor when pre-set limits are exceeded, or when a trip is manually initiated: remove heat from the reactor to a heat sink after reactor shut-down: activate any other safety-related system or equipment: provide power to the protection system. (bi Surety-related instrumentation. Instrumentation having a significant but indirect effect on r.uciear safety e.g.: — control systems whose failure can cause a demand on the protection system: instrumentation used to warn oi the onset of hazardous conditions or of conditions requiring manuai safety action: instrumentation for monitoring the protection system, reactor and plant variables and parameters; communications equipment for accident conditions: equipment for monitoring abnormal radioactive releases from the site. In carrying out an assessment oi the protection system ;he assessor should judge the extent to which the C-3
Annex l submission shows conformity with the principles in this section. Protective features of emergency cooling systems, essential supplies and containment are also dealt with in principles 38-106. 148-151 and 152-161 respectively and these should be read in conjunction with the principles of this section. Principles for the protection system 107 Adequate protective systems should be provided and. whenever fuel is in the reactor, they shouid be maintained at a level of readiness adequate to ensure nuclear safety. 108 The reactor and associated plant should be designed, constructed and operated so that the reactor can always be shutdown and held shutdown in a safe sub-critical state thereafter. 109 The reactor and associated plant should be designed, constructed and operated so that it can always be adequately cooled. 110 All those systems which are required to function and provide action in response to any specified fault should be identified in the submission. The aggregate of all such systems comprises a barrier or barriers for that fault. 111 For each specified fault it should be shown that adequate protection is provided and that such protection is capable of maintaining the plant in a safe state for as long as may be necessary following that fault. 112 No single failure within the protection system shouid prevent any protective action achieving its required performance in the presence of any specified fault or external hazard initiating a demand on the protection system. 113 For the purpose of initiating protection each fault sequence should be detected at the most appropriate point in the sequence and as directly as practicable. 114 The variables chosen as indicators of each postulated fault condition should be such as to enable the fault to be reliably and unambiguously detected. 115 The required performance of components, subsystems and systems should be stated and shown to be adequate for the purpose of providing protection. Limits should be defined outside which components etc., should not be operated and provision should be made to ensure that these limits are not infringed. It should be shown that the overall reliability of the protective system is adequate. 116 All variables to be used to initiate protective action should be identified and shown to be sufficient for the purpose of protecting the reactor. Appropriate and safe limits for these variables should be specified which are relevant to the state of the plant at any specified time, tt should be shown that the protective systems are designed to respond tb'rhe appropriate
variables within the above limits and that the resulting performance of the protective system is adequate. 117 Where a directly related variable cannot be used for the purpose of initiating protective action against a fault, a less directly related variable may be employed. In such oases it shouid be shown that the variable chosen to initiate protection has a known relationship with the main variable of concern and with the fault being detected. The physical coupling between the measured variable and the fault condition should be as close as practicable. 118 The final actions of the protection system should be achieved by means such that there is a known and direct relationship with the desired final objective. 119 Means should be provided to enable the necessary calibration and checks on the functioning of any measuring device used in a protection system to be carried out at appropriate times throughout the life of the plant commensurate with the reliability requirement. 120 When equipment has more than one function, one of which is to ensure nuclear safety, this equipment should be classed as protection equipment. The protective function should not be jeopardised by the other functions. 121 It must be recognised that unforseen plant or protection system faults or maloperations may occur. Protection system design should reflect this aspect by, for example, the provision of reasonably practicable diversity and redundancy, both within each system and in the nature of each input and output. 122 Diversity of fault detection and protection should be employed where reasonably practicable but where protection system reliability is required to be very high or when there is doubt about the reliability or effectiveness of a non-diverse system diversity should be introduced. 123 The protection system equipment should be so designed, laid out and sited that, notwithstanding the effect of plant faults, adequate protective action will be available. 124 The protection system shouid be automatically initiated. No operator action should be necessary in a timescale of approximately 30 minutes. The design should however be such that an operator can initiate protection system functions and can perform necessary actions to deal with circumstances which might prejudice the maintenance of the plant in a safe state but cannot negate correct protection system action at any time. 125 Only components having a proven reliability and performance should be selected for use in any protection svstem. C-3
Annex 1 126 Spurious operation of the protection system should not produce an unacceptable condition in the plant. 127 The minimum amount of operational protection equipment for which reactor operation will be permitted should be specified. Equipment being tested or maintained cannot be claimed as operational where the test or maintenance conditions put the plant into a less safe state. 128 Where a mechanism (including external hazards) can be foreseen which could invalidate more than one redundant or diverse protective function, action or channel, then its probability of occurrence should have an insignificant effect upon the combined reliability claimed for those functions, actions or channels. Additionally this should be applied to those mechanisms which could cause an initiating plant fault and failure of the associated protective functions. 129 Alarms should be provided to give warning that any safety-related system, component or parameter is at a pre-set limit of its acceptable operational state. Where reasonably practicable alarms should be initiated in the event of any unsafe failure of any element of a protective system. 130 Where required on nuclear safety grounds all protection system equipment including pipework and cabling should be segregated from all other equipment and its function clearly indicated. Where interaction or proximity to non-protection equipment or cabling is required each case should be justified. The segregation of equipment and cabling within the protection system should be such as to satisfy principle 128. 131 The design should be such that the means of access to all protection equipment can be physically controlled to limit access to an extent which ensures availability of the minimum amount of operational equipment referred to in principle 127. Instrumentation 132 Provision should be made in the form of indicating and recording instruments to inform the plant operators at all specified times ot' the state ot those items which have a significant influence on safety and on safety-related aspects of the overall plant state. Such provisions should include devices to give advance warning of unacceptable changes and rates of change and also alarms when set limits are reached. Sufficient information should be made available to the operator at all times to enable an accurate appreciation to be made of the plant state so that all actions necessary in the interest of safety can be taken promptly and effectively. Such instrumentation should as appropriate and where practicable be capable of monitoring, controlling and recording each parameter at all specified times. Provisions made to monitor, record and control the plant should be shown to be
2=66
effective at all specified times so far as is necessary for safe operation oi the plant. 133 The provision of control, monitoring and recording equipment should include equipment relevant to postulated fault conditions and should be suitable to enable the operator to assess plant state and take necessary control action during such faults. 134 There shouid be provided a suitable communications system to enable information and instructions to be transmitted between locations and to provide external communications with auxiliary services and such other organisations as may be required. 135 A reliable fire warning system should be provided for all parts of the protection system except where the design precludes a fire hazard. 136 The instrumentation provided to meet the requirements of this pan should enable an operator to take all necessary actions from a central control room. Adequate protection against radiation, contamination, toxic hazards and against plant faulrs shouid be provided to permit occupancy of the control room under plant fault or accident conditions without personnel being harmed or receiving radiation exposures in excess of the requirement of the radiological principles. 137 Instrumentation and control equipment should be provided at locations other than the main control room to enable the reactor to be manually shut down, maintained in a safe state and effective accident control undertaken should che central control room become inoperable or uninhabitable. 138 The minimum safety-related instrumentation for which reactor operation may be permitted should be specified. 139 All instrumentation should be of the highest quality appropriate to the duty. Evidence should be provided of its satisfactory performance under the worst environmental conditions anticipated. 140 The accuracy, stability, response time and range of all instrumentation should be adequate and appropriate for its required service at all times throughout plant life. 141 All safety-related instrumentation should be supplied from power supplies whose reliability is compa- • tible with the function being performed. In the case of monitoring, warning and communication functions this supply should be non-break. 142 Adequate means should be provided for the testing and calibration of ail safety-related instrumentation at any specified time without loss oi any essential functions.
Special principles for shut down systems The above general principles of protection should be C-3
Annex 1
acpiieu by :r.e assessor a> appropriate ro >y-.tem> a> MTOU:J ;he :'oilowin«j pnnc:pies :n addition: 143 Taking account oi appropriate requirements of these principles. :he shut down system should be capable of shutting down the reactor and holding it sub-critical with a margin of negative reactivity which should be avaiiabie at all specified times and which should allow for uncertainties in nuclear characteristics, perturbations in plant state stc. 144 The design or' the reactor should be such chat shutdown is not prevented by the other components oi the nuclear power plant or by mechanical failure, distortion, corrosion, erosion etc., or' plant compo- . nents or by the physical behaviour oi the reactor coolant, during normal operation or any postulated fault condition. 145 The design oi each shutdown system should be such that loss oi absorbing material due :o physical or chemical changes such as melting, boiling, leaking or • mechanical damage is either prevented or is kept within specified iimits so as not to lead :o an unacceptable !oss oi shutdown margin. 146 Retrievable shutdown devices should be capable of being tested and inspected in accordance with the requirements set out in the general principies. Nonrecrievable shutdown devices should be capabie oi being subject to such tests as are practicable in the reacior supplemented by proof and reliability tests in an appropriate facility out or" the reactor. 147 There should be supplied in the submission a design specification for the shutdown devices which should take into account: (a) allowances for changes in geometrical configuration due :o temperature, irradiation etc.: (b) the allowance for variations of neutron absorber concentration due ;o burnup. diffusion, deposition, corrosion ere: (c) the production oi capture or fission products within :he absorber assemblies: (d) the physical behaviour of the absorber assembly at all times throughout plant life: (ei allowance r'or reactivity changes in the >hutdown provision due to physical and chemical changes throughout piant life. At least one long-term shutdown system should not require an external energy source to maintain the reactor in a shutdown state.
C-3
XA0101495
Design and Safety Consideration in the High-Temperature Engineering Test Reactor (HTTR)
Shinzo SAITO, Toshiuki TANAKA, Yukio SUDO. Osamu BABA, Shusaku SHIOZANA, Minoru OKUBO
Department of HTTR Project Japan Atomic Energy Research Institute
C-4
ABSTRACT
The
budget
Test
Reactor
The
HTTR i s
for
construction
(HTTR) was r e c e n t l y a
test
reactor
with
of
the
committed thermal
High-Temperature by
the
output
Government of
Engineering in
30 MW and
Japan. reactor
outlet coolant temperature of 950 * C at high temperature t e s t operation. The HTTR plant uses a pin-in-block design core and will be used as an experience leading to high temperature applications. Several major important safety considerations design of the HTTR. These are as follows.
are
adopted
in
the
1) A coated particle fuel must not be failed during a normal reactor operation and an anticipated operational occurrence. 2) Two independent and diverse reactor shut-down systems are provided in order to shut down the reactor safely and reliably in any condition. 3)
Back-up reactor cooling systems which are safety ones are provided in order to remove redidual heat of reactor in any condition. 4) Multiple barriers and countermeasures are provided to contain fission products such as a containment, pressure gradient between the primary and secondary cooling circuit and so on though coated particle fuels contain fission products with high reliability. 5) The functions of materials used in the primary cooling circuit are separated to be pressure-resisting and heat-resisting in order to resolve material problems and maintain high reliability. The detailed design of the HTTR was completed accumulation of material data and component tests.
with
extensive
C-4
CONTENTS
1. Introduction
1
2. Safety Design Principles
2
3. Outline of Plant Design
3
4. Specific Considerations in t h e Reactor
4
Safety Design 4.1
Core Design
4
4.1.1
Fuel
4
4.1.2
Core Design
4
4.2
R e a c t i v i t y Control and Shut-down Systems
5
4.2.1
Control Rod System
5
4.2.2
Backup Reactor Shut-down System
5
4.2.3
Stand-pipe Fixing Device
6
4.3
Residual Heat Removal Systems
6
4.3.1
Auxiliary Cooling System
6
4.3.2
Vessel Cooling System
7
4A
Multiple
Barriers t o
Fission
Products Release
8
4.4.1
Reactor Coolant Pressure Boundary
8
4.4.2
Containment Vessel
8
5. Afterword
9
C-4
1. Introduction The High Temperature Engineering Test Reactor (HTTR) aims at establishing and upgrading the technology basis necessary for an HTGR, serving at the same time as a potential tool for new and innovative basic researches. The HTTR has a prismatic block type fuel core with 30 MW thermal output and outlet coolant temperature of 850 °C at rated operation and 950 °C at high temperature t e s t operation. The JAERI has carried out for many years development works on block type fuel, high temperature alloy, high temperature in-core instrumentations, high temperature components etc. and research works on reactor physics of a high temperature gas cooled reactor, heat transfer and fluid dynamics, fission products plate-out etc., in order to construct the HTTR which can supply high temperature coolant up to 950 ° C at the outlet of the reactor vessel for the first time in the world. This report describes designs and several major safety the design of the HTTR.
considerations in
C-4
2. Safety Design Principles The JAERI has set up following safety design principles for the HTTR referring the safety design criteria for LWRs taking into account the inherent safety characteristics as an HTGR and design requirements as a t e s t reactor. The major items of the principles are as follows. (1)
Coated particle fuel shall not be failed during normal reactor operation and an anticipated operational occurrence. The maximum fuel temperature including systematic and random factors should not exceed 1600°C even in an anticipated operational occurrence.
(2)
A reactor shall be shut down safely and reliably from any operation condition with a control rod system. A backup reactor shutdown system shall be provided independent from the control rod system.
(3) An accident of control rod ejection shall be avoided. (4) The residual core heat after reactor shutdown shall be removed safely and reliably for any anticipated operational occurrences and accidents. (5) A containment vessel shall be provided to prevent fission product release and excessive air ingress into the core in case of depressurization accident. (6)
The pressure of water in the secondary water cooling system shall be controlled lower than that of primary helium to prevent large water ingress into the core in case of rupture of heat tube in a pressurized water cooler. (7) The pressure of helium in the secondary helium cooling system shall be controlled slightly higher than that of primary helium to prevent fission products leakage from the primary system to the secondary system through cracks of the heat tubes in the heat exchanger. (8)
The pressure-resisting and heat-resisting functions of the structures where high temperature coolant is contained are separated in order to reduce mechanical loads on high temperature metal structures.
- 2 -
332-
C-4
3. Outline of Plant Design The HTTR consists of a core of 30 MW thermal output, a main cooling system, an auxiliary cooling system and related systems. The reactor core is contained in a steel vessel of 13.2 m in height and 5.5 m in diameter together with graphite reflectors and core support structure as shown in Fig. 1. Major specifications are listed in Table 1. The block type fuel of the HTTR is shown in Fig. 2. The main cooling system is composed of a primary cooling system, a secondary helium cooling system and a pressurized water cooling system as shown in Fig. 3. The primary cooling system is separated into two lines outside the reactor vessel. The heated helium gas is cooled in a He/He intermediate heat exchanger in one line or cooled directly in a pressurized water cooler in the other line. The heat is finally removed by an air cooler in both lines, although another pressurized water cooler is necessary after the intermediate heat exchanger in the first line. When the first line with heat transfer capacity of 10 MW is operated, the second line, which has heat transfer capacity of 30 MW, is operated at 20 MW. The intermediate heat exchanger is a vertical helically-coiled counter flow type heat exchanger in which primary coolant flows on the shell side and secondary coolant on the tube side as shown in Fig. 4. Materials for the pressure boundary below and above 440°C are 2 1/4 Cr-1 Mo steel and Hastelloy -XR, respectively. The heat transfer tube is as designed as to withstand the differential pressure of 0.3 MPa between the primary and secondary helium in normal operation condition, while it is designed for 4 MPa for short duration of depressurization accident of secondary helium cooling system. The pressurized water cooler is a vertical U-type heat exchanger. Primary helium gas coolant flows outside the heat transfer tubes and pressurized water inside the tubes as shown in Fig. 5. Heat transfer tubes are designed for the primary system pressure of 4 MPa. The pressure of water is controlled lower than that of primary coolant. Co-axial double pipes are used for transferring hot helium gas. The hot gas from the reactor flows inside the inner pipe, while cold gas of about 400 °C flows back to the reactor through the annulus between the inner and outer pipes. The outer pipe is designed as a reactor coolant pressure boundary, and the inner pipe is designed only to withstand the differential pressure caused by pressure drop due to coolant flow. - 3 C-4
4. Specific Considerations in the Reactor Safety Design 4.1
Core Design
4.1.1
Fuel
In order to reduce FP release from the core, the maximum fraction of defected fuel during fabrication including contamination in coating and compact matrices is limited to less than O.25C and no fuel failure during normal operation and anticipated operational occurrences is allowed. The failure mechanisms of the fuel considered during operation are (1) palladium attack on sillicon-carbide layer, (2) kernel migration, and (3) burst of coating layer with excess FP gas pressure. From the research and development works in the JAERI, it became clear that failures of coated particle fuel caused by the above-mentioned mechanisms can be kept out practically by limiting fuel temperature and maximum burnup lower than 1600°C ( Fig. 6) and 4 % FIMA, respectively. 4.1.2
Core Design
In order to maintain fuel temperature as low as possible keeping 2.5 w/cc of average power density and 950 °C of reactor outlet coolant temperature, following considerations are taken. (1) To increase effective core coolant flow by minimizing core bypassing flow and cross flow. (2) To flatten radial power distribution in the core. (3) To optimize axial power distribution in order to attain uniform fuel temperature. In order to minimize core bypassing flow, some sealing devices between permanent graphite reflectors and core restraint devices are provided. The core cross flow is reduced with low pressure difference between fuel columns. Coolant flow on a control rod is also limited. For the flattening radial power distribution in the core and optimizing axial power distribution, fuel enrichment is changed regionally. The HTTR core is designed with enough low fuel temperature as shown in Fig. 7, after above-mentioned efforts. _ 4 -
4.2
Reactivity Control and Shut-down Systems The power control and shut-down pairs control rods. When the center ation test, the control rod at the rods, backup reactor shut-down system the pairs of control rods.
4.2.1
of the HTTR are achieved with 16 or 15 column of the core is used for irradicenter is removed. Beside the control is provided with the same number of
Control Rod System
At the reactor start-up, the outermost three pairs of control rods are withdrawn at first as safety rods. All of other control rods are operated as power control rods. The control rod system can achieve subcriticality from any operation conditions and maintain subcriticality in the cold core conditions even when a pair of rods stick at the operation position. In case of the scram at high temperature operation, nine pairs of rods in the replaceable reflector region are inserted quickly and the rest rods are automatically inserted after the core is cooled to some temperature in order to prevent control rod cladding from being too hot above 900°C. A pair of rods are driven by one drive mechanism. The control rods are isolated from the drive mechnisms and inserted by the gravity when the reactor is scrammed. 4.2.2
Backup Reactor Shut-down System
The backup reactor shut-down system the reactor at any operation conditions core condition with shut-down reactivity shut-down system is operated only manually
has reactivity worth to shut down and keep it subcritical in the cold margin of 1S>. The backup reactor when required.
Absorber elements of the backup reactor shut-down system are boroncarbide pellets sintered with graphite powder. A valve at the bottom of the hopper, in which the absorbers are contained, is opened manually, and the absorbers fall into the separate hole in the same graphite block as control rods are inserted. - 5 C-4
4.2.3
Stand-pipe Fixing Device
A stand-pipe fixing device is provided at the which control rod drive mechanisms are installed, stand-pipes together with control rods from jumping pipe rupture accident. The stand-pipe fixing device rate in the accident keeping control rods away from flow.
4.3
top of stand-pipes, in in order to prevent up in case of standalso restricts gas flow floating up by the gas
Residual Heat Removal Systems The main cooling system removes the residual heat in the core in normal reactor shut-down condition. Besides the main cooling system, the HTTR has two other residual heat removal systems which are an auxiliary cooling system for the anticipated operational transients and for such an accident as coolant flow boundary is sound and a vessel cooling system for the accident in which the forced circulation of coolant can not be maintained because of damage in coolant flow boundary. The flow diagram of the auxiliary cooling system and the vessel cooling system are shown in Fig. 8.
4.3.1
Auxiliary Cooling System
The auxiliary cooling system automatically starts up when the reactor is scrammed in the accident as the core cooling by a forced circulation of coolant is possible, while the main cooling system is stopped. The auxiliary cooling system is a safety system with redundant dynamic components as gas circulators, water pumps and valves which are also backed up with emergency power supply. The residual heat of reactor core can be removed by the vessel cooling system without the auxiliary cooling system. The auxiliary cooling system, however, is needed from the viewpoint of operation flexibility, as the vessel cooling system takes very long time to cool down the core.
-
6'•-••
C-4
4.3.2
Vessel Cooling System
The vessel cooling system is used as a residual heat removal system when the forced circulation in the primary cooling circuit cannot be maintained due to the rupture of the inner pipe or both pipes in the co-axial double piping. The vessel cooling system is also a safety system equipped with two independent complete sets which are backed up with emergency power supply. It is operated even in normal operation in order to cool the reactor shielding concrete wall.
4.4
Multiple Barriers to Fission Products Release The HTTR has multiple barriers to coatings,
fission products release
besides fuel
which are a reactor coolant pressure boundary and a containment
vessel. 4.4.1
Reactor Coolant Pressure Boundary
The reactor coolant pressure boundary is designed as follows t o make i t an effective barrier to fission products release.
in order
The pressure boundary of low and/or medium temperature ( below 440 ° C), where design and fabrication with high reliability are possible based on the well-established technology, is designed for system pressure on one side with leaktightness. The pressure boundary of high temperature ( above system pressure is loaded only on one side is avoided.
440 ° C) in
which
The pressure loading on the boundary at high temperature as heat tubes in the intermediate heat exchanger is reduced by minimizing pressure difference between both sides of the boundary. The pressure of the secondary helium in which no fission product is contained, is, maintained a l i t t l e higher than that of primary helium to prevent fission products leakage into the secondary helium through cracks on the boundary.
- 7 -
C-4
The high temperature coolant from the reactor flows in the inner pipe of co-axial double piping with a l i t t l e lower pressure than the outside where the primary coolant cooled by heat exchangers and pressurized by gas circulators flows.
4.4.2
Containment Vessel The functions of the containment vessel in the HTTR are, (1) to contain fission products and (2) to limit amount of air which reactor core in an accident.
possibly
reacts with
graphite
in
the
There is no effective barrier to fission products release for the accident of primary pipe rupture which cannot be excluded from the HTTR safety evaluations, if there is no reactor containment vessel. The HTTR has a steel containment vessel in i t s reactor building and the reactor building serves a confinement which is called "service area". The service area is maintained at slightly negative pressure to the environment by ventillation systems in both normal operation and accident condition. The off-site radiation dose in such an accident as pipe rupture in the primary cooling system is remarkably reduced by the containment vessel together with the confinement. Furthermore, in the accident of primary pipe rupture, countermeasure to limit the amount of the air break-in to the is possible except for the containment vessel. The amount of graphite in the reactor core is limited in very low level with a containment vessel.
- 8 -
no effective reactor core of oxidation in the HTTR
C-*
5. Afterword
The HTTR is a high temperature gas cooled test reactor which has various aims and operational modes, while it can provide at the same time very high temperature coolant up to 950 "C at the outlet of the reactor vessel for the first time in the world. The JAERI has carefully thought-out the safety design principles for it as written in this report reflecting the results of R&D works on HTGR technology and experiences of LWR designing and operation. The JAERI submitted the safety analysis report of the HTTR to the Science and Technology Agency for safety review by the Government on February 10, 1989. The safety review by the Nuclear Safety Commission follows it. The installation permit will be issued in spring 1990 and i t will take about five years for the construction of the HTTR plant. The first reactor criticality will be attained in 1995.
- 9 -
C-4
S3 Table 1
Specifications of IITTR
Thermal power
30 MW
Outlet coolant temperature
850°C/950°C
Inlet coolant temperature
395°C
Fuel
Low enriched UO2
Fuel element type
Prismatic block
Direction of coolant flow
Downward-flow
Pressure vessel
Steel
Number of main cooling loop
1
Heat removal
IHX and PWC (parallel loaded)
Primary coolant pressure
4 MPa
Containment type
Steel containment
Plant lifetime
20 years
>-J
Stand pipe
Permanent reflector Replaceable reflector
Fuel element
Hot plenum block
Support post Lower plenum block Carbon block Bottom block Support plate Core support grid Auxiliary coolant outlet pipe Main coolant outlet pipe
Fig-. 1
Bird's-eye view of reactor vessel and core
C-4
0°
Fuel handling hole Fuel kernel
Plug
High density PyC SiC Low density PyC
Dowel pin
Fuel /compact
0.92mm B mm
Graphite 'sleeve
39mm
34mm
Fuel compact
Fuel rod
o
I,
Fig. 2
Block type fuel of HTTR
(p
IHX
Containment vessel Vessel cooling panel
Water pump,
Air cooler
PPWC : Primary pressurized water cooler PGC : Primary gas circulator SPWC : Secondary pressurized water cooler SGC : Secondary gas circulator AHX : Auxiliary heat exchanger AGC : Auxiliary gas circulator
I
Air cooler
Water pump
Fig. 3
I
Cooling system
: Intermediate heat exchanger
Secondory Helium (to Secondory PWC) Secondory Helium (from Secondory PWC)
Cold Header Primory Helium (to Primary Gas Circulator)
Primary Helium (from Primary Gas Circulator) Inner Shell Outer Shell
Tube Support Assembly
Central Hot Gas Duct (Center Pipe) Thermal Insulator Helically Coiled Heat Transfer Tube Hot Heoder
Primary Helium (to Reactor) . Primary Helium (from Reactor)
Fig. 4
Bird's-eye view of the He/He intermediate heat exchanger of HTTR -
14
-
C-4
Primary Helium (to Primary Gas Circulator in PWC Only Run)
Primary Helium (to Primary Gas Circulator in IHX/PWC Parallel Run)
Outer Shell
Baffle Plate Primory Helium (from Primary Gas Circulator)
Heat Transfer Tube
Thermal Insulator
Inner Shell
Primary Helium (to Reactor) Primary Helium (from. Reactor)
Tubesheet
Pressurized Water (from Pump)
Fig". 5
Partition Plate
Pressurized Water (to Air Cooler )
Bird's-eye view, of the pressurized water cooler of HTTR
- 15 -
C-4
(p i
i
i
i
1 r
i
i ^
i
i
i
i
i
i
i
i
i
i
i
i
i
i
i
I
II
r
i
r
i
i
i
i
i
i
i i
Healing condition Irradiation conditions
10
ramp rate fC/mirt)
Burnup (%FIMA)
Temp
fcf
c o
o
1
3.6
1250
D
•
5
3.6
1250
5
3.5
1510
0) 3
O
0
I I II
I
1800
1
1900
n
•
1I I
2000
I
2100
Heating temperature (°C) Fig. 6
Accumulated failure fraction of irradiated coated particles in ramp tests
I I I II
2200
CP 15.0
Operation time Maximum fuel Temp.
u o
660 EFPD 1326°C
10.0
o
•H •P -0 1
O
M-l
O
o;
5.0
o >
0.0 400
600
Fig. 7
800
1000 Fuel Temperature
1200
Volumetric fraction of fuel temperature in IITTR core
1400 (°C)
(p Cooling Panel
Cooling Tower
Cooling Tower
09 I
Vessel Cooling System
Vessel Cooling System
I I | AHX : Auxiliury Heat Exchanger A6C : Auxiliary Gas Circulator Water Pump
(100%X2) Auxiliary Cooling
I
System
Fig. 8
Residual heat removal systems
Air Cooler
XA0101496
THE DESIGN STATUS OF THE UNITED STATES DEPARTMENT OF ENERGY MODULAR HIGH TEMPERATURE GAS COOLED REACTOR
RAYMOND R.
MILLS,
JR.
DIRECTOR
MODULAR HIGH TEMPERATURE GAS COOLED REACTOR PLANT DESIGN CONTROL OFFICE
C-5
THE DESIGN STATUS OF THE UNITED STATES DEPARTMENT OF ENERGY MODULAR HIGH TEMPERATURE GAS COOLED REACTOR The U.S. Department Energy's Modular High Temperature Gas Cooled Reactor (MHTGR) is being designed using a systems engineering approach referred to as the integrated approach.
The top level requirement for the plant that it
provide safe, reliable, economical energy.
The safety requirements are
established by the U.S. Licensing Authorities, principally the Nuclear Regulatory Commission.
The reliability and economic requirements associated
with the top level functions have been established in close coordination and cooperation with the electrical utilities and other potential users and the nuclear supply industry.
The integrated approach uses functional analysis to
define the functions and sub-functions for the plant and to identifiy quantitatively how the various functions must be fulfilled.
The top four
functions associated with the MHTGR are:
1) Maintain Safe Plant Operation
Reliably maintain the functions necessary for normal plant operation, including the plant states of energy production,shutdown, refueling, and startup.
2) Maintain Plant Protection
Assume that, despite the care taken to maintain plant operation failures will occur, and provide design features or systems to limit plant damage within economic and safety limits.
3) Maintain Control of Radionuclide Release
Provide design features for systems to ensure control of radionuclides within safety limits in the event that Goal 1 and/or Goal 2 requirements are not met.
4)
Maintain Emergency Preparedness
Maintain adequate emergency preparedness to protect the health and safety of the public in the event that Goal 3 requirements are not met. C-5
Mo
Page 2 Goal 1 is to be achieved by highly reliable operation using well trained qualified personnel.
Goals 2 and 3 will be achieved through using inherent
characteristics and passive safety features.
Goals 1, 2, and 3 are to be
achieved so well that minimal reliance will need to be placed on Goal 4.
In addition to meeting all U.S. Regulatory Requirements this advanced reactor concept is being designed to meet the following requirements:
1. Do not require sheltering or evacuating of anyone outside the plant boundary of 425 meters as a result of normal or abnormal plant operation.
2. Do not require operator action in order to accomplish the above sheltering and evacuation objectives and the design must be insensitive to operator errors.
3. Utilize inherent characteristics of materials to develop passive safety features.
4. Provide very long times for corrective actions following the initiation of an abnormal event before plant damage would be incurred.
t
One of the important areas to the MHTGR passive safety characteristics in the event of off normal operations is the Reactor Cavity Cooling System (RCCS). The air cooled RCCS is designed to remove heat from the reactor cavity surrounding the reactor vessel in a passive manner through radiation and convection heat transfer.
Figure 1 provides an illustration of the
relationship between the RCCS and the reactor vessel.
The uninsulated reactor
vessel transfers heat by radiation to outside air through cooling panels located in the reactor cavity.
The reactor vessel also transfers energy
through conduction to the earth surrounding the reactor building.
Cool air
flows from the exterior of the reactor building as a result of natural convection through cooling panels which form a cylindrical area that is totally external to the primary coolant pressure boundary and surround the uninsulated vessel.
The cooling panels and associated ducting collect the
heat transferred from the vessel by radiation and natural convection and transport the heated air to the environment outside of the reactor building. In addition, the cavity cooling panels function to C-5
2f\A
Page 3 protect the concrete cavity wall from overheating during normal operation and provide an alternate method of decay heat removal in the event that forced cooling system functions are lost.
Figure 1 shows schematically the heat
transfer features which are described above.
The RCCS design has no valves or
active components which could malfunction to prevent achieving the design objectives.
The cooling panel surfaces function as a barrier to separate the
outside atmosphere from the atmosphere in the building and the reactor cavity. This separation minimizes the potential for off site radiological exposure which could occur during normal and abnormal operations.
During the present preliminary design phase of the MHTGR computer methods for analyzing the RCCS system will be evaluated and verified to assure that the computer code correctly performs the operations specified in a numerical model demonstrates substantially identical results when compared with known solutions to the problem from other sources.
Subsequent to the computer
methods verification, validation of the methods will be accomplished to assure that the model correctly represents the processes occurring in the RCCS and correlates with data such as experiments or reactor operation.
As part of the
process for the verification and validation of the computer codes and models the codes will be tested against benchmark solutions for the range of appropriate application.
Some appropriate benchmark solutions being
considered for use are comparison with hand calculations, comparison with other computer codes which have been verified and validated, and comparison with experimental or operational data.
Part of the method verification
process will include sensitivity analyses to evaluate the importance of the various physical phenomena to the method results.
The sensitivity analysis
will use an uncertainty band associated with the feature or parameter and a resultant uncertainty in the predicted result will be calculated by the code. Among the principal parameters which control the RCCS performance are the reactor vessel and panel emissivity, the convection coefficients in the reactor vessel cavity and in the panels, the environmental wind conditions, and the inlet and outlet structure configuration.
The impact of these
perimeters will be evaluated to determine the necessity of experimental testing of the reactor cavity cooling system.
During the current preliminary
design phase a determination will be made regarding the need for tests which may include separate effects tests, scale model tests, or full scale testing on the first MHTGR plant to be constructed. C-5
2X7-
Page 4 Another area of importance to plant reliability is the performance of main circulator.
The circulator is in an early stage of preliminary design and is
planned to be a two stage and axial flow machine.
The circulator will be of a
unique design using previous efforts to the maximum extent but modified to be specific for the MHTGR.
The circulator will be driven by an induction motor
of variable speed approaching four megawatts in size.
This unique design will
be capable of operating at speeds of approximately 4500 revolutions per minute at flow rates in the range of 160 Kg/sec.
Prior experience in using water lubricated bearings in circulators has led to the design selection of magnetic bearings for the MHTGR circulator.
In
addition to the 5 axial magnetic journal and thrust bearings a catcher bearing is also provided.
The roller type catcher bearing supports the circulator
shaft assembly when it is at rest and also will provide a back-up for the magnetic bearings in the event of a magnetic bearing malfunction.
An area of
particular design interest is the behavior of the rotating assembly shaft following the potential failure to function of the magnetic bearings. In conclusion, the preliminary design of the MHTGR continues to progress with many areas of great interest.
Of particular interest is the area of
technology development which is addressed in a separate paper.
We are looking
forward to finalization of the design and plant construction with optimism.
C-5
FIGURE 1: RCCS INTERFACE WITH REACTOR VESSEL
HOT AIR
• HOT AIR
COLD AIR
COLD AIR
COOLDOWN DECAY HEAT REMOVAL
C-5
XA0101497
HM NUCLEAR INSTALLATIONS INSPECTORATE IAEA Technical Committee Meeting Gas-Cooled Reactor Technology Safety and Siting' Dimitrovgred, USSR, 21-23 June 1989
HSE Health & Safety Executive
UK REGULATORY ASPECTS OF PRESTRESSED CONCRETE PRESSURE VESSELS FOR GAS-COOLED REACTOR NUCLEAR POWER STATIONS P S WATSON, HMNII C-6
OBJECTIVES :
1. Nil assessment of prestressed concrete pressure vessels. (PCPV)
2.
Prestress shortfall incident.
C-6
HEALTH AND SAFETY EXECUTIVE ORGANISATION :
HSE
INSPECTION
HMNII
POLICY
ASSESSMENT
ENGINEERING
SITE INSPECTION
PHYSICS
C-6
SAPs :
HMNII Safety Assessment Principles for Nuclear Power Reactors. ISBN 0 11 883642 0
HMNII Safety Assessment Principles for Nuclear Chemical Plant. ISBN 07176 01536
C-6
EXAMPLES OF SAPs, RELEVANT TO PCPV:
- use of best proven techniques
- use of recognised standards
- adequacy of margins
- inspection and maintenance
C-6
PCPV SAFETY ASSURANCE :
Pressure limitation by SRVs elastic behaviour beyond Pd liner, insulation, cooling water proof pressure test redundancy of prestress hypothetical ultimate pressure appointed examiner in service inspection
C-6
IN SERVICE INSPECTION OF PCPVs
Concrete surface examination Anchorage inspection Tendon load checks Tendon material examination Foundation settlement and tilt Long term deformation Vessel temperature excursions Coolant loss PVCW loss Top cap deflection C-6
HARTLEPOOL AND HEYSHAM 1 POWER STATIONS PRESTRESS SHORTFALL PROBLEM
Unusual features of vessel design Prestress arrangement Discovery of problem Investigations Nil report Follow up of recommendations Current position
C-6
PRESSURE VESSEL GENERAL ARRANGEMENT 25.908 m DIA
C-6
o o
u
"OOXO"
OAOO«OOAOO*' OAOO«5
-ooxoo-
01 V
LAN CN REACTOR PI15CAP
C-6
•o
TENDON ANCHORAGE AND PRE-STRESSING JACK
_
TENDON LOAD - TONS
ORIGINAL DESIGN PREDICTION
700 —
600 — DESIGN PREDICTION ALLOWING FOR SHORTFALL
500 —
LOAD CORRESPONDING TO '"30%*LOSS OF PRESTRESS
490
I 10
100
10000
1000
DAYS AFTER PRESTRESSLOG SCALE
C-6
•OG
MAIN RECOMMENDATIONS
1. At all concrete pressure vessel stations prestress systems should be calibrated in a manner which results in all load bearing components being loaded in a representative manner
2. At all concrete vessel stations load measurements during calibration should be verified by a redundant and diverse system.
C-6
XA0101498
INTERNATIONAL ATOMIC ENERGY AGENCY
Technical Committee
on Gas-Cooled Reactor Technology, Safety, Sitting
POSSIBLE APPLICATIONS OF HTGR's IN TURKEY
§.Metin ATAK
TURKISH ATOMIC ENERGY AUTHORITY NUCLEAR SAFETY DEPARTMENT
D-l
1. INTRODUCTION The factors that effect the cost of nuclear power plants are classified into three main groups : -
Technical Economical Political
Technical^Infrastructure : Some additional expenditure must be required for the unsufficient infrastructure in order to install Nuclear Power Plant. Roodways, railways and bridges should be either revised or rebuilt so that heavy components could be carried. Revisons on network or communication systems could be necessary. The QA/QC level of local manufacturing which may have contribution in plant should be raised. Organisation : Some organisational changes could be required to carry out a nuclear program. Licensing and Regulation activities should be performed independently from the owner of the plant. However, in some countries especially in developing countries newly embarking nuclear technology^one organisation acts both as utility and regulatory body. Training of the key personnel should be started in advance of plant operation. Operating personnel must be ready at the begining of operation. Network and Unit Power : Capacity of network in developing countries is generally smaller; than those in industrilized countries, so there is a limitation for the power of the plant which is connected to the small grid due to the grid stability problems. D-l
Lfo°\
Large nuclear power plants may be unsuitable, but in case of small plants specific cost per KWh will be increasing. Cost of'Local Contribution :
It's possible to have local contribution in construction, labor and some equipment for non-safety related systems. Cost of workers for the construction may be rather cheap than industrilised countries and it seems as an economical advantage however lack up qualified man power can cause some delays in mounting and some particular areas requiring special skill and qualification, so foreign manpower should be necessary. Foreign manpower always cost more in abroad therefore total labor cost isn't lower than in the industrilised
countries. For the first plant, local engineering
contribution will be limited.
Cares must be taken for the quality of local material and equipment, in some cases cost will be high due to the low quality material. For the aim of supporting the local manufacturing even it's more expensive (20-40%) local products can be preferred.
Financial Conditions : Long construction periods and need of large amount
of foreign credits
are major factors effecting the decission of NPP installation. As the operation of plant delays, interest paid for the foreign credit will reach to the high levels. By taking into account of other problems related on imported nuclear power plants and particularly the first one, modular advance reactor design will provide a chance for a developing country to embark nuclear field and adopt the new technology and increase the local contribition as well. To overcome those problems mentioned section 1, new reactors should represent the following advantages to a country newly embarking nuclear energy :
Inherent safety features and simple operation characters -
Providing nuclear technology without financial and technical risk
-
Adoptability to the local conditions.
-
For giving the human errors to the highest extend.
-
Rapid licensing procedure.
D-l
]
0
HTGR's seem one of the most attractive examples of new reactor designs which have the 'capability of meeting all these requirements mentioned above. Excellent safety characteristics and possibility of installing modular HTGR uits of
lower capacities near inhabited and
industrial centers predestinate the use of HTGR's beside electricity generation as heat source for urban heating and for process heat application. Civil works and erection HTGR's can be executed to a great extend locally even for the first plan. Fuel elements for initial years of operation have to be imported, in long run, one may introduce local fabrication of fuel elements if it is economically feasible.
In monetary terms, local costs may consitute approximately 1/3 of the total investment costs. 2.
POSSIBLE APPLICATION OF HTGR's IN TURKEY
:
Brief description of energy consumption pattern will give a feeling about the possible areas of HTGR's use.
Consumption of primary energy sources in Turkey in the period 19731986 are given Table 1. Turkey's energy sources are not sufficient to match the consumption. Nearly half of the consumption has to be covered by imported petroleum, hard coal, and recently natural gas. Beside non-commercial sources (Approximately 1/3 of the total production is provided by non-commercial sources) lignite and hydraulic energy constitute the main local energy sources.
Sectional energy consumption can be seen in Table 2. Energy Consumption in accomodations Lhas the greatest share. Consumption of conversion facilities (power plants, refineries etc.) are also relatively high.
So, Urban
Heating, Process Heat Applications, are the possible
areas of HTGR's usage.
D-l
2.1. URBAN HEATING : As mentioned above, energy consumptions in accomodations has the highest share in total (above 40 % and mostly is used for heating).
In rural areas, non-commercial energy, sources (wood, farm waste products) are used for heating. Commercial sources (petroleum products, coal, lignite) are mostly consumed in big population centers. Even in the bigncities, most of the buildings have individual heating systems. Limited application of centralized heating exists. However, centralised heating will take place in incerasing number due to the growing volume at new-constructions of accomodation blocks and air pollution in cities. Use of waste heat of thermal power plants for urban heating can be limited due to remote location of the lignite fuelled power plants. Other alternative may be using of natural gas and these -necessary work is being carried out at the present to use natural gas for heaving in highly populated and polluted cities in the noth-west, west and central Anatolia. If the high cost of Natural gas (imported from USSR) and related transport and distribution systems are considered; HTGR's can be another alternative to provide clean and economically feasible way of urban heating. HTGR's can be another alternative to provide clean and economically feasible way of
urban heating.
A feasibility study has been carried out in Hacettepe University for HTGR's (Ankara) and It'll be completed at the end of June. University has two campuses,
medical faculties are in the campus located in town and the other
faculties are in the other campus, namely Beytepe 20 Km away from the city. On the basis of -12 C° reference temperature, the total peak heat requirement of Bytepe Campus was estimated as 20-25-MWth by taking into account of future development of faculties,and labratories
etc.
In a preliminary evaluation, it was seen that total fuel cost per year is about 2 Million Dollars for heating with the half capacity operation of existing central heating units. On the other hand, Nuclear Engineering and Nuclear Physics Departments have intention of using the reactor for training and experiments because the D-l
HTGR design has the capability for this purpose. (Beam ports will be add the design, and another control room which is similiar to main control room will be installed to serve as a training facility of nuclear plant operation.) Although there is no need to load fuel for a long periods (more than 5-6 years). Once a year fuel refuelling will be chosen in order to provide a facility which represent a training oppurtunity for refuelling, fuel handling etc. 2.2
PROCESS HEAT APPLICATIONS :
Electricity and heat consumption of the major industries are shown in Table 3. As can be seen there electricity consumption is the highest for metallurgy sector, whereas a moderate consumption of steam has to be encountered. Sugar, paper and petrochemical industries have relatively high steam and space/water heat consumptions beside a rather high electricity consumption. Process heat applications will be discussed in more detail for petrochemical, paper and fertilizier industries.
Petkim, a govermental economic enterprise, owns two major petrochemical complexes, namely Yanmca with 420.000 tons/year capacity,, and Aliaga with 1.000.000 tons/year capacity. In big petrochemical facilities, approximately half of the electricity demand is utilized for motors, pumps, fans etc. the other half for chemical processes. Reliability and continuity of
elpctrical supply is very important
in petrochemical complexes, since interruption of the process can lead to defects in equipment and to deterioration of the required product quality.
Therefore in-house electricity generation is often necessary in case of unreliable
network as stand-by , Aliaga complex has in-house electricity
generation Yanmca complex has no electricity generation, but all' important pumps, compressors are driven primarily with steam.
Steam can be used for following purposes in a petrochemical complex:
D-l
To drive steam turbine -
Direct heating indirect heating
-
As an auxiliary material in the process. As drive for important pumps etc.
Steam is utilised for process heat at 4 different presure levels : Pressure (at) Very high pressure
135 41
High pressure Medium Pressure Low Pressure Turbine
Temperature (C°)
Quantity(Ton/hr)
17.5
535 330 250
27 113 326
4.5
195
335 205
Condensate Tntal
1006 tons/hr
Total electric power is about 140 MWe.
At Yaramca Complex
:
Approximately 350 tons/hr steam is generated at 40 at. Total electric power requirement is about 42 MWe of which the half is used for electrolysis of salt to generate Chlor gas. Total fuel consumption is at Aliaga 500 tons/year and at Yarimca 200 tons/year. From above information are can conclude that in a big petrochemical <; complex like Aliaga a combined heat/power station could be installed with as electrical power approximately 150 MWe and heat power 450 MWth
Process heat is also used in paper industry. Steam generation and the quantities and the condition of process steam are given Table 4.
Considering that hourly steam generation is arround 50-100 tons/hr one can drive that the required heat power is approximately 25-50 MWth .
D-l
So this demand can be met by low capacity heat/power units. Steam conditions in two fertilizer plants can are given Table 5. One can see that also in fertilizer plants smaller heat or heat/power units are required.
3.
CONCLUSIONS : Modular type new small plants may give a considerable opportunity to a developing country to embark in nuclear energy and to walk hand in hand with a new technology. Combined heat/power stations which can be installed near the facility or high population areas can be used as heat and power source in big accomodation centers and petnichemical complex.
Smaller units are suitable in acomodation centers with lower population and in fertilizer plants and paper mills etc.
HTGR's representing inherestly safe design and also ensuring the easy public acceptance might be used for these purposes.
D-l
TABLE:l CONSUMPTION OF PRIMARY ENERGY 50URCE5 IN TURKEY 10
Years Hard Coal
Lignite
Asphaltit
1973
2804
2292
124
1974
3069
2456
169
1975
3025
2692
1976
2954
1977
TEP (Tons Equivalent Petroleum)
N.Gas Petroleum
Hydraulic
Geothermal Wood
Waste Products
13119
582
4154
2299
-
13108
750
4350
2366
196
-
14243
1319
4369
2984
190
-
15887
1871
3085
3125
187
16
18233
1978
2827
3523
128
20
1979
2988
3537
87
1980
2732
4000
1981
2759
1982
3033
1983
Solar
Electricity Import
Per Capita Total Consumption (KEP) 25374
OLJ*-J
_
26269
673
2496
a
28348
707
4420
2638
29
30973
757
1915
4497
2741
42
33841
810
17981
2085
4575
2903
53
34094
800
29
15761
2299
4652
3119
90
325G2
240
20
16212
2535
4729
3363
115
33946
748 764
4209
241
14
16052
2818
4807
3439
139
34477
760
4596
370
41
17140
3165
5028
3569
152
37094
801
3213
5OB6
323
7
17706
2534
5126
3574
191
37760
799
1984
3464
6396
97
36
17735
2999
5
5177
3396
3777
7925
225
61
18165
2690
1
5210
3238
228 184
39532
1985
41477
820 842
1986
4006
8903
261
407
19614
2652
10
5271
3201
67
44393
882
_.
1
D-l
TABLE : 2
5ECT0RIAL ENERGY CONSUMPTIONS
1978
1973 3
(10 TEP)
Accomodations
10D59
Industry
5563
Transport
4500
1986 (10 TEP) «> 3
3
«>
47 26
12443
44
15087
42
8196
10978
6199
29, 22
3 2
1667 1020
31 19 5 3
00
(10 TEP)
Agriculture
763
21 4
As ratt material
454
2
933 751
21339
100
28522
100
35591
100
4034
-
5572
-
8801
-
25373
-
34094
-
44393
-
Total netto consumption Internal cons:' of conversion facilities Total primary energy consumption
Ref:1
6839
WEC Turkish National Committee, 1986 Energy Report, Ankara, September 1987
D-l
TABLE:3 ELECTRICITY AND HEAT CONSUMPTION PROJECTIONS BY SOME MAJOR IMDU5TRIE5
INDUSTRY
19B3
1985
1990
Electricity (106kuih) Steam (10 9 kcal) (Cinko-Hur§un Metal Sanayi A.S.)
971 969
1300 11<45
1500 1256
Electricity (10 6 kuh) Steam (10 9 kcal) Space/Water Heat (10 9 kcal) CERAMIC-t-CEMENT (T.Cimento ve TDprak Sanayi A.S.) Electricity (106kujh) Space/Water Heat (10 9 kcal)
87 2
99 50 2
193 100 2
^81 0,11
560 0,095
590 0,102
261 13^*2 11,7
287 17U5 9,5
t»30 2630 15
363 29,3
311 33,9
515 38
METALLURGY (Etibank)
US
PETRO-CHEMISTRY (Petkim-Yanmca) Electricity (106kujh) Steam (10 9 kcal) Space/Water Heat (10 kcal) FERTILIZER (T.GGbre Sanayi A.5.) Electricity (10 6 kuh) Steam (10 9 kcal) SUGAR (T.Seker Fabrikalan A.S.) Electricity (10 6 kuh) Steam (lD 9 kcal) g Space/LUater Heat (10 kcal) PAPER (T.Selluloz ve KaQit Fab.) Electricity (10 6 kuh) Steam (10 9 kcal) q
369 77^0 387
2877 125
375 8119
UU2 9680 ^90
123i* 3952 120
Space/Water Heat (10 kcal) Private Communication (Ministry of Energy and Natural Resources)
TABLE:4
STEAM GENERATION AND PROCESS STEAM CONDITIONS IN PAPER MILLS:
Paper Mills
Steam Generation Quantity Pressure Temperature (10 tons/year) (atu)
ro
Process Steam Conditions P=13 atii p=3, 5 - 4 , 5 gtu Quantity Temperature -, Quantity (lCPtons/vear) <°C) (ID tons/year)
Temperature
(PC)
Izmit
818
59
450
150
225
668
185
Qaycuma
515
59 59
205 -
170
315
165 -
350
Aksu
500 500
315
180
Dalaman
890
59
308
220
422
59
203
225
582 219
160
Afyon
450 450
180
Balikesir
495
63
450
60
435
160
Akdeniz
886
63
450
438
200
Kastamonu
75
19
Saturated
75
Saturated
448 -
160 -
Bolu
84
13
Saturated
84
Saturated
-
-
Total Ref:6
4500
38D-405
1483
3017
Private Communication (Turkiye Selluloz ve Kagit Fabrikalari isletmesi Genel Mudurlugu)
D-l
TABLE:5
STEAM CONDITIONS IN FERTILIZER PLANTS:
Fertilizer Plant
Steam Conditions Quantity(tons/hr) Pressure(atu)
SAMSUN
2x20
37
KUTAHYA Existing: Under Construction(x):
35 UQ
Temperature(°C) *»50
.
t»00 t»50
(x) Electricity generation beside heat supply.
tjlZ* P r i w a t e Communicatian (Turkiye Gubre Sanayi A.5.)
D-l
XA0101499 ABSTRACT AN ECONOMIC ASSESSMENT OF U.S. MHTGR DESIGN L. DANIEL MEARS General Manager, Gas-Cooled Reactor Associates
GCRA's economic goal for the U.S. MHTGR design is for the equilibrium plants to have at least a 10% power cost advantage over comparably sized, state-of-the-art coal plants. In addition, the designers are challenged to limit the overall financial risk to be on par with such a coal plant. During the past year, cost estimates and economic assessments have been updated in the U.S. MHTGR Program. Further, a major study has been completed adapting the MHTGR to a water desalination/cogeneration application. These results will be presented along with a discussion of the key GCRA design requirements that limit the overall financial risk to the prospective owner/operators of future MHTGR plants.
D-2
AN ECONOMIC ASSESSMENT OF THE U.S. MHTGR DESIGN L. DANIEL MEARS General Manager, Gas-Cooled Reactor Associates (GCRA) Presented at IAEA Technical Committee Meeting Dimitrovgrad, U.S.S.R. June 21-23, 1989 (Revised Issue - September 8, 1989)
Introduction Through GCRA, U.S. utility/users have established the following economic goal for the U.S. MHTGR reference plant design: "A goal shall be the development of a design that has an evaluated economic advantage of at least 10% in the 3 0 year levelized busbar cost of electricity relative to a comparable sized, state-of-the-art coal fired p l a n t while requiring equivalent institutional resources and posing equivalent financial risks." The term "equivalent institutional resources" means that approximately the same level of qualified personnel and organizational capability, functioning within similar corporate cultures, is required to successfully procure, license, operate, maintain and decommission the plants. The term "equivalent financial risks" means that approximately the same level of uncertainty exists for receiving an adequate return on investment, such that investors will be equally likely to invest in the MHTGR or coal plants. T h e intent of this economic goal is to guide the development of the MHTGR such that prospective owners would view the MHTGR as an attractive nuclear option that is competitive, and which affords ownership risks and . returns that are on par with coal plant alternatives.
-i-
D
-2
This paper presents a summary of the background, approach and the current estimate for the U.S. reference MHTGR electric generating plant. The plant is composed of two power building blocks producing a total output of 538 MWe. Each building block consists of two reactor modules and one turbinegenerator set. Additional economic results are given for a cogeneration application with a seawater desalination plant. Background Coal, oil and gas fired power plants have traditionally provided the backbone of U.S. electricity generation. Accordingly, the U.S. industrial infrastructure for procuring, licensing, operating and maintaining power plants evolved over many decades of successful experience with their use within the social, economic and political framework of the U.S. M o r e o v e r , the existing infrastructure also developed in response to local and regional needs for electrical power and is, therefore, quite diverse in terms of organizational styles and structures. However, prior to the introduction of nuclear power, such diversity presented no major difficulties in managing conventional power plants and was accepted as an effective means of assuring competitively priced electrical power. Nuclear energy differs from fossil fuel alternatives in that Federal regulations exist to provide "reasonable assurance that the health and safety of the public will not be endangered." Demonstrating compliance with these regulations has proven to be onerous to the industry. During the early 1980's, many in the U.S. began to question the compatibility of the current generation of LWR plants with existing U.S. institutions. For example, the U.S. Congress' Office of Technology Assessment related the need for significant changes in the technology and management of nuclear power plants to the size and complexity of then current LWR plants and the exacting regulatory process. -2-
•16
D
~2
The NRC grants an operating licensing on the basis of demonstrated financial and technical capability and an acceptable plant safety analysis. The philosophy behind the regulatory process is one of assuring that the plant is operated and maintained such that the assumptions of the safety analysis (equipment performance, operator action, etc.) are valid throughout the life of the facility. When a system is identified as safety-related for the purposes of regulation, it carries with it not only capital and operating cost premiums, but the multiplicity of risks that arise from assuring and documenting that the requisite level of reliability has been provided in each step of development from design through power generation. The discipline and vigilance required for the "exacting" chore of establishing and continuously verifying the ongoing safety of the plant, in accordance with regulations, distinguishes the organizational requirements and corporate cultures of nuclear from fossil institutions. The management of nuclear institutions must recruit, train, organize and direct specialist personnel for the design, fabrication, construction, operation and maintenance of nuclear installations and assure the accurate and auditable communication of requirements between their institutions over the life of the project and operating life of the plant. Because their performance is under continuous scrutiny and linked to the plant operating license, fluctuations in performance translate into business risks. These risks are compounded when the underlying regulations are subject to change. Within the commercial power industry, the discipline and vigilance required of the performing organizations is unique to the nuclear energy option, as is the financial risk associated with fluctuations in organizational performance. The owner/operating entity is at the end of the trail of quality requirements and is responsible for implementing the -3-
D
"2
results
of
the
design
operational
licensing
risks,
the
and
have
entity.
These risks
construction
requirements.
risks
process
and
to
largely
the
been
program
The cumulative
organization borne
financial
itself,
by the
through in
this
owner/operating
contribute heavily to the stagnation of
the nuclear power industry in the U.S. On the practical experience
with
adjustments
level of u t i l i t y operations, the frequent
nuclear
plants
necessary
disproportionate utilities
have
"culture"
within
to
strain found
that
accommodate
the
necessary
to
organizations
that
disruptive culture
establish
order
"institutional related
to
a
many
a separate
in which the
necessary
We have termed the
an owner/operating entity may have to endure a
transformation
in
placed
As a result,
discipline and vigilance can be fostered. prospect
organizational
them has
on management.
it
their
is
to
its
organization
accommodate
risk."
the
of
a
nuclear
While precedence
traditional
utility
and corporate
to
as
date has been
role,
avoidance of this type of risk is crucial if
option
even
greater
Independent Power
Producer (IPP) arrangements flourish. The
MHTGR offers
relative coated
to
fuel
its
nuclear
particle.
containment
concept
scrutiny
be
to
a
superior
"containment
competitors,
with
Conceptually,
will
shifted
permit from
the
plant
the
the
concept,"
high
coated
focus
of
design,
quality particle
regulatory
construction,
operation and maintenance to fuel manufacture.
If this
shift
in regulatory scrutiny can be accomplished, the owner/operator will
be relieved
of
a substantial
body of
the
institutional
risks that plague current plants. In
implementing
translate
inherent
regulations
that
this
and
are,
goal,
passive insofar
MHTGR development safety
characteristics
as possible,
U.S. business practices
for
emphasis
MHTGR Program has
in
the
U.S.
consistent
conventional plants.
-4-
been
must into with
Up to now, on
gaining D-2
acceptance of the MHTGR safety concept. However, in addition to the p u b l i c a c c e p t a n c e benefits that accrue to owner/operators for enhanced safety of nuclear plants, increased safety is even more significant if it translates into decreased business risks. The following section discusses how certain key features of the MHTGR support the overall economic goal. Approach to Achieve Economic Goal To achieve the economic goal, several key features of the MHTGR, plus key institutional arrangements, must be carefully optimized in order to adequately compensate for lost economies of scale. The first is the full utilization of the inherent and passive safety characteristics. This feature: •
Eliminates the need for many high cost safety systems (e.g. containment structures, auxiliary power sources).
•
Allows the fabrication of many systems to industrial (non-nuclear)
standards
(e.g. the plant
control
system). •
Allows the minimization and physical separation of the nuclear island and therefore the minimization of costly nuclear grade construction.
•
As a result of all the above, provides the potential for easing the plant licensing and certification process, further supported by the testing capability of the plant design. (Precluding the need for off site public evacuation and sheltering is a key prospect of this eased licensing.) ,
D-2 -5-
•
Provides the potential for overall simplification of operation and maintenance activities, including the operational licensing requirements.
Table 1 compares the safety related structures and systems proposed for the MHTGR plant to a recent, current generation LWR plant. This comparison highlights the basis for expected reductions in safety/licensing related costs for MHTGR. The second feature is the optimization of modularization and factory fabrication techniques that will enhance standardization and thereby: •
Minimize plant-specific design and licensing costs.
•
Minimize field construction schedule and time related costs (i.e. interest during construction).
•
Maximize factory related learning benefits to reduce capital costs.
•
Maximize the plant staff loading efficiency and related learning benefits to reduce training, operations and maintenance costs.
The modular nature of the power increments plus the short construction schedule provides important flexibility in responding to uncertain load growths. Further, the multi-module nuclear island and multi-turbine energy conversion area enables a zero planned outage of the total plant. Table 2 compares the major heavy equipment, commodities and selected craft labor on a per MWe basis for the MHTGR, a "Better Experience" PWR (typical of upper range of experience with current PWRs) and a comparably sized coal plant. Whereas the vessel tonnage comparison penalizes the MHTGR relative to -6-
TABLE 1 COMPARISON OF SAFETY RELATED STRUCTURES AND SYSTEMS (2) STRUCTURES: • REACTOR BLDG • REACTOR AUXILIARY BLDG • REACTOR SERVICE BLDG
CONTAINMENT BLDG AUXILIARY BLDG CONTROL BLDG FUEL BLDG DIESEL GENERATOR BLDG MAIN STREAM SUPPORT STRUCTURE CONDENSATE TANK FOUNDATION ESSENTIAL SPRAY PONDS & INTAKE STRUCTURE REFUELING WATER TANK FOUNDATION
SYSTEMS; REACTOR EQUIPMENT - FUEL ASSEMBLIES - CONTROL ELEMENT ASSEMBLY - CONTROL ELEMENT DRIVES - CORE SUPPORT STRUCTURES
• REACTOR SYSTEM - FUEL ELEMENTS - CONTROL RODS - CONTROL ROD DRIVES - CORE SUPPORT STRUCTURES - UPPER PLENUM THERMAL PROTECTION STRUCTURE
• PRIMARY SYSTEM COMPONENTS - REACTOR VESSEL & SUPPORTS - STEAM GENERATORS & SUPPORTS - SAFETY & RELIEF VALVES - REACTOR COOLANT PRESSURE BOUNDARY PIPING - MAIN STEAM/FEEDWATER ISOLATION VALVES - PRESSURIZER & SUPPORTS - REACTOR COOLANT PUMPS & SUPPORTS
• VESSEL SYSTEM - REACTOR VESSEL & SUPPORTS - S/G VESSEL & SUPPORTS - PRESSURE RELIEF VALVES - CROSS DUCT VESSELS - STEAM GENERATOR ISOLATION VALVES
• SAFETY INJECTION & SHUTDOWN COOLING SYSTEM CONDENSATE STORAGE FACILITIES AUXILIARY FEEDWATER SYSTEM MAIN STREAM PIPING IN CONTAINMENT CONTAINMENT SPRAY SYSTEM ESSENTIAL SPRAY POND SYSTEM ESSENTIAL COOLING WATER SYSTEM
REACTOR CAVITY COOLING SYSTEM
J^MHTGR PSID, TABLE 3.2-4 (2) VERDE FSAR, TABLE 3.2-1 -7-
D-2
TABLE 1 COMPARISON OF SAFETY-RELATED STRUCTURES AND SYSTEMS continued
ESSENTIAL UNINTERRUPTABLE POWER SUPPLY SYSTEM (120V AC)
• ELECTRIC SYSTEMS - 4.16kV CLASS IE AC - 4 80V CLASS IE AC - 12 0V VITAL AC
ESSENTIAL DC POWER SUPPLY SYSTEM
• CLASS IE DC EQUIPMENT • DIESEL GENERATOR & FUEL SYSTEM
• PLANT PROTECTION & INSTRUMENTATION SYSTEM
• PLANT PROTECTION SYSTEM - REACTOR PROTECTIVE SYSTEM - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM • SAFE SHUTDOWN INSTRUMENTATION AND CONTROL SYSTEMS • ESSENTIAL CHILLED WATER SYSTEM • HVAC • CHEMICAL SYSTEM
AND VOLUME
CONTROL
• FUEL HANDLING & STORAGE • FUEL POOL COOLING SYSTEM
& CLEANUP
• NUCLEAR COOLING WATER SYSTEM • RADIATION MONITORING SYSTEM • CONTAINMENT BUILDING COMBUSTIBLE GAS CONTROL SYSTEMS • SAMPLING SYSTEM • GASEOUS RADWASTE SYSTEM
-8-
D-2
Table 2 COMPARISON OF COMMODITIES, CRAFT LABOR AND MAJOR EQUIPMENT QUANTITIES
UNITS(1)
MHTGR NOAK PLANT
PWR-BE 1144MWe PLANT
COAL 488 MWe
MAJOR EQUIPMENT: REACTOR PLANT VESSEL (INCL. STEAM GEN.)
TN/MWe
10
N/A
FORMWORK
SF/MWe
1511
1648
1530
STRUCTURAL STEEL
TN/MWe
24
6
30
REINFORCING STEEL
TN/MWe
28
18
7
EMBEDDED STEEL
TN/MWe
1
1
1
STRUCTURAL CONCRETE
CY/MWe
232
116
115
PIPING, NUCLEAR GR
LB/MWe
315
1916
0
PIPING, INDUSTRIAL GR
LB/MWe
3725
6094
8571
WIRE AND CABLE
FT/MWe
5546
4352
5451
WIRE AND CABLE DUCT
FT/MWe
808
588
907
BOILERMAKER
MH/MWe
326
583
1566
CARPENTER
MH/MWe
1226
1156
713
ELECTRICIAN
MH/MWe
2305
1925
2400
IRON WORKER
MH/MWe
1334
1149
1111
MILLWRIGHT
MH/MWe
407
169
336
PIPEFITTER
MH/MWe
1352
2644
4035
OTHERS
MH/MWe
2065
3028
3324
: MH/MWe
9014
10654
13484
COMMODITIES:
SELECTED CRAFT LABOR:
TOTAL CRAFT LABOR
(1) SF =.SQUARE. FEET, TN = TON, CY = CUBIC YARDS, LB = POUND, FT = FEET, MH = MANHOURS — Q _
D-2
the PWR, counter gains are made for other areas, particularly the nuclear grade (and high cost) piping and the associated valving. The respective comparisons in Table 2 provide insight to the MHTGR's safety simplicity and modularity gains in achieving competitive economics. The third feature is the stringent operational reliability requirements posed by GCRA. For example: •
The design should facilitate in-service inspection and on-line maintenance and repair, where cost effective and within the economic goal.
•
Further, the plant design must allow for the removal and replacement of all equipment, including the reactor and steam generator internals and the primary system vessels.
•
An overall equivalent availability (capacity factor assuming a utilization factor of one) target of 80%, with the goal of achieving an economically optimum value through the design development process.
•
Forced outages are limited to 10% of the equivalent unavailability, plus events leading to forced outages of six months or more duration should be limited to 10% of the allowed 10% equivalent unavailability (1% total).
The fourth feature requirements posed by potential for plant equipment malfunctions. •
is the stringent investment protection GCRA that essentially eliminate the loss due to operator errors and/or For example:
The cumulative evaluated plant damage per year for all accidents/events when averaged over the plant lifetime shall not exceed the owner's annual property damage insurance premium. -10D-2
•
To assure that the likelihood of a class shutdown of a l l MHTGR p l a n t s i s remote, the probability of exceeding any safety related design condition (i.e. exceeding the design limits of a safety related component or system regardless of damage) shall not exceed 10 per plant-year.
The fifth feature deals with institutional development underway or envisioned within the U.S. MHTGR Program. For example: •
One or more vendor entities are envisioned that will provide a fixed price, turnkey contract for the entire plant (or at least the nuclear island) plus related performance warranties.
•
New generating company e n t i t i e s are likely to be established for much of the broad deployment of any new base load c a p a c i t y in t h e U.S. for the foreseeable future. The roles of the u t i l i t i e s , vendors and other interested parties are evolving.
•
Nuclear operating companies are also likely to evolve t h a t w i l l c e n t r a l i z e key l i c e n s i n g , t r a i n i n g , o p e r a t i o n s , maintenance, and technical support a c t i v i t i e s for the shared benefits of multiple plants.
Economic Assessment Framework In addition to the preceding f e a t u r e s , there i s a framework of detailed groundrules, assumptions and judgement estimates that must be developed with care and consistency, particularly for comparative cost estimates. Key examples for the MHTGR estimate include:
-11-
•
Three time-dependent, reference plant estimates are maintained: the initial Lead Plant; the second R e p l i c a Plant; and N t h - o f - a - k i n d (NOAK) or equilibrium plant-typically the eighth plant.
•
Estimates for three additional size-dependent plants are extrapolated from the reference 4 reactor module-2 turbine (4x2) plant, namely a twin reference plant (8x4) ; a half size (2x1) plant; and a single reactor plant (lxl) with a matching size turbine.
•
The NOAK and Replica Plants are based on an economically optimum deployment sequence leading to a 44 month construction schedule and a 77 month project schedule (through commercial operation) . The Lead Plant is deployed in two phases over a project interval of 103 months and a construction interval of 70 months. Phase I deploys one reactor module and one turbine-generator set. After one year of demonstration testing, Phase II deploys the other three reactor modules and the other turbine-generator set.
•
Plant startup schedules are targetted for the year 2000 for the Lead Plant, the year 2005 for the Replica Plant and the year 2010 for the NOAK plant.
•
Unless otherwise substantiated by specific data, direct costs learning benefits through the NOAK plant are conservatively limited to field labor and reactor equipment with learning factors of 97% and 95%, respectively. Material plus conventional components and systems are conservatively assumed to have achieved the full commercial learning experience. Equivalent learning benefits through the NOAK plant for the indirect cost elements vary, but on average correspond to a factor of 91%. -12-
D-2
•
Unless otherwise substantiated by specific data, nominal contingencies are applied at a rate of 25% for nuclear grade costs and 15% for industrial grade costs.
•
A c a p a c i t y f a c t o r of 8 0 % is u s e d for the multi-modular, multi-turbine NOAK MHTGR plant. The same capacity factor (80%) is used for generation cost comparisons with alternative plants. The Lead and Replica MHTGR Plants are assumed to have slightly lower capacity factors, 77% and 78.5% respectively.
•
Comparisons are made with a range of coal costs. A representative reference coal price of 1.55$/MBtu ('88$ and delivery) is applied with a constant real escalation rate of 1.2% per year.
•
The financial parameters and site related costs are representative of a private utility based owner entity with nuclear experience.
Cost Estimate Results Costs have been developed to design, construct, operate and maintain reference MHTGR power plants and a comparison of the costs has been made primarily with coal plants. The costs were developed in general conformance with the U.S. Department of Energy cost estimating guidelines for advanced nuclear technologies. Costs were developed by General Atomics for the reactor plant equipment. Costs for the other equipment, field labor, and field material necessary to construct the nuclear island were developed by Bechtel. Costs for all the equipment, field labor, and field material necessary to construct the energy conversion area were developed by Stone & Webster. GCRA has developed the owner's cost, integrated the cost estimates and performed the cost assessments. -13-
D_2
Figure 1 displays the major components of total plant capital costs for each of the three reference plants under study. The first-of-a-kind development and certification costs are included with the Lead Plant costs. Progressive comparisons of the estimates in Figure 1 highlight the relative and incremental cost reductions expected for the Replica and NOAK p l a n t s through learning and factory throughput gains for the direct costs, certified design replication gains for the indirect costs, and the reduced schedule impact on the interest during construction costs. Table 3 presents the 2 digit account summary of the NOAK MHTGR plant capital cost estimate. The nuclear grade and i n d u s t r i a l grade components of the cost estimate are delineated and illustrates that approximately one half of the plant capital costs may be treated as a modern fossil plant. Had the total plant been treated as nuclear grade, the plant costs would have increased approximately 20%. As discussed earlier, the O & M activities for the MHTGR are e x p e c t e d t o b e n e f i t from the i n h e r e n t and p a s s i v e characteristics. Also discussed were the O & M benefits that derive from modularity, such as the efficient manpower loading on the refueling and maintenance activities for the multiple reactors and turbines. In addition, the generic HTGR features associated with the slow response to transients leads to simplified control systems and less demands on the operator for upset conditions. Further, the inert helium coolant and the excellent retention of fission products in the coated fuel p a r t i c l e are expected to lead to simplified maintenance, component repairs and replacement as well as low radiation exposure levels for the related plant staff. Accordingly, the staffing cultures and levels for the MHTGR plant have various bases for paralleling fossil plants. I t is noted, however, t h a t the realization of t h i s potential will continue to r e q u i r e strong involvement by u t i l i t y personnel with enlightened O & M management experience. -14-
D-2
0-
NUUKL 1
MHTGR PLANT TOTAL COSTS (CAPfTAL + DEVELOPMENT, JANUARY 88$)
o CD
a
TABLE 3
NOAK MHTGR PLANT CAPITAL COST ESTIMATES (MILLIONS OF 1988$)
NUCLEAR GRADE COST
INDUSTRIAL GRADE COST
TOTAL COST
0
2
2
67
43
110
267
1
270
TURBINE PLANT EQUIPMENT
1
128
129
ELECTRIC PLANT EQUIPMENT
19
35
54
MISCELLANEOUS PLANT EQUIPMENT
4
18
22
MAIN CONDENSER HEAT REJECTION
_JO
22
22
358
249
609
CONSTRUCTION SERVICES
40
39
77
AE HOME OFFICE ENGINEERING
39
15
54
FIELD OFFICE SUPERVISION
18
13
31
0
_iI8
118
97
185
280
- TOTAL $
456
434
891
- $/kW(e)
849
808
1657
123
65
187
- TOTAL $
580
500
107,9
- $kW(e)
1078
929
2007
LAND & LAND RIGHTS STRUCTURES & IMPROVEMENTS REACTOR PLANT EQUIPMENT
TOTAL DIRECT COST
OWNER'S EXPENSES TOTAL INDIRECT COST BASE CONSTRUCTION COSTS
CONTINGENCY TOTAL OVERNIGHT COST
143
INTEREST DURING CONSTRUCTION TOTAL CAPITAL COST
- TOTAL $
1222
- $k\V(e)
2273
-16-
D-2
An assessment of O&M requirements and costs was developed for the reference MHTGR plant by a programmatic task force familiar with nuclear generating plant O&M requirements and the MHTGR design. Table 4 presents the resulting estimate for the onsite staff and related costs for the NOAK MHTGR reference plant. Note that the estimated manpower on a per MWe basis is approximately half that experienced by the recent large LWR plants in the U.S. Table 5 presents the summary of the total annual O&M costs that includes a fixed and variable cost for maintenance materials, supplies and expenses, and offsite technical support plus overall administrative and general (A&G) costs. It is noted that the expected cost savings approach of a central staff supporting multiple plants has not been applied for such estimates. Table 6 provides MHTGR fuel cycle cost data. The parameters were developed within the overall U.S. advanced reactor program and are given for reference. The MHTGR fuel fabrication costs were developed by General Atomics and represent the graduation from a low throughput pilot-scale f a b r i c a t i o n plant to a fully loaded commercial-scale fabrication plant. The resultant fuel cycle costs follow. Figure 2 presents the busbar generating cost estimates for the three reference plants under study. A decommissioning increment is included at a nominal cost of approximately $80 M ('88$). The respective comparisons highlight the overall cost reductions expected after introduction of a Lead Plant to the NOAK (equilibrium) Plant. The economic competitiveness of the MHTGR is sensitive to several of the assumed reference parameters. Two of these are the cost of coal and the coal cost real escalation rate. The MHTGR economic advantage is given in Figure 3 as a function of the assumed coal cost and the real escalation of the coal cost. Figure 4 illustrates the sensitivity of the MHTGR economic margin to percentage variations of the coal cost, _ 17 _
D-2
TABLE 4
ONSITE STAFF COSTS FOR REFERENCE MHTGR PLANT (JANUARY 1988 DOLLARS)
JOB TITLE .
SALARY ($/YEAR)
PLANT MANAGER'S OFFICE PLANT MANAGER ASSISTANT MANAGER TRAINING SAFETY AND FIRE PROTECTION ADMINISTRATIVE SERVICES HEALTH SERVICES SECURITY SUBTOTAL
104,000 72,800 50,960 42,640 28,080 28,080 24,960
1 1 5 1 25 1 34 68
OPERATIONS SUPERVISION SHIFT OPERATION SHIFT MAINT. SUPPORT SUBTOTAL
53,040 44,720 44,720
MAINTENANCE SUPERVISOR CRAFTS ANNUALIZED PEAK MAINT. QUALITY CONTROL WAREHOUSE SUBTOTAL
49,920 35,360 35,360 38,480 32,240
TECHNICAL AND ENGINEERING REACTOR ENGINEERING RADIOCHEMISTRY AND WATER CHEMISTRY ENGINEERING TECHNICIAN HEALTH PHYSICS SUBTOTAL
NUMBER
6 32 12 50
7 133 3 5 6
TOTAL ($/YEAR)
104,000 72,800 254,800 42,640 702,000 28,080 848.640 2,052,960 318,240 1,431,040 536.640 2,285,920
154
349,440 4,702,880 106,080 192,400 193.440 5,544,240
53,040 49,920
3 8
159,120 399,360
45,760 37,440 37,440
6 6 13 36
274,560 224,640 486.720 1,544,400
308
11,428,000
TOTALS WITHOUT PAYROLL TAX AND INSURANCE
1,143,000
PAYROLL TAX AND INSURANCE (at 10%)
12.6M$
TOTAL WITH PAYROLL TAX AND INSURANCE
-18-
D-2
TABLE 5
ANNUAL O&M COST ESTIMATES FOR REFERENCE MHTGR PLANT (JANUARY 1988 DOLLARS) NET RATING MW(e) CAPACITY FACTOR, % ANNUAL GENERATION, kWh/year ONSITE STAFF
538 8 0
POWER GENERATION COSTS (M$/YEAR) 12.6
ONSITE STAFF MAINTENANCE MATERIALS FIXED VARIABLE SUBTOTAL SUPPLIES AND EXPENSES FIXED VARIABLE PLANT CR AND REFLECTOR DISPOSAL SUBTOTAL OFFSITE TECHNICAL SUPPORT SUBTOTAL, POWER GENERATION COSTS FIXED VARIABLE SUBTOTAL
3.9 1.3 5.2 4.3 0.4 0.8 5.5 2.3 23.1 2.5 25.6
A&G COSTS (M$/YEAR) PENSIONS AND BENEFITS NUCLEAR REGULATORY FEES INSURANCE PREMIUMS OTHER A&G SUBTOTAL
3.2 1.0 3.6 3.8 11.6
TOTAL O&M COSTS (M$/YEAR) FIXED VARIABLE TOTAL MILLS/kWh
(
0
Q
3.77 X 10 y 308
34.7 2 5 • 37.2 10.0
TABLE 6
FUEL CYCLE COSTS DATA FUEL CYCLE COST PARAMETERS REAL ESCALATION (%/YR)
1987 PRICE u 3 o 8 , $/ib CONVERSION, $/kg U ENRICHMENT, $/kg SWU WASTE DISPOSALS, mills/kWh
23 8 109 1
2.0 1.0
-1.7 0
MHTGR FRESH FUEL FABRICATION COSTS (1988S/ELEMENT) ITEM INITIAL CORE AND RELOADS 1-3 RELOADS 4-6 RELOADS 7-end
LEAD
REPLICA
NOAK
34, 600 30, 000 10, 500
30, 000 10, 500 10, 500
18, 200 10, 500 10, 500
3 0 -YEAR LEVELIZED FUEL COSTS (1988$)
YEARS (LEAD) (REPLICA) (NOAK) JOAK (EQUIL. FAB)
2000 2005 2 010
FUEL (MILLS/kWm 5.62 5.62 5.71 5.71
FAB (MILLS/kWh) 6.16 4.68 3.50 2.65
,-20-
WASTE DISPOSAL (MILLS/kWh) 1.0 1.0 1.0 1.0
TOTAL TOTAL (MILLS/kWh^ ($/MBtu 12.8 11.3 10.2 9.4
1.44 1.27 1.15 1.05
D-2
NUUKE-2 MHTGR LEVELIZED BUSBAR COSTS (JANUARY 1988$)
1
2 30 -
10 -
a
X7~7\
CAPrTAL
FIGURE 3 MHTGR ECONOMIC ADVANTAGE VS COAL COSTS (80% CAPAcrrr FACTOR. 2010 STARTUP)
COAL REAL ESCALATION RATE - 2.0%
2 S I to I
M
m a. P
-15
1
1.30
1.35
• 1.40
• 1.45
i
i
i
i
1.50
1.55
1.60
1,65
COAL COST. '66 $/MBTU
1.70
1.75
1.80
FIGURE 4 MHTGR ECONOMIC MARGIN SENSITIVITY TO CHANGES IN REFERENCE PARAMETERS
MKTOR CAP FAC (SOX)
z o
COAL COST ($1^5/MBTli) COAiESCAL
o o
I
MHTGR OftM MHTGR FUEL
i
I
MHTGR CAPITAL ($122214) bJ Q.
-20 H
-30
-25 PARAMETER PERCENT CHANGE o
coal real escalation rate and the major MHTGR reference parameters. This figure shows that the capital cost and the plant capacity factor have the most pronounced effect on the MHTGR economic competitiveness. Figure 5 applies the NOAK MHTGR generating plant estimates over a range of plant sizes. The nomenclature designates the reactor x turbines combination that produce the indicated plant output. Coal plant generation cost estimates are included for the same range plus point comparisons for gas fired combined cycle plants. Key observations from this figure include: •
The reference (4x2) 54 0 MWe MHTGR plant continues to be evaluated as competitive with comparably sized coal plants.
•
A large (8x4) MHTGR plant has only a 4-5% evaluated economic gain relative to the reference plant due to the diminishing on-site learning and shared cost benefits. The large MHTGR plant maintains the evaluated advantage relative to comparably sized coal plants.
•
On the other hand, a half sized (2x1) MHTGR plant has approximately a 15% evaluated economic disadvantage relative to the reference MHTGR plant and some noticeable loss in competitiveness relative to the comparably sized fossil plants. However, for high fossil fuel cost regions in the U.S., this size plant may be attractive.
•
To no surprise, the one module (lxl) MHTGR plant is evaluated to be noncompetitive for any practical U.S. site. However, it may be competitive in other countries, depending on the availability and cost of the alternative technologies and related fuels. -24-
D-2
-o-
FIGURE 5 EQUILIBRIUM PLANT POWER COST PROJECTION 80 -
1x1
I
70 -
CO 00
60 -
v
(2010 STARTUP, 80% CAPACITY FACTOR)
2xVV
I to I
8
\
# " ^ ^ ^ ^ . ^ —
COAL / _ ^
^
/
10
"
MHTGR/
50 -
"
—
4
•
_
^
_
—
"
—
-
—
•
—
. 8x4
m
a N
-- GAS/OIL COMBINED CYCLE PLANTS
• 40 -
4x2 = 4 REACTORS AND 2 TURBINES, ETC REF!ERENCE COAL COST: $1.55/MBTU IN 88$ WTTH 1.2% REAL ESCALATION REFI[RENCE GAS/OIL COST: $3.00/MBTU IN 88$ WITH 2X REAL ESCALATION 1990 ON
30 -
I
i
i
I
4
i
i
i
6 PLANT SIZE (100 MWe)
o N5
i
8
1
I
10
i
12
In addition to the all-electric generating application, various cogeneration applications have been evaluated by the U . S . MHTGR Program. During the past year, a major cog e n e r a t i o n / s e a w a t e r desalination application study was completed. Major new variables that were considered included: •
Ownership arrangements of the overall plant and the prospect of different arrangements for the MHTGR plant versus the desalination plant.
•
The bases for the cost of steam to the desalination plant. Namely, is the cost of electricity based on market value, an allocated cost base or the nominal value of an all-electric plant.
•
The availability, quality and cost of blending water plus the quality requirements of the product water.
Figure 4 provides a nominal joint product cost curve that illustrates the tradeoffs when determining the relative costs of product blended water and busbar generation for an NOAK (equilibrium) plant. The products produced are 466 MW of electricity and 147,000 acre feet per year (181 million cubic meters per year) of blended water. The ownership arrangements behind this curve are based on public utility ownership of the desalination plant versus the relatively higher cost of financing for the private utility ownership of the MHTGR plant. Blended water was assumed to be available at 130$/Acre-Foot(AF) (105$/thousand cubic meters) with an impurity level of 1500 PPM total dissolved solutes (TDS) . Resultant product water quality requirements were 500 PPM TDS. As illustrated on Figure 4, applying an all-electric plant rate of 5.2 cents/kwh yields a product water cost in the range of
420$/AF
(340$/thousand
cubic meters). -26-
For the primary D
"2
desalination regions of interest in the U.S., this is a competitive water cost compared to other major sources of new water supply projected for the turn of the century timeframe. Conclusion Although the MHTGR is at an early stage of preliminary design, the ongoing economic and risk assessments continue to show promise and potential for meeting GCRA's goal for evaluated cost competitiveness as well as the demands on institutional resources and exposure to financial risk.
-27-
D-2
XA0101500 Nuclear Research Center Jiilich Institute for Reactor Development Prof. Dr.-Ing. Heiko Barnert Box 1913, D-5170 Jiilich Federal Republic of Germany
May 4, 1989
The HTR, Applications, Economics and Environmental Aspects Barnert, H., Nuclear Research Center Jiilich Schad, M., Lurgi Frankfurt Candeli, H. Interatom, Bensberg
Paper to be presented at the Technical Committee Meeting on Gas-Cooled Reactor Technology, Safety and Siting Dimitrowgrad, USSR 21-23 June 1989
Abstract: The High Temperature Reactor (HTR), as the only nuclear system producing high temperature heat up to 1000 °C, offers a wide variety of applications. Beside electricity production, via steam turbines and in future via gas turbines, there- is: District heat with high efficiency, long distance energy for urban energy supply, high pressure injection steam production for enhanced oil recovery, medium range temperature heat direct application in chemical and related industry and last not least, high temperature application for the refinement of fossil energy carriers. Recent results of studies and programmes will be presented: Near term applications are identified, e.g. refineries and alumina industry with smaller HTR units. Another large market is the production of hydrogen, methanol and ammonia on the basis of natural gas, the relevant technology has been developed up to the pilot scale. The refinement of fossil energy carriers, in particular of coal, is subject of the R + D programme in the cooperation between German industrial companies and the Nuclear Research Center. The results are very promising and will be explained in detail. This programme will be continued. Objectives are: improvement of the technology and of the economics as well as environmental aspects, e.g. the reduction of emissions of carbondioxid. The topics of the programme deal with the different apparatus, e.g. steam methane reformer, steam coal gasifier, intermediate heat exchanger and last not least the process heat HTR.
D-3
The HTR, Applications, Economics and Environmental Aspects 1.
Introduction: HTR Company founded
With the date of May 8, 1989 the HTR-GmbH has been founded as a Siemens/ABB Gemeinschaftsunternehmen for the High Temperature Reactor (HTR). This cooperation in the reactor industry of the Federal Republic of Germany can be understood as an important signal in the development of the HTR towards commercialization: The HTR for electricity production and cogeneration is an industrial product. The HTR-GmbH offers the HTR-technology in three variants: The HTR-500, the HTR-Modul, and the GHR-10, fig. 1, lit. 1. The HTR-500 is based on the THTR-300 in Hamm-Uentrop and will be used for the production of electricity in larger grids. The HTR-Modul-power plant with small and medium power at one side is based on the experimental reactor AVR in Jiilich and will be used for the production of electricity, process steam and
HTR - GmbH HTR-500 HTR-Modul
founded ,
THTR-300 AVR
Electricity
'X
Electricity
sn}ller
Process Steam grids Heat
GHR-10
District Heat Text16E
1: HTR-company founded: Principal industrial products of the HTR for electricity production and cogeneration D-3
- 2 heat, respectively for the production of electricity in countries with smaller grids. The GHR-10 is used for the production of district heat. Beside these possibilities there exists a wide variety of applications of the HTR in the heat market. In the following there will be given an overview with some remarks on different aspects, and there will be given a detailed description of new proposals in a second paper. 2. HTR Applications: Electricity and Heat 2.1 Physico-chemical Principals due to Temperature The HTR, as the only nuclear systems producing high temperature heat up to 1 000 °C, offers a wide variety of applications in energy conversion processes, because the heat produced in the nuclear fission process is high temperature heat. This heat can in principal be used to increase the heat capacity of heat carriers and to increase the "chemical energy" of chemically reacting systems.
HTR applications, phisico-chemical principles Electricity
Steam Gas
HTR
Turbine
District Heat,Steam
Heat Carriers
Enhanced Oil Recovery
Separation
Desalination of Seawater
Chemical Processes
^ 2 ™6~^" ^ 2 ^
Refinement of Fossils
CH+H2O - * - C O + 3 / 2 H 2
Thermal Loop Thermochemlcal Cycle
4 +H 2 O-*— CO+3H2
H2O—*~H2+1/2O2
Fig. 2: HTR applications, physico-chemical principals of applications and typical examples D-3
- 3 2.2 Short Description of Principals and Processes An overview on the wide variety of applications is given in fig. 2. Via the increase of the heat capacity of working fluids and via the turbine cycle nuclear energy is converted into electricity, fig. 2, upper part. For the steam turbine this is existing technology; the gas turbine offers the possibility of an improvement. In both cases cogeneration is possible for the production of district heat and of process steam. For many processes the production of heat carriers, e.g. in the form of process steam, is an important task. An example for a direct application is enhanced oil recovery by injection of high pressure steam. Another example of the application of heat carriers is the separation of substances, e.g. the desalination of sea water, fig. 2, middle part. A third class of applications applies the increase of the "chemical energy" of systems. Many processes in the chemical industry and related industry, lit. 2, require the application of heat of higher temperature. A classical example is the production of ethylene by the splitting of ethane. Another large area of application is the refinement of fossil energy carriers: Coal, heavy oil and natural gas may be converted into more valuable products for the energy market. These processes can be operated "allo-thermal", that means in such a form that high temperature heat is required, fig. 2, lower part. Process class no. 4 is the "thermal loop", better known as the "EVA/ADAM-system". Here the chemical reaction system of the steam reforming is used for the transportation of energy from a larger nuclear heat plant to consumers of electricity, process steam and district heat, fig. 2, lower part. Last not least, there can be mentioned class no. 5: the thermochemical cycle for the production of hydrogen and oxygen from water, txg. 2, lower part. This process, as well as the electricity production is CARNOT-limited. D-3
- 4 2.3 A Conclusion from the Logistic Growth Law The systemcatic of the "logistic growth law", being applicable in many cases, has been used to analyse the introduction of nuclear energy into the electricity market, lit. 3. Results of this analysis are presented in fig. 3 in form of a FISHER and PRY-transform. It describes the penetration of nuclear power for some countries of the world and gives the value of the capacity of saturation in GWe as well as the time-constant of this penetration. There has been drawn a conclusion, and that is: "According to these charts, nuclear power growth will stop around 1995, indicating the end of the Kondratiev cycle", lit. 3, page 24. Another conclusion may be, that nuclear energy may find new markets in the non-electrical market.
F
TT.2 F 10
1940
1950
1960
1970
1980
1990
2000
Pig. 3: Penetration of nuclear power in world nations with saturation in GWe and time-constant D-3
- 5 3.
HTR-Process Heat, a Future Option
3.1 Status of the Art in General R & D-work for nuclear process heat applications, in particular for the refinement of coal, has been performed in the Federal Republic of Germany since of 15 years in the projects "Prototype Plant Nuclear Process Heat (PNP)", lit. 4, and in the project "Nuclear Long Distance Energy (NFE)", lit. 5. The status of the R & D-work is as follows: 1) The license-ability of a process heat plant, consisting of a reactor plant and of a coal gasification plant, has been evaluated by legal bodies with positive results. 2) The qualification of the metallic high temperature materials is very advanced; for the nuclear reformer and for the intermediate heat exchanger materials lifetimes of more than 100,000 hours have been achieved, the test programmes are going on. 3) The newly developed material AC 66 for the helium-heated gas generator of the steam/coal-gasification plant withstands hard corrosion conditions. 4) Design and calculation methods for high temperature heat components are available. 5) The nuclear reformer and the intermediate heat exchanger have been tested in several modules in large experimental plants and in the pilot plant scale. 6) The hydrogen/coal-gasification process has sucessfully been tested in a semi-technical and a pilot-plant scale. 7) The steam/coal-gasification process, non-catalytic and catalytic, has successfully been tested in a semi-technical plant scale. 8) The AVR reactor in Jiilich has demonstrated the production of high temperature heat in the form of helium with the mean outlet temperature of 950 °C. 9) The THTR-300 in Hamm/Uentrop is demonstrating the large industrial applicability of the HTR technology. D-3
- 6 3.2 The Nuclear-Heated Steam Reformer in Particular A very important component for nuclear process heat applications is the helium-heated steam reformer. There have been developed two versions of helium-heated steam reformers, the NFE- and the PNP-version. Both have been tested in the large experimental facility "EVA/ADAM II" of the Nuclear Research Centre Jiilich, lit. 6. The construction of the experimental reformer for the PNP-project is shown in fig. 4. Primary helium with the mean temperature of 950 °C enters the facility, heats the tubes and is thereby cooled to 700 °C. Primary helium of 300 °C is used to realize "coaxial design". The process gas, a mixture of methane and steam, enters the tubes from the top and reacts almost completely. The end-concentration of CH. is e.g. 5.6 %. The important results of more-year experimental work are - the principals for the design of helium-heated reformers are known and have been verified and - computer programmes for the modulation of the operational behaviour are available.
3.3 Future Development Programme "Nuclear Process Heat" The partners of the PNP-project are continuing their R & Dwork on the application of nuclear process heat in particular for the refinement of coal, lit. 1. Due to the drop of the oil price the effort in this programme is reduced. The programme is planned for 10 to 15 years. On the one-hand side this indicates that there will be an economic future. On the other hand side the uncertainty in the development of the prices of conventional energy carriers has to be considered. The programme is organized in three phases: D-3
- 7 -
Spaltgassammelraum 460 °C Spaltgas ProzeBgas 330 °C
ProzeBgassammelraum Tragplatte Isolierung Rekuperator 300 °C P r i m a r _ 700 °C helium Zwischenplatte
Spaltrohre RohrfUhrungsplatte
300 °C Primarhelium 950 °C
Fig.
4: Design of the test facility for the PNP-bundle of the helium-heated steam reformer D-3
- 8 Phase 1: Development of concepts and evaluation Phase 2: Improvement of technical viability Phase 3: Development and design work for the first plant. The programme is broad and covers all main parts of a plant for the refinement of coal using HTR-heat, fig. 5.
Future
R+D
: Nuclear Process Heat
group
1
Steam Reformer
2 3 4 5 6 7 8
Hydrogen/Coal-Gasification Steam/Coal-Gasification Process Heat HTR Gas Cleaning Materials Safety and Licensing Potential of lmprovement,Market
T^ZE
Fig. 5: Topics of the future R & D-programme on nuclear process heat applications Group Group Group Group
1: 2: 3: 4:
Helium-heated steam reformer: new. app., higher T Hydrogen/coal-gasification: higher throughput Steam/coal-gasification: new app., higher T Process heat-HTR
Group 5: Gas cleaning: high T desulphurisation Group 6: Materials programme: cheaper materials, AC66 Group 7: Safety and Licensing: process heat applications Group 8: Potential of improvement, market, market penetration.
- 9 -
The topic "Process Heat-HTR" means the following: 1) Increasing of the outlet temperature to 1 000 - 1 110 °C. 2) Increasing of the inlet temperature to 350 - 500 °C. 3) Decreasing of the helium pressure. 4) Decrease of the effort on, omitting the intermediate heat loop. 5) Dedicated pressure containing, new arrangements of components. The increase of the outlet temperature and of the inlet temperature allows in total a better application of the heat in the chemical processes for the refinement of fossil fuel energy carriers, because this results in positive effects on the chemical equilibrium and on the kinetics of the chemical reaction. A typical example is the increase of the reaction rate of coal per HTR thermal power in the steam/coal-gasification by a factor of 2 by the increase of the process temperature by 50 K from e.g. 800 °C to 850 °C. The decrease of the helium pressure allows an equivalent decrease of the pressure of the chemical process, which has a positive influence on the chemical equilibrium. Of course the optimum value has to be found at, may be, around 20 bar. The intermediate heat loop may in future be not necessary because the retention qualities of the fuel element is excellent and because further improvement of materials and apparatus may allow heating by primary helium, e.g. by improved methods for inspection. The dedicated pressure containing and new arrangements of components should result in a reduction of costs due to less importants of pressure in process heat applications.
D-3
- 10 4.
Economics, the Values of Resources
4.1 Direct Competition The economic competitiveness of the HTR for electricity production has been shown by studies of the reactor industry together with the utilities, lit. 8 and 9. These evaluations have been done in comparison to a large light water reactor under German conditions. The general results are: The HTR is competitive. The economical competitiveness of nuclear process heat applications have been evaluated for nuclear coal gasification under the consideration of the potential of improvement of the technology, lit. 10. Further information are included in lit. 2. The general results is: There is economical competitiveness compared to conventional processes and - at higher prices of conventional energy carriers - there is competitiveness in the market. Of course, all these conclusions depend on the future development of nuclear energy and on the success of the continuing R & D-work. 4.2 The value of a Natural Resource Beside the "direct competition" it may be worthwhile to consider a second criteria in economics. This is the value of a natural resource. The resources of fossil energy carriers are relatively large and will be depleted only after several decades. Nevertheless they are limited. Therefore it may be an economic point of view, to expect a maximum of profit from a certain natural resource. Obviously and in principal natural resources are used for the built-up of infra-structures. For the explanation let's take an example. There may be a natural gas field with the capacity of lTWy (Tera-Watt-year). That
D-3
- 11 amount of energy is equivalent for the operation of an energy production of about 30 GW with the capacity factor of 0.8 for 40 years. At the market place Europe this resource has different values depending on the kind of product, which is produced from the resource and, of course, the market conditions. Typical examples are given in fig. 6, lit. 11. The example considers the import of products after transportation over a distance of about 2 500 km with a power input equivalent to about 30 gW. Depending on the product and on the conversion process the market value of the natural resource is as follows, fig. 6. If the resource is sold as natural gas the market value is 300 x 10 9 DM, if it is converted into methanol by an autothermal process the market value is 450 x g 10 DM, and if it is converted into methanol and ammonia by an allo-thermal process with HTR-process heat, the market value g is about 7 50 x 10 DM. So by conversion the market value is increased to 150 % respectively to 250 %. In this calculation
Natural
R e S O U r C e S , Increasing their values
Natural Gas Field of 1 TWy (30 GW x 0.8 x 40y)
Market Value
109DM
as
1) Natural Gas 2 ) M e t h a n o l (autothermal)
3) Methanol+Ammonia allothermal NG=1ODM/GJ;
with
300
100
450
150
750
250
HTR
Liquids= 500 DM/t
Fig. 6: Natural resources, increase of their market values with the example of natural gas for a natural gas field D-3
- 12 the following specific market values, typical for the Federal Republic of Germany in 1986/87/88 have been taken: Natural Gas = 10 DM/GJ and methanol and ammonia (average) = 500 DM/t, lit. 11. The first increase is achieved by the conversion of the natural resource "natural gas" in the more valuable product "methanol" by an auto-thermal conversion process including cheaper transportation as a liquid. The second improvement of the market value is achieved by the application of nuclear energy in the form of high temperature heat from the HTR. The chemical principals of the conversion of natural gas into the liquid products methanol and ammonia are explained in fig. 7. Natural Gas CH. + water + air is converted into methanol CH-OH and ammonia NH-.. If the process is operated "allo-thermal", that means using high temperature heat from the HTR, all CH. can be converted in CH^OH in principal and the hydrogen surplus can be converted into ammonia. In an auto-thermal process (without HTR) the product yield is about 0.6, in the allo-thermal process (with HTR) the product yield is 1.06. So the pro-
Methanol and Ammonia from Natural Gas 28CH 4 +24H 2 O+2(O 2 +4N 2 )
=$> 2 8 C H 3 O H + 1 6 N H 3
+HTR Product Yield
without HTR with HTR
0.60 . 1.06
Product Yield = H (Product) / H (Feed) TextKE Fig. 7: Methanol and ammonia from natural gas, chemical principals and product yields of different processes D-3
- 13 duct yield is increased by the factor of 1,77. The product yield is the ration of the heating value of the product compared to that of the feed, fig. 7. In the context of discussions on technologies to mitigate the CO?-problem an interesting proposal has been made, lit. 12. It is the export of Hydrogen from the Sovjet Union to Western Europe. The conversion process is the production of H 7 from Natural Gas by steam reforming, using HTR heat, and the separation of C 0 ? . The C0« may be used for enhanced oil recovery or may be stored in expired gas fields. In total this is C 0 ? free application of natural gas made effective by the application of nuclear energy from the HTR.
5. Environment and Safety In the scientific discussion on the future development of nuclear energy there are proposals to increase the safety of nuclear energy, lit. 13. To some extend this is because of the "nuclear controversy". Recently this discussion got an additional driving force by the climatic problems due to the emission of carbondioxide. In a paper on "carbondioxide and inherently safe reactors" by A.M. Weinberg, lit. 14, it is stated: "Carbondioxide is emerging as the world's central environmental concern. This development poses a dilemma for environmental activists: Nuclear energy, which especially after Chernobyl, has been regarded by many activists as an environemntal abomination, nevertheless is one of the few energy sources, that do not emit C O " . In discussing requirements for future reactors the following conclusion is made: "An acceptable core-melt probability seems to be an overriding necessary condition. At the Rasmussen fiducial rate of ca. 10 -4/reactor year, a carbon sparing energy system with about 5 000 reactors would suffer, on average, a core-melt every other year. The apriori core-melt probability would have to be reduced by at least /r
two, and possibly three three orders of magnitude, say to 10 two, and possibly 7 even 10~ /reactor year".
or D-3
- 14 In addition to the reduction and limitation of the risk there is the additional possibility of the reduction and limitation of the damage as for example with the user-requirement "no sheltering, no evacuation" of US-utilities, lit. 15. The most promising official results in that direction is realized by the HTR-Modul, lit. 16, which recently has been discussed in an licensing procedure in the Federal Republic of Germany. As an example for the limitation of the damage the following typical result is given from a safety evaluation on the HTRModul, lit. 16, p. 50. In the accident "loss of pressure in the primary circuit with subsequent core heat-up" the accident dose for "total body" for adults is 0,0003 Sv, equivalent 0,03 rem. This value is by a factor of about 15 lower than the limit values after article 28 of the German Radiation Protection Ordinance and by a factor of about 3 less than the yearly natural background. In addition to that encouraging state of the art in safety and in limitation of damage it is, of course, possible to develop further improvements. Such an improvement would be the "fire protection" of graphite, the main construction material of the fuel elements and of the core of the HTR. In that contexts it is important to remember that this material graphite - being at the same time the moderator for the neutrons - is high temperature stable and therefore one reason that high temperature heat can be produced. Initiating work on fire protection has been done in the Kurchatov Institute, lit. 17, on the deposition of carbon in the pores of graphite. Further proposals have been made on other materials, e.g. Silicon Carbide SiC, other techniques, e.g. thin layers, lit. 18, and so on. Protection of graphite is used meanwhile in conventional techniques, there it should be possible to develop finally such an improvement for the graphite of the HTR.
D-3
References 1)
ABB/HRB und Siemens/Interatom: HTR-GmbH jetzt gegründet, Siemens/ABB-Gemeinschaftsunternehmen für den Hochtemperaturreaktor, gemeinsame Presseerklärung, Frankfurt, 8.5.1989.
2)
Schad, M. et al.: Nutzung der Prozeßwärme des HTR in der chemischen und verwandten Industrie, Projektstudie HTR Prozeßwärme-Einkopplung, Teil 1, Dezember 1988.
3)
Marchetti, С : On society and nuclear energy, historical analysis of the interaction between society and nuclear technology with examples taken from other innovations, final report for contract no. PSS 0039/A between IIASA and the European Atomic Energy commission, Dec. 1988.
4)
Brandtstätter, A., Jansing, W.: Results of the PNP-Project, 9th International Conference on High Temperature Gas-Cooled Reactor, Coal and Nuclear Power for the Generation of Electricity and Gas, Dortmund, FRG, 27.-28. Oktober 1987.
5)
NFE: Nukleare Fernenergie, zusammenfassender Bericht zum Projekt Nukleare Fernenergie (NFE), Kernforschungsanlage Jülich GmbH, Jül-Spez-303, März 1985.
6)
Nießen, H.F. et al.: Erprobung und Versuchsergebnisse des PNP-Teströhrenspaltofens in der EVA-II-Anläge, Jül-2231, August 1988.
7)
PNP-Partner and Ruhrkohle Öl und Gas GmbH: F+E-Arbeiten zur Verbesserung und Absicherung der technischen Auslegung und der Wirtschaftlichkeit und zur Vorbereitung der Markteinführung der nuklearen Kohleveredlung, Phase 1: Konzeptentwicklung und Bewertung, 1989-1992, Oktober 1988,. PNP-Partner: Bergbauforschung GmbH, Hochtemperatur Reaktorbau GmbH, Gesellschaft für Hochtexnperaturreaktor-Technik mbH/Interatom GmbH, Kernforschungsanlage Jülich GmbH, Rheinische Braunkohlenwerke AG.
8)
Arbeitsgemeinschaft Hochtemperaturreaktor (Hochtemperaturreaktorbau GmbH, Ruhrgas AG/Ruhrkohle AG, Vereinigte Elektrizitätswerke Westfalen A G ) : HTR-500 Vorprojekt-Untersuchung, Zusammenfassung, Juli 1984.
9)
Arbeitsgemeinschaft Hochtemperaturreaktor (HochtemperaturReaktorbau GmbH, Ruhrgas AG/Ruhrkohle AG, Vereinigte Elektrizitätswerke Westfalen A G ) . HTR-Modul Vorprojektuntersuchung, Zusammenfassung, November 1984.
10) Barnert, H., Singh, J., Nießen, H.F., Neis, H. , Hohn, H.: Potential-Studie zur Kohleveredlung durch WasserdampfKohle-Vergasung (WKV) mit Hochtemperaturreaktor-(HTR)Wärme, Jül-2131, Mai 1987.
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11) Barnert, H.: Synthesegas aus Erdgas mit HTR-Wärme, in Terhorst, W.: Erdöl und Erdgas in der Kernforschungsanlage Jülich, Tagungsbericht des Arbeitsseminars in der Kernforschungsanlage Jülich, Jülich am 1. und 2. Juli 1986, Jül-Conf-58, Juli 1986. 12) Marchetti, C : How to solve the CO~-Problem without tears, 7th World Hydrogen Conference "Hydrogen today", Moscow, September 25-29, 1988. 13) Nucleanics Week: Outlook on Advanced Reactors, a special report to the readers of nucleanics week, page 1-20, March 30, 1989. 14) Weinberg, A.M.: Carbondioxide and inherently safe reactors, International Conference on Enhanced Safety of Nuclear Reactors, Institute for Technology and Strategic Research, The Georg Washington University, Washington, D.C., August 10, 1988. 15) Gas-Cooled Reactor Associates, GCRA: A utility User Summary Assessment of the Modular High Temperature Gas-Cooled Reactor Conceptual Design, GCRA 87-011, Revision, November 1987. 16) Siemens/Interatom: Hochtemperaturreaktor-Modul-Kraftwerksanlage, Kurzbeschreibung, November 1988. 17) Chruljow, A.A., Momot, g.W., Kozirew, M.M., Schmeljow, W.P.; presented by Grebennik, V.: Einfluß der Glübedingungen (Temperatur, Luftatmosphäre) auf die Dichtigkeit von BE-Modellen des HTR, Seminar im Rahmen der wissenschaftlich-technischen Zusammenarbeit zwischen dem Staatskomitee für die Nutzung der Atomenergie der UdSSR und dem Bundesministerium für Forschung und Technologie der Bundesrepublik Deutschland, Kernforschungsanlage Jülich GmbH; 7.-11.3.1989. 18) Schulten, R.: Die globale Bedeutung der HTR-Prozeßwärme, Jahrestagung Kerntechnik '89, Düsseldorf, 9.-11. Mai 1989.
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XA0101501
PROBLEMS OP ATTRACTING NUCLEAR ENERGY RESOURCES IN ORDER TO FROYIDE ECONOMICAL AND RATIONAL CONSUMPTION OP FOSSIL PUELS E.K. Nazarov, A . T . N i k i t i n , N.N. Ponomarev-Stepnoy, A.N. P r o t s e n k o , A.YOt, S t o l y a r e v s k i i a n d N.A. Doroshenko State I n s t i t u t e of Nitrogen Industry 50, Chkalova Street, Moscow, 109028, USSR I.V. Kurchator I n s t i t u t e of Atomic Energy Kurchatov Sg., Moscow, 123182, USSR Abstract
Depletion of fossil fuels resources and gradual increase in cost of their extraction and transportation to the places of their consumption put forward into a line of the most urgent ; tasks the problem of rational and economical utilization of : fuel and energy resources as well as introduction of new energy sources into various sectors of national economy. The nuclear energy sources which are widely spread in ; power engineering have not yet used to a proper extent in the sectors of industrial technologies and residential space heating which are the most energy consuming ones in the national ; economy* The most effective way of solving this problem can be the development and commercialization of high temperature nuclear reactors, as the majority of power consuming industrial processes and those involved incheraico-thermal systems of distant heat transmission demand the temperature ofa heat carrier generated by nuclear reactors and assimilated by the above processes to be in the range from 900° to 1000°C, Wide interest to the problems of hydrogen energy engineerIng in the earlier 1970's was raised by the world energy crisis, especially, in regard to hydrogen based fuel and energy resources, intensified search of new convenient, economical and pollution free energy carriers, necessity of utilizing new optional renewable energy sources for solving problems of fuel and energy balances, and demand for development of resourceand energy-saving technologies. While developing the strategy for solving the above problems in the Soviet Union, based on the worldwide established concept of selecting hydrogen as a promising and universal energy carrier of the future which has some advantages even over electricity, we have found important to concentrate our aiain efforts, first of all, on the use of widely adopted renewable energy sources, particularly nuclear energy ones, as !well as on comprehensive development of new efficient processes for hydrogen production involving more widely spread raw mateD-4
- 1 -
rials such as water and coal. In conformity with the strategy adopted in the Soviet Union the problem was named "Nuclear-hydrogen power engineering and technologies". Within the framework of the adopted programme the country has achieved certain progress in establishing the concept of new reaction units as well as in improvement of the existing and newly designed technological processes. The results achieved appeared in periodical publications and were reported at various conferences, seminars and congresses both in our country and abroad /1-9/. Based on analytical research performed by us and taking into account the present state of art we are of opinion today that the new alternative energy sources, e.g. nuclear ones, will be introduced into industry through a number of such main steps as: - Development and mastering of stable operation of high-temperature nuclear reactors generating high potential heat, the heat carrier temperature being in the range from 900 to 1000°C; - Search of rational combination of process achievements and technical solutions of the mastered and perspective productions with specific character of discharging heat from nuclear reactors; - Utilization of nuclear reactor energy for partial (approximately by half) substitution of organic raw materials to be presently combusted to meet the power demand of existing and commercialized production plants; - Complete substitution of organic raw materials with nuclear fuel, both as energy and feed, on the account of producing gaseous hydrogen containing mixtures mainly from water by electro-, thermal- and plasma-chemical methods recovering energy from high temperature reactors; - Review of conditions and development of organizational and engineering solutions to be acceptable for implementation of nuclear energy in commercialized processes. The work on engineering high temperature nuclear reactors has been carried out for nearly three decades. During this period the concepts of their application and main problems of their engineering have been drastically revised. Initially these works were mainly oriented to use high temperature nuclear reactors at the nuclear power plants equipped with steam turbines and, for the long term, gas turbine nuclear power cycles. In such case the heat carrier temperature at the outlet of an active zone is in the range from 600° to 800°C. It is known that presently the nuclear power plants equipped with the high temperature reactors with the capacity of about 300 1SW (e) each have been designed and successfully operated in Port-St.Vrein (U.S.A.) and Schmehauzen (PRG). The high temperature nuclear reactors were supposed to be - 2 -
469
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applied in high temperature technologies requiring heat of 9001000°C and higher after their adoption in power engineering, meanwhile, both in the USSR and abroad the expediency of constructing such reactors was purely connected with their adoption in technologies associated with energy sources. The attention and efforts of scientists, engineers and designers were concentrated on the development of the high temperature nuclear reactors to be cooled with helium under the pressure of 40-50 bars. In such reactors practically non-activated helium having good neutron- and thermal-physical properties is used as a heat carrier, graphite as a moderator and main material for constructing an active zone and uranium and thorium - as fuel. The reactors can provide large depth of burning out, low value of fuel component in energy cost and high degree of safety thanks to negative temperature coefficients and relatively high heat capacity of an active zone. Meanwhile, as a rule, one considers an integrated layout of equipment of HTGR within the shell made of prestressed reinforced concrete which is at the same time assigned a mission of biological protection. Recently the interest has greatly grown to the "module" concept of HTGR when the reactors are prefabricated within metal shells to a maximum extent of readiness. However, one needs to keep in mind that in spite of this concept attractiveness for many fields of activity (to meet the commitments of small developing countries in power engineering and process engineering, to provide heat and energy for distant uneasily accessible regions etc.) the reactors of maximum readiness (from the point of the shell construction and transportation to the consumer) will have maximum capacity of about 250-300 MW (th) and only half of this capacity will come to that high temperature portion which is highly required for high-temperature industrial technologies. Such capacity is not sufficient for the majority of large-tonnage energy-consuming plants and then the need will arise to install two or three reactor modules for energizing one industrial plant. Such low capacity of one reactor is not compatible with the future concept of scale adoption of nuclear energy into commercial technologies that will be discussed below. Progress in developing many major components of H3?GR including high-temperature nuclear fuel (fuel microjiairtswith a multilayer ceramic envelope which confines the fission fragments at a temperature of up to 125O°-145O°C and possibly higher) and in mastering helium handling procedures etc. gave a hope for realization of the HTGR in the nearest future. At the same time it is known that today, no matter how many programmes were established in various countries, no reactor, unit has been constructed for supplying high potential heat to industrial technologies. In our opinion there are several reasons for loosing the rate of realization of the programmes which were developed and announced in various countries. The first and main reason is -. 3 -
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perhaps the change of the situation at the world fuel and energy market, especially, as to petroleum and gas. Second, general tendency for changing views on structural development of the components of power engineering, particularly nuclear, after a number of accidents at the nuclear power stations (NPS), especially, in Three-Mile-Island and Tchernobyl. Third, revealing of some difficulties in achieving high temperatures of a heat carrier and particularly in designing suitable construction materials, heat transferring and perceptive equipment. In our opinion the above reasons are temporary in character and the need for development and commercialization of the . high temperature reactors for supplying energy to the industrial processes will become more acute from year to year towards the end of the century. Therefore, the Soviet Union continues to attach a great importance to the works associated with the development of a pilot nuclear energy-technological complex to be equipped with a high temperature reactor VG-4OO. The main data on performance of such complex and the programme of step-by-step achievement of the required temperature level were earlier issued in publications /10-12/. It is true that since the first publication / 1 / the approach to the technological part of the complex has been re-examined and at present the priority is given to the complexes associated with hydrogen conversion from natural gas instead of originally designed thermo-electrochemical hydrogen production in a sulphuric acid cycle. In the majority of the publications one can find a wide range of particular fields of adopting the high-temperature nuclear reactors in industrial technologies: - chemical industry (synthesis of ammonia, methanol, hydrogen etc.), - petrochemical industry (pyrolysis of oil and byproducts of its recovery and refinery), - ferrous and non-ferrous metallurgy (production of reducing gases and perhaps heating systems for high-temperature technological plants), - transformation of organic energy carriers and production of artificial ones (coal gasification and upgrading, hydrogen production from water, production of synthetic natural gas etc.), - establishment of centralized industrial and public heatsupply systems with long distance heat transportation (100 km and over) in a chemically fixed form based on a closed cycle of reversible chemical reactions, - energy accunulation by its conversion into chemically fixed forms with its further discrete consumption to meet a variable part of the load curve. The relative importance of the above directions varies from one country to another. In the USSR conditions the toppriority task which deserves utmost attention and effort is - 4 -
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perhaps the use of nuclear energy in the chemical industry. There are some reasons for this. First, the attraction of nuclear energy resources is very urgent for chemical technology as this field of industry (particularly in production of ammonia and methanol) is a well established, large-scale and intensively developed consumer of such valuable resources as natural gas (in 1985 the above productions consumed over 30 billion cubic meters of natural gas and by 2000 this figure can be increased by 2-2,5 times). Second, it is the chemical technology where the processes are developed, assimilated and implemented which make a technological basis both for production of reducing gases in ferrous and non-ferrous metallurgy, coal gasification, pyrolysis of oil and by-products of its recovery and refinery, transportation and accumulation of heat in a chemically fixed form, and production of hydrogen from water by means of electro- and thenno-chemical processes. That's why the progress in implementation of nuclear energy into chemical technology (particularly into a high-temperature, energy-intensive process of catalytic steam conversion of methane) is an important technological requirement and may become an engineering basis for nuclear energy practical adoption in other fields of industry (Table 1). Third, gained in chemical industry the experience in developing, constructing, assimilating and operating high-temperature technological plants is a basis for the development of chemical production associated with nuclear energy. Por instance, the reaction tubes of ammonia and methanol production which have a service life of about 100 000 hours (10-12 years) at a temperature of up to 950°C and pressure drop of 40 bars (when they are preheated with combustion gas) and duration service of 8000 hours per year can be an engineering and material basis for developing new technological plants to be heated with the help of nuclear reactors. In Pig. 1 there is shown a principal energy-recovery technological flowsheet of one alternative pilot nuclear-chemical ammonia production (about 900 000 tonn NEL/year), which is designed based on a modular lay-out when the nuclear based methane production is located in direct proximity to the process equipment of an ammonia synthesis plant, on the common site. Such design can still be justified for the first pilot nuclear energy-technological complex. Strictly speaking, even with such a lay-out, based on current fire-, toxic- and explosion requirements, it is necessary to solve a problem of spacing at proper distances a nuclear reactor and heat-consuming equipment (methane converter, heat exchangers etc.), as well as the construction sites of a nuclear energy-technological plant (NETP) and downstream sections of the ammonia production. Moreover, due to a pilot character of NETP and step-by-step assimilation of the required temperatures - 5 -
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of a heat carrier the task is raised to provide flexibility in self-consistent operation of power engineering and chemical process parts of the complex. Figures 2, 3 and 4 show several alternatives of feeding converted gas generated by NETP to downstream ammonia synthesis, and Figure 5 shows an alternative of an integrated complex layout. A number of new engineering problems have to be solved in every alternative: the most serious of them will perhaps be those connected with distant transportation of hot gaseous and vapour media under high pressure and development of adequate control and gate valves. The data which are available in the world scientific, technical and patented publications look encouraging for successful solution of the above task /13-15/. Meanwhile, such technical solution cannot be accepted as a typical one for scaling-up the nuclear reactors into the high-temperature industrial technology as it has some shortcomings. Prom a point of increasing economical efficiency of the nuclear-chemical production a non-optimized character of a modular design can be caused by several reasons. Single heat capacity of a nuclear energy-technological complex (even for such perspective productions as ammonia one with the capacity of up to 1 million tons per year) amounts to at least from 500 to 600 WW (th), which is acceptable but not yet optimal even in case of specialized high-temperature nuclear reactors. The increase of this capacity by 2-4 times could significantly reduce (by 20 to 30% and over) specific capital investments per one energy-technological complex. Autonomous location of a nuclear energy-technological plant (like any other separately located nuclear reactor of small capacity) results in the fact that the auxiliary facilities (such as a gas distribution system; control and treatment of process effluents and wastes; handling of non-irradiated fuel; holding, treatment and handling of wasted fuel; burial of radioactive wastes etc.) are non-optimal from economical point of view (i.e. specific capital investments have become higher). In case several NETPs are located on one common site, the specific capital investments to be required for these auxiliary facilities (based on one unit of the reactor installed capacity) could be appreciably reduced due to the use of some facilities (without their enlargement) for attending a set of NETPs (transportation systems, holding basins, container park etc.). The worldwide capacity utilization rate of the nuclear reactors to be installed in the nuclear power stations amounts averagely to 0.7-0.8 (6000-7000 hr/year), which is much lower the utilization rate of chemical plants where it usually amounts to 0.9 (8000 hr/year). If high-temperature nuclear energytechnological plants have the utilization rates equal to those of the nuclear reactors of HPS (nuclear power station) then under modular construction of the nuclear-chemical production they would become unprofitable in case of process equipment shutdown, - 6 -
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In perspective the expansion of ammonia, methanol etc. production can be effected in two routes: expansion and retrofitting of the existing plants (main portion of increment) and construction of a number of unique large-tonnage plants with summary capacity (based on product) of 5-10 million t/year. Modular construction of the nuclear-chemical plant will affect the realization of the both routes negatively. Along with the above mentioned economical losses the autonomous location of several nuclear-based chemical plants within the battery limits of a large-tonnage chemical complex inevitabely leads to transportation of radioactive materials along the territory of the chemical complex (delivery of fresh nuclear fuel to NETP of every production, disposal of the wasted nuclear fuel and radioactive wastes etc.), which will make the facilities to be required for anti-radiation protection significantly more complicated and expensive. Introduction of one or several nuclear-chemical production plants to be integrated with nuclear and chemical equipment on modular basis into the existing complexes, along with the above mentioned economical losses, will lead to dissipation of nuclear reactors along the territory of the USSR, demand of additional transport streams of both fresh fuel (to the chemical complexes) and highly active wasted nuclear fuel and radioactive wastes (from the chemical complexes) along the regions of highly densed population. This is undesirable. Therefore, the progress in large-scale implementation of high-temperature nuclear reactors into industrial technologies greatly depends on possibility of selecting such forms of arranging the nuclear based chemical productions which like in power engineering would permit to split a nuclear based technological section (methane conversion; and process sections (downstream stages of production) of the nuclear based chemical complex, to enlarge summary and single capacities of the hightemperature reactors within the scope of a nuclear energy based technological complex by means of parallel service of several process plants and complexes /16/. In Figure 6 it is shown a principal energy/process flowsheet of a nuclear based ammonia production where a nuclear based technological section and a process section are located on different sites and heat in a fixed form is transported for long distances. The process and power facilities of the site are linked with three "cold" gaseous streams (one - converted gaseous mixture to be used as a feedstock and two streams are transporting for long distances heat in chemically fixed form based on a closed cycle of reversible chemical reactions). Table 2 shows the data which are demonstrating the change in losses and capacity of HETP in the scope of a nuclear based chemical complex producing 900000 tons of ammonia per year and possibilities of raising single train capacity and a number of UETP on the site of the nuclear based process section of the complex. The data of Table 2 and Figure 7 say that inspite of - 7 -
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additional power losses the alternatives of the separate location of nuclear based process section and purely process sections provide the possibility of improving technical and economical conditions of the nuclear energy based chemical production. Engineering and feasibility study has shown that such arrangement of nuclear energy based chemical production would be effective from economical point of view in case the total length of the gas pipeline connecting the nuclear based process section and process sections of the complex will be up to 5001000 km. Meanwhile, such arrangement discovers purely new opportunities: - nuclear energy based conversion unit can render service to a number of chemical complexes, - the existing chemical complexes can be modernized and expanded by means of step-by-step connection of their units to the nuclear energy based conversion unit, - the nuclear energy based conversion unit which is feeding a number of chemical complexes can have a summary heat capacity of the nuclear reactors of about 10-30 million kilowatts, - the nuclear reactors integrated with NSTP can have a single train heat capacity of over 500 MW (ih), - additional methanators can be installed on the chemical complex site and then the heat power stations and boilers which were designed to meet power requirements of the complex may be shutdown thanks to the expansion of the capacity of the nuclear energy based conversion unit, - adoption of the methanators operating at the temperature levels of 300-400° and 6OO-7OO°C enables to construct nuclear heat chemical stations to meet both the power requirements of the complex and public space heating. The flowsheet of the nuclear energy based conversion unit and T-Q diagram of converted gaseous mixture heat recovery is presented in Figure 8. In view of the aaid above the nuclear energy based conversion -units may become an universal tool for efficient largescale implementation of new fuel resources not only into ammonia production but into all the spheres of high-temperature industrial technologies which are based on the methane conversion: ferrous metallurgy, coal gasification, centralized supply of low and medium potential heat for industrial and public consumers, generation of electric power in random fluctuating mode /Fig.9/. Meanwhile, the nuclear energy based conversion units may have a summary rated capacity of (10-20)* 10" kW and over and service the large geographical and economical regions.
1,
REFERENCES Legasov 7,A. et al. "Nuclear energy", vol. 45, No. 6, December, 1978. - 8 -
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2.
3.
4.
5. 6.
7.
8.
9. 10.
11. 12. 13. 14. 15. 16.
Dollezhal N.A., Koryakin J.I., Nazarov E.K. et al. "Application of nuclear reactora in high-teraperature energy-recovery industrial processes", Report IAEA-CN/36/339, Zalzburg, 1977. Spielrain E.E., Malyshemko S.P., Kuleshov G.G. "Introduction into hydrogen power engineering", Moscow, Energoatomizdat, 1984. Nazarov E.K., Stolyarevsky A.J. "Energy/technological application of high-temperature nuclear reactors", In: "Nuclear/hydrogen power engineering and technology", issue 3» Moscow, Atomizdat, 1980, pp. 58-128. Belousov I.G., Legasov V.A., Rusanov 7.D., "A plasmochemical concept for thermochemical hydrogen production", Intern. J. Hydr. Energy, 1980, vol. 5, pp. 1-6. Legasov V.A., Vaker A.K., Denisenko V.P. "Plasma radiolysis of gaseous carbon dioxide with strong current beam of relativistic electrons", Dokl. AN USSR, 197,8, vol. 243, No. 2, p. 323. Borisov E.A., Trusov G.N. , Shiryev 7.K. "Perspective evaluation of therrao-cheraical cycles in hydrogen production from water", "Voprosy atomnoy nauky i techniky", ser. "Nuclear/hydrogen engineering and technology", 1977, issue 2(3), p. 34. Belousov I.G. "Theory of thermal methods for hydrogen production from water", In: "Nuclear/hydrogen engineering and technology", 1980, p. 172. Legasov V.A. et al. "Reports of the Third International Symposium on plasmochemistry", Limozh, p. 5.18. Ponomaryev-Stepnoy H.H., Protsenko A.N. et al. "Nuclear/ Hydrogen Power Engineering and Technology", issue 1, Moscow, Atomizdat, 1978, p. 80. Mitenkov P.M., Koshkin J.N. et al. "Uuclear/Hydrogen Power Engineering and Technology", issue 2, Moscow, 1979» p. 73» Golovko B.P., Dorofeev A.M. et al. "Nuclear/Hydrogen power Engineering and Technology", issue 5, Moscow, 1983, p. 123. Patent 7RGt 2 946 511, 1979. Patent FRG, 3 147 725, 1981. Patent FRG, 3 146 305, 1981. Nazarov E.K., Tchernyaev V.A., Radchenko S.V., Smirnova E.S. "Voprosy atomnoy nauky i techniky", ser. "Nuclear/ Hydrogen power Engineering and Technology", Moscow, Zniiatorainform, issue 2, 1985, pp. 3-10.
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TABLE I : NUCLEAR ENERGY-TECHNOLOGICAL PROCESSES BASED ON METHANE CONVERSION
Designation
1 Reduotion of natural gas consumption
o i
Production type
Product
Gas (heavy oil) to be substituted with NETP*, billion m3/ year 3 4 Steam conver- Ammonia, 1.7 sion of memethathane to nol, produce synthehydrogen tic higher alcohols, hydrogen Direct reSponge 1.5 duction of iron
Million t-equivalent/ year
Economical efficiency
Potential volumes of substituted fuel under NETP adoption,million t-equivalent/year
5 2.1
6 Production cost is reduced by 10-15%
1.8
Not estimated
2-3
(20 million tons of sponge iron)
1.7-2.0
Not estimated
5-6
(30-40 million tons of castiron)
7 30-50 (NETP capacity being 15-20 million tons of nitrogen fertilizers)
ore Ore reduction in blast furnaces
Castiron
1.4-1.6
to be cont'd o
Table I cont'd
1 Production of synthetic fuel
Coal gasification
Synthetic gas
3-3.2
Replacement Chemo-thermal Peak elec- 1.9-2.1 of gas/oil accumulation trie fuel of energy energy
Chemo-thermal transmission of energy for long distanoes
Dec entra- 1,3-1.6 lized public and industrial heat supplying
2.6-3.8
Effective at 60-70 natural gas price - 40-50 roubles/tequivalent coal
(at gasification of 50 million t)
2,4-2.1
Not estimated
32-35
(installed capacity of flexible NPS - 20 GW)
1,6-1.8
More economical than NHS** located at a distance of 20-30 km from heat 6 onsumers
80-120 (50% of decentralized heat supply in European part of USSR as perspective)
* NETP - Nuclear energy-technological plants; the figures are in conformity with the NETP capacity to be consumed by processes - 2 000 MW (th); *• NHS - Heat supplying nuclear station based on nuclear reactors adopted in industry.
o p
TABLE II: POWER LOSSES OP NUCLEAR ENERGY BASBD CHEMICAL COMPLEXES AND POSSIBLE CONCENTRATION OP NETP CAPACITI UPON CHANGING PROM AN INTEGRATED LAYOUT TO SEPARATED ONE Siting of nuclear/process and process units
Integrated modular layout (Pig. 1)
Separate layout
Specific heat capacity of (•
NETP reactor, M W (fch)/O.9»1O° tons NH, per year Reactor capacity enlargement under a separated layout, MW (th)/O.9-1O6 tons NH 3 per year Portion of reactor capacity enlargement which can be used for public space heating, MW (ttO/O.9-106 tons NH- per year
520*
720'
0
200
0
Summary capacity of NETPs located on one site, including actual one,
M l CttO Required heat capacity of the nearest consumers, MW (th)
4320-8640 520
3120-6240
0
600-1200
Maximum single capacity of
NETP reactor, MW Cth) Admissible number of reactors to be located on one site *
90-120
520
2160
1
6-12
assumes a nuclear reactor capacity which covers process requirements.
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SJ
Pig* 1. - Ammonia produotlon flowsheet with utilization of nuclaar reactor heat, 1 - turbo-compressors of natural gaa air and hydrogen/nitrogen mixture, 2 - heat exchangers, 3 - steam generators and steam superheaters, 4 - desulphurization area, 5 • ateam reforming of methane, 6 - high-temperature nuolear reaotor, 7 - steam/air reforming of methane, 8 - shift conversion area, 9 - oarbon dioxide removal area, 10 methanation, 11 - ammonia synthesis loop0
?ig. 2. - Ammonia product ion flow/sheat with utilization of nuclear reactor heat and transportation of hot converted gas after steam reforming to distant consumers. 1 - high-temperature nuolear reactor, 2 - hot helium pipeline, 3 - reformer, 4 - steam/ gas mixture preheatera, 5 - steam generators, 6 - steam superheaters, 7 - boiler feed water preheat era, 8 - drum separators, 9 - turbo-compressor of natural gas, 10 - pretreatment area, 11 - flame preheaters of natural gas and air, 12 - raw natural gas, 13 - process air, 14 - steam/air convertor, 15 - shift converter, 1st stage, 16 - shift converter, 2nd stage, 17 - mathanator, 18 - ammonia synthesis loop, 19 - steam turnines, 20 - steam primary reforming area*
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o
Fig. 3. - Ammonia production flowsheet with utilization of nuclear reactor heat and transportation of converted gaa after the 1st stage 00 conversion to distant consumers*. Symbols are similar to those of Pigure 2.
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M 4 - 5 -15- 5
g 4» - Ammonia production flowsheet with utilisation of nuclear reactor heat and transportation of precooled converted gae after methane steam reforming to distant coneumera. Symbols are aimilar to those of Figure 2.
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1
1
1
r
i
L.
Pig. 5. - Example distribution or a nuclear power ata~ tion and process production on a plot-plan. 1 - hot helium pipelines, 2 - thermo-conversion units, 3 pipeline bridge, 4 - ammonia production*
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g 6. - Ammonia production flowsheet with utilization ot nuclear reactor heat and cheraothermal transmission of energy to distant consumers, I - high-temperature nuclear reactor, 2 - hot helium pipeline, 3 - methane oonvertor, 4 - steam/gas mixture preheater, 5 - steam generators, 6 - steam superheaters, 7 superheated steam, 8 - drum separator, 9 - turbo-ootnpreasors, 10 - pretreatment unit, II - fired preheaters, 12 - raw natural gas, 13 - gas after raethanators, 14 - preheater of cooling water, 15 - preheater of non-deaerated water, 16 - deaerator, 17 preheater of deaerated water, 18 - boiler feed water, 19 - methanatora, 20 - steam/ air methane oonvertor, 21 - 1st stage CO shift convertor, 22 - 2nd stage GO shift convertor, 23 - ammonia synthesis loop. D-4
10,0 Fig. 7. - Comparative efficiency of servicing the ammonia production in a self-sustained (modular) arrangement of nuclear based technological production and in case of separated location of nuclear baaed conversion plant, 1 - production baaed on organic fuel (0.45 - 0.9) mln tona HHo/year, 2 - modular nuclear energy based technological production (0.45 mln tons HHVyear; Q » 240 MW (f), 3 - modular nuclear energy based technological production (0,9 mln tons HHVyear; Q - 480 MW (f), 4 - separated nuclear energy based technological production with utilization of a nuclear energy based conversion plant without recovery of low-temperature heat, 5 - the same with low-temperature heat recovery (10.45 9.0) mln tons HHyyear; Q - 345-276OMW (f).
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t>-
I fc-
Fig» 8» — Flowsheet of nuclear based conversion plant and T-Q diagram of recovering heat from converted gaseous mixture, 1 - nuclear reactor, 2 - hot helium pipeline, 3 - methane convertor, 4 - steam/gas mixture preheater, 5 - steam generator, 6 - steam superheater, 7 - steam to turbo-compressor, 8 - drum separator, 9 - turbo-compressor, 10 - gas pretreatment, 11 - fired preheater, 12 - make-up gas, 13 - gas after methanators, 14 - cooling water preheater, 15 - preheater of non-deaerated water, 16 - deaerator, 17 - preheater of deaerated water, 18 - water, 19 - "the curve of cooling converted gas.
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4-3-Z-/ L_ 5 8 -7 - 4 - 5-6 /3
J
I
1
i
9
\
5-6-5-4-7-8
T
/s -*»•
•
—
*
-
5
8
T
5
\
6
5
T
Fig. 9. - Centraliued system for supplying «n«rgy to the industrial complexes from an interindustrial nuolear energy baaed conversion plant. 1 - air separation plant, 2 - steam/air methane convertor, 3 - atearn/gas mixture preheater, 4 - steam generator, 5 - methanator, 6 - steam superheater, 7 - gaseous mixture preheater, 8 - water preheater, 9 - gas pretreatment area, 10 - fired preheater of process gas, 11 - gas heating furnaces, 12 - nuolear energy based conversion plant, 13 - methanol production, 14 - production of synthetic natural gas (hydrogenation), 15 - hydrogen production, 16 - ammonia production, 17 - sponge iron production, 18 oil refinery. D-4
IHIIIM XA0101502
ASSESSMENT OF THE LICENSING ASPECTS OF HTGR IN YUGOSLAVIA
Z.Varaidinec INSTITUT ZA ELEKTROPRIVREDU-ZAGREB Zagreb, Socialist Federative Republic of Yugoslavia
INTERNATIONAL ATOMIC ENERGY AGENCY Technical Committee Meeting on Gas-Cooled Reactor Technology, Safety and Siting Dimitrovgrad, USSR, 21. - 23. June 1989. D-5
Abstract
This paper deals not only with the licensing procedure in Yugoslavia, but also reflects the Utility/Owner approach to the assessment of the licensability of the HTGR during the site selection process and especially during bid evaluation process. Besides the description of the existing procedure which was implemented on licensing of LWR program, the assessment of some licensing aspects of HTGR has been presented to describe possible implementation on licensing procedure.
Contents 1. Introduction and background 2. Nuclear licensing procedure in Yugoslavia 3. Applicable Codes and Guides for HTGR 4. Licensing during bid evaluation process from Utility/Owner point of view 5. Licensing aspects of HTGR 6. Conclusion 7. References
D-5
1. Introduction and background The "Institut za elektroprivredu" (IE) was founded in 1953. as an organization for research and development in electric power industry and electric power generation and supply. The IE carries out research related to energy problems and works on practical technical problems of electrical power plants and their equipment. IE covers different areas of planning, construction and operation of power plants and energy distribution network. The activities of IE in energy field are mainly oriented towards planning of the development of the complex energy system of Yugoslavia as a whole, republics and regions. For the performance of above mentioned activities the IE avails with improved modern methods for: - energy supply grid analysis and demand forecast, - determination of optimum energy structure, modeling of energy and cash flows. In the field of forecasting of electric energy consumption IE has mainly performed projections of energy needs for the Electric Utility of Croatia (ZEOH) and for the electric energy system of Yugoslavia. Since the very beginning IE has been engaged in the resolution of problems connected with energy supply and in particular electric energy supply of the country. In the field of planning, construction and operation of the thermal power plants IE is particularly active in the following fields: energy supply in urban areas, - planning, design and construction of thermal power plants in the electric energy system, - site selection & environmental impact studies and investigations of conventional and nuclear power plants. IE has been also engaged in the studies for optimization of production capacities and transmission networks. This has resulted in improvement of existing and development of new methods for research and gained experiences qualify the IE to perform the Nuclear power plant construction planning. In the area of environmental studies IE is actively engaged in coordination of environmental studies for all planned NPP in SR of Croatia. Several programs supported by IAEA are carried out by IE related to the risk-benefit analysis and comparison between conventional and nuclear power plants. 2. Nuclear licensing procedure in Yugoslavia According to the current practice in licensing process it can be assumed that the procedure for obtaining the license involves the the following : 1. 2. 3. 4. as
Including the NPP in general urban plan Site permit Construction permit Operating License it is schematically shown in fig.l. D-5
The early phase of licensing procedure is inclusive of the nuclear power plant site in The General Regional Plan of the State. This is the responsibility of the State Commission of the Civil construction ,Urban planning and Environmental protection as a State Licensing Authority. The Applicant for the nuclear license applies to the State Licensing Authority of the state where the plant will be located. After reviewing the documentation of the all site investigation, in standardized format, according to the requirements of Regulations, Preliminary Site Report is prepared and submitted with requirement for including the facility into The General Regional Plan. This phase consists of discussion with the local authorities which includes public hearing and possible additional local requirements to the Applicant. After reviewing, the documentation is submitted to the State Parliament, for the approval to include the facility and its site into the General Regional Plan. The process of issuing the site permit is performed on specific Federal regulations which includes study work relevant for site verification from the nuclear safety aspects. Format of proposed documentation is standardized, mainly on U.S. NRC RG 1.70 and IAEA recommendations. During the licensing procedure The State Licensing Authority uses outside technical experts and technical experts agencies for reviewing and advising on reactor safety, generic safety issues and radiological protection. This corresponds, for example, to U.S. Advisory Committee on Reactor Safeguards. The procedure for issuing the construction permit is under responsibility of State Commission of the Civil construction, Urban planning and Environmental protection, and issuing of the operational license is under responsibility of State Commission for Industry, Energy and Mining, which also provides supervision of the plants in operation. However, the Federal Government has the responsibility and authority under Nuclear Law to ensure state compliance with the law and to coordinate the activities of the other federal states to obtain consistency. 2.1 Preparatory phase of nuclear power plant planning performed by Utility/Owner The preparatory phase of the Nuclear power construction planning consists of the following activities, which are based on practice of Utility in Socialist Republic of Croatia , and mainly carried out by IE: A - Planning of Nuclear Power Plant Construction - optimization of the long-term energy strategy with forecast of energy consumption and determination of optimal energy sources and source distribution study for any region or a complex urban structure -The Feasibility Study Report for implementation of nuclear power plants in electric generation system of any given region, choice D-5
of unit size, technical aspects analysis, nuclear cost estimates, generation costs, total costs estimates and comparison of alternative sources, financial review, project development, domestic participation -optimization and development of complex centralized district heating systems using nuclear or conventional power sources B - Nuclear Power Plant Site Selection and Environmental Studies One of the earliest stages of a NPP construction planning is site selection study with respect to: - characteristics of the site influencing a NPP design i.e.: seismic characteristics, extreme meteorological characteristics, flooding probability, external man induced effects etc. -characteristics determining the impact of a nuclear power plant on its environment i.e.: population density distribution, atmospheric characteristics etc. (This study work results with Environmental Report of standardized format according to Yugoslav regulations, mainly based on U.S. documents, USNRC Reg.guide 4.2, USNRC NUREG-0555 and IAEA recommendation.) -other characteristics of site having impact on design of a nuclear power plant, its economic basis or effects on the environment. Site selection for nuclear power plants is performed phases:
in three
1-NPP oriented Site Survey for any region. Based on the defined elimination criteria with included safety, economical and ecological aspects a number of potential microlocations is determined. Definition and collection of early site data. 2-Site Qualification process with evaluation of several potential microlocations. Establishment of standard report form for geological, seismological, hydrological, meteorological and population data ,to be used throughout the site approval process. 3-Site Confirmation process with collection of all necessary data to obtain a site permit for preparing the construction of a nuclear power plant. Detailed field investigation and analysis to assess all of the site characteristics and their influence on the design basis for a NPP. Determination of detailed environmental impact in the phase of site selection. C - Optimization and Selection of Technical Characateristics and Parameters for NPP Equipment and Fuel Cycle This phase is performed before preparation of Bid invitation documentation in order to assess the technical characteristics and parameters of NPP components of different type. During this phase particular attention is given to various types of nuclear power plants in order to assess their differences in technical characteristics and parameters, which will be used for preparation of Bid Invitation Specification. D-5
D - Preparation of bid invitation documentation and bid evaluation This phase includes preparation of Bid Invitation Documentation for construction of a NPP with various contracting combinations such as turn key or split package. Specific requirements of the site , requirements of the respective Regulatory body, as well as the requirements and specific conditions of the electric Utility/owner are incorporated in Bid Invitation Documentation. There is a great interest of domestic industry for participation with a foreign partner in NPP construction and production of components, and according to the prior experience on NPP Krs"ko and up to date studies of potential capabilities of domestic industry it can be approx. 70 % (estimated for PWRs NPP).
2.2 Applicable Codes and Guides for HTGR All domestic regulations that apply to that type are valid. It is also expected that the regulations of the Supplier's country will be partly adopted, considering the requirement that offered design must be licensed in Supplier's country. Also valid would be IAEA , Design for safety of Nuclear Power Plants, A Code of Practice, Vienna, 1978., which can be applied to HTGRs without greater difficulties. It may be pointed out that Yugoslav nuclear regulations are based on IAEA-ICRP recommendations and U.S. regulations with regard to NPP KrSko, except that they are particularly somewhat stricter. This particularly applies to 10CFR20. The design criteria and the technical codes must be supplemented with regard to the requirements of HTGRs. Efforts in this direction and experience in the application of the relevant codes to HTGRs , which are mostly prepared for Light Water Reactors,have to be adopted mainly from the countries which have already licensed HTGRs. It can be concluded that the design criteria and technical codes can mainly be applied on HTGR.
2.3 Licensing documentation According to the Yugoslav Atomic Law, licensing comprises following main steps - documents : - Site selection and Environmental studies - Site permit - Construction of the NPP - Construction Permit - Commissioning and Operating of NPP - Operating License - Decommissioning - Decommissioning License Documentation is prepared to serve as an overview of all site investigations performed in accordance with Standard format for Environmental Impact Study.
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The volume of investigations is based on regulations taken . over from the recommendations of IAEA and mostly U.S. ( 10CFR100 and Reg.guide 1.4). This is the case mainly due to fact that our first NPP is PWR - 632 MW ( Westinghouse design). One of the basic assumptions for a quality licensing process is establishing a good communication between the Utility and Regulatory body. This can be ensured through careful preparation of materials needed in the licensing procedure in standard format. The basis for determining the format was U.S. regulation as follows: -USNRC Reg.Guide 1.70:Standard format content of safety analysis report for NPPs, rev.3,1978. ,- USNRC Reg.Guide 4.2:Preparation of environmental reports for NPPs,Rev.2, 1976. ,USNRC,NUREG75/087:Standard review plan for the review of safety analysis reports for NPPs, LWR, edition 1975. , - USNRC,NUREG0555:Environmental standard review plans for the environmental review of construction permit applications for NPPs, 1979. Also in use IAEA No.50-SG-G2: Information to be submitted in support of licensing applications for nuclear power plants,1979., which is suitable even it is not very precisely elaborated. On the basis of the aforementioned documents, the Standard Format for the Environmental Impact Study (Rev.l, 1985.) was proposed by IE and approved by the State Licensing Authority. In the case of NPP Kr§ko, U.S. regulations were used throughout and the documentation was elaborated and reviewed according to current standard format available in U.S.A. at that time, which resulted in submittal of Preliminary Safety Report and Final Safety Analysis Report. One of the main differences from U.S. licensing procedure is that we have requirement for issuing the Site permit which as a formal document does not exist in U.S. 3. Licensing during bid evaluation process from Utility/Owner point of view
Starting off with the fact that the licensing is the responsibility of the Utility/Owner, in the BID evaluation stage the special criterion is the assessment of licensability of the offered designs. In this, the basic assumptions are that the plant must be Iicensable in the country of the reference plant, and that the plant must be licensed in Yugoslavia. Assessment of the licensability for the nuclear power plant from the BID consists of evaluating the following licensing aspects: - licensing assistance as bidder s contribution to prepare licensing documents - licensability - codes,standards & regulations - inputs to PSAR,FSAR,Environmental Report - accident analysis - unresolved safety issues - The Reference plant D-5 6
During the bid evaluation phase of the licensing aspects it is particularly important to devote attention, analyze and assess the Supplier's approach related to: -licensing documents (PSAR,FSAR) -analysis of the safety report contents -Bidder responsibilities for the development of safety reports -evaluation of the type of safety analysis and studies to be furnished by the bidder for the PSAR and FSAR (safety classification of structures,systems and components in relation to their safety function,safety concept application,plant operating conditions ;normal-abnormal-accident, safeguard system design capabilities, deviations from the Reference Plant that can affect plant safety, shielding design,ALARA design) -Number of revisions for safety report documentation included in the scope -Supplier's responsibilities in preparing answers to questionnaires from licensing Authorities -Planning and scheduling proposed by Supplier to develop his licensing scope of services Approach with the Reference plant for licensing has advantages , for example the offered design refers to the specific plant as a pattern for defining safety criteria, but on the other hand with such approach there is a problem because of interface problems with different sites characteristics. In this light the Reference design as a reference concept can offer: envelop of latest design of same generation, take advantages of all modern designs and is more flexible. 4. Licensing aspects of HTGR The basic requirement that is applied in our country is that the offered design of the power plant is not contrary to the IAEA regulations or Supplier's country regulations and that it is licensable in the Supplier's country. In the BID document the principle of the reference power plant would naturally be used. In accordance with the established procedure in evaluation of potential sites for a NPP , where the evaluation was done for PWR, BWR and CANDU, HTGR would have the same treatment naturally taking into account the corresponding source term. The effect and evaluation of the designed layout from the aspect of the seismic effects is one of the decisive factors in assessment of safety characteristics of a NPP, which could have a significant impact in the licensing process involving HTGR. Furthermore , the collected data from HTGR power operation would be used to assess the probability of occurrence according to models of probabilistic risk This ,however, requires a great deal of data concerning of commercial plants of HTGRs.
plants in accident analysis. operation
D-5
Finally, the need for a complex safety analysis seems obligatory by means of which a detailed comparative analysis of advantages of HTGR type of power plant in relation to other alternative commercial power plants could be carried out. As an example , the permitted radioactive releases to the environment defined by our regulations are considerably stricter than values prescribed in 10CFR20 and 10CFR50,App.I, and lower than IAEA ICRP Pub.30, making the data on releases stated in Ref.5 regarding MHTGR very interesting. High degree of passive safety features and exclusion of any evacuation requirements for HTGR power plants, increases the number of potential sites and influences the utilization of existing ones, especially because of the standardization of reactor unit and conventional BOP with no requirements regarding safety related tasks. If this is applied to MHTGR, assuming an acceptable approach of the Supplier judged according to other criteria, the conclusion is, that high level standardized design of NSSS offers many advantages from the aspect of licensing in relation to LWR.
5. Conclusion HTGR represents a very interesting option particularly for developing countries and relatively developed industrialized countries, especially if it is considered that the status of already achieved quality standard , safety standard and degree of proveness of LWRs is the minimum requirement and scale for the HTGR layout. The basic advantage of this second generation nuclear power system is remarkable for its safety characteristics which do not depend on active engineered safety features or human actions for safety of the public or the investor. It can freely said that a high degree of licensability of HTGR exists in Yugoslavia. Here , it has also to be mentioned that by the Federal Regulation the construction of the NPPs in Yugoslavia is suspended for the time being. However this law does not extend to research and progress in nuclear energy by the still interested electric utilities. From the aspect of licensing it is believed that in case HTGR type is selected, promoted by the correct support of the Supplier, HTGR would successfully pass the licensing procedure. In any case this would cause certain moves to be made regarding the existing regulations. The design criteria complying with the valid regulations regarding PWR should definitely be supplemented with regard to the requirements of criteria for HTGRs especially on HTGR design related to pre- and inservice inspections,the criteria for shutdown systems ,residual heat removal systems, reactor coolant boundary and reactor containment.
8
D 5
"
8. References
1. Gas cooled reactor Associates, Modular Project Definition Study, GCRA 86-010
HTGR
Demonstration
2. Gas cooled reactor Associates,Utility user Requirements Document for the MHTGR Demonstration Project, GCRA 86-002. 3. A.J.Neylan,The Modular High-Temperature Gas-Cooled Reactor (MHTGR) GA Technologies Inc., Presentation at TCM IAEA, oct.1986.held in Juelich 4. N.Malba§a,D.Subasic, Usporedba regulative U.S. i evropskih zemalja iz podru£ja lociranja nuklearnih elektrana, Institut za elektroprivredu-Zagreb,1984. 5. A.J.Neylan,D.A.Dilling,J.M.Cardito, Environmental Aspects of MHTGR Operation, GA-A19439,General Atomics,San Diego,CA Sept.1988. 6. K.Hofmann, Comparison of regulatory aspects in countries, Rheinisch-Westfaelischer TUV, Essen, 1986.
different
D-5 9
Ftdiril Hiniitry of Enirgy, Industry md Mining
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XA0101503 SUBSTANTIATION OF CHOICE OF THE MAIN PHYSICAL AND THERMOHYDRAULIC PARAMETERS OF REACTOR PLANT SMALL POfcER HTGR.
INITH
AFANASYEV V.N., GOLOVKO V.F., ZHIGULSKY V.N., KARPOV V.A., KIRYUSHIN A.I., KUZAVKOV N.G., SUKHAREV YU.N. Abstract
AT PRESENT THE HIGH TEMPERATURE HELIUM COOLED ARE THE ONLY TYPE
OF
REACTORS
ALLOWING
TO
REACTORS
INCREASE
THE
COOLANT TEMPERATURES AT CORE OUTLET UP TO 950 C. THE PERSPECTIVENESS POSSIBILITY OF
OF
PROVIDING
THESE
THE
REACTORS
ECONOMY
OF
LIES
IN
NATURAL
THE
ENERGY
SOURCES BY THEIR USE IN POWER-INTENSIVE CHEMICAL-ENGINEERING PROCESSES SUCH
AS,
FOR
EXAMPLE,
COAL
GASIFICATION,
METALLURGY, PRODUCTION OF MINERAL FERTILIZERS, HYDROGEN ETC. CONCURRENTLY
THE
REACTORS
ADVANTAGES OF BEING ABLE
TO
OF
ENSURE
SPECIFIED BY THE CORE AND COOLANT
THIS
TYPE
THE
HAVE
INCREASED
INHERENT
THE SAFETY
PROPERTIES.
A LARGE MEASURE THESE ADVANTAGES ARE REALIZED IN
THE
POWER PLANTS THE HIGH
ALLOWS
SAFETY
DEGREE
UF
WHICH
IN
SMALL TO
DISPOSE THEM IMMEDIATELY ON SITE OF THE INDUSTRIAL PLANTS OR IN THE VICINITY OF POPULATED AREAS.
1. STATEMENTS BEING THE BASIS OF DEVELOPMENT OF SMALL POWER REACTOR PLANT (RP)
THE DESIGNING IN THE USSR OF
SMALL
HELIUM COOLED REACTOR WAS PRECEEDED BY
POWER-PLANT THE
WITH
DEVELOPMENT
0-6
OF
UP.
Br-mt>
THE
RP
SHOULD
PROVIDE
THE
CHEMICAL-ENGINEERING PROCESSES WITH HIGH-POTENTIAL HEAT
(UP
TO V0w C) AND
FOR
ALSO
PRODUCE
HIGH
PARAMETERS
HEAT
GENERATING THE ELECTRIC POWER WITH HIGH EFFICIENCY. THE PLANT IS DESIGNED FOR SOLVING THE MAIN RESEARCH AND ENGINEERING PROBLEMS OF CREATING THE HTGR PLANTS AT STEP SIEP DEVELOPMENT OF ELEMENTS AND THE HIGH PARAMETERS
STEAM
EQUIPMENT
FOR
STUDYING THE SPECIFIC QUESTIONS OF HTGR REACTORS SAFETY
AND
ALSO FOR DEMONSTRATION
AND
OF
HIGH
GENERATING HEAT,
ADVANTAGES ASSOCIATED
AND
FOR
BY
POTENTIAL
TECHNICAL
WITH
THE
CHEATING ON ITS BASIS THE
FEASIBILITY
INCREASED
COMMERCIAL
SAFETY/
PLANTS
OF
FOR
DIFFERENT
PUWER-ENG1NEEKING PURPOSES. ACCOUNTING FOR THE SHORT ASSIGNED PERIOD OF RP CREATION AND IDE AVAILABLE WORK IN
PROCESS
Bf-40«i HP, THE USE OF THE
MAIN
ON
THE
DEVELOPMENT
COMPONENTS,
EQUIPMENT
OF AND
DtSlbU DECISIONS, REALIZED IN THE Br-400 DESIGN, IS PLANNED. IT WILL ENABLE ALSO TO EQUIPMENT PRODUCTION
DEVELOP AND
TO
THE CHECK
CYCLE UP
OF
THE
THE MAIN
Br-100 DESIGN
DECISIONS AT SMALL POWER RP OPERATION. (-1OREOVER,
THE
SUBSTANTIATION
AND
CHOICE
OF
BLOCK
DIAGRAM, CHARACTERISTICS AND PARAMETERS OF THE MAIN
PRIMARY
CIRCUIT ELEMENTS AND ALSO THE RP SAFETY WERE
ON
BASED
THE
FOLLOWING PRINCIPLES: 1. ABSENCE OF CPS RODS INSERTED INTO THE PEBBLE BED. d. MAXIMUM FUEL ELEMENTS TEMPERATURE IN
THE
EMERGENCY
SITUATION NUT MOKE fHAN 1600 C. 3. VESSEL'S CONDITIONS SHOULD
TEMPERATURE NOT
EXCEED
IN 300
THE S,
NORMAL AND
IN
OPERATION EMERGENCY D-6
CONDIriUNS - 400 C WHICH
ENABLES
TO
USE
THE
STEELS
PRODUCTION TECHNOLOGY DEVELOPED BY HOME INDUSTRY. H. GRADUAL POWER AND HELIUM TEMPERATURES INCREASE
FROM
7b0 C UP TO 950 C IN THE REACTOR: AT THE FIRST STAGE - POWER BEING UP TO 200 Mln, TEMPERATURE - UP TO 750 C, HEAT IN STEAM GENERATOR; AT THE SECOND STAGE -
POWER
REMOVAL
EQUALS
200 MW, TEMPERATURE FROM 750 C UP TO 950 C.
AT
A HIGH - TEMPERATURE
EXCHANGER
INTERMEDIATE
INSIALLED BEFORE STEAM GENERATOR. AT
HEAT THE
FIRST
STANDARD PLANTS PROTOTYPES OF NUCLEAR HEAT AND AND AT THE SECONO STAGE - THE PROTOTYPES
THIS
OF
TO
STAGE IS
STAGE
THE
POWER
PLANT
NUCLEAR
POWER
PROCESS PLANT WERE DEVELOPED. b. PRESSURE IN THE PRIMARY CIRCUIT WAS TAKEN
EQUAL
a.9 MPA WITH ACCOUNT OF THE WORK IN PROCESS ON RESEARCH
TO AND
DEVELOPMENT.
2. CHOICE AND SUBSTANTIATION OF CORE POWER AND GEOMETRY
PROCEEDING FROM THE NECESSITY TO PROVIDE
THE
TEMPERATURE IN NORMAL OPERATION CONDITIONS NOT
VESSEL'S
HIGHER
THAN
300 C AND DUE TO USE OF THE MAIN Br-400 PLANT EQUIPMENT, THE FOLLOWING PARAMETERS OF PRIMARY HELIUM FOR THE
SMALL
POWER
PLANT WERE TAKEN: T1
- 300 C; T 2
= 750 - 950 C, PRJESSURE - 4.9 MPA
THE REQUIREMENTS TO
EMERGENCY
PROTECTION
AND REACTIVITY COMPENSATION SYSTEM, IMPOSED
BY
RELIABILITY REGULATIONS
FUR THE PLANT WITH INCREASED SAFETY, NECESSITATE TO USE
THE
REACTIVITY COMPENSATION MEMBERS, THE ACTUATION OF WHICH D0E3
5D0
NO I RtQUIRE ENERGY ABSORBER RODS
SOURCES.
AND
THIS
SPHERICAL
C0NDI1I0N3 SYSTEMS
THF
OF
USE
OF
REACTIVITY
COMPENSATION (SSRC), LOCATED IN THE SIDE REFLECTOR
CHANNELS
AND DROPPING UNOER GRAVITY IN EMERGENCY SITUATIONS. THE STUDIES OF CPS MEMBERS EFFICIENCY IN SIDE REFLECTOR SMOKED, THAT THE
REACTOR
ACCOUNT OF CALCULATION REQUIREMENTS MAY BE
MAXIMUM
UNCERTAINTIES
COMPENSATED
REFLECTOR ON CONDITION THAT EXCEED 3 M. AS DESIGN
REACTIVITY
BY
THE
AND CPS
CORE
STUDYINGS
MARGIN
REGULATIONS
MEMBERS
DIAMETER
SHOW,
24
WITH
CPS
IN
SIDE
WOULD
NOT
RODS
WITH
ABSORBEk DIAMETER 110 MM AND 22 SSRC CHANNELS WITH 26» X MM SECTION, FILLED WITH 6-10 MM
DIAMETER
ABSORBER
60
SPHERES
WITH NATURAL BORON CARBIDE CAN BE LOCATED IN SIDE REFLECTOR. THE DEPENDENCE OF CPS
MEMBERS
EFFICIENCY
UN
DIAMETER
OF
THEIR LOCATION IS SHOWN IN FIG.l. AT RELATIVELY SMALL CORE DIAMETER - 3 M IT IS
POSSIBLE
TO COOL DOWN THE CORE OR MAINTAIN IT AT REQUIRED TEMPERATURE LEVEL (AFTER BRINGING THE REACTOR
INTO
SUBCRITICAL
STATE)
WITHOUT ENGINEERED COOLING DOWN NEANS EVEN AT COOLANT IN IHI3 CASE THE
RESIDUAL
REMOVED TO REFLECTORS AND
HEAT
FROM
FURTHER
THE
PEBBLE
THROUGH
THE
LOSS. BED
IS
VESSEL
TO
REACTOR WELL COOLING SYSTEM. THE ENSURANCE OF RESIDUAL HEAT
KEMOVAL
AT
POSTULATED
FAILURE OF ENGINEERED HEAT REMOVAL SYSTEMS IS
SPECIFIED
THE REQUIREMENTS UF USSR CODES
THE
AT
DESIGNING
RP
WITH
INCREASED SAFETY. THE POSTULATED FAILURE OF ENGINEERED
HEAT
REMOVAL SYSTEMS PRESUMES THE HOT HELIUM LOCALIZATION THE GRAPHITE CORE FOR MINIMIZING
THE
POSSIBILITY
TRANSFER TO RP LOAD - CARRYING METALWORKS.
50A
BY
THIS
WITHIN OF
IS
HtAT MOSTLY D-6
FAVOURED BY MODULAR
RP
ARRANGEMENT
EXCLUDING
NATURAL
CIRCULATION DEVELOPMENT OVER THE PRIMARY CIRCUIT PATH AND BY PLACING THE CORE INTO
THE
CLOSED
SUCH DESIGN DECISIONS
PROMOTE
THE
BARREL
VOLUME.MOREOVER,
DECREASE
OF
PRODUCTS RELEASE OUTSIDE THE PRIMARY CIRCUIT AND AIR INGRESS AT
VESSELS'
FISSION LIMIT
THE
AS
THE
LOSS-OF-TIGHTNESS.
PRELIMINARY CALCULATIONS SHOWED, FOR SUCH AN ARRANGEMENT ThE NECESSARY CONDITION OF NON-EXCEEDING THE
MAXIMUM
FUEL fEMPERATURE 160* C, WHEN THE
APPRECIABLE
FISSION PRODUCTS RELEASE FROM THE
FUEL
ALLOWABLE
INCREASE
ELEMENTS
DOES
OF NOT
OCCUR, IS THE RESTRICT ION OF MEDIUM LINEAR LOADING ALONG THE CORE HEIGHT EQUAL TO - ?A MW/M. AT CORE
DIAMETER
EQUAL
TO
3 M THE SPECIFIC MEDIUM-VOLUME POWER DENSITY WILL ACCOUNT TO - 3.0 MW/M CORRESPONDINGLY. THE RP HEIGHT IS CIRCULATUR (MGC),
DEFINED
STEAM
UNAMBIGUOUSLY
GENERATOR
HIGH-IEMPERATURE INTERMEDIATE
HEAT
(SG),
OPtKATlUU CONDITIONS
OF
CHARACTERIZED BY REDUCED
AND MGC
CH03LN AS
FLOWRATES
AND
A
PART
REGIME
WITH
THE
AND KNOWN
DIAMETER. OF ARE
VALUES' RANGE EXCLUDING THE MGC SURGING. THIS CHARACTERISTIC OF RP OPERATION
BT
(HTIHX)
WITH
CORE
WHICH
MAIN RP
EXCHANGER
ALSO BY THE ADOPTED SCHEME OF BLOCK LAYOUT HYDRAULIC CHARACTERISTICS
BY
RP LIMITED
IS
ARh BY
SPECIALLY AND
OF
HELIUM TEMPERATURE INCREASE AT THE CORE OUTLET UP TO 950
C.
SURGING AREA BOUNDARY (SEE FIG.2) WOULD
AS
BE
HTIHX
THE
REPRESENTED
SQUARE DEPENDENCE:
-7 A1 Ti+2.75
0
D-6
WHERE AP,
, P - GAS CIRCULATOR HEAD AND CIRCUIT PRESSURE, MPA;
G
- HELIUM FLOrtRATE, KG/S;
If
- HELIUM TEMPERATURE AT CORE INLET. THE ANALAGOUS SGUARE DEPENDENCES IN THE FLOWRATE
FOR RP NOMINAL
OPERATING
CONDITIONS
RANGE
CHARACTERIZE
THE
HYDRAULIC RESISTANCE OF PRIMARY CIRCUIT MAIN COMPONENTS: THE CORE 3 M ACROSS
WHERE
H
- CORE HEIGHT, M;
PRIMARY CIKCUI1
PATH
WITH
8G,
HTIHX,
AND
CORE
SEPARATE VESSELS
..„„ ,.
P3
STEAM GENERATOR
WHERE
T5
= 75«> C
- HELIUM TEMPERATURE AT SG INLET;
HIGH TEMPERATURE INTERMEDIATE HEAT EXCHANGER
.
.
.
.
5
,
5
•
.
.
"
'
ACCOUNTING THAT
* 750 C
SOS
D-6
IN
IT
FOLLOWS FROM THE PRESENTED DEPENDENCES
H
-
THE OBTAINED RATIO SHOWS THAT THE OPTIMIZED CORE HEIGHT DOES NOT DEPEND ON CIRCUIT PRESSURE.
IT
FOLLOWS
FROM
PLOIS OBTAINED FROM THIS RATIO AND PRESENTED IN FIG.3
THE THAT,
RP OPERATING REGIME WITH HELIUM TEMPERATURE, AT THE OUTLET OF 950 C IS THE LIMITING ONE WHEN DETERMINING THE CORE THEREBY, THE CORE
HEIGHT
AT
THE
CHOSEN
HEIGHT.
HELIUM
INLET
TEMPERATURE OF 300 C WOULD BE NOT MORE THAN 9.5 M. TAKING INTO ACCOUNT THE DEFINITE CORE ALLOWABLE AVERAGE LINEAR LOAD $ 21
HEIGHT
AND
MAXIMUM
MW/M,THE
THE POWER
THAT CAN BE REACHED AT A SMALL POWER RP AMOUNTS TO 200 MW, BUT, SUCH
POWER
IS
REACHED
ALSO
DEPENDING
POSSIBILITY OF REQUIRED HEAD PROVISION, THE UPPER WHICH IS REALIZED
AT
MAXIMUM
(6000 RPM) AND MAXIMUM DRIVE
GAS POWER
CIRCULATOR (5.25MW).
ON
THE
LIMIT
OF
REVOLUTIONS TO
A
GREAT
EXTLNT fHESE LIMITS WOULD DISPLAY DURING THE PLANT OPERATION AF FIRST STAGE WITH HELIUM TEMPERATURE OF 750 C AT THE
CURE
OUTLET AND WITH HTIHX IN PRIMARY CIRCUIT. HYDRAULIC RESISTANCE OF
PRIMARY
CIRCUIT
CURE SIZES (D=3M, H=9-10M) FOR TfcO VARIANTS
OF
WITH
CHOSEN
COMPONENTS'
LAYUUT (WITH HTIHX OR WITHOUT IT) IS EXPRESSED AS A FUNCTION OF HELIUM FLOWRATE. 3.
f D-6
8
WHERE B = (0.475 Jo
+ 0.741 T.
+ 0.209 T,
+ 389)
+ 0.741 J^
+ 0.129 1^
t 345)
WITH HTIHX B = (0.395 T £
WITHOUT HTIHX. THEN, ACCOUNTING THAT N^
= G C , (^
n
- 121.5 Q,
- 101 \[b.072 U* + ^ H
- T,('
),
THE DEPENDENCE fa AND N^ ON POWER ONE COULD PRESENT AS FOLLOWS:
~ <1 WHERE T o
=
T4
+ 273, (Q - VOLUMETRIC FLOWRATE).
s M3.To- 8 N^ = 2.078 10 --3 — — o -
WHEKE |^ = 0.61 (TENTATIVELY, IN THE RANGE OF FLOWRATES INVESTIGATED IS ASSUMED TO BE CONSTANT). DEPENDENCES' PLOTS BASED ON
THE
EXPRESSIONS
DURING THE VARYING OF CIRCUIT PRESSURE FROM 3.9 TO
OBTAINED 5.9
MPA
ARE PRESENTED IN FIGS 4 AND 5. THE FIGURES SHOW THAT THE GAS CIRCULA1OR PROVIDES THE NECESSARY FLOWRATES IN THE RANGE
OF
VARIABLE REVOLUTIONS (300....6000 KPM) AT RP POWER OF 200 MW DURING ALL OPERATING STAGES. BUT, GAS CIRCULATOR MOTOR POWER IS DEFICIENT DURING THE OPERATION AT THE
FIRST
STAGE
WITH
HELIUM TEMPERATURE
OUTLET
AND
WITH
OF
750
C
AT
CORE
INSTALLED HTIHX. IN THIS CONNECTION THE RP OPERATION IN THIS
SOS"
REGIME
D-6
IS
POSSIBLE EITHER WITHOUT HTIHX OR WITH HTTHX BUT WITH CIRCUIT PRESSURE INCREASING UP TO 5.2 MPA. SO, TAKING
INTO
ACCOUNT
THE
. RP
PURPOSES,
MAIN
STATEMENTS ON SAFETY AND MAXIMUM USE OF Bf-q00 RP COMPONENTS IN A SMALL POWER PLANT THE MAXIMUM POWER SHOULD BE TAKEN
NO
MORE THAH 200 MW WITH CORE DIAMETER OF 3
OF
M
AND
HEIGHT
9-10 M,
3. FUEL LOADING PARAMETERS SUBSTANTIATION
THE MOST IMPORTANT CORE CHARACTERISTICS
SUCH
AS
FUEL
COST COMPONENT, MAXIMUM FUEL TEMPERATURE, MAXIMUM REACTIVITY MARGIN, DETERMINING
THE
REQUIRED
MEMBERS WERE TAKEN
INTO
ACCOUNT
EFFECTIVENESS FOR
OF
DETERMINING
CPS
OF
THE
OPTIMUM FUEL LOADING. NUCLEAR RATIO N C /N O (FUEL LOADING FUEL ELEMENTS), FUEL ENRICHMENT IN U-235 AND KERNELS DIAMETER
WERE
THE
VARIED
FUEL
PARAMETERS
ACCOUNT WHEN CHOOSING THE FUEL LOADING. WHEN
INTO
PARTIES
TAKEN
INTO
ANALYZING
THE
FUEL LOADING THE ENRICHMENT WAS VARIED BETWEEN 6.5% AND 17%, URANIUM LOADING PER ONE FUEL ELEMENT - BETWEEN 6.15 G AND 12 G, KERNEL DIAMETER - BETWEEN 200 MC.AND LOADING LOWER THAN 6.15
G
800
(CORRESPONDING
ELEMENTS LOADING) WAS NOT CONSIDERED AS
MC. TO
IN
FUEL BURNUP SUBSTANTIALLY DROPS AND FUEL
THE
FUEL
Bf-400
FUEL
SUCH
ELEMENTS
CASE
THE
SPECIFIC
CONSUMPTION RISES. TWO CIRCULATION SCHEMES - OTTO AND MULTIPASS TO 15 PASSES) BASED INVESTIGATED.
CPS
ON
TABPOUJ
MEMBERS
/I/
PROGRAM
EFFECTIVENESS
(FROM
COMPLEX WAS
10 WERE
ESTIMATED
ACCORDING TO'WI'MS - &H /I/ PROGRAM. D-6
306
10 THE INVESTIGATION OF THE FUEL COST COMPONENT DEPENDENCE
HAVE
WIDE
RANGE
SHOWN, OF
THAT
C T VALUE
NUCLEAR
RATIOS
= 35
THE
FUEL
VARIES
THE
LOADING.
VALUE
DIFFERENT ENRICHMENT. SO WHEN CHANGING FROM 6,5% SPECIFIC
DEMAND
IN
OF
N c /No ALSO
VARIES SLIGHTLY, BUT IT IS DIFFERENT FOR FUEL LOADINGS
IN U-235 TO 10%, THE
/N
DETERMINING
THE
SPECIFIC DEMAND IN NATURAL URANIUM DEPENDING ON
Nc
WITH
ENRICHMENT
NATURAL
URANIUM
DECREASES BY 25%. THE MOST CONSUMPTION IN
SENSITIVE THE
IS
REACTOR
THE
VALUE
GREATELY
OF
FUEL
ELEMENTS
DEPENDING
ON
LOADING PER ONE FUEL ELEMENT AND ON THE ENRICHMENT BE CONSIDERED AS THE MAIN CHARACTERISTIC FOR THE
F^EL
AND
MAY
CHOICE
OF
THE EFFICIENT LOADING A3 IT DETERMINES THE INTENSITY OF F^EL ELEMENTS FABRICATION AND THE REQUIREMETS FOR THE CAPACITIES. FIG.6 SHOWS, THAT THE SPECIFIC CONSUMPTION OF FUEL DECREASES BY A FACTOR OF 1.5
WHEN
LOADING (THAT IS FROM 6.15 G TO 9
CHANGING G
OF
TO
ELEMENTS THE
URANIUM
PER
ELEMENT) AND BY A FACTOR OF 1.8 WHEN THE ENRICHMENT FROM 6.5% TO 10*. THE MINIMAL CONSUMPTION IS
HEAVY F^EL
CHANGES
REALIZED
WHEN
THE LOADING PER ONE FUEL ELEMENT IS (12 - 13)G. SAID CHARACTERISTIC
FOR
OTTO
AND
MULTIPASS
SCHEMES
DIFFERS SLIGHTLY. THE AVERAGE FUEL BURNUP
DEPENDS
ON
U LOADING PER ONE FUEL ELEMENT AND INCREASES BY A FACTOR
OF
1.7
(FIG.7)
1.8 WHEN CHANGING FRON 6.5%
MULTIPLE CIRCULATION CHARACTERISTIC BY 5%
OF
FUEL
SLIGHTLY
TO
ELEMENTS
10%
ENRICHMENT.
INCREASES
THIS
10% IN COMPARISON WITH OTTO CYCLE.
D-6
£0}
11 IF THh VALUES
OF
FUEL
LOADING
CORRESPOND TO THE MINIMUM FUEL
PER
ONE
ELEMENT
FUEL
ELEMENT
CONSUMPTION
= 300, G - 12G), THL AVERAGE BURNUP DECREASES BY (15....2*)% IN COMPARISON WITH ITS MAXIMUM VALUE. BESIDES LARGE LOADINGS OF FUEL PER FUEL ELEMENT, AS IT KILL BE SHOWN BELOW, LEAD TO THE CONSIDERABLE
RISE
OF
MAXIMUM
REACTIVITY
MARGIN
(ESPECIALLY IN OTTO SCHEME), THAT IS WHY THE CHOICE OF
SUCH
FUEL LOADING IS UNPRACTICAL. AS THE INCREASt OF LOADING PER ONE FU.EL ELEMENT AND THE ENRICHMENT RISE PRACTICALLY INDENTICALLY INFLUENCE UPON
THE
SPECIFIC FUEL ELEMENTS CONSUMPTION, BUT THE ENRICHMENT
RISE
FROM 6.5% TO 10X LEADS TO THE SAVE FUEL BURNUP INCREASE
(BY
A FACFOfc OF 1.8) THE LATTER IS THE MOST PREFERABLE. BESIDES, THE USE OF 10% ENRICHMENT IS MORE PREFERABLE
(IN
MULTIPASS
SCHEME) FROM THE POINT OF VIEW OF FUEL TEMPERATURE
DECREASE
(SEE FIG.8) AS WELL. THE USE OF
17%
ENRICHMENT
PROVIDES
THE
DECREASE OF FUEL ELEMENTS CONSUMPTION BUT IT
ADDITIONAL
LEADS
TO
THE
CONSIDERABLE INCREASE OF FUEL TE^PERAIURE AND TO THE RISE OF PUWLK DISTRIBUTION IRREGULARITY. THAT IS WHY Z,- = 17% IS
NOT
CONSIDERED LATER ON. THL INVESTIGATION OF
FUEL
KERNEL
DIAMETER
INFLUENCE
UPON THE CORE CHARACTERISTICS SHOWS, THAT THE FUEL CONSUMPTION POORLY DEPENDS ON
SAID
PARAMETER
ELEMENTS
AND
MAXIMUM
FUEL ELEMENT POWER DENSITY AND FUEL TEMPERATURE CONSIDERABLY INCREASE AT THE KERNELS DIAMETER FEATURE OF THE LARGE
FUEL
RISE.
PARTICLES
THE IS
ONLY
THE
POSITIVE
DECREASE
OF
REACTIVITY EXCURSION AT THE WATER INGRESS INTO THE CORE. THE STUDY OF
THE
FUEL
LOADING
AND
FUEL
ENRICHMENT D-6
£08
INFLUENCE UPON
MAXIMUM
FUEL
REACTIVITY MARGIN
SHOWS,
LOADINGS, (10-12)
G
OF
TEMPERATURE
THAT U
WHEN
PER
AND
CHANGING
ONE
FUEL
TO
IN OTTO SCHEME. MAXIMUM REACTIVITY MARGIN, STEADY
-
OPERATIONAL REACTIVITY
MARGIN
FOR
REACTIVITY EXCURSION
IN
STATE
CASE
FUEL
ESPECIALLY
INCLUDING POISONING
REACTOR OF
HEAVY
ELEMENT,
TEMPERATURE DECREASE BY (2®....25)% (SEE. FIG.8),
TEMPERATURE EFFECT AND
MAXIMUM
TOTAL EFFECT,
CONTROL
STEAM
AND
GENERATOR
LOSS-UF-INTEGR'ITY, CONSIDERABLY RISES WHEN CHANGING TO HEAVY FUEL LOADINGS (SEE FIG.9). FIG.10 SHOWS THE DEPENDENCE
OF
THE
EFFECTIVENESS
OF
CONTROL MEMBERS IN SIDE REFLECTOR FOR THE CORES WITH VARIOUS FUEL COMPOSITION. WHEN CHANGING TO HEAVY FUEL
LOADINGS
THE
RODS EFFECTIVENESS INCREASES BUT NOT SO APPRECIABLY
AS
MAXIMUM REACTIVITY MARGIN RISES. THE
LOADING
USE
OF
FUEL
THE
WITH 1«% ENRICHMENT AND MULTIPLE CIRCULATION OF FUEL THROUGH THE CORT APPEARS TO BE MORE FAVOURABLE. TAKING INTO ACCOUNT THE RESULTS OF THE PARAMETRIC ST^DY AS WELL AS THE DISCREPANCY OF THE REQUIREMENTS DETERMINED BY THE PROBLEMS OF ECONOMIC CHARACTERISTICS OPTIMIZATION AND BY THE PROBLEMS OF SAFE AND RELIABLE REACTOR EXPEDIENT TO CHOOSE
FUEL
LOADING
CHARACTERIZED
- 500:6E>0 (G-U - 6 - 7G/ FUEL ELEMENT) : 10% ENRICHMENT IN V5 AND U0«> ( J'DIAMETER OF 500 MC.
FUEL
ELEMENTS
DESIGNING
NUCLEAR
8.5G/CM3 ) CIRCULATION
BY
IT
N C /NO
RATIO, FUEL
IS
-
D
KERNEL
SCHEME
MULTIPASS.
D-6
IS
o
CONTROL MEMBERS EFFECTIVENESS VERSUS THE DISTANCE PROM CORE SURFACE (UlflPORM LAYOUT)
o 5> CO H
s 01 09
8 o
Pig.
1. D-6
•UNIVERSAL GAS CIRCULATOR CHARACTERISTIC
H,M.cr.He
30 Fig.
SAA
2.
D-6
PERMISSIBLE CORE HEIGHT VERSUS PEDKKY-CIROTIT HELIUM PASUMETERS WHEH USIHO MGC, SG AlfD ?~W-
400 RP.^
6 TH,M
Tz
tk-750% C6re
-SSO'C
r Pig. 3.
D-6
MGC REVOLUTIONS VERSUS RP POWER
3,5Ma MPa
6000 • -
5000
3000 2000
Jig.
4. D-6
MGC DRIVE POWER VERSUS RP POV
Hi
3,9MUa
Fig. 5.
D-6
SPECIFIC FUEL ELEMENI CONSUMPTION VERSUS FUEL LOADING
0TT0
scheme
— Ml I rib Multipass scheme o o
2
o •H
0 fv
03+>
7
f$& fuel el.
Fig, 6, D-6
BUR1MJP VERSUS FUEL LOADING
/I?/? OTTO scheme
pjflAj Multipass scheme
g
5
el.
Fig. 7.
D-6
tvs Fuel temperature, relative units
p.
j
CO
tsi
vn
a
os o
a> <+
O
1-3
fl>
O
I
MAXIMUM REACTIVITY MARGIN VERSUS FUEL LOADING (NORMALIZED ON
FOR
u
/H
= 600, z
=10%)
in •H
> -P
H
OTTO scheme
5 -P •r4
A//7/43 Multipass scheme withou\,
without
HiO
-P O
J4
Fig. 9.
D-6
THE EFFECTI1
JS OF CPS RODS IN SIDE REFLECTOR VERSUS FUEL LOADING (NORMALIZED ON EFFECTIVENESS AT N c /N u
600, ZK « 10%)
Zs-/O%
MFIA5 Multipass scheme Qflfft OTTO scheme
F i g . 10. D-6
13
ЛИТЕРАТУРА
1. СА8АНДЕР В.П., САРЫЧЕВ Б.А. МЕТОДИКА РАСЧЕТА ВЫХОДА РЕАКТОРА ВТГР В СТАЦИОНАРНЫЙ РЕЖИМ РАБОТЫ НА ЗАДАННУЮ СХЕМУ ПЕРЕГРУЗКИ. МАТЕМАТИЧЕСКИЕ МОДЕЛИ ЯДЕРНОЭНЕРГЕТИЧЕСКИХ УСТАНОВОК. ПОД РЕД. В,В.ХРОМОВА, ЭНЕРГОАТОМИЗДАТ/М.,1983. 2, ASKEW J.R., PAYERS F.J. KEMSHELL P.B. A GENERAL DESCRIPTION OF THE LATTICE CODE WIMS, JBWES, OCT. 1966, P,56fl.
D-6
XA0101504
RADIATION RESISTANCE OF PYROCARBOH-BO1TDSD FUEL AND ABSOEBIHG ELEMENTS FOR HTGR V.A.Gurin, Yu.F.Konotop, N.P. Odejchuk, S.D.Sliirochenkov, V.K.Yakovlev, Kharkov Institute of Physics and Technology, Ukrainian SSR Academy of Sciences, 310108 Kharkov, USSR N.A.Aksenov, V.A.Kuprienko, I.G.Lebedev, B.V.Samsonov V.I.Lenin Nuclear Reactor Research Institute, 433510 Dimitrovgrad, USSR
On choosing the type of the reactor for a nuclear power plant, problems of nuclear and radiation safety are becoming at present of primary importance, particularly, for the regions with a high density of the population. The analysis of the design and experimental v/ork made in our country and abroad shows that high temperature gas-cooled reactors with a helium coolant (HTGR) best of all satisfy the requirements /1-6/. Early in the development of a high-temperature trend in atomic-power engineering in the Soviet Union it was planned to construct two plants with a HTGR, namely, a chemical-nuclear power plant with a VGR-50 reactor and an industrial nuclear-power plant with a VG-400 reactor /2,7/. Later, in order to try out the core components and the operation-support facilities for this new type of the reactor, it was decided to construct the experimental plant with a low-power high-temperature reactor (VGH). Spherical uranium-graphite fuel elements (60mm in outer diameter) comprising fuel zones (50 mm in diameter) with coated fuel particles (CP) dispersed in them are supposed to be used in the HTGR core. E-l
SZA
2.
A unique technology for producing spherical pyrocarbon-bound fuel and absorbing elements of a monolithic type has been developed at the Kharkov Institute of Physics and Technology. It is based on binding porous media with pyrocarbon produced in the pyrolysis of hydrocarbon gas /8,9,10/. In accordance with the proposed method, pyrocarbon wraps around each particle of a graphite powder, CP or a boron carbide particle until there is a free access for gas to the porous billet (Pig. 1). The main indicator of normal operation of HTGR fuel elements is their ability to confine fission products < (PP) throughout the operating period. The release
a
of gaseous fission products
"\ (GPP) at temperatures up to HOO°C and 8 to 15% PIMA burnup should not exceed
Pig. 1. Structure of the GSP graphite.
R/B
10 5 .
To substantiate the serviceability of spherical fuel and absorbing elements, we have undertaken extended reactor tests of CP, fuel pellets, small- and full-scale spherical fuel elements, and also stnall-scale absorbing elements. During tests we investigated: - radiation resistance of CP with, different types of fuel; - influence of the CP design on GPP release; -influence of nonsphericity on CP serviceability} -dependence of GPP release from fuel elements on the thickness of fuel-free cans;
Bit-
E-l
3. - a confining role of a pyrocarbon binder as a factor capable to diminish the rate of GPP release; - radiation resistance of spherical fuel elements down to the ultimate burnup; - radiation resistance of spherical absorbing elements down to rated fast neutron fluences and boron burnups. The CP were irradiated in the state of free filling at temperatures between 1200 and 1500°C to a~7% PIMA burnup. The main characteristics of some OP batches are given in Table 1. The irradiation parameters and measured rates of GPP release (R/B ratios) are listed in Table 2. As seen from Table 1, the tested CP differed in both the design of their protective coatings and the type of the fuel employed. During reactor tests all CP remained intact. It is also found that under the test conditions indicated in Table 2 the rate of GPP release depends on the CP design rather than on the type of the fuel employed. In the process of operational development the CP design underwent essential changes. At the first stage, the CP had four main fragment-confining coatings, namely: (i) a buffer coating, (ii) a dense pyrocarbon layer, (iii) silicon carbide coating, and (iv) an outer dense pyrocarbon layer. Later, the inner dense pyrocarbon layer was replaced by a combined (PyC + SiC) coating, and in present-day practice the outer layer is also made as a combined coating. Pigure 2 shows the rates of GPP release from CP of different designs irradiated under the same conditions. As seen from the the figure, the replacement of dense pyrocarbon layers by combined
5lb
E-l
Table 1. Main characteristics of some CP batches CP batch
Kernel Kernel material diameter
21-21-08-79
U0
3 400-630
Enrichment in
Coating Coating Coating densit r material i thickness (g/cm 3 )
U-235 (%) 4
5
6
21
PyC PyC PyC
PyC
1.1 1.5 1.7 3.2 1.1 1.5
28 14 56 80 10 18
Pyc
1.1
50
PyC PyC PyC
1.4 1.9
9 63
PyC
1.2 3.16 1.2
15 70 15
Pyc
1.7
65
Pyc
1.1
35
PyC
1.5 2.4
7
SiC Pyo
21-I-X-80
21
U0
Sic
21-3-X-81
U0
400-630
21
PyC+SiC
36-3-X-81
U0
400-630
36
400-630
21
110 80 55
3.16
PyC PyC
1.01 1.5 2.4
45 20
1.3
3 100 3 35
PyC PyC U0
7
SiC PyC
PyC+SiC PyC SiC
21-4-X-32
(^m)
PyC PyC PyC+SiC
SiC PyC PyC
1.7
3.18 1.2 1.8 1.1
1.5 2.4 3.18 1.1 1.8
70
42 15 70 80 3 100 E-l
5. 1 36-4-X-82
2
uo 2
3
4
5
6
7
400-630
36
PyC Pyc PyC+SiC SiC Pyc Pyc
1.1 1.5 2.4 3.18 1.1 1.8
40 25 105 60 3 35
36-5-X-S2
uo 2
400-630
36
Pyc Pyc PyC+SiC SiC" Pyc Pyc
1.1 1.5 2.4 3.18 1.1 1.8
28 15 63 63 3 42
21-4-X-82
uo 2
400-630
21
PyC Pyc PyC+SiC SiC PyC PyC
1.1 1.5 2.4 3.18 1.1 1.8
42 15 70 80 3 100
45-I-X-82
uo 2
400-630
45
PyC PyC PyC-fSiC SiC PyC PyC
1.1 1.5 2.4 3.18 1.1 1.8
70 50 85 60 3 112
45-2-X-83
uo 2
400-1500
45
PyC Pyc SiC PyC
1.0 1.8 3.18 1.8
55 30 120 40
36-6-X-83
UC1T
400-630
36
PyC PyC PyC SiC Pyc
1.1 1.5 1.8 3.18 1.1
40 25 35 140 3
E-l
1
2
3
36-13-X-84 (Th,U)0 2 Th/U=3
400-650
4 36
5
6
7
Pyc PyC PyC
1.0
35 7 60
SiC
Pyc 21-9-X-84
36-16-X-85
U02
UCH
800-1000
450-550
21
36
PyC PyC
36-17-X-86
U02
450-650
36
600 + 5 0
21
S%G
UCN
450-600
36
70 14
SiC
3.16
PyC+SiC
2.4
35
Pyc pyc
1.1
80 10 60
PyC PyC
PyC PyC
PyC PyC PyC+SiC SiC
36-22-X-88
70
80 105
PyC+SiC SiC PyC+SiC 21-10-X-87 (U,Th)0 2
1.1
100
PyO+SiC
PyC+SiC SiC PyC+SiC 36
1.8 3.18 1.8
1.5 2.4
PyC+SiC SiC PyC+SiC 36-18/1-X-86 TJO2 450-650 Al o 0- -4.65% 3 % S i 0 o - 3.5
1.5
PyC PyC
1.5 2.4 3.16
2.4 1.1
1.5 2.4 3.16
2.4 1.1
1.5 2.4
65 35 42 7 50 63 35 80 10
2.4
70 65 70
1.1
42
1.5 2.4
10
3.16
3.16 1.1
PyC+SiC
1.5 2.4
SiC
3.16
PyC+SiC
2.4
56 70 45 15 50 70 40 E-l
7.
36-21-X-88
UO,
500-600
36
PyC PyC PyC+SiC SiC PyC+SiC
1.1 1.5 2.4 3.16 2.4
60 20 58 70 50
36-23-X-88
UOr
600+25
36
PyC PyC PyC+SiC SiC PyC+SiC
1.1 1.5 2.4 3.16 2.4
50 18 60 80 47
36-24-X-88 continuous deposition
UO,
576+50
36
PyC PyC PyC+SiC SiC PyC+SiC
1.1 1.5 2.4 3.16
60 15 55 70 50
PyC PyC PyC+SiC SiC PyC+SiC
1.1 1.5 2.4 3.16 2.4
PyC PyC PyC+SiC SiC PyC+SiC
1.1 1.5 2.4 3.18 2.4
50 63 42
Pyc PyC PyC+SiC SiC PyC+SiC
1.1 1.5 2.4 3.18 2.4
56 10 50 60 42
•36-25-X-88
36-26-X-89
3 6-27 -X -89 continuous deposition
UO,
'UO,
UO,
576+50
500+50
500+50
36
36
36
2.4
60 15 60 70 50 50 10
E-l
8, Table 2. Irradiation parameters and measured GPP release from CP.
GPP release ( 85m Kr) (R/B.105)
Irradiation temperature
Burnup
21-21-08-79
1200
7.9
5.0
21-1-X-80
1200
6.9
6.7
21-3-X-81
1200
3.0
2.4
36-3-X-8I
1250 1500
2.4 2.9
0.8 1.18
21-4-X-82
1250
1.8
1.1
36-5-X-82
1250 1500
4.6 2.7
0.51 0.49
45-1-X-82
1250
2.4
0.75
45-2-X-83
1250
3.0
0.72
36-6-X-83
1250 1500
2.7 6.00
0.11 0.81
36-13-X-84
1250
4.3
0.93
21-9-X-84
1250
5.8
0.13
36-16-X-85
1500
7.1
0.14
36-17-X-86
1250 15OO
2.6 5.8
0.10 0.13
36-18/1-X-86
1250
2.5
0.09
GP batch
tt.FIMA)
E-l
S2JS
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z ptre t saqoq.Bq a o j q.Bqq. u a a s sx
C 9T
qq.Tr/A a:D JO sq-saq. j:oq.OBao; j o sq-insacc a q j •lunxtiBjn .fiq sSuxq.Boo aAxq.oaq.ojcd j o uox^BtixwBq.1100 iBxq.Bds paonpaa: ^xi^Tq-^aesa TIB UX paq.Xnsaj s x q i ' j a ^ B x paxjanbxxopnasd * qq.x/A ocoq.OBaj: aqq. apxsq.no XBAOuiaa Jxaqq. puB uoxq.xsodap Suxq.Boo pauxqmoo guxanp sxosoaaB uinxuBorn j o ..SuxqoBax suxjcoxqo,, puB saxq.o:adoa:d a,inq.oruiq.s oxq.sx jaq.OB.iBqo qq.oq oq. anp aq ^BUI sxqj] *Q_0L • L ~ oq. UMOP
QJ)i»3
~ WOJTI
9
^
g S
r
0)
TO DO
-Z •suSxsap q.uao:ajjxp jo saxoxq.iBd paq.Boo
*6
01
s-
10, Table 3. Results of reactor tests of CP with different degrees of nonsphericity.
Batch
Burnup
number
(% FIMA)
1
2
Irradiation temperature
R/B»10 + 5
(°C)
max' min
2.9 3.2 2.9
1200 1200 1200
1.0 - 1.05 1.0 - 1.2 1.05 - 1.2
3.9 7.0
1.0
1200 1200 1200
1.0 - 1.05 1.0 - 1.2 1.05- 1.2
1.9 4.8 7.8
1200 1200 1200
1.0 - 1.05 1.1 1.0 - 1.2
1.6 2.7 2.4
6.9 1.0 1.0 1.0
3
FGP ( 85 Kr) release
Nonsphericity degree,
3.0
2.0
lower (by a factor o f ~ 1 . 5 ) . This is probably due to a lower uranium content in the combined coating. The irradiation parameters and the measured rates of GPP release from some batches of small- and full-scale spherical fuel elements are given in Table 4. Table 4 Irradiation parameters and measured rates of GFT? release from some batches of small- and full-scale spherical fuel elements Irradiation temperature
Burnup (% FIMA)
(°C), fuel element centre
2
3
4
1 2
2
14.4 12.9
3
2
4
3 3
600...1350 500... 750 180... 460 1250 1250
VJl
release, R/B» 10^ under burnup
"5
1
5
GFP (85Kr)
VJl
3xp. Number number of fuel elements in experiment
2.0
1.0 1.0 1.0
13.0 13.0
0.5 -
Fuel kernel material
Type
of
CP
%
6 10.0 1.0 1.0 6.2 -
7 U0
2 uo 2 uo 2
uo 2 UO,
8 1 1 1 1 1 E-l
11, Table 4 (cont) 1
2
3
6 7
2
650...1050 1250 400... 900 750...1100 640...1250 1400 1200 1200 1250 1400 1250 1250 1250 1250 1500 1500 1500 1250 1250 1250 1250 1250 1250
8
9 10 11 12
13 14 15 16
17 18
19 20 21 22
23 24 25 26 27 28
3 10 8
7 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3
4 21.8
27 2.7 13.8
3.8 13.3 14.0 8.0 14.0 15.3 15.0 16.0 17.5 18.0 17.0 10.0 18.0
16.1 17.4 12.9
9.8 17.5 16.0
5
6
7
8
0.6 0.1 1.0 1.0 10.0
1.0 1.0 1.0 -
0.5 -
4.8 -
0.2 0.2 0.18 0.2 0.13 0.22 0.18 0.18 0.15 0.12 0.11 0.1 0.12 0.1 0.07
0.9
uo 2 uo 2 uo 2 uo 2 uo 2 uo 2 uo 2 uo 2 uo 2 uo 2
1 1 1 1 2 1 1 1 1 2 2 2 2 1 1
0.6 0.84 0.81 0.33 0.63 0.78 1.0 0.48 0.27 0.16 0.15 1.0 0.25 0.2
U0
2
,uo2 uo 2 UCN ITCH
uo 2 ucir
2
3 3
UCN
uo 2 uo 2
2
3
(Th,U)02
uo 2 uo 2
2* 2
3
Notes; 1: PC design with PyC-SiC-PyC 2: PC design with (PyC+SiC)-SiC-PyC 2*: PC design with (PyC+SiC)-SiC 3: PC design with (PyC+SiC)-SiC-(PyC+SiC) In the process of fuel element irradiation, two special features were established: (i) at the initial stage of irradiation (up to 4% PIMA) the GPP release from fuel elements is always lower than that from the CP, on the basis of which the fuel elements are made; (ii) there is a jump in the rate of GPP release after reaching a 4 to 5% PIMA burnup.
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5%A
12.
It is of interest to note that the jump interval depends neither on the neutron fluence nor on the CP breakage, but is determined only by the burnup level, i,e., by the PP storage. Besides, the comparison between the rates of GPP release from small-scale (2.5 mm can thickness, exp.114, Table4) and full-scale fuel elements (10 mm can thickness, exp. 1T13» Table 4 ) , where the same batch of CP was employed, reveals a tendency to the reduction in the GPP release rate as the can thickness grows. This is obviously seen from Pig. 3» which shows the rate of GPP release from one batch of CP irradiated in the state of free filling (1), and in the composition of small- (2) and full-scale (3) fuel elements.
j
2
-5
id6
j i
,
i
1
i
i
t
„
-,
„
1
~
i
i
1
1
i
.
_
9
W
H
i
/2 15 Butnup, ^ Fit**
Pig. 3. Release rate of GPP from CP (1), small- (2) and full-scale (3) fuel elements. To study the influence of the can thickness on the GPP release, we have made model fuel elements comprising uncoated uranium dioxide kernels,~3 mm in diameter. The thickness of the graphite GSP can varied from 9.96 to 7.12 mm. The parameters of these model fuel elements and the results of their tests are presented in Table 5. E-l
13. Table 5. Model fuel element parameters and test results Burnup
Irradiation
thickness (mm)
(% FIMA)
temperature
03 03
5.47 7.12
03
2.95 3.82
03
GSP can
GFP release, R/B. 10 3
(°c)
Burnup (%FIMA) <4 >5
1100 1100
1.3 1.1
4.1 3.6
1100 1100
0.8
2.7 1.4
0.5
It is seen from the table that, similarly to the case of irradiating small- and full-scale spherical fuel elements, this experiment also shows two stages of GFP release: one at a burnup o£^A-%t and. the other - at higher burnups. Besides, the GFP release rate depends on the thickness of the GSP can. This dependence is shown in Fig. 4. It is seen that the GFP release value decreases as the can thickness increases. -2
ia
In the present experiment it has been unambiguously established that a 1 mm increase in the thickness of the graphite GSP can
-5
to
reduces the GFP release by a factor of
1.3.
Thus, owing to its structure, to
5 tO can thickness, mm
the pyrocarbon-bound graphite GSP acts as a sufficiently efficient
Fig. 4. Release rate of GFP from barrier to GFP and may provide model fuel elements. 1: Jfo FIMA burnup; 2: 4% FIMA burnup.
an
additional sealing of process
impurities, including uranium aerosols. E-l
14. As mentioned above (Fig. 3 ) , at the early stage of irradiation to burnups of~ 4% PIMA, the GPP release from fuel elements turned out to be by a factor of 15 to 20 lower than that from the CP of the PyC-SiC-PyC design, which constituted the fuel elements (Pig. 3 ) . At present, the employment of CP with combined coatings made it possible to reduce the GPP release from them down to (0.9—1 )»1O" (see Tables 1 and 2 ) . In this case, the GPP release from fuel elements at the initial stage of irradiation appeared comparable and, after a 5% PIMA burnup, even greater than from the CP used in these fuel elements (CP batch 36-18/I-X-86, Table 2, and experiment 28, Table 4 ) . This indicates that for hermetically sealed CP the uranium impurity in the graphite matrix becomes the main source of GPP. A sharp enhancement in the rate of GPP release after a burnup of 4 to 5% PIMA. may be attributed to some faults arising in a pyrocarbon binder, and as a result, a loss of sealing of uranium inclusions present in the matrix graphite. It may also be assumed that the pyrocarbon binder may efficiently confine the GPP only until a certain level of fission gas pressure, which corresponds to a 4-5% PIMA. burnup. Experimental verification of the serviceability of spherical pyrocarbon-bound absorbing elements of a monolithic type under reactor irradiation was performed at temperatures between 400 and 1250 C. The boron content of the natural composition in the absorbing elements was varied from 0.1 to 5.0 g. The greatest fluence of fast (E^O.IMeV) neutrons was about 1.0*1022 n/cm2. With this fluence value there was actually a complete burnup of boron-10. We studied the influence of the granulometric composition of the powder filler and its type, and also of heat treatment condi-
15. tions on the radiation resistance of the absorbing elements. The after-reactor examination has shown that under irradiation in a wide range of temperatures all absorbing elements, similarly to fuel ones, undergo an isotropic shrinkage, whose value is independent of the filler type (CP or boron carbide powder) and is weakly dependent on the irradiation temperature and neutron fluence. The shrinkage of spherical fuel and absorbing elements as a function of fluence is shown in Pig. 5. An insignificant straggling in the shrinkage values is evidently due to different densities of the spherical elements.
5372 OA^flfl
m
1250 53°
/?oo •
—
-
-
—
^
~—-—_ 920 A /\6iO /200 ^10
.
A920 ^750
920 ^—A-^&920
w Qin
mo o 750 ^—
2,0
0,5
^
f /02tc-m2
. 5. Dimensional changes of spherical fuel and absorbing elements versus fluence of neutrons. Q : fuel elements; ^ : absorbing elements. Figure 6 shows shrinkage values of fuel and absorbing elements of densities between 1.78 and 1.8 g/cnr* irradiated at ~ 12OO°C to fluences^ times greater than the rated ones. It is seen that the decrease of the diameter of these elements in the range of fluences from 1.0-1CT
to 9.6-1CT1 is approximately the same and
is found to be 0.4 to 0.6%. The change in the granulometric composition of the filler in the range between 0 and 630
m and the rise of the finite ternE-l
16.
E>,Q,/rieV
.0
-0,5
0
A
So"
o —
o
°o
"~"^———— —— 0—
A£ n_
7X
A
-1,0 ~
Pig. 6. Dimensional changes of fuel and absorbing elements of densities between 1.75 and 1.82 versus neutron fluence. Q: fuel elements; /j\ : absorbing elements.
perature of heat treatment up to 16OO°C produced no effect on the shrinkage value. The static strength of spherical fuel and absorbing elements is not affected by irradiation, and is about-^45 kK at a density of ^ 1.85 g/cnr3. Conclusion On substantiating the serviceability of spherical pyrocarbonbound fuel and absorbing elements under reactor irradiation conditions we have established that: - the replacement of dense pyrocarbon coatings in CP by a combined (PyC+SiC) coating reduces the GPP release rate by nearly an order of magnitude; - the increase in CP nonsphericity causes the enhancement of the GPP release; - pyrocarbon binder of the matrix graphite acts as an additional barrier to fission products, and a 1 mm increase in the thickness E-l
17. of the uranium-free can reduces the GFP release by a factor of ^1.3. The mentioned advantages are peculiar only to the fuel elements made by the GSP technology. During irradiation, spherical fuel and absorbing elements experience an isotropic shrinkage. Thus, the relative change in the diameter was 0.4 to 0,6% for the fluence range from (2.0*101^ to 9.6*10
)n/cm . The static strength of the spherical
elements remains the same. The results of the reactor tests of spherical fuel elements at
1500°C and with the burnups exceeding the rated values by
a factor of 2.5 or 3» and also of spherical absorbing elements at temperatures 400 - 125O°G with fluences of fast (E>0.1 MeV) neutrons -~ 1•1022 n/cm 2 indicate a high radiation resistance and reliability of the designed spherical pyrocarbon-bound fuel and absorbing elements. References 1. Ponomarev-Stepnoj N.N., Protsenko A.N., Grebennik V.N. Status of HTGR development in the world (in Russ.). Vopr. At. Nauki i Tekh., ser. Atomno-vodorodn. ehnergetika i tekhnologiya, 1984, iss. 2(13), pp. 3-11. 2. Grebennik V.N. Status of HTGR development in the USSR (in Russ.). Atomno-vodorodn. ehnergetika i tekhnologiya, iss.5. Moscow, Ehnergoatomizdat publ., 1983, pp. 106-118. 3. Karpov V.A. Nuclear safety of HTR with spherical fuel elements (in Russ.). Atomn. tekh. za rubezhom, 1979, 1111, pp.15-18. 4. Glushkov E.S., Grebennik V.N., Kosovskij V.G. et al. Physical aspects of nuclear safety for the VGR-50 device (in Russian). Vopr. At. Nauki i Tekh. Ser. Atomno-vodorodnaya ehnergetika i tekhnologiya 1983, iss.3("i6), pp. 51-53.
18.
5. Legasov V.A., Demin V.F., Shevelev Ya.V. Fundamentals in the analysis of safety in nuclear power-engineering (in Russian). Atomno-vodorodnaya ehnergetika i tekhnologiya, iss.7, Moscow, Ehnergoatomizdat publ., 1986, pp.61-104. 6. Grebennik V.1J. Status of HTGR development in the world (in Russ.) Atomno-vodorodnaya ehnergetika i tekhnologiya, iss.7, Moscow, Ehnergoatomizdat publ., 1986, pp. 147-159, 7. Glebov V.P., Grebennik V.N., Ponomarev-Stepnoj N.K. et al. HTGR (VGR-50) of the chemical-nuclear power plant (in Russian), Atomno-vodorodnaya ehnergetika i tekhnologiya, iss.5, Moscow, Ehnergoatomizdat publ., 1983, pp.118-123. 8. Zelenskij V.F., Gurin V.A., Konotop Yu.F. et al., Development of pyrocarbon-bound fuel and absorbing elements of the monolithic type for HTGR (in Russian). Atomno-vodorodnaya ehnergetika i tekhnologiya, iss. 5» Moscow, Ehnergoatomizdat publ., 1983, pp. 213-225.
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XA0101505 INTERNATIONAL ATOMIC ENERGY AGENCY
622-I3-TC-389.26
Technical Committee Meeting on Gas-Cooled Reactor Technology, Safety and Siting
Dimitrovgrad, USSR 21-23 June 1989
Convened by the International Atomic Energy Agency (IAEA)
THE MATERIALS PROGRAMME FOR THE HIGH-TEMPERATURE GAS-COOLED REACTOR IN THE FEDERAL REPUBLIC OF GERMANY: Status of the development of high-temperature materials, integrity concept, and design codes H. NICKEL Kernforschungsanlage Julich GmbH and RWTH Aachen E. BODMANN Hochtemperatur-Reaktorbau GmbH, Mannheim H.J. SEEHAFER INTERATOM GmbH, Bergisch-Gladbach
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THE MATERIALS PROGRAMME FOR THE HIGH-TEMPERATURE GAS-COOLED REACTOR IN THE FEDERAL REPUBLIC OF GERMANY: Status of the development of high-temperature materials, integrity concept, and design codes H. NICKEL Kernf o r s c h u n g s a n l a g e J i i l i c h GmbH and RWTH Aachen E. BODMANN Hochtemperatur-Reaktorbau GmbH, Mannheim H.J. SEEHAFER INTERATOM GmbH, Bergisch-Gladbach
Abstract THE MATERIALS PROGRAMME FOR THE HIGH-TEMPERATURE GAS-COOLED REACTOR IN THE FEDERAL REPUBLIC OF GERMANY: Integrity Concept, status of the development of high-temperature materials and design codes During the last 15 years, the research and development of materials for high temperature gas-cooled reactor (HTGR) applications in the Federal Republic of Germany have been concentrated on the qualification of hightemperature structural alloys. Such materials are required for heat exchanger components of advanced HTGRs supplying nuclear process heat in the temperature range between 750° and 950°C. The suitability of the candidate alloys for service in the HTGR has been established, and continuing research is aimed at verification of the integrity of components over the envisaged service lifetimes. The special features of the HTGR which provide a high degree of safety are the use of ceramics for the core construction and the low power density of the core. The reactor integrity concept which has been developed is based on these two characteristics. Previously, technical guidelines and design codes for nuclear plants were tailored exclusively to light water reactor systems. An extensive research project was therefore initiated which led to the formulation of the basic principles on which a high temperature design code can be based.
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•o
1.
INTRODUCTION
In the Federal Republic of Germany, the helium-cooled high-temperature reactor (HTGR) and the fast breeder reactor are being promoted as advanced nuclear reactor systems. The HTGR core development is based on spherical fuel elements forming a pebble bed. Following favourable experience with the operation of a small experimental reactor, the AVR with 15 MWel / I / , a 300 MWel prototype reactor, the THTR 300 was constructed and became operational in the mid-eighties. It has exhibited good service behaviour and has confirmed the safety characteristics of the system. At present, two HTGR concepts are being followed up by German companies: a small HTGR modul type (approx. 80 - 100 MWel) with a steel pressure vessel for steam cycle as well as for nuclear process heat applications (PNP), and a medium-sized reactor (approx. 550 MWel) with a prestressed concrete pressure vessel for electricity generation. Extensive effort has been devoted to the qualification of materials, concentrated on these two German advanced HTGR concepts. The materials for the pebble-bed HTGR have been defined and are qualified for the projected operating conditions . The manufacturing procedures for the required product forms have been proven /1-9/, and the semifinished products and the fabrication methods are available. This statement stands without any reservation for the components for a steam-cycle HTGR with working temperatures up to 750°C as is demonstrated by the commissioning of the THTR. The spherical fuel elements and the structural graphite have been developed and qualified, although long-time irradiation experiments and tests of simulations of emergency conditions are still going on /1-6/. Some restrictions, however, must be made for components in a nuclear process heat plant with the highest working temperature (up to 1000°C). The continuing structural materials research work is now concentrated on the verification of the integrity of components over the envisaged service life. This paper will be concerned with metallic materials for structural components of steam cycle and process heat HTGRs and with the HTGR integrity concept and the development of design codes.
2.
METALLIC MATERIALS FOR STRUCTURAL COMPONENTS
2.1 Alloy selection For the selection and qualification of metallic materials, the following properties have been taken into account: -
-
strength and ductility for components which can be designed with time independent materials data; creep resistance and microstructural stability for those components which must be designed using time-dependent design values; corrosion resistance in the service atmospheres; hot and cold formability and weldability; possibility of non-destructive testing during manufacture and in the plant; sensitivity of materials to neutron irradiation induced embrittlement for those components which are in the core region.
In designing the THTR, metallic materials were used for which wide experience was already available from conventional power plants. Typical alloys are the ferritic steel X20 CrMoV 12 1 for low temperature regions
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of the steam cycle system and X10 NiCrAITi 32 20 (Alloy 800) for components of the steam generator and the superheater piping. For the application of an HTGR plant for nuclear process heat, it is necessary to use alloys which are creep resistant up to high temperatures. The lifetime of the heat-exchanging components of a nuclear process heat plant primarily depends on the creep strength. From a number of possible creep resistant nickel-base materials the alloy NiCr 23 Co 12 Mo (Alloy 617) has been qualified for the intermediate He/He heat exchanger (IHX) and the steam reformer (RSO) subjected to the highest operating temperatures, and the materials NiCr 22 Fe 18 Mo (Hastelloy X) X10 NiCrAITi 32 20 (Alloy 800 H) X20 CrMoV 12 1 have been qualified for components operating in the lower temperature range. Besides the evaluation of the commercially available Alloy 617 for the high-temperature region, the separate alloy development for heat exchanging components resulted in a new Fe-Cr-Ni alloy called Thermon 2.4972 (tradename of Thyssen-Edelstahlwerke, Krefeld), for which creep rupture data up to 20000 h is now available.
2.2 Material qualification for nuclear process heat plant components The main important components in a nuclear process heat HTGR plant are the intermediate heat exchanger (He/He IHX), the steam reformer (RSO) for the methane reforming system, and the steam generator. The basic work has been concentrated on the following items: -
-
creep and creep rupture; high-cycle fatigue; low-cycle fatigue, without and with hold times; influence of ageing on short-term properties; creep ratcheting; creep buckling; environmental effects.
The investigations have been carried out by the joint effort of the partners in the German HTGR materials working group.
2.2.1 Creep behaviour Creep data for different heats of the alloys Alloy 617 and Alloy 800H have been established up to test durations of approximately 50000 h for Alloy 617 and to 100000 h for Alloy 800 H. In this range it is not necessary to differentiate between the results in air and helium as shown in Fig. 1 for Alloy 617. For this alloy the data from the qualification programme has enabled a design for a He/He-IHX up to 100000 hours at a materials temperature of 950°C, for steam reformer tubes up to 140000 hours at 900 °C, and for steam generator tubes up to 350000 hours at 750 °C 11-12/. E-2
ir 700 750 800 850
degrees C degrees C degrees C degrees C
900 950 1000 1050
degrees C degrees C degrees C degrees C
i— ...,
helium
700 750 800 850 ....
ios
"TT
degrees C degrees C degrees C degrees C
1.
degrees C degrees p degrees 0 degrees C j \'i" s ,o 5
,oo:
time (h)
FIG.
900 950 1000 1050
time (h)
Creep strain limit and stress rupture strength of NiCr23Col2Mo (Alloy 617) in air and simulated HTGR helium
2.2.2 Fatigue behaviour Cyclic stressing of components can be caused by rapidly changing primary loads, e.g. vibrations (high cycle fatigue, hef), or by thermally induced alternating strains (low cycle fatigue, lef). The thermally induced alternating strains are simulated in lef tests. The aim is to determine the number of cycles to rupture or the number of cycles to crack initiation as a function of temperature and strain amplitude. Figure 2 shows the lef behaviour of Alloy 617 in air and simulated reactor primary coolant (HTGR helium), which contains the impurities H2O, CO, CH4, H2, and N2 in the ubar range for the temperature range from 750° to 950°C /13-15/. The results obtained suggest that a higher number of load cycles to failure can be expected for the base material in HTGR helium than in air. 2
rantje for design curves
cycles to failure FIG.
2.
Results from lef test on NiCr23Col2Mo (Alloy 617)
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2.2.3 Creep-fatigue interaction Fundamental investigations of creep-fatigue interaction of Alloy 617 at 950°C in HTGR helium shows the strong influence of hold-times /16/. A hold time irrespective of its position in the cycle always reduced fatigue life in comparison to continuously cycled tests at lower strain rates but of equal cycle duration. Tensile holds were found to be more damaging, than compression holds. In spite of these results, a simplified model can be used for components under realistic process heat conditions. From the experimental data for the creep and fatigue behaviour, design curves have been obtained and also have confirmed the validity of the linear life-time fraction rule for creep and fatigue (Fig. 3 ) . D
= D c + D f with D ^ 1
D c = E ti/t R i where D = Dc = tf = =
lifetime exhaustion factor, exhaustion due to creep, actual service duration at a± and temperature T^ allowable time at Oj_ and temperature T^
and Df = E where = exhaustion due to fatigue = actual number of cycles at strain range Ae and temperature T^ = allowable number of cycles at strain range Ae^ and temperature
fatigue design
-5MPa -106cycles
FIG. 3.
Scheme of creep-fatigue analysis (values for NiCr23Col2Mo , (Alloy 617))
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6 The analysis of component life for the high temperature PNP-components cannot only be treated by elastic analysis; inelastic analysis is necessary to confirm a realistic service life. The strain range for fatigue of the heat exchanger piping is calculated to be less than 0.03 % which can be neglected for life-time calculations. For steam reformer tubes, during operating time of 140000 hours 160 start-up and shut-down cycles are assumed, which gives a running time of 875 hours between shut-downs. Taking into consideration the relaxation of stresses induced by temperature gradients through the tube wall, the lifetime calculation gives the following result (Fig. 4) I'12,16/. Assuming that each cycle produces the same amount of consumption of lifetime as the first cycle, the lifetime exhaustion factor D = 1 allows 455 start-up and shut-down procedures, more than twice the number postulated for the operation time.
LOAD SITUATION
, Relaxation equation (uniaxial) modified relaxation equation
r n *65mm'
I additional axial stress ' of 3 N/mmJ
0,06 875 T
0,05
c
E0.03 0,02 0,01 0,01
FIG. 4.
0.1
10 timeh
100
1000
10000
Creep relaxation of thermal stress and resulting strain behaviour at the inner surface of the tube NiCr23Col2Mo (Alloy 617)
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SLtS
2.2.4 Properties after long-time ageing After long-time ageing at high temperatures, the alloys generally undergo structural changes resulting in a decrease in deformability at room and intermediate temperature. Figure 5 gives an example for the Alloy 617, aged at temperatures of 900°C and then rupture tested at different temperatures. There is a reduction in fracture elongation due to thermal ageing up to test temperatures of about 800°C.
•8
A aged 1000 h D aged 300h 0 aged lOOh o sol. treated 600
FIG. 5 .
800 1000 temperature °C
Influence of ageing at 900°C on fracture elongation of NiCr23Col2Mo (Alloy 617)
2.2.5 Creep ratcheting Creep ratcheting effects on the He/He heat-exchanger components cannot be handled by using any calculation methods given in the ASME-CC N47. An inelastic analysis using finite element programs is very expensive and time consuming. Therefore, the degree of ratcheting due to hot streaks in the coolant helium has been experimentally investigated. It has been found that /17.18/: -
creep ratcheting may occur even if the primary stress is low (Fig. 6 ) ; but the accumulated ratcheting strain remains below the permitted strain limits.
2.2.6 Creep buckling To achieve the planned operation time of 100000 hours, it is necessary to keep the loads as small as possible. The used pressures acting inside and outside on the tube are in the same range of 40 to 60 bar; therefore normal operation conditions are characterized by low load-controlled
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X10 Ni Cr At Ti 32 20 H
/N mm "2
A •
Th = const. = 950°C ratcheting
.25
200 time/h FIG. 6.
300
400
Creep curve under axial primary stress (A) and superimposed alternating secondary stress (i.e. ratcheting, •)
primary stresses even for the hot test parts of the tubes. Under accident conditions it may happen that at very high temperatures load-controlled stresses considerably exceed the normal operational level, and buckling of tubes under external pressure may occur. This has been investigated in experiments, the results of which have been compared with theoretical calculations based on Hoff's model and on finite element analysis /12,19/. For example, the dimensioning of the hot header of a heat exchanger is determined by the postulated emergency condition. The header must be safe against creep buckling. In the case of a helium/helium heat exchanger for the postulated emergency condition, loss of pressure in the secondary circuit exposes the components to an outer pressure of 40 bar at 950°C. This exposure may cause a loss of geometrical stability. The judgement of evaluation component behaviour must be done by inelastic analysis, in which the deformation with increasing time must be considered.
tube with external pressure
time to collapse
d5 max
shape factor; 10-
J
max
n = 7 Apo= 60 bar
k = «r 15 h- 1 0.10.01 FIG.
7.
0.1
1.0
10 100 time/h
Time dependence of tube ovality E-2
A parameter study of creep buckling for the hot header of a He/He heat exchanger has been performed /111: dimensions were 437.5 mm/125 mm; external pressure 42 bar, T = 957°C. Figure 7 shows the increase of ovality as a function of time. According to this calculation, a total exposure of 20 hours for an initial ovality of 1 % may be tolerated. In reality, the temperature will decrease immediately after the emergency event. If the cooling rate is 5 K/h, failure due to creep buckling is not expected. Summarizing the creep buckling phenomenon, it should be noted that in the experiments both the duration of stressing at high temperatures and the initial ovalities exceeded by far the conditions expected for real components. Moreover, the observed collapse of tubes does not actually represent a safety related problem, but rather a question of availability since all the experiments up to now have shown that the leak tightness is maintained despite tube collapse through buckling /19/.
2.2.7 Environmental effects The selected material should withstand the environmental attack. In practice the corrosion problem can be principally solved by the following methods: -
control of control of control of reduce the
the chemistry of the service atmosphere; corrosion attack by a wall thickness allowance; surface condition or application of protective coatings to attack of the material.
Based on the results of the materials evaluation for HTGR process heat the first two principles are used. For example, the steam corrosion in the secondary circuit is controlled by appropriate conditioning of the feed water.
FIG. 8.
Schematic representation of the modified stability diagram of chromium for identification of the corrosion behaviour in different atmospheres. Ill indicates the region of acceptable behaviour of the alloy. a c : carbon activity of the atmosphere P Q 2 : partial pressure of oxygen of atmosphere PCO : minimum partial pressure of carbon monoxide for stable behaviour of the alloy
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10 As mentioned above heat transferring gas in the primary circuit, helium, is slightly contaminated by H2O, CO, CH4, H2 and N2 in the ubar range. Chromium is the metal element governing the corrosion reactions of the impurities with the Fe- and Ni-base alloys. No deleterious high temperature gas corrosion effects have been found at temperatures up to 750°C. At higher temperatures, compositions of the impurities with excessively high or very low carbon activities may cause carburization or decarburization, respectively (Fig. 8 ) . By keeping the CO partial pressure of the cooling gas helium in a well defined range, unacceptable carburization/decarburization reactions can be avoided. The impurity concentrations in the primary circuit helium may lies within this range without the need for conditioning /20/.
2.3 Material for control rods The control rods are essential components for the safety of an HTGR plant. These components are exposed to irradiation by thermal and fast neutrons. The effect of fast neutrons on material properties is limited to temperatures up to about 450°C. At higher irradiation temperatures, however, the helium atoms, produced by thermal, epithermal and fast n,a-reaction in boron and nickel, decrease the ductility of steels and Ni-base alloys. This high temperature embrittlement increases with increasing neutron fluences and temperature. Up-set conditions in advanced large HTGRs may lead to short-term temperature excursions up to about 850°C in the control rod materials. Therefore, the selection of material for the control rod tubes is of crucial importance. In a screening programme a number of high temperature steels and Nibase alloys have been irradiated at 400, 500 and 600°C and were examined in post-irradiation tensile and short-term creep tests. As reference material a thermo-mechanically treated optimalized version /21/ of 1.4981 (DIN X8CrNiMoNb 16 16) was chosen because of its superior post-irradiation ductility properties, i.e. total elongation and creep rupture elongation (Figs 9, 10) /21/. The irradiation tests are continuing to accumulate a fluence of 10^° m~^, the expected maximum fluence of control rods in large HTGR plants.
2.4 Continuing research 2.4.1 Constitutive equations Constitutive equations are the basis for inelastic analysis of the component behaviour under complex loading and for complex component geometry. Particularly the inelastic analysis at very high temperature is a large field of theoretical and experimental investigations /12,22/. The existing estimates for constitutive equations are divided in the two principle categories:
-
models separating time independent plasticity and time-dependent creep unified models which define only an inelastic strain.
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11
973K J
38
a 25
n
2 !
24
19
L
a.
I • 1
II
0.5-
i
0.1-
FIG. 9.
1.4981
1.4981 KA2
1.4876
^4603
1.4981
1.4981 KA2
I
1.4876
I 2.4603
Creep rupture strain of various metallic high temperature alloys before and after neutron irradiation ("("th = 3x10^-* m~^) and short fracture lifetime (<100 hrs) S unirradiated @ irradiated Irradiation experiment 200
100 80 60
HTR-K1
HTR-K2
HTR-1 HTR-2c HTR-2AB
T
V Material 1.4981-KA optimized
u c 40
I
20
1.4981
£
10 8 6
O creep rupture elongation O hot tensile rupture elongation
10 23
r* 10 za 1 Thermal neutron fluence
FIG. 10. Creep rupture strain of stainless steel 1.4981 (X8 CrNiMoNb 16 16) after neutron irradiation (irradiation temperature: 400°C; test temperature: 850°C)
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BSD
12 In the work for the HTGR beside the analytical approach, Norton's creep law in the three-dimensional formulation for stationary creep is used and the applicability of the ORNL-model as well as the Interatom model is under consideration. Our theoretical and experimental work is, for example, directed to combining internal pressure, tensile and torsion load on tube components (see 2.4.2). Summarizing the results, the multiaxial creep of Alloy 617 tubes can be described mathematically by the von Mises' theory on using Norton's creep law as the constitutive equation for stationary creep. The theoretically derived formulas for the stress/strain/time behaviour give an acceptable approximation to the observed deformation behaviour of tubes under multiaxial and complex loading conditions.
2.4.2 Multiaxial testing Creep laws and design values are derived from creep rupture tests on uniaxially loaded specimens at temperatures between 700 and 1050°C. The validity of the data and parameters have been proved for tubes under multiaxial loading. For the tests semi-finished products such as rods, IHX tubes (22 mm x 2.2 mm) and RSO tubes (120 mm x 10 mm) were used all fabricated from the same master heat of Alloy 800 H to the same specification concerning the final heat treatment and grain size distribution. The creep properties of standard creep specimens from rod material, of IHX tubes, and of RSO tubes under constant tensile load are compared in Fig. 11. For the IHX tubes, the beginning of secondary creep took place at about 150 h, whereas the two other materials exhibited a very limited region in primary creep and a similar behaviour in the second stage. Using Norton's creep law, with n and k values, derived from this experiment, the creep behaviour of RSO tubes could be sufficiently approximated. For the IHX tubes, however, the simplified mathematical description is not acceptable, due to the extended primary creep range.
calculation
X -. rod specimen • IHX tube o : reformer tube
X 10 NiCrAITi 32 20
J
I 00
125
150
175
200
225
time / h FIG. 11. Creep curves for axially loaded rod specimen (x), IHX-tubes (...), RSO tubes (o) at 950°C, X10NiCrAlTi3220, a = 30 N/mm 2
SS4
E-2
13 300 250 200
SZ
CD
150 100 50
CD •+-• Q.
0
tension
internal pressure
tens.+ int. press.
tens.+ torsion
tens.+torsion int. press
type of loading FIG. 12. Comparison of the lifetime of tube specimens under various loading combinations, X10NiCrAlTi3220, cry = 30 N/mmm 2 , T = 950°C The poor approximation of the experimentally obtained creep behaviour in the case of the IHX tubes may be explained by the fact that in these tubes (2.2 mm wall thickness) the influence of the surface region may significantly influence the deformation behaviour. In another series of experiment with IHX tubes, the following loading conditions were applied and the time of failure determined: internal pressure and tension; tension and torsion; tension, torsion, and internal pressure. In all cases the von Mises' deviatoric stress (stress intensity) was Oy = 30 N/mm 2 . The time to failure in short time tests with a failure mode of deformation rupture of tube specimens under multiaxial loading depended in the case of IHX tubes on the proportion of main deformation direction. Figure 12 compares the time to failure of these IHX specimens (failure means for internal pressure conditions leakage and for axially loaded tubes fracture. Summarizing the available data on the deformation behaviour of high temperature components under upset and postulated emergency conditions, it has been proved that the deformation of a thick-walled RSO tube can be well described by the use of the proposed mathematical model, but this is poor agreement for the thin-walled IHX tube under multiaxial loading, because the influence of surface effects is stronger as with RSO tubes. 2.4.3 Fracture mechanics investigation The fracture mechanics calculation is used to evaluate a detected or postulated flaw in the component. The analysis should result in the proof that at least within the next inspection period or the total service life, the crack will not grow to a critical size. In order to clarify the problems of transferability of fracture mechanics data from small specimens to tubes in the temperature regime above 700°C, creep crack growth and fatigue crack growth experiments have been performed with standard specimens (1"CT, 1/2"CT, 1/2"CCP) and with reformer tubes (120 mm o.d. and 10 mm w.t.) /25/. The tubes were stressed by cyclic tensile load or by static
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14 tensile load and, additionally, superimposed internal pressure. The crack growth was monitored by the DC potential drop technique. Figure 13 shows the fatigue crack growth data for the different standard specimens and steam reformer tubes of Alloy 800 H. The average curves for different specimens geometries are in good agreement. Above an initial build-up range the curves follow the Paris relationship = C • AK T m dN Depending on the change in the material properties produced by different load conditions during pre-cracking, major deviations between tubes and CT specimens were found for the threshold values. Summarizing, the fatigue crack growth in Alloy 800 H can be described by the linear elastic AK approach, up to temperatures of 850°C. Studies of the creep crack growth properties for steam reformer tubes show that the crack propagation is definitely C*-controlled /25-27/'. This means that creep crack growth can be described by the C* integral. 1E-02T
6 7 1E+01
2
3
4
4K1[MP«m1/2J
FIG. 13. da/dN vs. AKj-curves for different specimen geometries of X10NiCrAlTi3220 3.
THE INTEGRITY CONCEPT
A basic principle in every nuclear safety philosophy envisages a staggered system of fission product and activity barriers so that in the case of an accident no release of fission products and radioactive waste occurs /28/. The barriers are of different significance for different types of reactor. Figure 14 shows the fission product barriers of the HTGR. In the light-water reactor, LWR, emphasis is placed on the pressurized enclosure of the primary circuit (reactor pressure vessel, large pipes, etc.) and on the reactor containment. In order to satisfy the stringent nuclear safety requirements, the concept of basic safety was developed for the LWR, particularly for the pressurized enclosure. This ensures a
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15
Grophite shell matrix
coated particle
• Fraction of defective .. particles £ W * operational temperature < 1250 ° C
pyrocarbon
• Increasing permeability >l600°C total particle failure
» 2500 °C • Mechanical stability
FIG. 14. Fission product barriers of the HTGR and retention properties of the fuel element high quality level for the pressure-retaining components of the primary and secondary circuits, ranging from the choice of material, design and calculation up to including fabrication. Corresponding to basic safety, a similar protection concept was to be created for high-temperature reactors /12/. The pressurized containment of the HTGR differs to a greater or lesser extent, depending on the reactor concept, from that of the LWR. In the case of the THTR 300 and HTGR 500, for example, the containment consists of the prestressed concrete reactor pressure vessel, with large cavern closures. The walls of the steam generator tubes are operated at high temperature (above 400°C) so that time-dependent materials behaviour (creep processes) must be taken into consideration. In particular for the heat-exchanging components, it must be remembered that the secondary side of all reactor concepts operates at the higher pressure so that leakages lead only to an ingress of foreign matter (water, process gas), which may affect the reactor core by chemical reactions or reactivity changes. However, the fact that the spherical HTGR fuel element has proved to be very resistant under accident simulation situations has to be borne in mind. On the basis of these facts, an independent HTGR integrity concept is being formulated, taking the special properties of the HTGR into consideration. Having a heterogeneous structure, the pressurized containment of the HTGR primary circuit no longer represents the most important factor in safety considerations. In contrast, the fission product barriers in the fuel element have become more prominent. Even in the case of a considerable leak in the pressurized containment (e.g. rupture of the largest helium duct), not only the fuel elements but also the core geometry remain intact. Coolability is ensured even if a complete depressurization results. An increased release of fission products is - if at all - imagin^ able only in hypothetical core heat-up accidents of medium-sized HTGR in a depressurized state after prolonged downtimes.
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16 The HTGR integrity concept includes the following groups of components: components with a barrier function; components whose failure would impair the barrier function; components whose failure would result in an ingress of foreign matter into the primary circuit. The integrity concept requirements for the component groups are to be differentiated as follows. The components with a barrier function include the fuel element, the pressurized containment of the primary circuit, and to a lesser extent the reactor building, to which a retentive function is attributed. As far as the fuel element is concerned, it is required that the barrier function should be maintained during normal operation and upset conditions. As far as the pressure vessel is concerned, the pressure-retaining and sealing function must be ensured in normal operation. Under upset conditions, large-scale failures must be avoided: the structure must remain intact and leakages must be restricted to such an extent that other components are not unduly influenced. The reactor building does not have such a great significance as the containment of the LWR. Its major tasks involve protection against external impacts (aeroplane crashes, explosion shock waves). Components whose failure would impair the security of the barrier function (e.g. shut-down and residual heat removal components) must remain operational and structurally stable in normal operation and during upset conditions. In the case of components whose failure could result in ingresses of foreign matter, large-scale failure must be ruled out, and it must be possible to keep small leaks under control.
4.
GUIDELINES AND DESIGN CODES
4.1 Status for the High Temperature Regime The objective of a structural design code is to establish a common understanding of basic principles as a guidance for the designers and manufacturers of components and as a help for the users, inspection institutions, and local authorities. Most of the industrial countries have their own tradition and legal status with regard to structural design codes, design regulations, material selection and specifications, and requirements to meet the specific service condition and assurance of high reliability /29,30/. For conventional power stations and pressure vessels there are many different material codes, which differ slightly in the definition of -
-
design pressure; design temperature; time and temperature dependent design stresses; stress and geometrical design safety margins.
The chemical and petro-chemical industries use in the design rules for pressure vessels. Design in the steam and gas turbine field is based on internal specification of the manufacturer. All these rules are based on a considerable background of experience and performance in the past. The planning, construction, commissioning and operation of nuclear facilities must proceed in accordance with a nuclear design code. The
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17 German Atomic Energy Act /32/ and the ordinances based upon it, e.g. the Radiation Protection Ordinance /33/, are also strictly applicable for HTGR (see Fig. 15). The basic safety principles for the environment and facility are laid down in the Safety Criteria for Nuclear Power Plants issued by the Federal Ministry of the Interior (BMI Criteria /34/). They apply to all reactor types, but have been formulated especially for LWR. Only partial aspects of the guidelines of the Reactor Safety Commission /35/ for pressurized water reactors or the accident guidelines /36/ are applicable to HTGR, for example, the design of the reactor pressure vessel of the HTGR Modul.
/Atomic^ Energy \ / Act \ / e.g. Radiation ^ / Protection % / Ordinance ^ / Safety Criteria for % / Nuclear Power Plants ^ / RSK-Guidelinesfor ^ /
/ / / /
PWR/BWR e.g. TRD / AD / DIN / KTA-Rules Specifications for Components and Systems
Law
Ordinances Safety Codes and Administrative Regulations
V \. ^^ ^
Technical Rules Specifications for Licensing
FIG. 15. Legal safety standards and design codes for nuclear power stations The analogous application of BMI Criteria to HTGR largely refers to regular examinations, the shut-down systems, the residual heat removal systems, the pressurized containment of the coolant, and the safety containment. In order to avoid the problems involved in analogous application in the licensing procedure for the THTR 300, a draft of safety criteria for HTGR /37/ and a redraft of the BMI Criteria /38/ which is to include the HTGR, were prepared in the early 80s. Planning principles /39/ have been drawn up especially for the THTR 300 licensing procedure which were approved by the relevant authorities involved in the licensing procedure. Experience with this interim solution has been on the whole positive. However, this positive experience should be converted into clear, generally valid criteria, rules and guidelines. The rules of the Kerntechnischer AusschuB (KTA - Nuclear Safety Standards Commission), which pay little attention to HTGR, are used for the detailed design of nuclear power stations. In the licensing procedure, aspects of the HTGR not included in the nuclear design code are laid down in specifications which may be orientated towards the conventional design
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18 code, e.g. towards DIN standards^-) or AD codes of practice^). For nuclear power plants with service temperature ranges in which time-dependent properties and creep must be considered, the American ASME Code, Case N 47 /40/, the French "RCC-MR Code" /41/ and the codes recently developed in Great Britain /42/ have been applied, mainly for the liquid metal fast breeder reactors. These foreign design codes have to some extent been consulted. The KTA design code is well advanced. It contains nuclear rules solely applicable for LWR, rules applicable to all types of nuclear power plant, and a few rules specially compiled for HTGR facilities. The latter refer to the thermal and thermohydraulic reactor design. In order to accelerate the inclusion of the HTGR in the KTA design code, the KTA subcommittee "High Temperature Reactors" was established in 1984. The first step undertaken by this sub-committee, namely the examination of existing rules and draft rules for their applicability to the HTGR, has been completed.
4.2 Research Project "Design Criteria" The lack of nuclear rules and guidelines described above led to delays in the licensing procedure for the THTR 300. There is therefore a need for KTA design rules for the wide range of typical HTGR components. Endeavours aimed at forming the basis for an HTGR design code stretch back to 1979 /43/. In a special research programme sponsored by the Ministry of the Inferior /44/, the experience gained from the PNP (Nuclear Process Heat) and HHT (HTGR with direct cycle helium turbine) projects was considered with the emphasis on metallic components operating at temperatures above 800°C. At the same time, a probabilistic risk study /45/ was carried out to provide a survey of the HTGR accident topology, which revealed the safety properties of the HTGR. On the basis of this preliminary work, a comprehensive research programme was initiated in 1984 on "Design Criteria for High Temperature Metallic and Ceramic Components, and for Prestressed Concrete Pressure Vessels of Future HTGR Plants" /12,46/, and sponsored by the Federal Ministry of Research and Technology. The above mentioned research programme was divided into four areas:
-
Part Part Part Part
A: B: C: D:
Technical Safety Boundary Conditions Metallic Components Prestressed Concrete Reactor Pressure Vessel Graphitic Reactor Components
In Part A, the HTGR integrity concept is formulated, which is based on the special properties of the HTGR. The role of the fuel elements that remain stable up to the highest temperatures was especially important. Proposals for the safety classification of components were developed taking into account the radiological effects of a failure.
DIN = p_eutsches Institut fur Normen AD = Arbeitsgemeinschaft Druckbehalter
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19 Part B deals with metallic components for service at temperatures above 400°C, based on information from the initial project /44/. This opened up new aspects since the range of applications for the HTGR includes components operating at temperatures up to 950°C. The time-dependent behaviour of the materials has to be considered here. An analysis of the mechanical behaviour requires detection methods no longer covered by elastic calculation procedures. Inelastic verification methods are required to evaluate the various loading conditions, which means a procedure of design by analysis. The decisive phenomena have been recognized, and methods for their treatment indicated. As mentioned above, in other countries there are design codes going beyond LWR applications, in which components operating at temperatures above 400°C are treated. However, the ASME-CC-N-47 (USA) and the RCC-M (France) are designed to cover fast breeder reactor components and therefore only involve partial aspects of HTGR components. HTGR plants of limited power capacity up to about 200 MWel can still make use of steel pressure vessels which may be designed for the most part according to the LWR pressure vessel principles. For larger capacity plants, a prestressed concrete pressure vessel is more advantageous (Part C of the programme). Its special feature is the separation of enclosure and pressure-retaining functions. Its safety benefits have already been presented. In the formulation of design codes, extensive regulations in the civil engineering sector may be utilized, and in particular experience with the British advanced gas-cooled reactors (AGR). The state of the art concerning graphitic and ceramic core components (Part D of the programme) permits the formulation of a design code since studies undertaken in the sixties with respect to stress calculations, influence of neutron irradiation and corrosion by foreign media in accident and normal operating situations have been comprehensively evaluated. These investigations carried out in a total of 24 sub-groups have been documented in more than 200 individual contributions. The programme ran from 1984 to early 1988. From these investigations, the following HTGR design code titles were derived:
-
Metallic HTGR components (KTA 3221) Safety requirements for the design of prestressed concrete pressure vessels for HTGR (KTA 3231) Ceramic components in an HTGR pressure vessel (KTA 3232)
The prepared reports /I2/ have been submitted to the Nuclear Safety Standards Commission which will decide in the near future whether drafting of the code rules may begin. A workshop was held in early 1989 presenting the results at an international level /46/. REFERENCES HI
HI
13/
AVR - 20 Jahre Betrieb, ein deutscher Beitrag zu einer zukunftsweisenden Energietechnik, Symposium Aachen, 17./18.5.1989, Verlag des Vereins Deutscher Ingenieure, VDI Berichte 729, IS3N 318-090729-0 Kirch, H., Brinkmann, H.U., Nabielek, H., Bestrahlungserprobung von HTGR-Komponenten, Stand und zukiinftige Anforderungen, Proc. The HFR Petten, Coll., Prospects and Future Utilization, Petten, April 20.111.1.1989 Nickel, H., Long-term testing of HTGR fuel elements in the Federal Republic of Germany, KFA-Jiil-Spez. 376, 1986
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/5/
/6/
/7/
/8/
/9/
/10/
/11/
/12/
/13/
/14/
/15/
/16/
/17/
/18/
/19/
"Special Issue of Nuclear Technology on Coated Particles", ed. R.G. Post, K. Wirtz, T.D. Gulden, H. Nickel, 35, No. 2 (1977) 205-573 Schulze, R.E., Schulze, H.A., Rind, W., Graphitische Matrixwerkstoffe für kugelförmige HTR-Brennelemente, Ergebnisse der Materialentwicklung und der Bestrahlungserprobung, KFA-Bericht Jül1752, 1981 Nabielek, H., Schenk, W., Heit, W., Mehner, A.W., Goodin, D.T., The performance of high-temperature reactor fuel particles at extreme temperatures, Nucl. Techn. JH (1989) 62-81 "Special Issue of Nuclear Technology on High Temperature Gas Cooled Reactor Materials", ed. R.G. Post, K. Wirtz, H. Nickel, P.L. Rittenhouse, T. Kondo, 66 (1984) 11-702 Nickel, H., Ennis, P.J., Schubert, F., Schuster, H., Qualification of metallic materials for application in advanced high temperature gas-cooled reactors, Nucl. Techn. 58 (1982), 90-106 Nickel, H., Schubert, F., Schuster, H., Evaluation of alloys for advanced high temperature reactor systems, Nucl. Engin, and Design 78 (1984) 251-265 Diehl, H., Bodmann, E., Alloy 800: New stress rupture and creep data for pressurized components in high-temperature reactors, Proc. 14. MPA-Seminar, 6./7. Okt. 1988 Bodmann, E., Diehl, H., Blume-Firla, I., Demus, H., Service conditions and relevant properties of HTGR metallic materials, IAEA Specialists' Meeting on High Temperature Materials for Gas-Cooled Reactors, Cracow, June 20-23, 1988 Design criteria for high temperature metallic and ceramic components and for prestressed concrete pressure vessels of future HTGR plants, Endbericht zum Verbund-Forschungsvorhaben des BMFT: Auslegungskriterien für hochtemperaturbelastete metallische und keramische Komponenten sowie des Spannbeton-Reaktordruckbehälters zukünftiger HTR-Anlagen (August 1988) Meurer, H.P., Breitling, H., Dietz, W., Influence of hold-time and strain rate on the LCF behaviour of Alloy 617 at 950°C, in K.T. Rie (ed.), Proc. 2nd Int. Conf., Low Cycle Fatigue and Elasto-plastic Behaviour of Materials, München, 7.-11.9.1987 Diehl, H., Sonsino, С М . , Mean stress and notch effect in HCF experiments on Alloy 617 between 600 and 950°C, Third International Conference on Biaxial/Multiaxial Fatigue, April 3-6, 1989, Stuttgart Diehl, H., Blume-Firla, I., Mergler, W., Alloy 800: Low cycle fatigue curves as the basis for design against fatigue of HTGR components, Proc. 14. MPA-Seminar, 6./7. Okt. 1988 Rao, K.B.S., Meurer, H.P., Schuster, H., Creep-fatigue interaction of Inconel 617 at 950°C in simulated nuclear reactor helium, Mat. Sei. and Eng. A104 (1988) 37-51 Bieniussa, K., Breitbach, G, Over, H.H., Penkalla, H.J., Seehafer, H.J., Life-time and creep ratcheting calculation of two typical HTGR components, 2nd International Seminar on "Standards and Structural Analysis in Elevated Temperature Applications for Reactor Technology", Venice, 14-17 Oct. 1986 Zottmaier, R., Over, H.H., Schubert, F., Nickel, H., Untersuchungen zum Kriechratchetingverhalten von Rohrproben, KFA-Jül-2019, Sept. 1985 Breitbach, G., Achmed, К., Schubert, F., Nickel, H., Creep collapse of tubes at high temperatures under external pressure, Proc. 10th Intern. SMIRT-Conf.,* Anaheim, Cal., August 1989 E—2
21
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/21/
/22/
/23/
/24/ /25/
/26/
/27/
/28/
/29/ /30/
/31/ /32/
/33/
Quadakkers, W.J., Schuster, H., Corrosion of high temperature alloys in the primary circuit of high temperature gas cooled reactors, Part 1 : Theoretical background, Materials and Corrosion, 16 (1985), 141-150 Quadakkers, W.J., Part 2: Experimental Results, Materials and Corrosion 16 (1985), 335-347 Thiele, B.A., Schubert, F., Bendick, W., Weber, H., Improvement of the resistance to neutron-induced high temperature embrittlement of austenitic steel X8CrNiMoNb 16 16 by thermomechanical treatments, Materials for Nuclear Reactor Core Applications, BNES, London, 1987, p. 141-146 Penkalla, H.J., Schubert, F., Nickel, H., Experimentally verification of the application of creep laws on multiaxial and complex loading conditions on Alloy 617, Proc. 10th Intern. SMIRT Conf., Anaheim, Cal., August 1989 Schubert, F., Rodig, M., Nickel, H., Mult.i,axial loading tests of high temperature reactor components, Nucl. Engin. and Design 98 (1987) 359-366 Bodmann, E., Breuer, H.J., Raule, G., Rodig, M., Material behaviour under complex loading, in cit. 7, 667-674 Rodig, M., Kienzler, R., Nickel, H., Schubert, F., Fatigue and creep crack growth in methane reformer tubes at temperatures above 700°C, Nucl. Engin. and Design 108 (1988) 467-476 Riedel, H., Wagner, W., Creep crack growth in NIMONIC 80 A and in 1Cr-1/2Mo-steel, Proc. 6th Int. Conf. on Fracture (ICF6), New Delhi, 1984 Hollstein, T., Kienzler, R., Fracture mechanics characterization of crack growth under creep conditions, J. of Strain Analysis ^3 (1988) 87-96 Nickel, H., Hofmann, K., Wachholz, W., Weisbrodt, I., The helium-loaded high-temperature reactor in the Federal Republic of Germany, safety features, integrity concept, outlook for design codes and licensing procedures, Proc. IAEA Symposium on Regulatory Practices and Safety Standards for Nuclear Power Plants, Munich, Nov. 7-10, 1988, IAEA-SM-307/31 "Design of High Temperature Metallic Components", Applied Science Publishers, London, ed. R.C. Hurst (1984) Nickel, H., Schubert, F., Breitbach G., Development of structural design codes for helium cooled high temperature reactors in the FRG, Proc. Third Intern. SMIRT Post Conference, Seminar on Construction Codes and Engineering Mechanis, Los Angeles, August 21/22, 1989 Proceedings First Intern. SMIRT Post Conference Seminar on Construction Codes and Engineering Mechanics, Paris, August 1985 Gesetz iiber die friedliche Verwendung der Kernenergie und den Schutz gegen ihre Gefahren (Atomgesetz) - Act on the peaceful use of nuclear energy and the protection against its hazards (Nuclear Energy Act) Publication of the revised text of the Atomic Energy Act of 15.07.1985 with the amendment on the Federal Law Gazette I no. 8 of 21.02.1986 Verordnung iiber den Schutz vor Schaden durch ionisierende Strahlen (Strahlenschutzverordnung (StrlSchV)), Regulation on protection against damage caused by ionising rays (radiation protection ordinance) of 13.10.1976, last amended on 20.05.1981, Bundesgesetzblatt I, No 19.
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/35/
/36/
/37/
/38/
/39/
/40/
/41/ /42/ /43/
/44/
/45/
/46/
Der Bundesminister des Inneren, Sicherheitskriterien fur Kernkraftwerke (Federal Minister of the Interior, Safety criteria for nuclear power plants, Bundesanzeiger No 206, 03.11.1977 RSK-Leitlinien fur Druckwasserreaktoren, RSK guidelines for pressurised water reactors, 3rd edition of 14.10.1981, as amended on 21.03.1984, Bundesanzeiger No 104, 05.06.1984 Der Bundesminister des Inneren, Leitlinien zur Beurteilung der Auslegung von Kernkraftwerken mit Druckwasserreaktoren gegen Storfalle im Sinne des § 28 Abs. 3 StrlSchV - Storfall-Leitlinien, Federal Minister of the Interior, Guidelines on the assessment and design of nuclear power plants with pressurized water reactors with regard to accidents in the meaning of Article 28, para. 3 StrlSchV, Accident Guidelines, Bundesanzeiger No 245, 31.12.1983 TUV-Arbeitsgemeinschaft Kerntechnik West, Sicherheitskriterien fur Anlagen zur Energieerzeugung mit gasgekiihlten Hochtemperaturreaktoren; Safety criteria for energy producing plants equipped with gas-cooled high-temperature reactors, Draft September 1980, Essen, 15.09.1980 Der Bundesminister des Inneren, Sicherheitskriterien fur Kernkraf twerke; Federal Minister of the Interior, Safety criteria for nuclear power plants, Draft as of 21.05.1984 HKG and BBC/HRB, Planungsgrundlagen fur die Errichtung des 300 MWel-THTR-Prototyp-Kernkraftwerks Hamm/Uentrop as of 22.06.1976 ASME Boiler and Pressure Vessel Code, Case N 47-21, "Class 1 components in elevated temperature service, Division 1", American Society of Mechanical Engineers (1981) R. Noel, The French RCC-MR-Code "Rules for design and construction of LMFBR components", in /31/ Rose, R.T., Tonkins, B., Townley, C.H.A., Recent UK research and the development of high temperature design methods, in /31/ Nickel, H., Schubert, F., Schuster, H., Very high temperature design criteria for nuclear heat exchanger in advanced high temperature reactors, Nucl. Engin. and Design 9h_ (1986) 337-343 Nickel, H., et al., Erarbeitung von Grundlagen zu einem Regelwerk iiber die Auslegung von HTR-Komponenten fiir Anwendungstemperaturen oberhalb 800°C, Jul-Spez.-248, Marz 1984 KFA-ISF/GRS, Sicherheitsstudie fur HTR-Konzepte unter deutschen Standortbedingungen (Safety Study for HTGR Concepts under German Site Conditions), Jiil-Spez.-136, vol. 1, 1981 Proc. of the Workshop on Structural Design Criteria for HTR Editors: G. Breitbach, F. Schubert, H. Nickel, KFA-Bericht, JiilConf-71, April 1989
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XA0101506
Research and Development Programs for HTGRs in JAERI
Isoharu NISHIGUCHI and Sinzo SA1TO
Department of HTTR Project Japan Atomic Energy Research Institute
Presented at the IAEA Technical Committee Meeting on Gas-Cooled Reactor Technology, Safety and Siting Dimitrovgrad, USSR, June 21-23, 1989
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-•'
•
Since 1969, JAERI has programs
A B S T R A C T .
.•
•
.-..:...•
conducted research and development
for High-Temperature
Gas-Cooled
Reactors
High Temperature engineering Test Reactor (HTTR), first High Temperature Gas Cooled Reactor
(HTGR).
(R&D)
And
the
which will be
the
(HTGR) in Japan, is under
licensing process now C1J. • In
this
following
paper,
some
of
results
of
R&D
are
outlined
fields which are closely connected with
in
the
the HTTR design,
that is, i) fuel, ii) nuclear design, iii) thermal-hydraulic design, iv) graphite structure and v) high temperature metal structure. In
the
performed fuel
field
to develop
(cpf).
of
fuel,
extensive
the fabrication
investigations
have
technology of coated
In parallel, data of coated
been
particle
fuel particle failure and
fission product release in in- and ex-reactor experiments as well as mechanical properties data
were obtained and irradiation tests have
been
Gas
done
using
the
Oarai
Loop
No.l
(OGL-1)
to
verify
the
integrity of mass-produced fuel. Concerning conducted using
the
nuclear
design,
critical
experiments
the Very High-Temperature Reactor Critical
were
Assembly
(VHTRC). Also carried out were hydrodynamical and thermal experiments using the Helium Engineering Demonstration Loop (HENDEL). On
the graphite structures which compose
the reactor
internals,
design criteria have been developed based on ASME B5PV Code Section III Div.2, subsection CE and design data have been accumulated on
a
domestic graphite material. High
temperature
subjects of RSD for HTGRs.
metal
structure
is
also
one
of
major
Hastelloy XR, which is a modified version
of Hastelloy X, was developed and various tests have been conducted
E-3
which
include creep
tests, creep-fatigue
design criteria and allowables. Component
tests, etc. to
establish
tests of the Intermediate
Heat Exchanger (IHX) have been also performed.
1. Introduction The High Temperature engineering Test Reactor (HTTR), be
which will
the first High Temperature Gas Cooled Reactor (HTGR) in Japan, is
under licensing process now [1]. The HTTR consists of a core of 3OMWt, a main cooling circuit, an auxiliary cooling circuit and related systems as shown schematically in Fig.l. reactor
Table 1 summarizes major design parameters of the HTTR. The
pressure
vessel
is 13.2m high
and
5.5m
in diameter
and
contains the core, graphite reflectors, core support structures and the core restraint mechanism as shown in Fig.2. This
paper
presents
a
brief
overview
of
RfiD programs
in
connenction with the HTTR design.
2. Fuel design As is shown in Fig.3, a fuel element assembly of the HTTR is made up of fuel rods and a hexagonal graphite block.
The fuel consists of
TRISO coated particles of low enriched uranium oxide whose average enrichment
is
about
6%
particles are dispersed
and
the kernel
diameter
is
600/im.
The
in the graphite matrix and consolidated
to
form a fuel compact. These compacts are contained in a sleeve to form a fuel rod and these fuel rods are contained within vertical holes of a graphi te block. In the fuel design of HTGRs, it is very important to retain fission
products
within
particles
so
that
their
primary coolant may not exceed an acceptable level.
release
to
the
From this point
of view, in the HTTR safety criteria, the failure fraction- of asfabricated
fuel
coating
layers is limited
to 0.2 '/. and
the fuel
temperature is limited below 1495°C under normal operating conditions and below 1600°C under abnormal transient conditions in order to avoid E-3
additional fuel failures d u r i n g o p e r a t i o n . Extensive fabrication
investigations
have
been
performed
t e c h n o l o g y of coated p a r t i c l e fuels
to
develop
(cpfK
the
In p a r a l l e l ,
coated fuel p a r t i c l e f a i l u r e and fission p r o d u c t r e l e a s e b e h a v i o r has been
investigated
out-of-pile
by
in-pi Ie, e.g., Oarai Gas Loop No.1
e x p e r i m e n t s . D a t a on
fuel
properties
( O G L - 1 ) . and
have beerr o b t a i n e d
for the t h e r m a l - h y d r a u l i c d e s i g n and safety a n a l y s i s .
In
the
performed
OGL-1,
irradiation
mainly
to
tests
investigate
of the
the
HTTR
:
fuel
integrity
have
under
been normal
operating conditions [2]. The OGL-1
is a h i g h - p r e s s u r e
in-pile gas
loop installed
r e f l e c t o r zone of the Japan M a t e r i a l T e s t i n g R e a c t o r ( J M T R ) . 1
the
The O G L -
was put into s e r v i c e o p e r a t i o n in March 1977 and has b e e n o p e r a t e d
for more than 2 2 , 0 0 0 h o u r s . A flow diagram of Fig.4. of
in
the OGL-1- is s h o w n
The c o o l a n t gas p r e s s u r e is 3MPa, the m a x i m u m gas
1000°C,
thermal
lxlO13/cm2s,
and
fast
neutron
/cm2s
and
respectively.
The c o n d i t i o n s and r e s u l t s of irradiation conducted
temperature
are 6 x l O 1 3
fluencies
in
in the OGL-1 are listed
in Table 2 [ 3 ] .
r e l e a s e rate to b i r t h rate ( R / B ) of in F i g . S [ 3 ] .
As is o b s e r v e d ,
s h o w i n g a good p e r f o r m a n c e and
tests of the H T T R fuel
88
The c h a n g e of the
K r d u r i n g i r r a d i a t i o n is s h o w n
the R/B is almost c o n s t a n t that there is no s i g n i f i c a n t
throughout increase
of the fuel f a i l u r e f r a c t i o n . The
fuel
investigated
behavior
under
accidental
conditions
by c o n s t a n t - and r a m p - t e m p e r a t u r e
has
heating
also
been
tests on
the
irradiated c o a t e d p a r t i c l e s [ 4 ] .
3. Nuclear des\gn The
flow
of
the
HTTR
nuclear
calculation
is
shown
in
Fig.6
schematically:;
a computer
neutron spectrum
cods,
of a fuel cell
DELIGHT
is used
to' obtain the
and to produce group constants
based
on the nuclear data from E N D F / B - 3 , - 4 .
The calculation of a control
rod
whi ch- i s ; based
cell
is performed
dimensional
b y the TWOTRAN-2
transport ••theory.
CITATION-IOOOVP,:which
on'" i the two-
is a vectorized
version of the C I T A T I O N . [ 5 ] is used: to cal cul ate -the • three dimensi bnal Gore performance.
:- : , ••
••- Accuracy- of the nuclear experimental
data
obtained
.'
codes
using
Assembly
(SHE)[63
and the Very
Assembly
(VHTRC)
which
';?.•••
has been
• • • •
• . •. •
examined • by . various
the S'emi-Homogeneaus High
Temperature
was reconstructed
from
specifications of VHTRC are shown in Table 3.
.
-Experimental
Reactor
of a movable
half
and- a fixed
half
Critical
:the: SHE. K
A schematic view of the VHTRC;is shown -in Fig. 7. composed
,. ' .;-.
both
Main :?
;
The VHTRC is
are made up
of
hexiagonal graphite ^bl ocks in the horizontal position and insented in holes of ;blocks are fuel rods, control, rods, safety rods and heaters. A fuel rod contains fuel compacts in which TRISO coated par t i d e s
•
.
Major.items of code verification are
.,
- :
a) the effective multiplication factor, b ) the c o n t r o l
.
rod reactivity worth,.
• _ • ..
c) the burnable poison rod reactivity worth, d) the power distribution and
.. • . • •
e) the temperature c o e f f i c i e n t . As factor
an e x a m p l e , obtained
of. the
codes
core
diameter
.
ranging
from
.• ,.., effeetive : multiplication in the VHTRC is
fuels are combined wi th
104
results are in very good agreement with.the experimental The
errors
estimated
by these
;,
•
and experiments
shown in Fig.8 in which 2, 4 and 6 % enriched effective
...... :
.
a comparison
by the. above
-^
;
comparisons
between
calculated
results. experimental
E-3
and calculated results are reflected in the nuclear design of the
HTTR. 4. Thermal-hydraulic design As pointed out in Section 2, in the HTTR safety criteria, the maximum fuel temperature is limited under the normal operation and the abnormal transients condition.
Thus, the core should be designed so
as to maintain the sufficient core flow rate and to keep the maximum fuel temperature as low as possible, with the structural features of the core and fuel are taken into account. hydraulic
design
is
shown
in
Fig.9.
The flow of the thermalFlow
and
temperature
distributions of coolant in the core are calculated by the FLOWNET in which the flow network model is employed.
A flow network consists of
branches and nodes which simulate various flow paths in the core such as main flows, leakage flows through the gaps between the permanent reflectors and between FLOWNET
and
the
power
fuel blocks.
Based on the result of the
distribution
obtained
by
the
nuclear
calculation, the temperature of the fuel is estimated by the TEMDIM code, in which compacts and a graphite sleeve of a fuel rod are modeled
by coaxial
cylinders and
the maximum fuel
temperature is
calculated with systematic and random errors considered. Various R8D programs have been performed to verify the thermalhydraulic design.
Appearing below is a research program in which the
leakage flow rate within the core support structure is estimated using the
Helium
experimental
Engineering
Demonstration
Loop
(HENDEL)
results are compared with the analytical
and
the
ones of the
FLOWNET. The HENDEL was constructed to perform large scale demonstration tests of high-temperature components for the HTTR in March 1982. A schematic flow diagram of the HENDEL is shown in Fig.10.
The HENDEL
consists of the Mother (M), Adapter (A) and Test (T) sections. The Mother and Adapter (M+A) section circulates helium gas at a flow rate of 4 kg/s, a pressure of 4 MPa and at a maximum temperature of 1000°C. The M+A section has been operated for more than 10,000 hours since 1982.
The test section is made up of the Fuel Stack Test Section (Ti
test section)[7,8] and the In-core Structure Test Section section)C9].
(T2 test
A general view of core bottom structure of the T2 test
section is shown in Fig.11.
The Ti test section has been in operation
since March 1983 and the T2 test section since June 1986 and various data have been obtained in the fields of
thermal-hydraulic
and high
temperature structural design. Shown
in Fig.12 are experimental
results of leakage flow rate
from outside of the permanent reflector blocks into the hot plenum of the T2 test section as a function of the pressure difference between the hot plenum and the outer side of the permanent blocks at inner and outer
temperatures
results
obtained
950°C by
and
400°C,
the FLOWNET
respectively.
are shown
also
The
and
analytical
an
excellent
agreement between experimental and calculated results is obtained. We note references C7-10] for the results of R&D in the field of the thermal-hydraulic design.
5. Graphite structure design The reactor core is composed of fuel blocks, control rod guide blocks and replaceable reflector blocks and is supported by the core support structures of graphite and metal.
They are bound by the core
restraint mechanism as shown in Fig.2. A fuel block is graphite hexagonal right prism as shown in Fig.3 and is made of the IG-110, isotropic fine-grade graphite.
Control rod
guide blocks and replaceable reflector blocks have the same external shape as the fuel blocks and are also made of the IG-110. E-3
The
graphite
reflector the
b l o c k s , hot
thermal
grade
core
support
plenum
insulation
medium-to-fine posts
support
is
made
b l o c k s , seals, k e y s ,
b l o c k s . They are made
grained
which
structure
molded
are made
of
of
graphite
IG-110 and
up
of
permanent
support
posts
the PGX,
except
structural
seals,
the middle
and
keys
layer
and
of
the
insulation which is made of carbon. In
order
to
assess
the
integrity
of
the
above
graphite
structures, a design code has been developed and design data have been accumulated on In
the
components
the IG-110 .
code and
the graphite
core
support
components
components
and
are
categorized
different
into
core
allowances
adopted in view o£ the differences of their functions as summarized Table 4.
are in
•
The design code is based mainly on the ASME C E Code, however, the ASME
CE
buckling
Code
is
modified
limit
and
oxidation
regarding
the
ef f ects Cl 1]-.
bi-axes
failure
A comparison
theory,
between
the
JAERI code and the ASME code is given in Table 5. The strength
of oxidized
graphite, for example, is specified
in
the code as f o l l o w s : The
region
should
be
region
where
based
on
tensile
where
deemed the
as
having
burnoff
the reduced strength
amount
no
oxidation load
is below
strength
decrease
of
exceeds
carrying
the
capacity
8 0 % , the allowance
which
of grade
is obtained IG-110
8 0 % , burnoff and is
for
calculated
by experiments.
is shown
the
as a function
The of
burnoff in Fig.13 [12] . .
the
In order
to develop
research
work
temperature properties, so on.
Young's
has
the design been
modulus,
low cycle fatigue
criteria
carried impact life,
We note ref.Cll] for d e t a i l s .
out
and
on
strength,
irradiation
their
allowables,
including fracture
the
high-
mechanics
creep properties
and
6. High temperature metal
structure
High temperature metal structure is also one of major subjects of R&D for HTGRs. In the HTTR, a He/He intermediate heat exchanger
(IHX)
of•lOMWt is used as shown in Fig.l and the heat tubes and the central gas duct of the IHX constitute a part of the pressure boundary of the primary coolant at a temperature of about 900°C. A bird's eye of the IHX is shown in Fig.14.
Hastelloy X R C 1 3 ] , which is a modified version
of Hastelloy X, is used for the very-high
temperature
structures
in
the IHX. A structural design code for the HTTR was developed by the JAERI based
on
both
the Elevated-Temperature
Monju
( E T S D G H 1 4 ] and
fatigue, fracture
Structural
the ASME Code Case N-47 and
Design
Guide
for
tests for creep,
toughness, corrosion and other critical
items have
been undertaken to accumulate design data of Hastelloy XR [ 1 5 ] . Component
tests
of
the
Intermediate
Heat
Exchanger
(IHX)
have
been also performed. In the
IHX,
the pressure of both primary and secondary gas is
about 4 MPa and their pressure difference is very small in the normal condition.
If the secondary gas were lost by an accident, however,
the tubes and header would be subjected to the external primary gas of 4
MPa
and
the
possibility
of
the
creep
collapse
would
arise.
Therefore, the experiments have been and being performed to evaluated the integrity of the tubes subjected
to external pressure [16] and a
simplified method for the prediction has also developed [ 1 7 ] . Another important potential failure mode of the IHX is the creepfatigue
failure
induced by cyclic relative displacement
between
the
hot header and tubes due to the difference of thermal expansion. In order the structural
to verify the design criteria for the creep-fatigue scale, a IHX structural
model
has been made and
in the
E-3
test apparatus will be completed in July 1989. a hot
header,
installed
8 helical
tubes and 8 connecting
in an electrically heated retort.
shown in Fig.15.
The model consists of
The
tubes, which
are
test apparatus
is
Each connecting tube is attached to the hot header
in one end horizontally and the other end is extended to the upper loading grid vertically.
These tubes are subjected to both repeated
loading by the actuater and an internal pressure at a temperature of 950 °C.
Deformation and the life obtained by the experiment will be
compared
with predictions by inelastic analysis and the JAERI code.
7. Concluding Remarks Some of RSD results were provided in the field of fuel, nuclear and thermal-hydraulic design, graphite and high temperature structures in connection with the HTTR design.
RSD of the HTGR, however, extends
over far wider range than outlined in the above. An overview of the HTGR RSD program may be found in the annual report of JAERI [18].
10
References Cl] S. Saito, 'Present Status of HTGR Development Program in Japan,1 presented
at
the
11th
international
Conference
on
the
HTGR,
Dimitrovgrad, USSR, June 19-20, 1989. [2] K. Fukuda et al., 'HTTR Fuel Irradiation Performance and Fission Product Behavior under Normal and Transient Reactor Conditions,' presented
at
the
JAIF-GKAE
Seminar
on
Fuel
Elements
and
Fuel
Composition of HTGR, Tokyo, Japan, Oct. 20-22,, 1987. C3] K. Fukuda, Private communication, 1989. [4] T. Ogawa et al., 'HTGR Fuel Behavior under Accident Conditions,' presented
at
the
JAIF-GKAE
Seminar
on
Fuel
Elements
and
Fuel
Composition of HTGR, Tokyo, Japan, Oct. 20-22, 1987. [5] M.
Takano
Neutronic
et
al.,
Design
'Analysis
Codes
for
of
SHE Critical
Experimental
Very
Experiments High
by
Temperature
Reactor,' J. Nuclear Science and Technology Vol.22, pp.358-370, 1985. [63 T.B. Fowler and D.R. Vondy, 'Nuclear Reactor Core Analysis Code, CITATION,' ORNL-TM-2496, 1969. [7] S. Maruyama et al., 'Experimental Hydraulic
Performance
HENDEL single-channel
of
the Fuel
Studies on Stack
of
the Thermal
and
the VHTR, Part
I:
tests with uniform heat flux,' Nucl. Eng.
Des. Vol.102, pp. 1-9, 1987. [83 S. Maruyama
et al., 'Experimental
Hydraulic Performance of HENDEL multi-channel
the Fuel
Studies Stack of
on
the Thermal
and
the VHTR, Part II:
test rig with twelve fuel rods,1 Nucl. Eng.
Des. Vol.102, pp. 11-20, 1987. [9] K. Kunitomi Supporting
et al., 'Thermal and Hydraulic Tests the Development of
in HENDEL T 2
the Core Bottom Structure of the
High Temperature Engineering Test Reactor (HTTR),1 Nucl. Eng. Des. Vol.108, pp. 359-368, 1988.
11
E-3
[10] S. Maruyama et al., 'Verification of In-Core Thermal and Hydraulic Analysis Code FIOWNET/TRUMP for the High Temperature Test Reactor Topical
(HTTR) at JAERI,' to appear at Fourth
Engineering International
Meeting on Nuclear Reactor Thermal-Hydraulics,
Krlsruhe,
FRG, Oct. 10-13, 1989. [11]
T.
Iyoku
Components
and of
S.
Shiozawa,
HTTR,'
'Design
presented
at
Criteria
the
Workshop
for on
Graphite Structural
Design Criteria for HTR, Julich, FRG, Jan. 31-Feb. 1, 1989. [12] S. Yoda et al., 'Effects of Oxidation on Tensile and Compressive Deformation
Behavior
for
Nuclear
Grade
Isotropic
Graphite,'
International Carbon Conf. 84, Bordeaux France, 1984. [13] M. Shindo and T. Kondo, Proc. Conf. on Gas-Cooled Reactors Today, Bristol, UK, Vol.2 ,p.l79, 1982. [14] K.
Ii da et al . , 'Simplified Analysis and D e s i g n ; f o r
Temperature Components
Elevated
of Monju,' Nucl. Eng. Des., Vol.98, pp.
305-317, 1987. [15] Y. Muto-et al . , 'The present status of Research and Development works for the Preparation of the High Temperature Design Code,' presented at the Workshop on Structural Design Criteria for HTR, Julich, FRG, Jan. 31-Feb. 1, 1989. [16] I. Nishi-guchi, 'A study on Creep Deformation Behavior of the Heat Tubes of a He/He Heat Exchanger Subjected to External Pressure at High Temperature,' presented at ASME B&PV Code Task Force on Very High Temperature Design, New York, Feb. 12, 1987. [17] I. Nishiguchi et al . , 'A Simplified Method for Predicting Creep Collapse
of
a Tube
under
External
Pressure,'
submitted
for
publication. [18] Dept. of Power Reactor Projects, JAERI, Present Status of HTGR Research & Development, JAERI, 1989. 12
Table 1 Major design parameters of the HTTR
Thermal power
30 MW
Outlet coolant temperature
850°C/950°C
Inlet coolant temperature
395°C
Fuel
Low enriched UO2
Fuel element type
Prismatic block
Direction of coolant flow
Downward-flow
Pressure vessel
Steel
Number of main cooling loop
1
Heat removal
IHX and PWC (parallel loaded)
Primary coolant pressure
4 MPa
Containment type
Steel containment
Plant lifetime
20 years
14
E-3
Table
Fuel Assem. Number
2
Number of compacts
Irradiation
test
conditions
Fast neutron fluence (Max.)
Burnup
[n/m*.
tMWd/t]
andresults [3]
Fuel
temp.
Ave.
E>29fJ]
1
54 60
2 3
20 60 60 20
4 5 6 7 8 9» 10»
60 20 60
1.2xlO24 2.0xl02< 23
8.9xlO
2
2.3xlO < 3.8xlO24 4.1xlO23 2
1.6xlO < 1.2xlO2< 2.8xlO2<
20
» f a b r i c a t e d by s c a l e - u p
5,800 8,500 4,800 18,000 30,000 3,900 12,000 9,100 24,000 26,000
facilities
par t i d e
failure
(OC)
Max.
Coated
after Irr.
fraction before Irr.
1490 1520
1190 1220
2.7x10-5
3.7xlO"6
1.4xlO-5
1.0xl0"<
1370
1240
7.1xlO-<
8.5xlO-<
1400 1410 1530 1430
1120 1230 1410
7.3x10-5
l.lxlO"4
3.1x10-3 9.6x10-5
1.9xlO-3 5.0x10-5
1230 1300 1260 1280
7.1x10-5
2.7xlO-5
1,. l x l - 0 " <
5.9xlO-s
1.4X10-3
8.7x10"*
4.9xlO-<
2.5x10-4
1440 1390 1550
T a b l e 3 M a i n s p e c i f i c a t i o n of t h e VHTRC
Specifications
Hems Core Shape
Horizontal Hexagonal Prism Sprit in Two Half Assemblies on Tables
Core Dimensions
Core Across, 2.4 m Axial Length, 2.4 m
Maximum Thermal Power Output
10 W
Maximum Core Temperature by Electric Heating
21Q'C (800*C for Single Fuel Rod)
,
Fuel Element Type
Fuel Compact of Coated Particles Dispersed in Graphite Matrix
Enrichment
2,4 and S w t %
Maximum U-235 Loading
10.4 kg
Moderator and Reflector
Hexagonal Prism Graphite Blocks
Control Rod
Cd Cylinder Sheathed with SUS Motor and Pneumatic Drive, 2 sets
Safety Rod
Cd Cylinder Sheathed with SUS Pneumatic Drive, 8 sets
Inslrumentation
Six Neutron and One Gamma Ray Monitors
16
Table 4
Comparison between the core and the core support components of the HTTR
Core component
Main component
Replaceability Irradiation effects Design life
I
• • • •
Fuel block Graphite sleeve Control rod guide block Replaceable reflector block Routine Major 3y
Core support component • Hot plenum block • Permanent reflector block • Core support floor thermal insulation layer • Support' post Difficult Negligible 20y
Table 5 Comparison between JAERI Design Code and ASME CE Code
Core support component
I terns
JAERI's criteria
ASME CE Code
Failure theory
Maximum principal stress + modified Coulomb-Molt r theory
Maximum principal stress theory
Buckling limits
RanKin-Gorden type
Karman type
Pure shear stress limits
Considered
Not considered
Oxidation effects
Considered
Not specified
Specified
Not completed
Quality control Stress evaluation method Stress category Safety factor Minimum ultimate strengtn Core component
Same for both Fundamental concept is same as the core support component with some exceptions (safety factor, irradiation effects ,., etc.)
Not specified
IHX
Containment vessel
: Intermediate heat exchanger
PPWC : Primary pressurized water cooler
Vessel cooling pone!
PGC
: Primary (as circulator
SPWC : Secondary pressurized water cooler
Water pomp3
Air cooler
SGC
: Secondary {as circulator
AHX
: Auxiliary heat exchanger : Auxiliary gas circulator
AGC
Air cooler Reactor
4MPa
30MW 20/30MW Water pump
3.5MPa
F1g.1
C o o l i n g system of the HTTR
19
E-3
Stand-pipe Control rod
Replaceable reflector Permanent reflector
Reactor pressure vessel Hot plenum
Core
Support post Core restraint mechanism
Carbon block
Auxiliary coolant outlet pipe
Support plate Diagrid Main coolant outlet pipe
F i g.2
B i r d ' s - e y e view o fthe reactor vessel and core
20
SSA
°T
Fuel handling hole Fuel kernel High density PyC SiC Low density PyC
Plug
Dowel pin
Fuel ./compact
0.92mm 8 mm
Graphite sleeve
39mm
3 4 mm
Fuel compact
Fuel rod Fig.3
Block type fuel of the HTTR
(SECONDARY SYSTEM) COOLER BLOWER
UCL
IN-PILE TUBE
REGENERATIVE HEAT EXCHANGER
COOLER HEATER
(PRIMARY SYSTEM) . He CIRCULATOR
-W—*
to to
CORE
TEST SPECIMEN -"""^ PRE-CHARCOAL TRAP
v REGENERATIVE /LIQUID Hi HEAT EXCHANGER VSUPPLY SYSTEM,
MOLECULAR CHARCOAL TRAP SIEVE TRAP
Fig.4
rr.
Ti SPONGE TRAP ,
-+--IXJ
JMTR
(He SUPPLY SYSTEM)
Flow diagram of the OGL-1
00
No.5
No.9
(3.1x10-3)
(1.4x10-3)
o tt) ID
10 -5
No.3
(7.1xl0-<)
at
No.10
(4.9xlO'M
c o to CO
o
No.6
No.7
(9.6xl0-5)
(7.1xlO"5) No.4
.No.8
10-7
No. 2
(7.3x10-5)
(1.4x10-5)
4
(l.lxlO* ) No.l (2.7x10-5)
I lxlO<
,
I 3xl0<
2xlO<
Burnup
[MWd/t]
n
Fig.5
R/B of 8 8 K r d u r i n g I r r a d i a t i o n . F i g u r e s In p a r e n t h e s e s show f a i l u r e f r a c t i o n a f t e r I r r a d i a t i o n [ 3 ] .
{Oimens1ons and material properties of fuels, icontrol rods, graphite blocks, etc.
Calculation of Atomic Density of each material
Calculation of group constants of fuel block and reflector blocks
Calculation of group constants of control rods
DELIGHT
TWOTRAN
Group constants for the core calculation
Reactor core analysis CITATION -1OOOVP *
« Vectorized version of the CITATION Fig.6
Flow of the HTTR nuclear design
24
Fixed half assembly
Movable half assembly
Steel frame
CRD,
driving mechanism (SRDM)
Heat insulaling cover
Neutron source guide tube
Control rod driving mechanism (CRDM)
01
Fixed side table I
SF1D hanger
\ /Movable side table Heat insulating table
F 1 g. 7
B i r d ' s - e y e v i e w o f the V H T R C
1.02
1.01
1.00 u
0.99 ±0.01 Ak 0.98
0.97
0.97
0.98
0.99
1.00
1.01
1.02
E x p e r i m e n t
Fig.8
Effective multiplication factor. experiment and c a l c u l a t i o n .
26
Camparison between
Nuclear design
Dimensions and material properties of fuel, the core structure, the support structure, etc.
V Power and fluence distributions
Calculation of flow rate distribution and coolant temperature
FLOWNET Calculation of Fuel temperature
TEMDIM
Distribution of the temperature of the core structure, support structure
D1str1buat1on of feu! temperature and evaluation of the maximum temperature
F1g.9
Flow of the HTTR thermal-hydraulic design
27
S&B
E-3
Adapter section : A
Mother section : M
400'C 0.4kg/s
M, Healer |400"C 4 . l M P a
Circulator
— Test section : T
Test section
Cooler 0. 2MW
1000'C Heofer
Heater
Circulotor 370'C 4 kg/s 4 . I M P 0
Heater
Fig.10
Test section (T 2 )
F l o w d i a g r a m of the H E N O E L
28
CORE BARREL
SEAL PLENUM BLOCK PLENUM SIDE INSULATION BLOCK
SUPPORT
PERMANENT REFLECTOR BLOCK PLENUM 8OTT< BLOCK
CORE RESTRAINT BAND SIDE RADIATION SHIELD BLOCK
CARBON Bi
PRESSURE VESSEL LOWER END BLOCK
SUPPORT PLATE DIAGRID
OUTLET GAS DUCT —
HOT HELIUM GAS
—
COLD HELIUM GAS COLD HELIUM GAS LEAK .
Fig.11
Bird's-eye section
view
of the core
29
bottom
structure
of T2
E-3
test
_ io'
i
i
I
VI
en
Inner temp, = 950 ° c outer temp. = 400 • c
—
—
y Exp.
o
Cal.
io 2 1 to
1
1
10'
10'
Pressure difference CKPa]
Fig.12 Relationship between leakage flow rate and pressure difference between Inner and outer of the permanent reflector blocks.
30
2 S o
z 10
20 BURNOFF
30
-1.0
( 7. )
6f : TENSILE STRENGTH OF OXIDIZED SPECIMEN
Fig.13
Dependence of s t r e n g t h on burnoff In u n i f o r m l y o x i d i z e d I G - 1 1 0 graphite
31
E-3
Secondary Helium (to Secondary PWC) Secondary Helium (from Secondary PWC)
Cold Header Primary Helium (to Primary Gas Circulator )
Primary Helium (from Primary Gas Circulator) Inner Shell Outer Shell
Tube Support Assembly
Centra! Hot Gas Duct (Center Pipe)
Thermal Insulator Helically Coiled Heat Transfer Tube Hot Header
Primary Helium (to Reoctor Primary Helium (from Reactor)
14
B1rd's-eye
view of t h e I H X 32
Actuater
Load Cell Helium Gos Supply Nozzle Connecting Rods Bellows
-Casing •Insulator ^Heater •Test Model
Fig.15
Test apparatus for the IHX structural test
33
E-3
XA0101507
BEHAVIOUR OF HTGR COATED FUEL PARTICLES AT HIGH-TEMPERATURE TESTS A.S.Chernikov,
R.A.Lyutikov,
V.V.Khroraonoshkin,
S.D.Kurbakov,
G.I.Soloviyov,
V.M.Repnikov,
USSR
ABSTRACT
At the temperature range 1200-2600 °C prereactor tests of TRISO fuel particles on the base of UQ . UC 0 2
and
x y
UO2+2A12O3. SiO 2 kernels, and also fuel particle models with ZrC kernels were performed. Isothermal annealings carried out at temperatures of 1400-2600
C, thermogradient ones - at
1200-2200 °C (A T = 200-1200 °C/cm). It is shown that at heating to 2200 °C integrity of fuel particles is limited by different thermal expansion of PyC and SiC coatings, and also by thermal dissociation of SiC. At higher temperatures the failure is caused by development of high pressures within weakened fuel particles. It is found that uranium
migration
from alloyed fuel (UCxO , UO2+2Al2O3.SiO2) in th<» process of annealing is higher than that from UO2.
1. Introduction HTGR safety is connected with possibility of coating leak tightness preservation of coated fuel particles (CP) under normal, as well as transitional operatioA conditions up to overheating of the core in hypothetical accidents (up to 2500 "C E-4
2. and more). That was the reason of investigation of high temperature influence on CP
characteristics [1-6].
The aim of the present work is investigation of influence of higher temperatures and temperature gradients on the. character of PyC interaction with U0 2 and dioxide, alloyed by aluminium silicates or uranium carbide, and also investigation of the coatings integrity preservation and main mechanisms of their failure.
2. Experimental procedure
Investigation was performed on CP of U0 2 , UO,+2A12O3. SiCT , UC^ 0 w (table 1 ) . All the coatings were deposited in a fluidized x y
bed. Pyrocarbon coatings were of two types: high temperature (HTI), produced by decomposition of CH^, and low temperature (LTD, produced by decomposition of C3H
kernel. CP-simulators with ZrC kernels and the
same set of coatings (BISO and TRISO type) were tested for comparison of the oxide fuel interaction with PyC. For testing at temperatures of 2250-2500 °C the mixture, consisting of 100 CP (U02kernels and LTI coatings) and graphite press-powder, was moulded into pellets.
E-4
Table 1 Characteristics of initial CP Batch
Parameter
LTI
HTI 1. Kernel
UC x 0y
UO.
UO 2
1.1. Density, g/cm
3
10.4
10.4 510
1.3. Nonsphericity coefficient
1.04
2. Layer PyC-1 2.1. Density, g/cm
0.98-1.05
2.3. Technological layer PyC 2.3.1. Density, g/cm3
2
2
11.6
1.2. Size (d), pm
2.2. Thickness (6), /um
UO +2A1 0 .SiO 3
9.8
100
1.5-1.6
2.3.2. Thickness (6),^m
20
3. Layer PyC-2 3.1. Density, g/cm3
1.80-1.85
3.2. Thickness ('S),t~im 4. Layer SiC 4.1. Density, g/cm 4.2. Thickness
m
50
1.80-1.90 60
3.20
3.20
50
55
5. Layer PyC-4 1.80-1.84
5.1. Density, g/cm 5.2. Thickness 5.3. BAF
m
50 1.10-1.20
1.80-1.95 70 1.03-1.05
6. Coated fuel particles 6.1. Nonsphericity coefficient
1.04
1.04
E-4
4. Table 1 „ . Parameter
Batch
HTI
LTI
6.2. Load to fracture, kg
5-6
6.3. Leakage of fission gas products at weak irradiation, 10*
2,-3
7-8
< 2
Moulding pressure did not exceed 300 kg/cm . The following thermal treatment was performed at 1800 °C for lh. The tests of CP, loosely packed and consolidated in pellets, were performed in the vacuum 5.10
- 5.10
Pa with a mass-spectrometric control
of residual medium. CP of every studied batch were charged (up to 200 CP) in thin-wall graphite ampoules, which were put in the isothermal zone of the graphite heater. CP were heated to the temperature 1400-2600 °C at a rate of 5-60 °C/s. The isothermal exposure was in the range 10-2000 h for the temperature range 1400-2000 °C and 0.5-5 h at the temperature more 2000 °C.The exposure for CP in pellets did not exceed 30 h at 2250 - 2500 °C. Prereactor thermal gradient tests were performed on batches, containing 100-200 CP, at the temperature 1200-2200 °C and the temperature gradients 200-1200 °C/cm in the vacuum 5.10
Pa, and
using fuel compositions without coatings, imitating the kernel behaviour in tightless CP, in the atmosphere
of noble gas.
E-4
5. Coatings failure at heating or isothermal exposure was registered with a mass-spectrometer at a short-term increase of the gas pressure. Investigation of degree of fuel interaction with coating materials at annealing was carried out by comparison of weight losses of CP batches, strength characteristics change.migration of fuel and coating components. The latter was registered by methods of absorption contact X-ray radiography, ceramography and X-ray microanalysis.
3^ Experimental results and discussion
At the annealing to 2000
C for all investigated batches
CP strength losses were not observed. Over 2000 °C strength of CP, containing a SiC layer (fig 1), decreases sharply and doesn't depend upon the kernel content [8]. Mass losses of TRISO CP correlate with their strength change (fig. 2 ) . The mass of all types of CP does not change up to 2000 °C. At temperatures over 2000 °C CP with a SiC layer intensively lose their mass. At the same time the mass of BISO CP with gas-tight PyC coating on U0 2 remains unchanged up to the temperature 2450-2500 At temperatures over 2000
C.
C SiC layer decomposition
was characterised by irregularity of material evaporation from the surface and porosity development in the coating volume [3], The same change in SiC at annealing was noticed in [5, 10]. Under the considered test conditions visible changes
E-4
6. in PyC microstructure were not observed. However, the optical anisotropy coefficient of highly dense PyC, beginning with ~ 2200
C increased with temperature test. The more its
initial value was, the more it increased. The application of the method of absorbtion contact micro- X-ray radiography allowed to estimate the temperature of fuel interaction with PyC and the character of fuel migration in the coating without CP failure (fig. 3 ) . CP having passed the tests below the temperature of the failure beginning, i.e. intact CP, were examined at this stage (table 2 ) . Under the same test conditions (temperature, annealing time) for TRISO CP uranium migration rate decreased in U(CxO ) - U0 2 + 2Al2O3.SiO2 - UO
2
In BISO CP with intact PyC coating the uranium migration from UO after tests in pellets at 2500
C for 20 h was not registered.
The most essential distinctions in uranium migration in PyC were observed at temperatures over 1800 °C, when microdefects can appear in a SiC layer due to its decomposition and evaporation and silicon migration in PyC (fig. 4) and absence of migration in an intact BISO CP at 2500
C. The uranium migration rate may be is
connected with a different reduction degree to dicarbide of the fuels considered, when the same quantity of fission gas products are released from CP through microdefects, formed in the coatings before their complete destruction. Formation of a fluid phase, containing uranium compounds (e.g., the melting point of UO2-SiO2 eutectic is 1650 °C) , can be an additional reason of the increased uranium migration in a dioxide with aluminosilicate
E-4
(p OO
7. Table 2 The temperature of beginning of PyC interaction with fuel and the character of its migration in coatings of tight coated fuel particles
Material Kernel
UO
Coatings
Time of isothermal annealing, h
LTI-SiC HTI-SiC
Temperature, C
2
2200
Migration depth 5-10 /urn with formation of a symmetric UC ring
1000
1600
8
1300
Loosening of the kernel-PyC boundary. Migration depth 10-25 turn. Migration up to SiC, increased uranium concentration at SiC.
1000
1600
60
1800
PyC
UC 0 x
LTI-SiC y
(15 and 50% UC)
UO 2 +2Al 2 O 3 .SiO 2
LTI
(Aluminosilicate 5 mass %, Al/Si=l)
Character of migration
Loosening of the kernel-PyC boundary, migration depth 20-25 fjm Uranium migration up to SiC
additive. In an oxicarbide fuel dicarbide phase formation by the reaction [11]:
=
0C
i.« 0 O.O3
with an insignificant quantity of CO, released within CP/ can be the same reason. Uranium migration relationship with formation of microdefects in SiC layer is confirmed by long-term pellets annealing at 2250 °C, when the uranium migration in the form of E-4
Go A
8. alternative rings was observed after 10 h testing (fig. 5e,f). Such a character of the uranium migration is the result of microdefects formation in coatings. When CP macrodefects are present at these temperatures, uranium migration differed both in rate (one order of magnitude higher) and appearance (as a rule the continuous front was present). Migration rate subsequently increased when CP temperature in a pellet increased to 2500 °C, defects in some CP began to develop already after 2 h annealing. Lamination of the outer PyC and subsequent SiC evaporation were characteristic both of CP loosely packed and consolidated in a pellet(fig. 5b, c ) . However, in the latter case such defects formation was shifted to larger values owing to matrix graphite banding influence (2500 °C, 10-20 h) . Uranium migration rate in coatings from neighbouring failed CP (fig. 5a, d) is noticed to be much greater
than
from their own kernel (fig. 5b, c ) . The following development of defects in coatings leads to their destruction. Such types of CP failure were marked: - chipping of an outer PyC part without carbide coating failure (fig. 6a); - equatorial chipping of outer PyC and SiC layer (fig. 6b); - formation of through craters up to the kernel (fig. 6c) or CP failure into plurality of fragments. In batches, containing failed CP, sample surface had a characteristic lustre. In the case of all coatings failure the colour change of the surface may be connected with ura-
E-4
9. nium dioxide evaporation (fig. 7 ) , and also incongruent evaporation of silicon carbide. The temperature of failure beginning of CP coatings (HTI, LTI) and fuel compositions (UO , OC 0 , 00 2
X
y
+ 2A1 0 . SiO , ZrC) 2
2
3
2
is shown in table 3. This temperature is characterised by appearance of at least one CP with a failed coating among 200 tested. It should be noted, that the heating rate of the sample within 5-60
C/s did not influence the temperature of
failure beginning. In the table it is seen, that in the investigated conditions the temperature of the CP failure beginning is defined only by presence or absence of SiC in CP. The type of pyrocarbon coatings and the content of fuel composition do not effect the temperature of the failure. Thus, under isothermal conditions oxide fuel interaction with a pyrocarbon and uranium migration begin after appearance of defects in coatings, through which
the
oxide carbon removal from
CP is possible. Displacement of consolidated CP in a graphite matrix inlarges the temperature and time range of SiC coating operation and, consequently, CP leak tightness as a whole. Uranium migration within the kernel becomes possible without loss of coatings leak tightness, when a temperature gradient arises in CP. This migration can occur owing to 0 2
or C0 2
thermal diffusion inside the kernel and to cyclic reactions of carbon masstransfer in a gaseous phase. The
indicated-
processes of carbon transfer from the hot side of CP on the cold one can be generally expressed by the fuel kernel migration coefficient (KMC) [12]. Results of KMC difinition for all tested E-4
10. Table 3 Temperature of the beginning of loosely packed coated fuel partucles failure. Material Kernel UO
UO
Temperature of failure, °C Coating HTI-SiC, LTI-SiC
2
HTI, LTI
2
UC 0 (15-50% UC) y UO x +2A1 0 .SiO (Aluminosilicate 5 wt %, Al/Si=l)
ZrC ZrC
LTI-SiC 'LTI-SiC
LTI LTI-SiC HTI, LTI
2250-2300 ~ 2400 2250-2300 2250-2300 2250-2300 2300 > 2600
types of fuel (fig. 8-10) agree well with data in [12, 13-15]. Our data on temperature dependence of KMC for UO 2 kernel in CP (fig. 3, curve 1) are rather close to KMC values given in [13] for fuel of different enrichment, excepting the results obtained in KFA (curve 8) and Belfonueleaire (curve 5 ) . Dependences of KMC=f(T), obtained for UO 2 kernels (fig. 8, curve 2 ) , are almost one order of magnitude higher than KMC for intact CP. Perhaps free migration of the formed carbon monoxide in experiments with kernels essentially accelerates carbon transfer from the hot side to the cold one. Carbon transfer processes in the gradient temperature field at oxide kernels annealing weakly depend on their content. Alloying of uranium dioxide by 5 wt% aluminosilicate additive with the ratio Al/Si = 1/1 practically doesn't
E-4
11. change the process activation energy (fig. 8, curves 2, 3 ) . Alloying of uranium dioxide by uranium dicarbide or monocarbide changes rather greatly (in 2-3 times) the process activation energy (fig. 10). Nevertheless total KMC values for these fuel types in the temperature range 1300-2000 °C are within the KMC values of the tightless CP with UO 2 kernel, limited by dotted curves of 95% probability (fig. 10). Taking into consideration approximately equal conditions of carbon transfer through the gaseous phase in these experiments, the exposed migration characteristics differences, perhaps, should be referred to peculiarities of carbon diffusion processes in fuel of various content. Nevertheless the main contribution in the kernel migration process under these conditions is given, perhaps, by the transfer through this gaseous phase. The amoeba effect of CP with an oxicarbide kernel depends on the phase content. When the kernel content has second phases, the migration of phase components can take place with different rates. The shifting of the main part of UCxO
kernel practically did not occur. Comparison of coefficient values in temperature depen-
dences of KMC of the investigated fuel types are given in table 4. Thus, on the base of gradient tests of CP with different types of the fuel, one can range it according to KMC at increased temperatures (1600-1800 °C) in such a way:
U0 2 tightless CP
E-4
12. Table 4 Pre-exponential multiplier and activation energy in the equation KMC = KMC
exp (-Q/R) O
Fuel composition
KMC ,
Q, cal/mole
Note
o
K. cm
C
uo 2 uo 2 uc 2
4.7x1Of2
-32000-4000
2. 1
-35000-4000
2.0xl07
-97000-10000
UO2+6%UC
1.8xlO~2
-20000^4000
tightless CP
U0 2 +2Al 2 0 3 .Si0 2
0.9
-30000-4000
tightless CP
tightless CP
4. Conclusion According to the obtained results the failure can be characterised by the
dynamics of CP
following
sequence of
the processes. At the annealing temperatures 2: 2000 °C intensive silicon carbide dissociation on C, Si, SiC2, Si2C begins. Fission gas product
pressure
at
the
temperatures
2000-2300 °C
is
about 10"" Pa, this cannot be the reason of outer PyC lamination and CP failure. On reaching the critical
tempera-
ture (it is different for every loosely packed CP, and it is in the range 2200-2300 °C) failure of separate CP begins
with
outer PyC lamination. This can be conditioned by the difference of thermal linear expansion coefficients of
SiC
and
PyC. An intensive diffusion of silicon in PyC coating greatly
E-4
13. influences its structural characteristic changes. As a result outer PyC lamination has an
irreversible
character and can
occur either without coating failure or with formation of through defects (cracks, chipping, etc.) in it. Depending on the character of outer PyC lamination in due course an incongruent evaporation of
SiC
mass-transfer take place either onto the inner outer PyC or its removal from CP
and
silicon
side of the
through a defect. Some
loosely packed CP with such through defects (with intact inner PyC and loss of SiC banding properties for a definite period of time) can remain tight to 2400 °C (beginning of BISO CP failure) or to 2500°C in a consolidated state in a graphite matrix. The obtained experimental results
allow
to
conclude,
that the CP failure in the course of high-temperature tests is a multi-step process. However, one can distinguish two decisive factors: first, degradation of coating properies (especially SiC) at high temperatures and, second, high gas pressure, developing under the coating. The calculations show that, for example, the pressure of CO within CP on UO (for U0 2 o o i ) at
the
temperatures
the developing pressures at 2200 CP, then these gas
2200
base is (2-4).10
Pa
and 2400 °C [17]. If
C are not dangerous for intact
pressures can lead to particle failure after
appearance of the mentioned above changes. Uranium migration from a kernel greatly influences character of CP failure. Uranium migration in the
form
of concentric
rings with the diameter being 10-15 t-na larger than the kernel diameter observed at initial stages is, perhaps, activated by
E-4
14. silicon penetration to the kernel surface under conditions, when SiC layer is not completely decomposed, and CP still remain tight. The
same
conclusion has been made earlier in [16].
From this point of view a greater migration of uranium from alloyed fuels
under temperature gradient conditions
adversely affect CP serviceability. On the base of the performed investigations and the obtained dependences KMC = f(T) within spheric fuel elements under of the reactor VG-400
being
the
kernel
using
migration
normal operation conditions
designed was calculated
(fuel
temperature 1250 °C, AT = 30 °C/cm, r = 1.95xlO? s, that is 2/3 of the total life-time of
fuel elements in the core ) .
The obtained results show, that UO2 kernel can shift less than by 10 pm.
References 1. Ikawa K., Kobayashi F., Iwamoto K. - J. Nucl.
Sci.
and
Technol., 1978, 15, N 10, p. 774-779. 2. Godin D.T. - J. Amer. Ceram.
Soc.,
1982,
65,
N
5,
p. 238-242. 3. Nabielek H., Naoumidis A., Goodin D.T. - Jahrestag. Kerntechn.83. Tagungsber., Berlin, 14-25 Juni, 1983,- Bonn, 1983, s. 589-592. 4. Schenk W. - Jahrestag. Kerntechn.84. Frankfurt, 22-24 Mai, 1984, Tagungsber. Bonn,1984, s. 203-206. 5. Ogawa T., Fukuda K. -
Nucl. Eng. and
Des.,
1986,
92,
N 1, p. 15-26. 6. Fukuda K., Kashimura S., Iwamoto K. - Trans. Amer. Nucl. Soc. , E-4
15. 1985, 50, p. 241-242. 7. Chernikov A.S.,
Permyakov
L.N.,
Mikhailichenko
L.I.,
Rurbakov S.D. - IAEA-TECDOC-436, Oct., 1987, p. 445-460. NTIS PCA23/MFA01.
Order
Number
DE88701817/JAW
(Conf.
8. Chernikov A.S., Mikhailichenko L.I., Orlov G.V.,
Kurba-
-8610160).
kov S.D. - JAIF-GKAE Seminar on Fuel Elements
and
Fuel
Composition of HTGR. October 20-22, 1987, Tokyo, p. 3.13.20. 9. Chernikov A.S., Kolesov V.S., Deryugin A.I. - In:
Spec.
Meet, on Fission Product Realease and Transport in Cooled Reactors. Gloucester, UK, October,
1985,
GasIAEA,
IWGGCR/13, p. 74-80. 10. Ogawa T., Fukuda K., Shiba K. - In [8], p. 3.81-3.100. 11. Heiss A. - J. Nucl. Mater., 1975, v. 55, p: 207-230. 12. Lindemer T.B. et al. -
J.
Amer.
Ceram.
Soc.,
1977,
v. 60, p. 5-14. 13. Homan F.J. et al. - JQ1.-1502, KFA, 1978. 14. Naoumidis A. et al. - Thermodynamics of nuclear materials, Vienna, IAEA, 1975, p. 173-186. 15. Wagner-Loffler M. -
J.
Nucl.
Technol.,
1977, v. 35,
p. 392-402. 18. Bens R., Naoumidis A. - J. Nucl. Mater.,
1981,
97,
P. 15-24. 17. Khromov Yu.F., Lyutikov R.A. - Atomnaya Energiya, • 1980, v. 49, issue 1, p. 28-30 (in Russian).
E-4
9.0 bQ
7.0
o 05
/////
W///////////M
U
5.0 o
3.0 1600
1800
2000
\
2200
Temperature, °C Fig. 1. Strength of TRISO coated fuel particles after isothermal annealing
4.0
1-3
1400
1800
2200 Temperature, °C
Fig. 2. Change of coated fuel particle mass during isothermal annealing: 1 - BISO OP with average-dense outer PyC layer; 2 - TRISO CP; 3 - BISO CP with highly-dense outer PyC layer
E-4
64 o
Pig. 3. Uranium migration during isothermal annealing of loosely packed coated fuel particles, X40: a - U0 2 initial coated particle; b - U0 2 , 2200 °C, 2 h; c - UCXO , 1600 °C, 2000 h; d - UCXO , 1800 °C, 60 h; e - U0 9 + 2Al p 0- r Si0 p , 1600 °C, 2000 h; f - U0 p , 22502300 °C, 2 h - lamination of the outer PyC and dissociation of SiC layer
E-4
•H CO
Pi O 02 •H +> +=> -H
f 2 O 0) Pi fH
o o
V SiC
Kernel
PyC
Layers
a
Kernel
PyC
SiC
PyC
Layers
Fig. 4. Silicon migration in PyC coatings and lamination of the outer PyC at coated fuel particle annealing: a - initial CP; b - annealing 2000 °C, 2 h; c - annealing 2250-2300 °C, 2h
E-4
64 L
Fig. 5» Uranium migration in coatings during isothe-rmal annealing of coated fuel particles(in graphite pellets) X40 a - a general view of a pellet; b - U0 5 , 2500 °C, 20 h; c - U0 9 , 2500 °0, 20 h; d - U0 9 , 2500 *G, 20 h; e - U0|, 2250 °C, 30 h; f - U0|, 2250 °C, 30 h
E-4
a, X100
*
•
b, X50
•
c, X50 Pig. 6. Types of failure of coated fuel particles
X100
X1000
Pig. 7. Evaporation of uranium dioxide at through failure of coated particle layers
E-4
I-"
6 n
10
5.0
Fig. 8. Dependence of IK>2 kernel migration coefficient upon temperature: 1 - UOp, fuel particles with 5-layer coatings; 2 - U0 2 , without coatings; 3 - U0 2 + 2Al20^.Si02 without coatings; 4 - UOg, coated fuel particles /15/; 5-9 - U0 2 , coated fuel particles /13/
E-4
10"
10-3 eo 1 0-4
B o
10" o -6 10 10'-7 -8 10
4.5
5.0
5.5
6.0
6.5
7.0
Pig. 9. Dependence of UC 2 kernel migration coefficient upon temperature: 1 - U0 2 without coatings; 2 - UC 2 , coated particle /15/i 3, 4 - UCp, coated particle /14/
E-4
10 -3 B
o .
o
10-4
10 -5
Pig. 10. Dependence of UC^.0 kernels (without coatings) migration coefficient upon temperature: 1
"
UO
1.3CO.17S
3 "
UO
1.3 C O.685
2
"
U0
1.23 C 0.25 ;
U0
1 ,01°0.51 ;
7 - U0 ? , coated fuel particles
E-4
XA0101508
PRESENT STATUS 07 KHTGR PROGRAM IN USA Cooplltd froa Contributions from the HHTGR Program Teaa
ABSTRACT the U.S. Department of Energy (DOE) Modular High-Temperature Gas-Cooled Reactor (MHTGR) program has produced a conceptual design which has been reviewed by the U.S. Nuclear Regulatory Commission (NRC). The results of the review were generally favorable, and the program team has now moved into the preliminary design phase. The program team consists of a nuclear island engineering (NIE) team, an energy conversion area (EGA) team, a design Integration organization, and a technology development teaa. Utility user requirements are provided by a utility organization which also participates in design and programmatic reviews/evaluations. This paper will review the direction and accomplishments of each participating
organization. INTRODUCTION High-Temperature Gas-Cooled Reactor
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information from Cha Large HTGR (LHTGR) program waa used wh«rt appropriate. Where design data vert misting, assumptions were mad*, and a technology program formulated to validate (or modify) the assumptions. A Preliminary Safety Information Document (PSID) was completed In 1986. This document, which was based oh the conceptual design of the HHTCR, was forwarded to the NRC in September 1986 to initiate an early dialogue with the NRC on this unique design. Substantial discussions with the NRC took place for the next two years, leading to a sound understanding by the NRC staff of the attributes of the MHTGR, and a wall-documented regulatory position to guide DOE in continued design and development of the MHTGR concept. A conceptual design for the MHTGR was completed in early 1988. The preliminary design phase began during the second half of FY 1988. The remainder of this paper will focus on the major facets of reactor design, licensing, and technology development. Program participants and their roles will be identified and discussed. DESIGN ACTIVITIES NUCLEAR ISLAND (NI) The nuclear island Includes the four reactor nodules, each producing 350 KW(t) (see Fig. 1 [Neylan 19882]). The modules are headered In pairs to feed two turbine generators of 300 MW(e) each. The turbines operate in parallel. All systems containing radlonuelides and all systems essential to nuclear safety are located within the NI. Therefore nuclear standards and practices ate implemented for the NI. The NX includes the reactor core,, the reflector, three steel vessels (the reactor vessel, the steam generator vessel, and the crossduct vessel), the shutdown cooling system (heat exchanger and shutdown circulator), the control rod drives, the reserve shutdown system, the main circulator and other components of the heat transfer system, and the reactor cavity cooling system (RCCS). ENERGY CONVERSION AREA (ECA) The ECA includes all those components not in the NI. The use of high levels of conventional standards and practice* will be utilized to meet requirements of the ECA. Included in the ECA are the turbine generator building, operations center, and cooling towers (see Fig. 2 [Neylan 1988*]), DESIGN INTEGRATION This important activity includes: Writing and implementation of program policies and procedures Maintenance of the Overall Plant Design Specifications 1
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Support to program development, planning and control Maintenance of eh« Summary Level Program Pl«n (SLPP) Program technical management support and deilgn evaluation Functional analyst* implementation Sptclfication of data storage and retrieval requirement* Integration of the utility/user requirement* Maintenance of licensing plan criteria, methods, and standards Planning and coordination of licensing submittals to the NRC and supporting NRC briefings Interface control Scheduling Quality assurance LICENSING ACTIVITIES The licensing objective of the MHTCR program Is to eventually obtain a certification rulemaklng for the MHTGR from the NRC. The current plan eo reach this objective includes three major steps:, (1)
Obtain a favorable SER on the MHTGR conceptual design,
(2)
Obtain a Preliminary Design Approval, and
(3)
Obtain a Final Design Approval for a standard MHTGR
These activities, in combination with experience gained from construction and operation of the first MHTGR, will form the basis for design certification. te is the policy of the NRC to Interact as early as possible with proponents of advanced reactor* (applicants, vendors, and government agencies) [see Commission's advanced reactor policy statement 51 £& 24643]. The NRC review of the MHTGR began In October 1986, and was completed In March 1988. The NRC review focused on three key documents prepared by the MHTGR Program. The documents were the Preliminary Safety Information Document (PSID), the Probabilistic Risk Assessment (PRA) document, and the Regulatory Technology Development Plan (RTDP). The RTDP includes the portions of the Technology Development Program (TDP) which focus on safety issues. The Issuance of an SER by NRC is Imminent, summarizing the results of the review, A preview of the SER content is contained in three papers presented In August/September 1988 (Williams 1988,' King 1988/ Rogers 1988 s ). Key points from these papers are outlined below: o
The MHTGR design reviewed by NRC was conceptual. Therefore, the NRC review concentrated on features and issues related to safety and viability. The General Design Criteria (CDC) used for LVRs were considered by NRC to insure that the MHTGR provided equivalent protection to the public. Several key policy issues arose during ,
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tha raviav, dua to tha . different approach by which tha MHTCR proposed to atat tha criteria. The NRC ataff made recommendations on tha four policy iasuea liatad below. Their recommendations are conditional on completion of needed R&D program*, successful resolution of aafety inua* identified, successful prototype testing at an isolated site, and favorable completion of the deferred review items. (1)
Selection of design basis events
(2)
Siting aourca term calculation and use
(3)
Adequacy of containment
(4)
Adequacy of emergency planning
Several Important areas were not reviewed (see Table 1 [Williams 1988s)). Review of these areas has been deferred to the future. NRC expects, and DOE committed to, timely submission of a revised RTDP and a document describing DOS'* plans for prototype testing. Final da termination of the llcensability of the MHTCR is contingent upon tha following: Satisfactory resolution of Issues identified in the SER. Table 2 (Williams 1988s) is a listing of areas where analysis, research and development, and testing results will be required. Completion of final design and licensing reviaw by NRC Successful design, construction, testing, and operation of a prototype reactor prior to design certification TECHNOLOGY DEVELOPMENT The KHTGR TDP was formulated during 1986 and 1987 to supports the design effort. Top-level regulatory (see Table 3 [Cunllffe 1988 ]) and user requirements (see Fig. 3 [Millunzl 19B81]) were established, and linked to design selections through a detailed functional analysis (see Fig. 4 [Millutui 19881]). Data accumulated during the past 25 year* from tha LHTGR technology program ware used to develop the KHTGR conceptual design. It was determined that additional design data, beyond what is available from the LHTGR program, are required for KHTGR-specific conditions. When suitable data were not available, designers used assumptions from data developed under LHTGR conditions, and then defined Design Deta Needs (DDNs) and Technology Development Needs (TDNs) to validate or modify the assumptions (sea Fig. 5 [Homan 19887]). The 1987 TDP describes programs E-5
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In fuel manufacturing, fuel performance, fitiion product behavior, graphite behavior, and structural materials performance. The slx>year program formulated in 1987 has been stretched somewhat in time due to funding constraints. In addition, several additional TDP areas have been added, including shielding analysis, physics methods validation, and thermal hydraulics validation.
UTILITY INTERFACE The design and technology development teams have interfaced extensively with the nuclear utility organizations in several areas; o o
Identification and development of top-level user requirements . Participation in Analysis and Trade Studies (see Fig. S)
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Economic analyses (LaBar 1988*)
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Evaluation of exiatlng experience base (Dllllng 19889)
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Evaluation of evolving technology (Gray 198810)
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Review of design and design concepts (EPRI 1988") PROGRAM PARTICIPANTS AND THEIR ROLES
DESIGN Two design teams have been formed. The NI design team consists of three prime contractors: General Atomics (GA), Combustion Engineering (C»E), and Bechtel National Incorporated (BNI). Stone and Webster Engineering Corporation (SWEC) is the prime contractor for the ECA team, and C*E is a subcontractor to SWEC in this area. The Plant Design Control Office (PDCO) serves as design integrator. CA has responsibility for design of the core, the control system, and several major components. CA is responsible for fuel fabrication, Including fabrication technology development. James Howden Company of Scotland is a subcontractor to CA for design of the main circulator and the shutdown circulator. C-E Is responsible for the design of the pressure vessel, the steam generator, the crossduct vessel, the steam generator vessel, and the shutdown cooling heat exchanger. In addition, C-E is also responsible for the plant supervisory control system. BNI is responsible for the RCCS design, shielding design, and all seismic analyala. 3
(oVL
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5VEC Is the prime balance«of-piant (BOP) contractor, and la responsible for designing the EGA, the operations center, and the Plant Control Data and Instrumentation System (PCOIS). Since many of the reliability/availability issues are related to BOP functions, SWEC has primary responsibility for reliability/availability assessments. SVEC is responsible for constructabllity evaluations, and contributes to economic analyses. PDCO coordinates and Integrates the NX and SOP design efforts. PDCO is responsible for schedule preparation and maintenance, configuration management, maintaining design discipline, and quality assurance. PDCO organizes and coordinates all design reviews. LICENSING Gas-Cooled Reactor Associates (GCRA) provided coordination from the KHTCR program side during the two*year review of the conceptual design by NRC. Technical support was provided by all other program participants. PDCO now has the lead for licensing coordination and NRC interactions. TECHNOLOGY DEVELOPMENT Oak Ridga National Laboratory (ORNL) haa the technical lead in technology development. The technology development program supports only the NI phase of design. DDKs and TDNs are prepared by the design organizations (primarily GA and C-E). ORNL responds to the DDNs and TDNs with experimental plans, which are reviewed by the designers. Several subcontractors are Involved in the technology program: o
CA performs specific tasks associated with fuel fabrication development, graphite testing, and application of design methods to predict the results of experiments. The design methods validation areas are primarily in the area of core physics, fuel performance, fission product behavior, and thermal hydraulics performance.
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Massachusetts Institute of Technology (HIT) is involved tn three technology-related tasks: (1)
Building a high-pressure fission product loop. This loop will be used to measure the behavior of fission products under temperature and pressure conditions which will be present in an operating MHTGR. The data from the loop experiments will be used to complete formulation of MHTGR fission product behavior models.
(2)
Control system design for the Direct-Cycle HTCR (HTGR-DC). This Is an advanced application of the MHTCR, expected to offer significant cost and performance advantages over the Steam-Cycle HTGR (HTCR«SC). MIT investigators have recently 6
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proposed solutions to two technical problems (recuperator performance and turbomachlnery size) which contributed to the demise of the HTCR-DC program In Germany, Switzerland, and the U.S. in the mid*197Os. (3)
Thermal hydraulics code verification and validation.
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Commissariat A L'Energle Atomique (CEA) is refurbishing the COMEDIE Loop in the Slloa Reactor at Centre D'Etudes Nuclealres De Grenoble (CEKG). The COMEDIE Loop was used during the 1970s for cooperative (CA/CEA) research in the area of fission product behavior for the LHTCR design. This loop will be used to validate the KHTCR fission product behavior models.
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C»E also has three tasks in the KHTCR technology development program: (1)
Evaluation of elevated temperature properties of the MHTGR pressure vessel materials. This task includes development of an elevated temperature code inquiry to be submitted to the ASMS code committee, interface vith ASME subcommittees, and participation in the US/FRG materials subprogram.
(2)
Elevated temperature testing and irradiation testing of MHTGR pressure vessel materials. This task includes evaluation of results from the ORNL creep program, and results from the Nil Ductility Transition Temperature (NDTT) shift program. This work is discussed in more detail in a separate paper at this workshop.
(3)
Steam generator seal development and testing. This task includes evaluation of steaa generator seal concepts and requirements, completing trade studies, selection of a reference seal design, update of shroud seal test specifications, and Initial design of test models and rigs.
UTILITY INTERFACE GCRA manages and coordinates the utility interface. GCRA is funded directly by participating utilities. GCRA has direct access to the Electric Power Research Institute (EPRI), utility executives, and operations managers, These contacts are a valuable source of information and data which are used to accomplish the tasks listed earlier. CONCLUSIONS The U.S. MHTGR program Is now in the preliminary design phase. The program has utilized a disciplined, functional analysis approach to develop a conceptual design, The conceptual design has been reviewed by E-5
the HRC, with favorable results. A well-integrated design team is in place, supported by « technology development organization* with an international flavor. When the design, licensing, and technology development programs were formulated in 1986, a six-year program was planned. Funding constraints have stretched the program schedule since 1986, but these three major activities still are closely coordinated and can deliver the first MHTCR by the «id- to late-l990s. REFERENCES
1. A. C. Millunzi and S. R. Penfieid, Jr., "Developing a Reactor for Today's Realities and Tomorrow's Needs," Proceedings of tha 23rd Interaoclaty Enerrv Conversion Engineering Conference (Volume 1), August 1988, pp. 479-482. 2. A. J. Neylan, D. A. Dilling, and R. Ng, "Designing a Reactor for the Next Generation," Ihijj., pp. 483*488. 3. P. K. Williams, T. L. King, and J. N. Wilson, "Results of * Preliminary Safety Review of the MHTCR," paper presented at Tenth International Conference on the HTGR, San Diego, California, September 20, 1988. 4. T. l>. King, "Safety Evaluation of the Modular High-Temperature CasCooled Reactor," Proceedings of yha 23rd Intcrsoelfttv Energy Conversion. Enf inhering Conference, (Volume 1), August 1988, pp. 495-497 5. K. C. Rogers, "Key Questions Facing the NRC on the MHTCR," paper presented at the Tenth Annual International Conference on the HTCR, San Diego, California, September 19, 1988. 6. J. C. Cunliffe and F. A. Si lady, "The Challenge of Licensing a Reactor With Passive Safety Characteristics," Proceedings of the 2^rd Interaocletv Energy Conversion Enginaer;Lnf. Conference (Volume 1), August 1988, pp. 489« 494. 7. F. J. Homan and A. J. Neylan, "MHTCR Technology Development Plan," tt. pp. 511-514. 8. H. LaBar and H. Bowers, "Economic Characteristics of a Smaller, Simpler Reactor," ikU., pp. 449-504. 9. D. A. Dilling and 0. Jones, "The Potential for Construction/Zone Outfitting of the MHTCR,* Ibid,, pp. 531-535.
Modular
10. S. Gray and C. Jones, "Magnetic Bearings: A MHTGR Design Selection With Broad Industrial Potential," Iktt. pp. 521-524. 11. EPRZ Review Meeting on the Modular High Temperature Cas-Cooled Reactor Program, July 12-14, 1988, San Diego, California. 8 E-5
LISTING OF MAJOR AREAS REQUIRING SUPPORTING ANALYSIS, RESEARCH, OR TESTING FUEL DESIGN FUEL PERFORMANCE MODELS FUEL PERFORMANCE STATISTICS FROM LABORATORY TESTING MANUFACTURING QUALITY CONTROL FUEL PERFORMANCE UNDER ACCIDENT CONDITIONS EFFECTS OF FUEL COMPOSITION ON PERFORMANCE EFFECTS OF EXTERNAL CHEMICAL ATTACK ON FUEL PERFORMANCE NUCLEAR DESIGN METHODS AND DATA VALIDATION UNCERTAINTIES IN NEGATIVE TEMPERATURE COEFFICIENT OF REACTIVITY CONTROL MATERIALS THERMAL AND FLUID FLOW DESIGN CORE FLOW DISTRIBUTION NOT STREAKS DIFFERENTIAL PRESSURES AND SHEAR FORCES DURING DEPRESSURIZATION EVENTS FLOW-INDUCED VIBRATION ON CONTROL ROD GUIDE TUBES REACTOR INTERNALS SEISMIC DESIGN AND FRAGILITY DATA IN-SERVICE DETERIORATION OF MATERIALS VESSEL SYSTEM ASME AND STAFF APPROVAL FOR ELEVATED TEMPERATURE SERVICE CATASTROPHIC FAILURE PROBABILITY NEUTRON IRRADIATION EFFECTS SEISMIC DESIGN, INCLUDING SUPPORT SYSTEM REACTOR CAVITY COOLING SYSTEM AND REACTOR CAVITY HEAT TRANSPORT DESIGN VESSEL HOT SPOTS IN-VESSEL CONDUCTION EMISSIVITIES EFFECT OF WATER VAPOR REPAIR AND RECOVERY MODELING CONSERVATISM AND SENSITIVITIES TO UNCERTAINTIES SEISMIC DESIGN AND FRAGILITY DATA REACTOR CAVITY TEMPERATURES DUCT AND CHIMNEY DESIGN HEAT TRANSMISSION TO THE EARTH RADIONUCLIDE CONTROL (SOURCE TERM, FISSION PRODUCT TRANSPORT) ASSUMPTIONS AND MODEL FOR BACK CALCULATION FROM SITE BOUNDARY OPERATIONS ADVANCED CONTROL SYSTEM DEVELOPMENT HUMAN FACTORS ANALYSIS CREW SIZE AND TRAINING WORKER EXPOSURE DURING MAINTENANCE PROTOTYPE PLANT TESTING
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GIG
SUMMARY OF AREAS CONSIDERED IN AARC REVIEW OF MHTGR AREAS REVIEWED FUEL DESIGN REACTOR PHYSICS REACTOR VESSEL PASSIVE HEAT REMOVAL SYSTEMS SAFETY ANALYSIS HEAT TRANSPORT EQUIPMENT COMPONENTS OF THE PRIMARY SYSTEM BOUNDARY INSTRUMENTATION CONTROL ELECTRICAL SYSTEMS SELECTED AUXILIARY SYSTEMS OCCUPATIONAL EXPOSURE HUMAN FACTORS SAFEGUARDS AND SECURITY SOME BALANCE OF PLANT ITEMS AREAS DEFERRED SEISMIC DESIGN RADIOACTIVE WASTE HANDLING SYSTEMS MECHANICAL EQUIPMENT DESIGN STRUCTURAL GRAPHITE COMPONENTS MODELING OF FISSION PRODUCT TRANSPORT NUCLEAR DESIGN PHENOMENA INVOLVING CHEMICAL PROCESSES FLUID FLOW DESIGN REACTOR INTERNALS VESSEL SYSTEM AND SUBSYSTEMS HEAT TRANSPORT SYSTEM AND SUBSYSTEMS SHUTDOWN COOLING SYSTEM AND SUBSYSTEMS REACTOR CAVITY COOLING SYSTEM REACTOR BUILDING PLANT PROTECTION AND INSTRUMENTATION SYSTEM PLANT CONTROL, DATA. AND INSTRUMENTATION SYSTEM MISCELLANEOUS CONTROL AND INSTRUMENTATION GROUP ELECTRICAL SYSTEMS SERVICE SYSTEMS STEAM AND ENERGY CONVERSION SYSTEMS OPERATIONAL RADIONUCLIDE CONTROL OCCUPATIONAL RADIATION PROTECTION EMERGENCY PREPAREDNESS ROLE OF OPERATORS SAFEGUARDS AND SECURITY PROTOTYPE PLANT TESTING SAFETY ANALYSIS TECHNICAL SPECIFICATIONS AND ADMINISTRATIVE CONTROLS QUALITY ASSURANCE
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SUMMARY OF TOP-LEVEL REGULATORY CRITERIA FOR THE MHTGR CRITERIA TYPE
CRITERIA
PUBLIC RISK
NRC POLICY STATEMENT ON SAFETY GOALS PROMPT FATALITY RISK < 1/10 OF 1% OF SUM OF PROMPT FATALITY RISKS DUE TO OTHER CAUSES. CANCER FATALITY RISK < 1/10 OF 1% OF THE SUM OF CANCER FATALITY RISKS RESULTING FROM OTHER CAUSES.
DOSE LIMITS FOR
10CFR20 LIMITS
• * °B REM W H ° L E B0DY < WS > /YEAR
• < 0.002 REM IN ANY ONE HOUR • < 0.1 REM IN ANY 7 CONSECUTIVE DAYS. 10CFR50 APPENDIX I LIMITS • LIQUID PATHWAY DOSE < 0.003 REM WS OR 0.010 REM • GASEOUS PATHWAY DOSE < 0.005 REM WB OR 0.015 REM • PARTICULATES DOSE <: 0.015 REM ANY ORGAN. 40CFR190 LIMITS • WHOLE BODY ANNUAL DOSE FROM ENTIRE FUEL CYCLE < 0.025 REM WB • < 0.075 REM THYROID. OR < 0.025 REM ANY ORGAN. DOSE LIMITS FOR ANTICIPATED OCCUPATIONAL OCCURRENCES
10CFR50 APPENDIX I DOSES ON AN EXPECTED VALUE BASIS
DOSE LIMITS FOR ACCIDENTS
10CFR100 GUIDELINES «<25REMWB • < 300 REM THYROID
DOSE LIMITS
EPA 520 PROTECTIVE ACTION GUIDES • SHELTERING: ONE TO FIVE REM WB; FIVE TO 25 REM THYROID • EVACUATION: > 5 REM WB; > 25 REM THYROID
cnrc TCD?IH£." o u t cUc! ancLTERING
OR EVACUATION
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ENERGY CONVERSION AHA
/
PERSONNEL SERVICES NUILOING RADIOACTIVE WASH MANAGEMENT WILDING REACTOR 'SERVICES WILDING
STANDBY POWER WILDING'
REMOTE SHUTDOWN WILDING .REACTOR AUXILIARY WILDING HEUUM STORAGE STRUCTURE
SWITCHYARD,
REACTOR MODULE (1OM) , MAINTENANCE HAIL
- NUCLEAR ISLAND
MHTGR Site Plot Plan
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TRANSMISSION SYSTEM NUCIEAR ISLAND
COOLING TOWER
'r^Wfy^W'jrw **4 I
CtM WATER
PUMP
REACTOR MOOULE NO. 2 TO: REACTOR MODULES > 0 . 1 AND 4
.FROM: POWER CONVERSION TRAIN 1 0 . Z
4 x 350 MW(t) MHTGR Energy Conversion Plant
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TOP-LEVEL FUNCTIONAL STRUCTURE
SAFE ECONOMIC POWER
MAINTAIN PLANT OPERATION
MAINTAIN PLANT PROTECTION
MAINTAIN CONTROL OF RADIONUCLIDE RELEASE
MAINTAIN EMERGENCY PREPAREDNESS
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MHTGR DESIGN APPROACH
USER REQUIREMENTS
TOP—LEVEL REGULATORY CRITERIA
INTEGRATED APPROACH
ENGINEERING PRODUCT PLANT DESIGN, ETC.
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RELATIONSHIP OF DESIGN DATA NEEDS TO DESIGN PROCESS DESIGN PROCESS REQUIREMENTS 12-
I
ANALYSES AND TRADE STUDIES 12-
RECYCLE TRADE STUDIES/ANALYSES EVALUATION: DOES DESIGN MEET ALL REQUIREMENTS
REEVALUATE WITH NEW DATA END
DO ANY ASSUMPTIONS REQUIRE VERIFICATION
\ _
"0
DDN
REQUIRED
YES PREPARE DDN (»)
PLAN AND CONDUCT TECHNOLOGY PROGRAM
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3. A
XA0101509
LEU-HTR Critical Experiment Program for the PROTEUS Facility in Switzerland R. Brogli, K. H. Bucher, R. Chawla, K. Foskolos, H. Luchsinger, D. Mathews, G. Sarlos, R. Seiler Paul Scherrer Institute Laboratory for Reactor Physics and System Technology Wiirenlingen and Villigen CH-5232 Villigen PSI Switzerland 16 June 1989
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LEU-HTR Critical Experiment Program for the PROTEUS Facility in Switzerland R. Brogli, K. H. Bucher, R. Chawla, K. Foskolos, H. Luchsinger, D. Mathews, G. Sarlos, R. Seiler 16 June 1989 Abstract New critical experiments in the framework of an IAEA Coordinated Research Program on "Validation of Safety Related Reactor Physics Calculations for Low Enriched HTR's" are planned at the PSI PROTEUS facility. The experiments are designed to supplement the experimental data base and reduce the design and licensing uncertainties for small- and medium-sized helium-cooled reactors using low-enriched uranium (LEU) and graphite high temperature fuel. The main objectives of the new experiments are to provide first-of-a-kind high quality experimental data on: 1) The criticality of simple, easy to interpret, single core region LEU HTR systems for several moderator-to-fuel ratios and several lattice geometries; 2) the changes in reactivity, neutron balance components and control rod effectiveness caused by water ingress into this type of reactor, and 3) the effects of the boron and/or hafnium absorbers that are used to modify the reactivity and the power distributions in typical HTR systems. Work on the design and licensing of the modified PROTEUS critical facility is now in progress with the HTR experiments scheduled to begin early in 1991 Several international partners will be involved in the planning, execution and analysis of these experiments in order to insure that they are relevant and cost effective with respect to the various gas cooled reactor national programs.
1 Introduction and Summary Gas cooled high temperature reactors (HTGRs or sometimes simply HTRs) represent a valuable option for the future development of nuclear technology. Their inherent safety characteristics due to unique features such as fission product barrier and structural integrity up to very high temperatures, high heat capacity in the core, etc., make them especially suitable for nuclear power and process heat production at sites close to densely populated areas. Although HTGRs have been extensively investigated in the past, the shift towards low enrichments and away from the mixed thorium/uranium fuel cycle as well as the introduction of new core materials (e.g. hafnium as burnable poison) reveals a lack of experimental data against which design and safety evaluation procedures can be validated. In addition, some effects such as reactivity increase caused by water ingress are more important in the smaller HTGRs of current interest. In order to cover this domain with experimental data and reduce the design and licensing uncertainties for small- and medium-sized helium-cooled reactors using low-enriched uranium (LEU) and graphite high temperature fuel, a series of critical experiments in the zero-power reactor facility PROTEUS is proposed.
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The main objectives of the new experiments are to provide first-of-a-kind, high quality experimental data on: • The criticality of simple, easy-to-interpret, single core region, low-enriched uranium (LEU), high temperature reactor (HTR) systems for several moderator-to-fuel ratios and several lattice geometries; • The changes in reactivity, neutron balance components and control rod effectiveness caused by water ingress into this type of reactor, • The effects of the boron and/or hafnium absorbers that are used to modify the reactivity and the power distributions in typical HTR systems. The new experiments have been approved within Switzerland and work on the design and licensing of the modified PROTEUS critical facility is now in progress. The Swiss contribution to the international LEU HTGR experimental program consists of the facility construction, licensing and operating costs as well as a portion of the scientific support staff. The Federal Republic of Germany contribution of the LEU pebble bed HTGR fuel needed for the initial experiments as well as considerable scientific support has been essential in the planning of these experiments. Work on the necessary international contractural and safeguards arrangements for the transfer of the fuel to PSI has been initiated. The HTR critical experiments are presently scheduled to begin early in 1991. Fuel from the LEU HTR experimental program in the AVR test facility of the KFA-Julich will probably be used for the first phase of the experiments. A series of two-dimensional discrete ordinates transport theory calculations have been performed for several configurations with this LEU AVR fuel which contains 6 grams of 16.7% enriched uranium per fuel pebble. The results indicate, that 5400 of these fuel pebbles should be sufficient for this initial phase, involving single core zone critical configurations with moderator-to-fuel pebble ratios ranging from about 1-to-l up to 3-to-l, pebble packing fractions from 0.6046 to 0.74 and core diameters from 1250 to 1500 mm. Experiments with lower C/U ratios are possible but will require either additional fuel pebbles or use of some of the existing PROTEUS UO2 driver fuel rods in more complex multi-zone systems. The experiments have been accepted as an International Atomic Energy Agency (IAEA) coordinated research program entitled "Validation of Safety Related Reactor Physics Calculations for Low Enriched HTGR's" in the framework of the Agency's Gas Cooled Reactor Working Group. In addition to the basic Swiss and Federal Republic of Germany cooperation, the Soviet Union and the Peoples Republic of China have already decided to participate and will supply some of the scientific manpower necessary to plan, execute and analyze these experiments and insure that they are relevant and cost effective with respect to the various gas cooled reactor national programs. Some other countries may also join in. The detailed experimental program is not yet completely determined with respect to just which experiments are done with what priority. As indicated above, a fairly wide range of possible configurations is possible within the overall envelope of the critical facility size and fuel availability constraints. It is expected that the cooperating international partners will meet in the near future to begin to define the details of the initial experimental program.
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2 HTR Reactor and Critical Experiment Status 2.1 HTR Status The high temperature gas-cooled reactor (HTGR) has received increased public attention in the recent past, not so much for its high-temperature-applications potential, as for its "inherent safety" characteristics. These stem from several broad design features, viz. the use of coated fuel particles embedded in a graphite matrix, a single phase, chemically inert coolant (helium), the availability of a large heat sink in the event of a loss-of-coolant accident, etc. Certain basic conditions need to be met in the physics design of HTGRs to ensure that the improved safety characteristics can indeed be realized, while maintaining acceptable economics and operational flexibility. Thus, an adequate shutdown margin (control absorber reactivity) has to be guaranteed under all normal and accident conditions, the temperature and power coefficients of reactivity have to be negative at each stage of the fuel cycle, and various conceivable reactivity and power distribution changes under accident conditions (e.g. water ingress) have to be kept within tolerable limits. An appropriate experimental data base is needed for the validation of LEU HTGR physics calculational tools from the viewpoint of practical design as well as licensing [20]. There are several HTGR specific considerations, viz. • the doubly heterogeneous nature of the fuel elements caused by the use of coated fuel particles, • the much lower degree of self-shielding for 238U resonance capture (relative to the light water reactor (LWR) systems for which the 238U neutron capture cross section data is usually evaluated), • the possibility of water ingress, • burnable poison effectiveness, • neutron streaming effects in coolant channels and void regions, • worths of control rods located in the reflectors, etc. With these specific features in view, experimental activity has been under way in Kernforschungsanlage Julich since the late sixties in the zero-power critical facility KAHTER and in the 15 MWe test reactor AVR. The culmination of the HTR development effort in Germany to date has been in the construction of the 300 MWe THTR power plant in Hamm-Uentrop which reached full power operation in Autumn 1986. The high-enriched, prismatic-block HTGR program in the United States has resulted in two zero-power critical experiment facilities [2,3,4] and two power reactors [5,6]. Of these, only the 330 MWe Fort St. Vrain (FSV) reactor which reached full power in November 1981 [7] continues to operate. The present U.S. HTGR program is oriented toward LEU prismatic block modular designs with "passive safety" features [8,9]. The Japanese HTGR research program is also oriented toward prismatic block type, low-enriched fuel [11].
2.2 Physics Experiments High-enriched uranium (HEU), i.e. 232Th-235U, has been the fuel of interest in most of the experimental pebble-bed HTR physics investigations carried out in Germany. In the past decade or ^
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so, however, non-proliferation considerations (the INFCE study), as well as the practical difficulty of commercially establishing a second fuel relative to the well established 238U-235U fuel used in LWRs, has led to the decision that all new HTR plants in Germany, Switzerland and the U.S. should use low-enriched uranium (LEU) fuel. With the U.S. HTGR experimental facilities dismantled many years ago, the KAHTER facility dismantled several years ago, the AVR reactor due to be shutdown soon, and the THTR and FSV plants continuing to operate on HEU fuel, there is an urgent need for an experimental facility to address certain open questions related to the physics, safety and design margins which would be available for HTR systems with LEU fuel. The importance of this is augmented by the fact that, in Switzerland in particular, there is a certain interest in relatively small, low-power district heating reactors [31,32,27]. Helium cooled heating reactors using HTR fuel tend to be undermoderated, high-leakage systems with special control needs and even larger uncertainties in the physics parameters. Criticality data for a mixed high- and low-enriched pebble-bed system has been generated at the 15 MWe AVR high temperature reactor in Germany in which about half of the original high-enriched fuel has been replaced with low-enriched fuel [12,13]. Criticality data for multizone systems in which external non-pebble driver fuel was used to maintain criticality has been obtained in the CESAR facility in France [14,15] and at the Technical University of Graz, Austria [16]. Criticality data on low-enriched prismatic and/or annular HTGR paniculate fuels was obtained in the NESTOR and HECTOR facilities at AEE Winfrith in 1966-69 [17]. Both the NESTOR and HECTOR experiments were multizone cores with external driver regions of non-HTGR fuel. The 20 MWt Dragon reactor experiment [18] at the Atomic Energy Establishment Winfrtih was built and operated in the 1960's under the auspices of the Organization for Economic Cooperation and Development (OECD). The Dragon reactor used high enriched annular fuel in prismatic fuel elements. The Japanese Atomic Energy Research Institute (JAERI) has also performed experiments since 1985 in the VHTRC (Very High Temperature Reactor Critical) facility[ll]. The main objectives of the Japanese program have been to obtain experimental data on critical masses, temperature coefficients of reactivity, neutron flux distributions and reactivity worths of boron burnable poison rods for a prismatic block fuel element design using 16 annular fuel rods of 2%, 4% or 6% enrichment per hexagonal block. As far as the JAERI experiments in the VHTRC facility are concerned, the main differences with respect to the proposed PROTEUS experiments are: 1) the VHTRC uses prismatic block fuel geometry so that different heterogeneity and streaming effects are encountered; 2) water ingress effects have not been investigated (it is now understood that water ingress simulation experiments in the VHTRC may begin in 1990); 3) absorber materials other than boron have not been used in the VHTRC; 4) the present PROTEUS experimental program does not include the temperature dependent measurement capability of the VHTRC facility; 5) measurements of neutron balance components (reaction rate ratios and koovalues) needed to complement the basic criticality data have not been made in the VHTRC. The Japanese plan to start the construction of a 30 MWt helium-cooled high temperature engineering test reactor (HTTR) using low-enriched UO2 fuel at the end of fiscal year 1989 with operation presently scheduled to begin in fiscal year 1995 [19].
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ess
3 Proposed Experiments 3.1 Clean Lattice Experiments The low-enriched, pebble-bed HTR critical experiments proposed for the PROTEUS facility are designed to complement the data base obtained from previous high-enriched HTR experiments such as the pebble-bed investigations in the KAHTER facility in Germany [21,51], the prismatic block experiments at General Atomics and at Battelle North West Laboratory [2,3,4] and the LEU HTGR experiments in the VHTRC facility in Japanfll] and to yield useful information for LEU HTR systems using either pebble or prismatic block fuel elements. In addition to experiments using the usual random packing of pebbles, experiments are planned with a hexagonal close-packed lattice having a packing fraction of about 0.74 as well as with an alternative hexagonal arrangement with a packing fraction of about 0.60. This will permit an experimental assessment of streaming effects between pebbles and in the case of the second hexagonal arrangement, allow easy access to the core center for reactivity worth and reaction rate measurements with minimal perturbation of the system. In principle, k^and the related reaction rate ratios are not affected by the pebble packing fraction. This means that we can choose any convenient geometrical arrangement of pebbles for the zero-leakage neutron balance measurements, provided that the system may be made critical and is large enough so that experimental corrections for reflector effects are acceptably small at the central measurement location. The basic results will be the critical masses and geometries, the experimentally measurable neutron balance components, inferred Rvalues, and neutron flux and fission rate distributions. The neutron flux distribution measurements will generate additional information on the neutron flux levels in the reflector regions and provide a basis to validate the computer models. An accurate prediction of the fast neutron flux in the reflector zones is an important factor in the prediction of the service life of the inner reflector regions. We expect that three single-zone cores with moderator-to-fuel pebble ratios of 3:1, 2:1 and 1:1 will be constructed with hexagonal close-packed geometry in the first phase of the experiments. The 3:1 and 1:1 cores will also be constructed as single-zone cores with stochastic (random) pebble-bed geometry and with the alternate hexagonal geometry for k^ and reaction rate ratio measurements.
3.2 Water Ingress The possibility of steam generator or liner cooling system leaks necessitates the consideration of accidental water ingress in HTGRs [9]. Most graphite moderated HTGR systems are significantly undermoderated for reasons relating to fuel cycle economics (the conversion ratio increases in undermoderated systems so that less fissile material needs to be supplied). This means that these systems will gain reactivity as moderator is added to the core. Arguments based upon the volume of the reactor core as compared with the entire primary cooling circuit and the amount of water contained in the steam generators and liner cooling systems as well as the heat capacity stored in the core are used to limit the amount of water that must be considered in accident analyses to a small fraction of the maximum void space in an HTR core (about 2.5% in the case studied by Hiibel and Lohnert) [23,24,10]. The reactivity increase caused by water ingress into an HTR core depends: • Strongly upon the moderator-to-fuel ratio;as may be seen in the data from Hu'bel and Lohnert [23] and Pelloni, et al. [25]. 5
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• Less strongly upon the reactor temperature (generally larger at high temperatures); and • Also depends upon the fuel type (HEU or LEU) and the burnup status. The only previous water ingress experiments that have been performed for pebble-bed reactor systems are the high-enriched experiments at Graz [16]. The Graz experiments used a relatively small amount of high-enriched pebble-bed fuel in a heterogeneous externally driven system and are thus not very useful in assessing the accuracy of water ingress calculations for low-enriched pebble-bed systems of representative size and neutron leakage. Pebble-bed HTRs usually have a strong reflector effect because of the undermoderated core with nearly 40% void coupled with a high density graphite reflector [26]. The high thermal neutron flux in the reflector compared with the adjacent core region enhances the worth of control rods located in the reflector regions. One of the effects of water ingress is to reduce the worth of the reflector and hence of any control rods or poison ball reserve shutdown sytems (KLAK) located in the reflector regions. Similar considerations apply to the modular prismatic block HTGR designs which usually have a relatively small core diameter in order to minimize the maximum fuel temperatures under accident conditions. All of these effects act to complicate the life of the HTGR designer who must consider all of these effects and their associated uncertainties and guarantee an adequate cold shutdown margin at all times. The use of a uniform hexagonal lattice and plastic foam inserts in the proposed PROTEUS experiments should allow very accurate simulation of water ingress effects and the development of an accurate experimental benchmark for use in validating the design calculations needed to insure the safety of low-enriched, graphite-moderated, pebble-bed systems. The columnar hexagonal lattice with a theoretical packing fraction of 0.6046 is desirable for access to the center of the core for reaction rate measurements without disturbing the entire bed. The changes in neutron balance components caused by water entry appear to be large enough to be experimentally measurable [28]. The use of a single zone critical core will allow the water ingress experiments in PROTEUS to provide information on changes in the reactivity worth of control rods located in the radial and/or axial reflectors in a straightforward and rigorous manner. We expect that at least two single-zone cores with moderator-to-fuel pebble ratios of, say 3:1 and 1:1 would be investigated in the water ingress studies. The same basic results as in the initial phase of the program would be expected along with measurements of the changes in control rod effectiveness and burnable poison material worths caused by simulated water ingress. Some information on the streaming of neutrons between pebbles and the effects of water ingress on them may also be able to be extracted from single-zone core simulated water ingress experiments in two different known, regular, reproducible lattice geometries.
3.3 Burnable Poisons An important difference between the HEU and LEU systems is linked to the use of burnable poisons in the initial cores to modify the changes in reactivity and power distribution during the transition to an equilibrium core. In the case of hafnium which has been used in Euopean HTGR designs, the use of low-enriched instead of high-enriched uranium causes added uncertainty because of the much larger overlap of the resonance capture in hafnium with 238 U as compared with the 232Th used in the high-enriched uranium reactor fuel cycle (resonance capture in hafnium is largest in the 1 to 10 eV energy range where resonance capture in 232Th is very small but resonance capture in 238 U is large). There is thus considerable incentive to investigate the interactions of hafnium and 238 U in pebble-bed reactor systems. Integral experiments with hafnium absorbers over a wide range fi
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of moderator-to-fuel ratios would also be quite valuable for assessing the accuracy of the presently available neutron cross sections for hafnium. We expect to conduct two series of experiments with hexagonal close packed pebble-bed geometries and moderator-to-fuel ratios of about 3:1 and 1:1, respectively. The number of absorber pebbles will range from zero to enough to fully utilize the available fuel supply and core volume. The primary interest is currently in hafnium absorbers although, at least one core with boron absorber pebbles will also be constructed. Studies of alternative poison nuclides (erbium for example) and geometries may lead to additional experiments. A combination of water ingress and the effects of hafnium absorbers has not been studied before to our knowledge. Such data would be of considerable interest in the design and licensing of gas-cooled district heating reactors in which the water ingress problems are likely to receive more attention than in an electric power generation reactor in which the higher core temperatures and power densities limit the maximum water density [23,24]. The basic critical loading and water ingress experiments require at least 5000 fueled pebbles (see the calculational results section). A much larger number of fueled pebbles or more complex driven configurations will be needed to approach power reactor absorber-to^fuel-to-moderator pebble ratios.
4
Calculational Results
4.1 Introduction Two-dimensional discrete ordinates transport theory calculations have been performed for LEU HTR PROTEUS configurations with LEU AVR pebble bed fuel containing 6 grams of 16.7% enriched uranium per pebble. Basic criticality results plus the reactivity changes associated with possible pebble bed densification (slumping), upper reflector collapse onto the pebble bed, removal of the last layer of pebbles, etc. were calculated. Two-dimensional transport theory calculations were also made for Core 13 of the present PROTEUS LWHCR experiments [33]. They were used to develop and validate an R — 0 geometry control rod model. Good agreement between the calculated and measured [52] worth of the 8 borated steel safety and shutdown down rods in PROTEUS LWHCR Core 13 was obtained (Calc/Exp = 0.93) when an experimentally determined axial buckling in the test zone region was used. This model was then used to compute the worth of the same 8 borated steel safety and shutdown rods as a function of the distance of the safety and shutdown rods from the core-reflector boundary in a 150 cm diameter LEU HTR PROTEUS configuration.
4.2 Method of Calculation and Fuel Specifications The same calculational methods and nuclear data libraries [30,29] were used as in a previous series of two dimensional calculations [36,1]. The previously used LEU AVR fuel pebble specifications were updated to values consistent with the quality assurance records for this fuel obtained from KFA Julich. The only significant fuel pebble specification change was an increase in the thickness of the outer unfueled shell from 5.0 mm to 7.2 mm. This change was based upon a verbal statement by Dr. H. Nabielek of KFA Julich [37]. It has a significant effect on the computed Dancoff correction factor calculation but only a small effect on the overall eigenvalue calculation.
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A small auxiliary code called GHR was used to compute atom densities for the PEBBLE code [38,39] and the MICROX-2 code [40,41,42]. Carbon-to-uranium and carbon-to-235£f atom ratios for various mixtures of the 16.7% enriched AVR fuel pebbles with pure graphite moderator pebbles are given in Table 1.
Table 1: Carbon-to-Uranium Ratios for 16% AVR Fuel M:F Pebble Ratio
C/U Atom Ratio
CP35U Atom Ratio
0 1 2 3 4
634 1258 1881 2504 3127
3757 7451 11144 14837 18530
The significance of these atom ratios for the proposed HTR PROTEUS experiments is due to the following considerations. The older high temperature helium-cooled reactor (HTR) designs were usually undermoderated with C/ 2 3 5 ^ atom ratios in the 4000 to 5000 range (C/U atom ratios in the 800 to 1000 range). The present HTR-500 pebble bed reactor design has a CP35U atom ratio near 10000 with about 8 grams of 10% enriched uranium per fuel pebble. Smaller helium-cooled pebble bed reactor systems such as the MODUL and the district heating reactors are presently being designed with C/235!!/ atom ratios approaaching 15000. The prismatic block modular HTR designs continue to use CP35U atom ratios in the 4000 to 5000 range. PSI edition 13 of the MICROX-2 code [40,41,42] was used to obtain 13 broad group (8 fast (E > 2AeV) and 5 thermal) modified P l s cell averaged, macroscopic cross sections in the Los Alamos "XSLIB" 6E12 format for use in the ONEDANT and TWODANT codes [43,44]. The MICROX-2 calculations used 193 group JEF-1 based nuclear data libraries prepared at PSI [45] with the NJOY/MICROR cross section processing system [46,47,48,49]. The PEBBLE code [38,39] was used to calculate the Dancoff correction factors for the LEU AVR fuel pebbles required by the MICROX-2 code. The pebble bed core region cross sections were obtained for the critical buckling. The reflector region cross sections were obtained with zero buckling. The actual measured neutron absorption cross section of the PROTEUS relector graphite (about 3.8 millibarns at 2200 m/s) was simulated by addition of an appropriate amount of 10B to the JEF-1 graphite data. The Los Alamos developed TWODANT code [44] was used to obtain critical pebble bed heights via dimensional searches and also for ordinary kejf calculations. The radial reflector was 64 cm thick graphite and the axial reflector was 60 cm thick graphite (both with 3.8 millibarn 2200 m/s capture) in all of the TWODANT calculations reported here. The height of the core cavity was fixed at 190 cm. A pure void was used in the core cavity above the pebble bed fuel. All of the TWODANT calculations used the P x , £4 approximation.
4.3 Basic Criticality Results Some basic R — Z geometry criticality results are given in Table 2. These results are for three different pebble bed core diameters (140,150 and 160 cm), one pebble bed packing fraction (0.6046) and two different moderator-to-fuel ratios (1:1 and 3:1). 8
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Table 2: Critical Height Results Filling Factor
M:F Ratio
Core Diameter (cm)
Core Height (cm)
Height-toDiameter Ratio
Number of Fuel Pebbles
Number of Mod Pebbles
0.6046 0.6046 0.6046
1 1 1
140 150 160
119.01 111.69 106.08
0.85 0.75 0.66
4958 5342 5773
4958 5342 5773
0.6046 0.6046 0.6046
3 3 3
140 150 160
179.06 164.87 153.63
1.28 1.10 0.96
3730 3943 4180
11190 11829 12540
4.3.1 Last Layer Worth, etc. The M:F = 3:1 case with core diameter 150 cm and core height 164.87 cm was then selected as a reference point for evaluation of the worth of the last layer of pebbles, etc. A pefj value of 0.007 AK/K was assumed. The following results were obtained: • AKeff = -0.01313(-1.88$) due to removal of the last 6 cm thick layer of fuel and moderator pebbles. • AKejf
— +0.00765(+1.09$) due to the addition of a 6 cm thick layer of 60% dense graphite.
• AKeff = -0.01349(-1.93$) due to the addition of a 6 cm thick layer of 60% dense layer of aluminum. • AKeff — +0.01053(+1.50$) due to removal of the air in the core and in the void (hohlraum) above the core. • AKeff — — 0.00021(—0.03$) due to removal of the water vapor appropriate to 50% relative humidity.
4.4 Miscellaneous Results Some two-dimensional calculations were performed to obtain information on the effects of certain hypothetical geometry changes on the reactivity of the proposed LEU HTR PROTEUS pebble bed critical assembly. The reference configuration for these calculations had a pebble bed core diameter of 125 cm, a moderator-to-fuel pebble ratio of 1:1, a pebble filling factor of 0.62, and a pebble bed core height of 128.15 cm for a height-to-diameter ratio of 1.025. This pebble bed core would contain about 4310 fuel and 4310 moderator pebbles. 4.4.1
Pebble Slumping Reactivity Changes
The number of moderator and fuel pebbles were held constant at the above values for pebble filling factors of 0.62 (reference case), 0.67 and 0.74 (core heights of 128.15, 118.59 and 107.37 cm, 9
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respectively) to estimate the effect of hypothetical changes in pebble bed density. The results are given in Table 3.
Table 3: Pebble Slumping Results
4.4.2
Filling Factor
M:F Ratio
Number of Fuel Pebbles
Number of Mod Pebbles
Core Diameter (cm)
Core Height (cm)
keff
0.74 0.67 0.62
1 1 1
4310 4310 4310
4310 4310 4310
125.0 125.0 125.0
107.37 118.59 128.15
1.0368 1.0162 1.0001
Void (Hohlraum) Region Worth
The reference case (0.62 filling factor) was then modified to remove the void region above the pebble bed to simulate the collapse of the upper reflector. The calculated eigenvalue without the 61.85 cm high void region between the pebble bed and the upper reflector was 1.0506 (+7.23 $ compared to the reference case).
4.5 Safety and Shutdown Rod Worth The R — 0 geometry control rod model that was validated with calculations for LWHCR Core 13 was then used to compute the worth of 8 standard borated steel PROTEUS safety and shutdown rods as a function of the distance of the safety and shutdown rods from the core-reflector boundary in a 150 cm diameter LEU HTR PROTEUS configuration with a M:F ratio of 3:1 and a filling factor of 0.6046. An axial buckling of 1.37 x 10~4cm~2 was used everywhere except in control rod holes which had zero buddings. A value of almost 22 $ (for all 8 rods) was obtained at the minimum possible distance of 3.75 cm between the rod centerline and the core-reflector boundary as shown in Table 4.
Table 4: Worth of 8 Borated Steel Safety and Shutdown Rods Distance of Rod From Core-Reflector Interface (cm)
3.75 5.0 10.0 15.0
K-Effective All Rods Out All Rods In
1.00441 1.00443 1.00458 1.00397
0.85181 0.86146 0.89263 0.91645
10
AK
0.15260 (21.80 $) 0.14297 (20.42 $) 0.11195(15.99$) 0.08752 (12.50 $)
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4.6 Discussion of Calculational Results It should be noted that the GHR/PEBBLE/MICR0X-2/TW0DANT calculational path does not as yet include any corrections for streaming between pebbles [21,22]. The effect of such streaming corrections is normally small. In the water ingress studies cited in [25], the correction for streaming between pebbles was found to increase k^ by a maximum of about 0.15%. The cavity above the pebble bed core is treated correctly in a two-dimensional transport theory code such as the TWODANT code used here (if a high enough SN value is used) without the usual diffusion theory boundary condition approximations [50]. The effect of higher than S 4 angular quadrature on the R — Z geometry calculations should be small but has not yet been checked in these studies. The results indicate that 5400 of these LEU AVR fuel pebbles should be sufficient for the initial phase of the planned LEU HTR PROTEUS experiments involving critical configurations and water ingress studies with moderator-to-fuel pebble ratios of 1-to-l, 2-to-l and 3-to-l [C/U ratios of 1258, 1881 and 2504 (C/U-235 ratios of 7451, 11,144 and 14837)] and the low density columnar hexagonal lattice (theoretical filling factor = 0.6046) that is most desirable from an experimental viewpoint. In the well moderated cases [moderator-to-fuel pebble ratio of 3-to-l (C/U atom ratio = 2500, C/U235 atom ratio = 14800)], the pebble bed core diameter needs to be at least 150 cm in order to obtain criticality with a reasonable core height. In the less well moderated cases [moderator-to-fuel pebble ratio of 1-to-l (C/U atom ratio = 1260, C/U235 atom ratio = 7440)] that are of particular interest in the water ingress portion of the experimental program, the pebble bed core diameter needs to be less than 150 cm in order to obtain criticality with a reasonable core height and number of fuel pebbles. We are thus led to consider a design with a variable active core diameter, with a maximum of about 150 cm. If we are to retain the safety and shutdown rods in their present position (in a ring of diameter 135 cm), which is logistically and economically desirable, these rods need to be located in noses or buttresses which protrude into the core cavity. The core diameter could be reduced to about 125 cm by filling in the regions between the noses with special graphite blocks or more simply by using only moderator pebbles in those regions. The presence of the noses will complicate the physics analysis somewhat but are actually prototypic of various real and proposed pebble bed HTR designs. In the PROTEUS experiments, the safety and shutdown rods are always be fully withdrawn when the system is critical so that no large undesirable experimental perturbations from the presence of the noses are to be expected.
5 PROTEUS Capabilities 5.1 Experimental Facility The PROTEUS reactor is a zero-power reactor that is characterized by a modular construction which allows rapid adaptation to new research conditions. The bottom-, side- and top-reflectors are composed of layers of nuclear grade graphite blocks and leave the inside of the test zone empty. The side reflector consists of three concentric rings of graphite blocks. The blocks in the outermost ring use wedges and bolts for greater structural stability. This type of construction allows the inner side reflector rings to be modified or removed as necessary for different experimental conditions. The PROTEUS facility has, in the past, been used mainly as a coupled reactor system, consisting of a central test zone driven critical by surrounding driver/buffer regions. In the seventies, the
physics of gas-cooled fast reactors was investigated in the test zone, while in the eighties light water high converter reactor (LWHCR) physics has been the subject of interest[33]. Both experimental programs have attracted considerable international interest, the ongoing PROTEUS-LWHCR Phase II experiments, for example, forming the most important single component in the joint R & D activities of PSI, the Kernforschungszentrum Karlsruhe and Siemens/KWU to establish an experimental data base for validating LWHCR design methods.
5.2 Experimental and Analytical Techniques 'The principal strengths of the PROTEUS group which make the facility well suited for the proposed LEU HTR experiments are: • Experience in defining an appropriate set of integral experiments for improving specific power-reactor design accuracies. • Experimental techniques developed over the years for measuring and inferring individual neutron balance components (k^, reaction rate ratios) referred to fundamental conditions [34,35], so that diagnostic information can be obtained with respect to discrepancies in reactivity predictions. The importance of the second point above lies in the fact that the various contributions to neutron balance play a different role in different situations (e.g., when nuclide density changes occur with increased burnup, or neutron spectrum changes upon water ingress). An understanding of the individual components is essential to rule out error cancellation and to allow a reliable extrapolation of experimental results to the power reactor situation.
5.3
Personnel
The experimental and analytical team at PROTEUS has generally consisted of about 4 to 7 physicists and 4 technicians, responsible for the planning, execution and analysis of the experiments on the one hand, and the operational safety and technical maintenance of the reactor on the other. This staffing level is necessary to allow the planning, preparation and execution of the experiments to proceed in parallel without excessive delays. It is also necessary to allow the analysis and evaluation of the experiments to proceed quickly enough to be able to check any doubtful results. About half of the total scientific staff will be provided by PSI and the other half of the staff will be provided by the other participating organizations. The responses received so far indicate that, in fact, foreign scientists will come and take active part in the experiments.
6 International Program This project is intended to serve as a focal point for international collaboration in HTGR research. We hope to bring together good scientists from the various gas-cooled reactor national programs, not only to help share the cost of performing such experiments, but most of all to provide a joint in-house peer review for the planning, execution, analysis and significance of the experiments and to share knowledge to the greatest possible extent. Two different international cooperation mechanisms appear to be available at the momemt. One such mechanism is the IAEA coordinated research program in the framework of the Agency's gas cooled reactor (GCR) working group which is already approved, the other is the on-going US/FRG/F/CH umbrella agreement for cooperation in GCR research and development. 12
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7
Conclusions
New LEU-HTR experiments in the PROTEUS facility are proposed. They will provide data on: • the criticality of simple, easy to interpret, single core region HTR systems for several moderator-to-fuel ratios; • the changes in reactivity and control rod effectiveness caused by water ingress; • the effect of boron and hafnium absorbers in a low-enriched HTR system. These experiments are needed to reduce uncertainties and to verify codes and data for design and licensing purposes. The PROTEUS experiments should for the first time, provide keff bias factors relevant to the LEU HTR fuel cycle and, in addition, shed light on the individual sources of error. The latter will be achieved by virtue of measured results for individual neutron balance components as well as the investigation of streaming effects. A wide range of conditions will be povered by the experimental program, thereby providing an adequate integral data base for the validation of LEU HTGR design calculations. The PROTEUS facility and its team are well suited to perform these experiments because they have much experience in accurate reactivity and reaction-rate measurements. An international cooperation with the Federal Republic of Germany, the Soviet Union, the Peoples Republic of China and possibly Japan and the United States of America, will provide a cost sharing and bring together the international knowledge in HTGR reactor physics ensuring the quality of the work.
8
Acknowledgements
A large number of people participated in the preparation of this report either directly or indirectly through their previous contributions to the planning of HTGR related experiments at PSI. The criticality and control worth calculations were performed with codes and data supplied by J. Stepanek, S. Pelloni, C. Higgs and P. Vontobel. K. Gmiir and H. Graf were quite helpful in discussions concerning the required modifications of the PROTEUS facility. Many people at ABB/HRB in Mannheim have provided information and support in planning these experiments.
References [1] D. Mathews, R. Brogli, R. Chawla and K. H. Bucher, "LEU HTR Experiments in the PROTEUS Critical Facility," PSI technical memorandum TM-41-88-24, 23 Dec, 1988. [2] J. R. Brown, et al, "Hazards Summary Report for the HTGR Critical Facility," General Atomic Report GA-1210, 1960. [3] C. R. Richey and T. J. Oakes, "Determination of Neutron Multiplication Factors as a Function of Temperature with the HTLTR," Nucl. Sci. & Eng., 47, 40 (1972). [4] M. H. Merrill, "Nuclear Design Methods and Experimental Data in use at Gulf General Atomic," Gulf General Atomic Report Gulf-GA-A12652(GA-LTR-2), 1973. 13
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[5] K. R. Van Howe and J. R. Brown, "Peach Bottom - Initial Loading to Criticality," USAEC Report GAMD-7351, Gulf General Atomic, 1966. [6] G. Ivens, W. A. Simon and K. E. Asmussen, "Verification of HTR Characteristics at the AVR and the Fort St. Vrain Reactor," GA Technologies Report GA-A17575, June, 1984. Presented at the Jahrestagung Kerntechnik 1984, May 22-24, 1984, Frankfurt, West Germany. [7] W. A. Simon and G. C. Bramblett, "Fort St. Vrain Reactor Performance and Operation to Full Power," GA Technologies Report GA-A16774, May, 1983. Presented at the Third Japan - U.S. HTGR Safety Technology Seminar, New York, NY, June 2-3, 1982. [8] A. J. Neylan, D. V. Graf and A. C. Millunzi, "The Modular High Temperature Gas-Cooled Reactor (MHTGR) in the U.S.," GA Technologies Report DOE-HTGR-87-088, August, 1987. Presented at the International Seminar on Small- and Medium-Sized Nuclear Reactors, August 24-26, 1987, Lausanne, Switzerland. [9] William Kerr, Chairman, U. S. Nuclear Regulatory Commission Advisory Committee on Reactor Safeguards, "Preapplication Safety Evaluation Report for the Modular High Temperature Gas Cooled Reactor," Letter to Lando W. Zech, Jr., Chairman, U. S. Nuclear Regulatory Commission dated October 13, 1988. [10] J. Szabo, E. Greenspan, A. Ketter and S. Ron, "Reactivity Effects of Water Ingress in HTGRs - A Review," IAEA-TC-610, November 1987. [11] Y. Kaneko, "Reactor Physics Research Activities Related to the Very High Temperature Reactor in Japan," Nucl. Sci. & Eng., 97, 145 (1987). [12] W. Schercr, H. Gerwin, T. Kindt and W. Patscher, "Analysis of Reactivity and Temperature Transient Experiments at the AVR High-Temperature Reactor," Nucl. Sci. & Eng., 97, 58 (1987). [13] P. Pohl, M. Wimmers, T. Kindt, W. Feltes, U. Schmid and H. Jung, "Determination of the Hot- and Cold-Temperature Coefficient of Reactivity in the AVR Core," Nucl. Sci. & Eng., 97, 64 (1987). [14] W. Scherrer, "Das kritsche HTR-Experiment CESAR-2," Report JU1-975-RG, Kernforschungsanlage Jiilich, Institut fur Reaktorentwicklung (1973). [15] H. Schaal, "Entwicklung eines Rechenverfahrens zur Bestimmung von effektiven Plutoniumwirkungsquerschnitten in niedrig angereichten HTR-Gittern und Test am Experiment CESAR-II," Report Jiil-1757, Kernforschungsanlage Jtilich, Institut fur Reaktorentwicklung (1982). [16] F. SchUrrer, W. Ninaus, K. Oswald, R. Rabitsch, Hj. Miiller and R. D. Neef, "Steady-State Neutronic Investigations to the Accident of Water Ingress in Systems with Pebble-Bed HighTemperature Gas-Cooled Reactor Fuel," Nucl. Sci. & Eng., 97, 72 (1987). [17] I. Johnstone and V. E. Delia Loggia, "Experimental Results from the UKAEA Reactor Physics Programme on Low Enrichment HTR Lattices at AEE Winfrith," D. P. Report 730, 1970. [18] B. G. Chapman, "Operation & Maintenance Experience with the Dragon Reactor Experiment," CONF 740501, Proc. ANS Topical Meeting on Gas Cooled Reactors, 1974. 14
(ok 2
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[19] "Present Status of HTGR Research & Development," Japan Atomic Energy Research Institute, January, 1988. [20] T. King, "Licensing of Advanced Reactors: Status Report and Perspective," Proceedings of the International Reactor Physics Conference, Jackson Hole, Wyoming, September, 1988. [21] R. Hecker and L. Wolf, "The Significance of Critical Experiments for Core Calculation and Design of High-Temperature Gas-Cooled Reactors Constructed on the Pebble-Bed Principle," Nucl. Sci. & Eng., 97, 1 (1987). [22] R. D. Neef, "Ausbreitung von Neutronenwellen in statistischen Schuttungen von kugelformigen Elementen," Report Jul-1273 (1976). [23] H. Hiibel and G. Lohnert, "Das Sicherheitskonzept des HTR-Modul, veranshaulicht am Beispiel des Wassereinbruchs in den Primarkreislauf," Report Jiil-Conf-53, Kernforschungsanlage Jiilich, Institut fur Reaktorentwicklung (1985). [24] D. Leithner, "Reaktivitatseffekte bei Wassereinbruch in der Kern eines Hochtemperaturreaktors," D. Leithner, Hochtemperaturreaktorbau GmbH, Mannheim, personal communication, 1988. [25] S. Pelloni, W. Seifritz, J. Stepanek and P. Stiller, "Parameterstudie zum Wassereinbruch beim HTR," EIR internal technical memorandum TM-22-87-31, 1987. See also; S. Pelloni, W. Seifritz, J. Stepanek, P. Stiller, W. Giesser and D. Leither, "Parameter Study on water ingress in a high temperature reactor," Kerntechnik 53 (1989) No. 3 [26] R. Chawla, "On the Minimum Overall Size of a Reflected Pebble-Bed Reactor," Nucl. Technol., 52, 306 (1981). [27] "GHR-PROJECTSTUDIE Heizwerk mit gasgekuhltem Hochtemperaturreaktor SICHERHEITSBERICHT der Anlage fur die Erzeugung nuklearer Nahwarme," Report EIR-TM-2386-06, Juli 1986. [28] R. Chawla, "On the Reactivity Effects of Water Entry in Pebble-Bed HTR Lattices," Ann. Nucl. Energy, 8, 525 (1981). [29] S. Pelloni and J. Stepanek, "Analysis of LWHCR-PROTEUS Phase H Experiments Performed Using the AARE System with JEF-1 Based Data Libraries, and Comparison with other Codes," PSI-Bericht Nr. 3, Paul Scherrer Institut (1988). [30] J. Stepanek and C. E. Higgs, "A General Description of AARE: A Modular System for Advanced Analysis of Reactor Engineering," Proceedings of the International Reactor Physics Conference, Jackson Hole, Wyoming, 1988 [31] H. Sommer and B. Ribbat, "HGR 10 MW: The Technical Concept of the Gas Cooled Heating Reactor," Proceedings of the First International Seminar on Small and Medium Sized Nuclear Reactors, Lausanne, Switzerland, August 24-26, 1987. [32] E. Teuchert, H. Gerwin, K. A. Haas, and H. J. Rutten, "Pebble Bed Reactor Under Peu a Peu Fuelling for Decentralized District Heating," ENC ' 86 Transactions, Geneva, June 1-6, 1986.
15
E-6
[33] R. Seiler, R. Chawla, K. Gmttr, H. Hager, H. D. Berger and R. Bohme, "Investigation of the Void Coefficient and other Integral Parameters in the PROTEUS-LWHCR Phase II Program", Nucl. Technol., 80, 311 (1988). [34] K. Gmiir, "Techniques of Reaction Rate Measurements on the PROTEUS Reactor," EIRBericht Nr. 529, 1984. [35] K. Gmiir and R. Chawla, "Reaktionratenmessungen am HTR-PROTEUS," EIR internal technical memorandum TM-PH-767, 1979. [36] D. Mathews, "HTR PROTEUS: Two Dimensional Calculations With LEU AVR Fuel," PSI internal technical memorandum TM-41-89-04, 8 Feb. 1989. [37] H. Nabielek, KFA Jiilich, personal communication, 27 Feb. 1989. [38] J. Stepanek, K. Herda and P. Koehler, "PEBBLE: Ein RSYST-Modul zur Berechnung der Entkommwahrscheinlichkeiten fur zweifach heterogene Anordnung des Brennstoffes," EIR internal technical memorandum TM-PH-690 (1977). [39] E. Teuchert and R. Beitbarth, "Resonanzintegralberechnung fur mehrfach heterogene Anordnungen," Report Jul-551-RG, Kernforschungsanlage. Jiilich (1974). [40] D. Mathews and P. Koch, "MICROX-2 An Improved Two-Region Flux Spectrum Code for the Efficient Calculation of Group Cross Sections," General Atomic Company Report GA-A15009 (1979). [41] C. E. Higgs, "Implementation of MICROX-2 with its Associated File Handling Programs and MICROBURN at EIR," TM-22-84-4 (EIR internal distribution) (1984). [42] D. Mathews, "PSI Edition 12 of the MICROX-2 Code," TM-413-88-04 (Eir internal distribution) (1988). [43] R. Douglas O'Dell, Forrest Brinkley, Jr., and Duane Marr, "User's Manual for ONEDANT: A Code Package for One-Dimensional, Diffusion-Accelerated, Neutral Particle Transport," Los Alamos National Laboratory Report LA-9184-M (1982). [44] R. E. Alcouffe, F. W. Brinkley, D. R. Marr, and R. D. O'Dell, "User's Guide for TWODANT: A Code Package for Two-Dimensional, Diffusion-Accelerated, Neutral Particle Transport," Los Alamos National Laboratory Report LA-10049-M,Rev. 1 (1984). [45] P. Vontobel and S. Pelloni, "JEF/EFF Based Nuclear Data Libraries," PSI Bericht Nr. 636 (1987). [46] R. E. MacFarlane, D. W. Muir, and R. M. Boicourt, "The NJOY Processing System, Vol I: User's Manual and Vol II: The NJOY, RECONR, BROADR, HEATR, and THERMR Modules," Los Alamos National Laboratory Report LA-9303-M (ENDF-324), Vols. I and II (1982). [47] R. E. MacFarlane and D. W. Muir, "The NJOY Nuclear Data Processing System Vol. Ill: The GROUPR, GAMINR, and MODER Modules," Los Alamos National Laboratory Report LA-9303-M (ENDF-324), Vol. HI (1986).
16
E-6
[48] S. Pelloni, C. E. Higgs, and J. Stepanek, "The NJOY Nuclear Data Processing System: The COLLGR, ECOMBR, SEPR, DICTR, MANAGR, EPLOTR, CPLOTR, and ACOMBR Modules," EIR-Bericht Nr. 566 (1985). [49] D. R. Mathews, J. Stepanek, S. Pelloni, and C. E. Higgs, "The NJOY Nuclear Data Processing System: The MICROR Module," EIR-Bericht Nr. 539 (1984). [50] H. Gerwin and W. Scherer, "Treatment of the Upper Cavity in a Pebble-Bed High-Temperature Gas-Cooled Reactor by Diffusion Theory," Nucl. Sci. & Eng. 97, 9 (1987). [51] V.Driike, W. Litzow and N. Paul, "Ergebnisse der kritisschen Experimente zur OTTO Beladung an der kritischen Anlage zum Hochtemperaturreaktor (KAHTER)," Kernforschungsanlage Jiilich Bericht Jiil-Spez-187, Dezember, 1982. [52] H. D. Berger, H. Hager, K. Gmiir and R. Seiler, "Auslegungsrechnungen und Inbetriebnahmeresultate fur den PROTEUS FDWR-Kern 13," PSI internal technical memorandum TM41-88-03, 25. August 1988.
17
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XA0101510
Present
Status
of
for
Re s e a r c h
and
Development
HTR i n Chi n a
Wang D a z h o n g Zhong Daxin Xu Y u a n h u l I n s t i t u t e of N u c l e a r E n e r g y T e c h n o l o g y Ts i n g h u a Un I v e r s 11 y
Beijing,
China
June
E-7
ABSTRACT The HTR R & D Project is being carried out in the relevant
institutions
in China. Some topics are covered such as. fuel element technology, graphite development.
fuel element handling system, helium technology.
fuel
reprocessing technology as well as HTR design study. Some results of HTR research work are described. In addition, to provide a test facility for investigation of HTR Module reactor safety and process heat application of HTR. a joint project on building a 10MW test HTR with
Siemens—Interatom.
KfA Juelich. and I.NET is going on. The conceptual design of 10MW test HTR has been completed by joint group. In parallel the application study of HTR Module is being carried out for oil industry, petrochemical industry as well as power generation. Some preliminary results of application study for example, for heavy oil recovery on Shengli oil field, process heat application in Yan shan petroleum company are described.
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1. The Status of HTR Program The research and development program of HTGR in China began in mid seventies. In the first phase begining from 1974. the target was to design and
construct a lOOMWt HTGR thorium thermal breeder with two-zone pebble
bed
core. In parallel with the design work of the reactor, a series of
research
work for HTGR development were carried out which included the
manufacture of
fuel kernels, coated particles and several kinds of
spherical fuel element, the characterization of graphite, the prestressed concrete reactor vessel technology, the HTGR components development, such as graphite core structure, control rod and its drive, charge and discharge of spherical fuel, helium circulator, steam generator as well as the reprocessing technology of thorium fuels. But the HTGR breeder project was stopped in end of seventies due to various reasons mainly
in financial
problem. The second phase of HTGR reserach and development program was in period of the Sixth Five Year Plant (1981 — 1985) . The State Science and Techno logy Commission gave support to continue some basic technology development and to investigate the possibilities of nuclear process steam and heat
for
industry application. The work was concentrated in certain areas: - to continue the coated particle fuel technology development and to improve some laboratory facilities. - to investigate the HTR module concept design and to develop design tools e. g. the computer
codes for phisics,
thermohydraulics and safety analysis. - Investigation of the application possibilities of HTGR for heavy oil recovery, chemical industry, refinary,
oil-shale
retorting and coal gasification. The present research and development program on HTGR (third phase) is covered in the
high technology research and development program.
Considering the development of nuclear energy in the next century in China, It is important to develop the advanced reactors which have iherent safety, economical viability and highly fuel utilization. The high temperature gas
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cooled
reactor
(HTGR)
is the one of these advanced reactors to be
selected to do research and development work for future applications. The main topics of this program include. - HTR fuel technology development, preparation and manufacture of fuel kernels, coated particles and fuel element in laboratory scale,
characterization and qualification as well as the
irradiation test of coated particles and fuel element. - Graphite, metallic and ceramic materials development. - Helium technology and components development. - Charging and discharging technique for spherical fuel elements. - Research work on the thorium-uranium fuel cycle and HTR fuel reprocessing technique. - Design study and safety analysis for modular concept of HTR. - Application study of the modular HTR for different potential users. For management and coordination of HTR research and development program a special unit was established. The institutions which take part in this program are the Institute of Nuclear Energy Technology (INET) .
Beijing
Institute of Nuclear Engineering (BINE) , SoutlHWest Center of Reactor Research others.
(SWCR)
and Shanghai Institute of Nuclear Research
(SINR) and
The aims of this program are to research and develop the key
technologies
and to investigate the possibility of
constructing
a
demonstration modular HTR plant. 2. The 10MW HTR Test Module Concept In order to introduce and develop the modular HTR technique in practice, a joint project on construction of a 10MW HTR Test Module which collaborated among the Institute of Nuclear Energy Technology
(INET) of Tsinghua
University PRC, Siemens-Interatom &nbH and the Nuclear Research Center Juelich
(KFA) FRG. is started.
The 10MW Test Module will be constructed at the site of INET in the north-vest of Beijing. The main object of the Test Module is to verify and demonstrate the
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relevant and unique features of HTR-Modular on a real nuclear test facility. Therefore, the aims for the Test Module have been defined jointly as follows: - The test Module will be designed for a wide range of possible applications,
e. g. electricity, steam and district heat
generation in the
first operation phese and process heat
generation, methane reforming in the second phase. - The relevant components can be tested and proven at nominal conditions, e.g. graphite core structures, steam generator, helium blower and fuel handling facility. - Verification of the inherent safety features of the HTR-Module such
as negative temperature coefficient of reactivity,
temperature
limitation due to passive decay heat removal and
limitation of power excursion due to water ingress. - The test Module is capable to withstand extremely high core temparature, so that fuel element mass-test could be carried out for nominal reactor conditions at temperature up to 1600'C. The conceptual design of the 10MW Test Modele was carried out jointly by Interatom and W E T in 1988. The main data of the Test Module are as follows: Maximum thermal power Average thermal power Primary heliwn pressure
20 m 10 MW 30 bar
Helium temperature
290X2/700X2 35 bar
Secondary steam pressure Life steam temperature
435 "C
Power density Core diameter
2 MW/m3
Average core height Height/diameter radio
176 cm 0.93
Number of fuel element
27.000
Heavy metal content
5 g/fuel element
Average burn-up Fuel element incore time
80.000 MWd/t 1078 EFPD
Loading scheme
OTTO
190 cm
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Fig. 1 and 2 show the vertical cross section of the reactor and the primary circuit of the Text Module. It has the following
important design
features: The reactor core and steam generator are housed in two separate steel vessels and positioned side by side in a staggered arrangement. The graphite cylinder core which holds the pebble bed has a diameter of 1900 mm and a height of approx. 2200 mm. This is adequate for the required volume of the active core which amounts to 5 m'. The new fuel elements are charged from the top of the core via five charging tubes. The fuel elements are removed from the core-bottom via a fuel element discharge tube with an inner diamenter 500mm. Decay heat is removable by surface coolers outside the reactor vessel. The surface cooler system is subdivided into two trains, it is sufficient to dissipate decay heat by means of passive heat transfer mechanisms to the simple surface coolers. The primary helim pressure of 30 bar is chosen, the inlet and outlet temperature of the core are 250*C and 700'C • A secondary pressure of 35 bar and a life steam temperature of 435'C are chosen so that the small standard industrial turbines can be used. The average thermal power of the Test Module plant is set to be 10 MW. In order to enhance the experimental flexibility the maximum thermal power output is set to be 20 MW. Some variants of the secondary heat sink have been evaluated. As sample fig. 3 shows the heatflow diagram for district heating and electricity generation with the maximum thermal power of 20 MW. The respective figures for the normal operation case (10MW) are given in brackets. The basic data of the steam generator are: Helium flow (shell side)
8.6 (4.3) k g / s
Helium pressure
30 bar
•
Helium temperature
250/700 *C
Water/steam flow (tube side)
7.5 (3.75) k g / s
Water/steam pressure
40/35 bar
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Water/steam temperature
150/435V
Steam generation rate
27 (13.5) t h
Fig 4 Shows the time schedule of the whole cooperation programme. The 10MW Test Module will take 5 years for design, construction,
installation
and commissioning. The basic design will be performed jointly after this project is approved by both Chinese and german governments. Major topics during the basic design phase are - Design of the Test Module with the aim of preparing adequate documents for the preliminary safety analysis report. - Start of research and development as well as training programmes. - Cost review for the Test Module programme with special emphasis on the definition of the local/import sharing. - First evaluation of application concepts in the field
of
electricity generation, cogeneration of electricity and process steam and tertiary oil recovery. A period of about 1 year is assumed for completion of this phase. 3. Application Study of HTR in Heavy Oil Recovery and Chemical Industry The heavy oil reserve is relatively rich in China. Since begining of 80', heavy oil recovery by injecting steam had been developed fields in order to increase the crude oil production.
in several oil But conventional
thermal recovery of heavy oil needs a great quantity of high temperature steam with high pressure. About 30-40% of produced crude oil would
be
consumed to supply such amount of injecting steam. In some pilot areas of thermal recovery, the injecting steam with temprature 355'C and pressure 170 bar is produced from the small oil-fired boilers. Therefore,
using HTR
instead of oil-fired boilers can save a great amount of crude oil. For the investigations on the use of the HTR in heavy oil recovery, the Shaajasi section of Shengli oil field has been selected to serve as reference case for this study. The initial 0IP in Shanjasi section is
—o—
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6 6 A 10 tons. expected.
the exploration is not finished and I00xl0 6 tons of 0IP is
The crude oil is very heavy and does not flow at reservoir
conditions. Therefore, the heavy oil has been extracted by the steam-soak process since October 1984. The production planning aims at an output of 1 million tons heavy oil per year with subsequent upgrading in a special refinery. The main aims of the application study are to find out. whether. - the physical properties of the Shanjasi reserovir are suited for a continuous steam injection, i.e. oil recovery by steam drive process. - the nuclear steam is economic compared to conventional steam generated by oil-fired boilers. The evaluation of the physical properties of Shenjasi reservoir with respect to the steam drive process was a major effort of the study.
For
calculation of the reservoir properties a numerical simulation computer code (NUMSIP model ) has been developed by INET. Based on the results of calculation and evaluation, a option of using 2 HTR with total thermal output 400MW for steam and electricity generation was proposed. As it is shown in Fig.5 the oil production capacity of 1 million tons per year (about 20,000 bbl./d) may last up to 13 years with steam soak and partly steam drive. Then steam soak will decrease, and the production by steam drive will be remained for other 20 years. It means in total a duration of about 33 years with a nearly continuous oil production of
0.5
million tons per year. According to preliminary investigations a ratio of 4 tons steam to 1 ton oil can be expected. With this boundary condition a steam
production
of about 2 million tons per
t / d = 2 5 0 t / h ) is needed.
.year
(about
6000
This steam amount can be generated by one
HTR-module with a capacity of 200 MW thermal.
Another HTR-Module is
necessary for electricity generation to meet the electricity demand in oil-field area.
Therefore, 2XHTR-module plant is proposed as a energy
source for the Shanjasi heavy oil-field. Fig.6 represents the flow scheme for such 2xHTR-roodule plant. It is proposed to interconnect the feedwater/steam circuits of both
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"
HTR-modules to get a cogeneration plant generating injection steam for heavy oil recovery and electricity for the Shengli grid. generates
Each steam generator
about 77 k g / s live steam with a pressure of 190 bar and
temperature of 530'C. From there the steam is distributed to injection wells and the turbine. The electrical output is about 75 MW. The steam
leaving
the turbine at different extractions is partly used for the preheaters to form a part of the feedwater. The other part of feedwater (with 70 kg/s) is fed from outside via a water treatment station. The economic assessment is performed on the basis of the dynamic cost calculation.
The most important results of this application study are as
following: - Recoverable oil by steam drive using HTR nuclear steam supply system: approx. 15 million tons in the total period of about 30 years. - Average oil production by steam drive: approx. 0.5 Million tons per year. - Additionally recoverable oil due to oil substitution by nuclear energyi approx. 4 million tons (25X10 obi) - 75 MWe of electricity could be supplied to the Shengii-oilfield grid. The electricity production amounts to 600 million KWh per year which is equivayalent to an oil consumption of about 140, 000 t/a. - Electricity produced by the HTR-module plant will have the near same price
(levelized cost) as produced by an oil-fired power
plant. The application study of use HTR in heavy oil recovery has been carried out jointly by INET, Beijing and KFA, Juelich, and supported by Shengli oilfield Co. in Chinese side and Siemens KWU/Interatom in German side. It is well
understood the shanjasi HTR application study is a preliminary
evaluation in the technical and economical possibilities. But it can be regarded as an example, and other heavy oil field with similar properties may also be possible candidates for HTR application project.
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6 Go
E-7
The similar investigation has also carried out for use of the HTR in Chemical industry.
The Yangshan Petrochemical corporation
located in
South-West of Beijing is selected as a reference user for this study. According to the
total requirement of steam in different pressure and
temperature ranges is approx. 730 t/h in summer and 1650 t/h in winter, and total annual consumption of electricity is about 1 billion kwh which should be partly supplied from own cogeneration power plants. 4XHTR-Module cogeneration
plant with thermal power output 800 MW and 2XHTR-100
cogeneration plant with thermal power output 500 MW have been proposed by the joint
application study between Chinese and german institutes and
reactor companies. More detailed economic and safety studies for application HTR in chemical industry will be continued.
E-7
G(oA
REKECT3R 9 0 0 [12.1
FUEL-ELEMENT LOADING t l » OUTER CORE )
CORE 8OTTOH gQBONATED CARBOH
rner
FIG. l
VERTICAL CROSS SECTION OF THE TEST MODULE E-7
FIG. 2
CROSS SECTION OF THE PRIMARY CIRCUIT OF THE TEST MODULE
E-7
35 bar 3304 kJ/kg 7,49 kg/t « 27 l/h
Steam Generator
4310 kW
40 bar I5O*C 634kJ/kg 7.49 kg/s j V Steam "SnrConverting
2,5 bar I27,43*C 2716,4 kJ/kg 7,1 kg/s
37,3 kg/s
US Valve Air Temperature 25.6*C
4.76 ^ar I5O*C 632 kJ/kg 7.49 kg/s
Feed- Woter Tank 4.76 bar I5O*C
District Healing Extraction Sleom 4600 kW
8 bar 25O'C 2950.4 kJ/kg 0.39 kg/s
2,4 bor 126.09*0 504.7 kJ/kg 7.1 kg/s
FIG. 3
4 bor 60*C 251.4 kJ/kg 125 kg/s
0,4 bor
HEAT FLOW DIAGRAM - D I S T R I C T HEATING AND POWER GENERATION E-7
°\ COOPERATION PHASES
BASIC DESIGN PHASE
DETAILED DESIGN PHASE
SITE APPROVAL
•END OF BASIC DESIGN PHASE COST REVIEW
PHASE OF TEST MODULE OPERATION
CONSTRUCTION PHASE
TIME (YEAR) TEST MODULE PROGRAMME MILESTONES
. COST REVIEW • CONSTRUCTION LICENSING • ORDER PLACEMENT COMPONENTS
START OF OPERATION
BASIC DESIGN DETAILED DESIGN LICENSING PROCEDURE PRELIMINARY SAFETY REPORT
COST ESTIMATION: • GERMAN PART • CHINESE PART
FINAL S'AFETY REPORT
UPOATEO FINAL SAFETY REPORT
CONSTRUCTION: • SITE OPENING •ORDER PLACEMENT • MANUFACTURING OF COMPONENTS • BUILDINGS • INSTALLATION • COMMISSIONING OPERATION: • PHASE I: WITH STEAM GENERATOR • PHASE II: WITH STEAM REFORMER TRAINING R+D APPLICATION STUDIES
FIG 4
OVERALL TIME SCHEDULE TEST MODULE CHINA E-7
Steam Soak Steam Drive Conv. Steam Drive HTR Waterflooding f106t/aJ
1.5
ID
••••••••••••••••••••••••••••••••v
2014
FIG. 5
2019
2024
PRODUCTION SCHEDULE AT SHANJASI
2029
2034 [year] E-7
HTR Module Shengli Oilfield ..}..
1 i
i (r
1
Ot 07 Oi
•::::M;r(i"::::::i::::z. :x.-Jr. -
• ft
M OS OS 0/ OS iK 10 u I? 15 14 15 16
Reactor flulldlng Anno to Oeactor ( S t a r t - u p / ^ H i t a o m
530 °C
HTR-ModJe
HTR-MotUe
70 kg/s
FIG. 6
HTR-2 MODULE S I T E PLAN A N D E-7 FLOW SCHEME OF POWER AND STEAM COGENERATI ON
XA0101511
THE PROSPECTS" OP HTR PLANT IN CHINA
Ye Liangcheng
Southwest Center for Reactor Engineering Research & Design May
1939
Chengdu, China
66*
E-8
The Prospects of HTH Plant in China
A working team was set up about one year ago for the feasibility study of construction of a MHTGR demonstration plant in Zhongqing . The first phase of the study has been completed indenpendently by. Chinese specialists. The report ( draft ) has been submitted to related orgnizations for evaluation. The main results of the study are :
1 , The great attraction of MHTGR to Zhongqing. - Zhongqing is the largest city in the southwest of China. The area ( including suburbs ) is 22 341 Km , and the population is 14.27 million. It is an important economy developing district. The shortage of electricity supply is very severe. As more than 40# of power is provided by hydropower, in the dry seasons the situation becomes more graver. Some factories have to stop work for
3-4
days each week. Therefore, it is absolutely necessary to construct new power plants. - Increase of coal production in Zhongqing is very difficult. Import of coal from outside districts is limited by the difficulties of transportation. As lack of coal, in the coal-fired power plants part of their units have to be stoped operation. The construction of new coal-fired power plants is almost impossible. - Development of hydropower may be possible, because southwest China abounds hydropower resources. However the more the portion of hydropower in the capacity of grid, the larger the shortage
of "
electricity supply in the dry seasons. 1 E-8
- Chinese government has decided to develop 600 M'ie PWR for its nuclear power plants in the future. As the financial resources of the central government is not plenty, the number of PWR which can be built by the end of this century will be limited, and all of them will be distributed in the east of China. The construction of 600 I'P.7e PWR plant in Zhongqing will be impossible. - Modular HTR has inherent safe property, it can be built close to city or to industry center..The small unit size is more suitable for local electricity grids. The small investment capital has great attraction for local users. Therefore, Zhongqing utility is interested in the construction of a MHTGR demonstration plant.
2, Yuzue Tuanbao— Prelimilary selected site. The working team has made an on spot survey of 8 possible sites. They collected all informations as they can do. After prelimilary evaluation, Yuzue Tuanbao may be the best site for a MHTGR demonstration plant. Its geological structure is stable..The distance from this site to city is about 30 km. It is located in east of Zhongqing and near Yangtze River. There is enough Cooling water resources and the heaviest component of the plant could "be transported through the River.
3, The prelimilary technical scheme of Zhonjqirig Plant. It was assumed that the major equipment for nuclear island will be introduced from oversea. The major equipment for conventional island will be supplied by Chinese industry. The architecture, equipment erection and most of services will be provided by China. 2 E-8
The available modular HTR designs today are 350 MWt MHTGR ( GA concept ) and 200MWt small modular HTR ( Siemens/ Interatom Conpet)• The suitable size of turbine-generator set for Zhongqing and most of Chinese local utilities may be 300 HWe. Therefore, there are two basic technical schemes for Zhongqing demonstration plant: —
2 modules ( 2 x 350 MVt ) + 1 TG- set ( 300 HWe ) .
—
4 modules ( 4 x 200 MWt ) + 1 TG set ( 340 MWe ) .
It has been pointful from prelimilary technical analysis tnat each scheme has its own advantages and both of-them will be feasible for Zhongqing plant.
4, The portion of equipment supplied domestically. Chinese power plant equipment manufactories have been able to supply 200 - 600 MWe conventional plant units, not only for inland, but also for export. The turbine-generator set for MHTGR plant is almost same as that for coal-fired plant. According to the prelimilary analysis-more than 90f$ of equipment for conventional island of Zhongqing MHTGR plant could be produced in China. Since China has been developing nuclear power,.a basic infrastructure for nuclear power industry has been formed. The technology and experiences of construction of PWR plant would be beneficial for developing HTR plant. Even for Zhongqing demonstration plant, the portion of equipment of nuclear island supplied by China may cost about 5 40 million. In China, A High Technology Programme, including development of HTR technology, is been implemented now. Some results of research work for HTR have been achieved. This would be a great promotion for
3 E-8
6}A
Zhongqing HTR demonstration plant. For example, the coated fuel particles have been produced in laboratories at Southwest Center for Reactor Engineering R & D and Qinghua University. Spherical fuel elements have also been made and tested. The particle properties, such as particle size, coating density, thickness and strenth, kernel metallography,and spherical element strenth have been examined. The results are very well. If
introducing
some special eauipment ,
the production capability of HTR fuel element would be further increased. We would be able to supply fuel elements not only for Chinese HTR plants, but also for that of foreign countries. It is very important to increase the portion of equipment supplied by China for
reducing
the need of foreign exchange and the
capital cost of HTR plant.
5» The results of economic estimation. An estimation of investment cost and electricity generation cost for Zhongqing MHTGR demonstration plant has been completed. The input data for nuclear island were based on the information from GA and Siemens /In teratom. The other data were provided by Chinese Orgnizations. The exchange rate for Chinese Yuan to US doller is: -1 $ = 3.74 y, to Germany Mark is: 1 DM = 2 Y. For 2 x 350 MWt MHTGR plant ( 300 MV/e ), the basic capital cost is: 3 212.47 M + ¥ 564.43 M or
Y 1355 M
or
V 3953 / kw
For 4 x 200 MWt plant ( 340 MV/e), that is; DM 513 M + ¥ 589.14 M
4 E-8
o
or
V 1615 M
or
V 4750 / kw
The Cost of MHTGR plant corresponds to that of the same size PWR plant built in China. Following commercialization of this HTR plant, its economic competitiveness will be reinforced and would be accepted by more utilities. These costs are higher than that for coal-fired plant in China. For example, the cost of MHTGR .is almost 2 times of that of import coal plant or 3.5 times of Chinese Coal plant. As the coal cost in China is very cheap today, the electricity generation cost would be much higher than that of coal plant. However, this situation will be changed in the near future. Since the coal supply becomes more and more difficult, the nuclear power would be '.a reasonable option. Zhongqing is going to make its contribution to construction of a MHTGrR demonstration plant and to promote the development of HTR. But as an utility, its main concern is shown for construction cost of the plant and electricity production cost. It w$ld be unreasonable for Zhongqing utility to bear all the technical and economic risks as well as the higher cost of a demonstration plant. All beneficiaries of the demonstration plant, especially the contractors arid subcontractors, should share the risks and make their economy contribution . They will share all the experiences and information of construction and ..operation of the .plan*.
6, The prospect of HTR plant in the future. Following economy reform in China, the development of electricity power will be localization. Many local utilities will choose small nuclear power plant. Since the beginning of this independent
5 E-8
study, many utilities, such as V/uhan City, Zaoyian City ( Hubei Province ), Nanjing City, Huiyan City ( Guangdong Province ), and Hainan Island, have shown their interests in developing HTR plant. The important step for opening up the HTR market is construction and operation of a demonstration plant. The best way for speeding up the construction of Zhongqing I'IHTGR demonstration plant
may be to strengthen the international cooperation. Pore
example, to form a joint venture by Chinese and Foreigners. As Chinese products and services are cheaper, the participation of China in HTR development would reduce the capital cost of the plant. More and more utilities, including many developing countries, would be able to accept this kind of small nuclear power plant.It could be expected that commercial HTR plants would be constructed in the world at the beginning of next century.
6 E-8 '
XA0101512 CZECHOSLOVAK APPROACH TO THE POTENTIAL OF HTRGs' INTRODUCTION L. Jakesova, M. Podest - Nuclear Research Institute, Rez near Prague V. Pinkas - Czechoslovak Atomic Energy Commission, Prague Abstract Czechoslovakia - a rather small country with developed utilization of nuclear electricity production in the Soviet types of PWRs - WWERs - is looking forward to the prospects of high-temperature heat provision for processes enabling better utilization of brown coal resources, in the first range with view to the ecology impact improvement. The introduction of the HTGRs could be the best way to solve these problems in the most convenient approach - i.e. through the international co-operation with the USSR. Key words: Nuclear Heat Application,--High Temperature Heat, HTGR reactor safety, coal gazification. Introduction Nuclear power has become an essential part of electricity supply in many countries, Czechoslovakia being one of them. At present, when the state of environment•is continuously worsening due to the burning of fossile fuels with high content of sulfur and nitrogen compounds (not forgetting radioactive elements impurities which are also released and as a result contaminate environment) the enhancing of nuclear power share appears to be the only reasonable way out. It should be stressed that burning any fossile fuels, even "pure" ones, means consumption of atmospheric oxygen and production of carbon dioxide leading to probably already advancing greenhouse effect. In spite of 26,7% of "nuclear" electricity in 1988 Czechoslovakia still has to use a big number of fossil fuelled plants. The biggest ones located in North Bohemian region burn brown rather poor quality coal from nearby mines. Chemical industry of the same area adds to the high level of pollution. It is clearly realized that for the improvement of existing conditions in North Bohemia industrial and mining region is necessary to find ecologically advantageous solution for the future utilisation of brown coal. As one of such solutions which could be introduced in co-operation with Soviet partners is production of synthetic gaseous or liquid fuel using local brown coal and high temperature heat generated by high-temperature gas cooled reactor. As alternative heat sources are discussed pressure water reactors and as alternative heat user - existing and eventual future chemical industry in the area.
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- 2 Technical answer to this problem if found in near future should make a necessary impact on ecological situation of North Bohemia. The Czechoslovak authorities which pay special attention to the area are currently informed about planned developments, similarly the local population is being kept informed. General criteria for HTGR introduction The main part of energy consumed in industrial countries goes to the energy demand of various industries. Direct (as fuel) and indirect (electricity, steam, hot water) consumption of energy resources by industry is more than 60% of the total requirements. Accordingly, industry's share in air pollution is about 50%. If HTRGs are to solve the problem of nuclear energy complex use, they could count on comparable demands of technological and electricity consumers. When evaluationg a possible scale of the introduction of nuclear power it is necessary to take into account temperature at which heat is actually required. The higher the temperature, the more difficult the heat production. Cumulative curves in Fig. 1 show the temperature dependence of the consumption of heat by whole national economy and by industry of the SSSR. We beleave that this type of dependence is common for all countries with highly developed heavy industry and heat consuming chemical industry. Rather different picture of heat consumption in various temperature ranges is in Fig. 2 which represents situation in countries resembling Switzerland. Nuclear heat applications could be devided into two basic categories - use of steam at the temperatures up to 55O*C and use of helium - up to 95O*C and higher. LWR's produce saturated steam at tempreratures below 300*C and pressres up to 80 barr (which can be transformed into slightly superheated steam with pressure up to 120 barr). HTGRs with primary coolant temperatures in the range of 650 - 75O*C (existing prototypes) produce steam with parameters similar to fossile fuelled plants steam, i.e. 54O*C and 180 barr. Even these parameters can satisfy the majority of industrial and chemical energy - technology complexes. As can be seen in Fig. 1 and as given by a number of published data, it is possible to cover about 3/4 of technological heat demands at the temperature up to 54O*C. First generation HTGRs are able to support various petrochemical separative processes and some of reforming technologies. The following advanced HTGRs which should be available after material problems will be definitively solved can be counted upon for coal gasification, metallurgy purposes etd. It can be said that HTGRs may become an universal source supplying different-temperatures heat as well as E-9
- 3 electricity with high degree of efficiency. When satisfying heat demands the essential part of costs is contributed by its transport. Shortening the distance between heat generating and heat consuming plants is always problematical. In the case of coal (or oil)fuelled heat generator the critical point is uniterrupted supply of voluminous fuel, nuclear plant's siting is given by licensed safety limits and conditions. Renewed attention to high-temperature reactors is explained by two main reasons: the first being their inherent safety properties, the second is connected with HTGR's ability to produce high temperature heat besides electricity. In this context the most important inherent (and favourable) properties of HTGRs' are enhanced passive safety and enhanced efficiency by electricity production. Modern nuclear'safety concepts are based on so-colled defence-in-depth which means that all safety related functions are always overlapping. Thus, any fault is compensated of its consequence is lessened by the next (overlapping) defence layer in such a way that the damage to the health of personal and public is excluded. Multilayer defence of any nuclear power plant comprises the number of barriers which prevent radioactivity release from the core during operation, all transients and anticipated accidents. They consist of: - nuclear fuel itself (with different ability to retain fission products), - cladding - walls of primary circuit - confinement of containment* From this point of view high-temperature reactors are very favourable because of a superb inherent safety level resulting from their excellent thermotechnical and neutronic properties. The main of these properties being: - low specific power of the core which corresponds to 5-10% of the relevant value for LWR's, - high heat capacity of graphite matix of the fuel - 500-1000 MJ/K. The heat capacity available in LWR in case of loss-of-coolant accident is about two orders of magnitude lower. From the moment the flow of the coolant to the core is interrupted to the beginning of severe damage to the fuel (no automatic corrective action) the time span in an HTGR is about 10 hours, in a PWR it takes only several tens of minutes. According to the model cumputations of a severe LOCA in WWER-440 (broken main coolant charging line and lost active emergency core cooling) fuel melting starts in 32 minutes. - negative temperature coefficient of reactivity under all conditions, E-9
en
- HTGR's fuel elements (here we speak about spherical fuel of German and Soviet existing and proposed reactors) have very low releases of fission products at all normal and off-normal events - 0,0001% of total fission products inventory. In Soviet VG-400 the major part of the release to the primary circuit, which is less than 1000-2000 Ci, are inert gases. In such a case even complete loss of the coolant is not going to endanger the safety dose limits. (That is a reason why in HTGR's designs no containment is included.) The CSSR projects for HTGR implementation It is safe to assume that HTGR with primary helium temperatures up to 75O°C will not present an unsurpassable technical challenge because of their similarity to the successful prototypes. What is not sufficently verified is actual heat utilization. Taking into account economical and technological possibilities we suppose that the development will start with medium temperature processes (maximum temperature below 550 C) and continue towards high temperature technologies - cool gasification. The first temperature region enables catalytic reforming of methan with subsequent transfer to other applications based on hydrogen. We do not expect that technologies like coal gasification will become industrial earlier than in the beginning of the next century. The main problems envisaged are construction materials, and actual design of chemical reactors. Such materials are in Czechoslovakia not available and their eventual development is believed to be unlikely. So, for the CSSR the only possibility to introduce HTGRs as well as following-up technologies is co-operation with the SSSR and eventual co-operation with other countries on some individual problems. Therefore, the realistic approach to the HTGR's introduction into Czechoslovak integrated power supply system could include an intermediate period when necessary high potential heat will be supplied by fossile fuelled advanced (with minimum impact on environment) power plant. The requisit condition of such an approach is that the actucal heat - chemical system is able to absorb the exchange of a classical plant for HTGR. The CSSR, thus, cannot separately develop all aspects of high-temperature reactor application for process heat purposes. On the other hand, our country has established the large and reliable nuclear engineering base which could be employed to benefit our common purpose - industrial HTGR.
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Fig.
1
STRUCTURE OF ENERGY CONSUMPTION IN THE USSR
CtP
0
200 400 600 800 1000
t,€C
1 - WHOLE NATIONAL ECONOMY 2 - INDUSTRY
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Fig. 2 The energy consumption in Switzerland as a function of temperature
22%
\
1
\
800-
\
Heat market
\
I
\
o oo
6004-.O
D
00
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ess
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i 40
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5560
45' KX>
ENERGY SHARES, %
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LESSING OF THE POLLUTION BURDEN TO THE ENVIRONMENT DUE TO THE SUBSTITUTION OF FOSSIL FUELLED PLANTS BY NUCLEAR
INCREASE OF THE NUCLEAR ELECTRICITY PRODUCTION
year
1990
1995
2000
2005
TWh/y
26,8
46,0
62,8
82,1
DECREASE OF POLLUTIONS' p.y
1
t
i
i : >
t : •
• : i
SO 2
(1O6 tons)
0,63
1,09
1,52
2,0
NO
(1O6 tons)
0,261
0,448
0,612
0,798
6,27
10,86
15,34
20,30
solid (10
tons
E-9
XA0101513
RESULTS AND FUTURE PROGRAMME OF HTR's STUDY Mursid Djokoleiono * Soedyartomo Soentono *
Presented at the Technical Committee Meeting on Gas-Cooled Reactor Technology, Safety and Siting June 2\ - 2 3 . 1989. Dmitrovgrad, USSR.
National Atomic Energy Agency, Indonesia
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RESULTS AND FUTURE PROGRAMME OF HTR'S STUDY Mursid Djokolelono Sudyartomo Soentono
Abstract Study on the application of HTRs for the enhanced oil recovery in the Duri oil field (Sumatra, Indonesia) was performed in 1986/1987. The economic and technological advantages over crude burning option- were identified. Crude oil prices, HTR capital costs, discount rates, company's income structure represented dominant parameters. Further sensitivity calculations on important economic parameters were obtained to reflect the condition of 1988. This nuclear option was also incorporated in the energy planning study for the whole Indonesia using the MARKAL model, and resulted the conditions of its applicability. The scenarios chosen in this MARKAL'study were high and low GDP growth rate, whereas the criteria chosen were the minimum cost with and without a predetermined policy of reduced domestic use of oil. In the high scenario the HTRs as well as the natural gas options could not compete against the low cost boilers with crude-oil fuel. But in the case of reduced domestic oil use the HTRs came out to supplement the crudeburnini? boilers starting in the sixth five year plan (1994 1999), even earlier then the natural gas option. The authors further discuss the industrial environment, in relation with the regional development, the possible local participation, as well as the plan to materialize the merits of this novel application. 1. Introduction A prefeas.ibi.1ity study of the application of HTRs in Indonesia, namely for the heavy oil recovery in the Duri oil field, was completed and reported to the Indonesian Government. The study was performed under the cooperation between KWU, Interatom of the Federal Republic of Germany and BATAN, BPPT, MIGAS, LEMIGAS, PERTAMINA of Indonesia. From the Duri steam flood requirement 4 units of 4 HTR 200-MWt modules were exercised, and positive results were obtained. Various advantages, economical and technological, on the use of nuclear steam supply systems over conventional crude-burning ones were identified (1). E-10
Later an economic study of only one unit of 4 HTR 200MWt module introduction, resulting in a similar positive results, was also presented by KWU (2). Whilst uncertainties associated with assumptions still have to be verified through a comprehensive feasibility study, some issues are hindering the decision to proceed. Among these are short term practice of oil production sharing contracts, actual low oil price*, optimistic view on the availability of other energy sources, institutional and safety aspects (3). The 1988 oil price situation and its declining tendency did not support the (micro-economic) profitability of this nuclear alternative. Further if the burned crude has the price of the well-head cost and the field electricity is provided by the field associated gas or by the crude in the same well-head cost manner, it is very difficult for any other energy source to compete (4). 2. Energy Resources of Sumatra Riau Province Sumatra is a large island among 13 thousand islands of Indonesia. The region is 0.47 million square kilometers, compared to 2.0 millions of Indonesian land and to 5.2 millions of total Indonesian land and sea. Out of to-day's Indonesian population, which is 179 millions, 37 millions (or 20 %) live in seven Provinces of Sumatra island. One of the Provinces is Riau, bordering with the Malaysia and Singapore, and it covers a small part of the eastern coast of Sumatra as well as archipelagoes spreading in the Malacca straight, on the Natuna and the South China seas. This province, as most other Sumatra provinces are, is endowed with large mines and energy resources. Tin, bauxite, oil and gas are the products of Riau province, which area totals to 94.5 thousand square kilometers. The Caltex Pacific Indonesia, operating a number of oil fields in this area, contributes the largest share of the whole Indonesian crude oil production. One of the oil fields is the Duri field, covering a productive area of about 100 square kilometers. Since the crude has high average gravity and viscosity (22 API, 120 cp), its exploitation requires enhanced oil recovery using the steam flooding, which had become the subject of this nuclear application study. Refineries are located on the coast (Dumai and Sungai Pakning), and are connected with pipeline system from the oil fields. The gas reserve in the Natuna islands represents the largest in Indonesia, although the exploitation needs special treatment due to substantial content of carbon dioxide. Tin is mined in the Karimun, Kundur and Singkep islands, as bauxite in the Bintan island.
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3 • The Nuclear Appl.icatio.n__S.tu.dy. As mentioned previously the Prefeasibility Study of HTR application in the Duri oil field for the steam flood project had identified various technological and economic advantages. The study started with a plan of oil field development, followed by technical solutions on steam distributions deploying centralized nuclear plants in place of evenly allocated crude boilers. Then came the assumption of economic parameters to assess the viability of the alternative. A very prospective results were obtained when oil price was taken to be consistently increasing (1 %/a and more, in real value, with 4 %/a general inflation rate) starting with 18 US$/bbl in 1987. The same positive results were obtained if either the nuclear plants were operated by the oil operator or by an independent company. The acquisition of an advanced technology could' be propelled by the saving of the burned crude. Later sensitivity calculations were performed to evaluate the influence of 1988 oil prices, which fell short behind the assumed figures. The result showed that, if the oil price was 16 US$/bbl and was increasing less than 2 %/a, the mere economic oonsiderat. ion would not be able to support the nuclear application. But in later years when oil price will rise due to resource depletion or escalation of production cost, then it will be too late, since already substantial amount of crude has been burned up. 4
• HTRs in the MARKAL Study
The Agency for the Assessment and Application of Technology (Indonesia) with the assistance of KFA Juelich had coordinated an energy strategy study utilizing the MARKAL model. In the model Indonesia is divided in four regions: Jawa, Sumatra, Kalimantan, and Other Islands, among which energy transports like pipelines, ships and high voltage lines are accounted. Two scenarios were formulated, i.e. the high and the low scenarios. For each of the scenarios two optimization cases were investigated: minimum cost case and reduced oil use case. The maximum saving of financial resources is obtained by applying the minimum cost objective function without any restriction on the domestic oil use. By this minimum cost case the result is an extended oil use on domestic market and hence in an accelerated depletion of the proven reserves. Therefore a cost minimization with the additional restriction, in which 15 % of the domestic oil use has to be saved, was made. Cost projections in real terms, rather than prices, are used in the study in order to express monetary expenditures. A
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Prices are taken only in connection with exports or imports. For oil export the price was assumed to be 14.5 US$/bbl in 1986, increasing 4.9 %/a in the high scenario and 3.0 %/a in the low scenario. But for the enhanced oil recovery the actual cost of burned cude are accounted excluding profits, taxes and subsidies. Three options of fuel supply for the Duri Enhanced Oil Recovery were considered: the crude oil, the gas transported through pipelines from Natuna, and the HTRs. The HTR's data were taken from the prefeasibility study on the application of HTR module for the Duri oil field (2). The conditions under which a 4x200 MWt.HTRs is competitive are shown in the following tables. Table 1: HTR Option for Duri Oil Field (PJ/a produced steam) i !
criterion/ Source/REPELITA VII VIII IX ! V VI case 89-94 94-99 99-04 04-09 09-14!
! High Seen. ! Minimum ! Cost
Crude Gas HTR
26.8 101.9 115.2 115.2 115.2! ~* —
Crude Gas HTR
26.8 _
t
! High Seen. ! Reduced Oil i Use
81.4 20.5
81.4 94.7 20.5
_ 94.7 20.5
94.7! 20.5! 1 I
Crude/Gas boiler cost : 262 $/Kwth REPELITA = Five Year Development Plan This shows that with only minimum cost as criterion, the crude-burning option is the only solution to the end of the time horizon. The fact that it consumes about one fourth of the crude produced does not matter as much. The future values put in the burnt oil cost are not high enough to balance with the high capital cost of HTRs. Further if the policy of reduced domestic oil use is applied the HTR contribution becomes supplemental, starting in the REPELITA VI (Sixth Five Year Development Plan). In the REPELITA VII the gas contribution starts to replace completely the crude-burning practice. It is obvious that 15 % domestic oil use saving, which is a conservation measure, can mean considerable additional costs. This was demonstrated in the sensitivity calculation shown in the Table 2. The introduction of HTRs in the Sixth REPELITA gives the same effect of the boiler cost raised to almost three times. E-10
6S6
Table 2: HTR Option in the Sensitivity of Boiler Costs, with Criterion of Minimum Cost (PJ/a produced steam) ! Boiler Cost Source/REPELITA VI VII VIII IX ! V $/KWth 94-99 99-04 04 09 09-14! 89-94 ' I
262
!
467
f
Crude Gas HTR
26.8 101.9 115.2 115.2 115 . 2 ,' _ _
Crude Gas
26.8 101 .9
HTR
•:•
—
™"
94.7 —
| t
94.7 — 20.5
94.7! 20.5! I I
t I
!
721
Crude Gas
26.8
HTR
81.4
94.7
94.7
94.7J
20.5
20.5
20.5
20.5! | 1
REPELITA = Five Year Development Plan 5. Concepts of Regional Development for Sumatra Sumatra is an island which has abundance natural resource covering various minerals as well as oil and gases. Therefore, there are many industrial zones having been developed in the future to optimize the wealth of the island. In this island there are three regions for industrial growing centres, i.e. two of them are containing various base chemical and metal industrial zones while the other one is planned to be the region for industrial growing centre based on key industry, see Figure 1. The vast industrial development in the -island is in harmony with the transformation toward industrialization now taking place in Indonesia aimed by the Government. Efforts have been carried out to deepen the industrial interdependency within and among the zones in the industrial regions so that various industrial trees can emerge and grow well. The industrial region will also be developed based on key industries. The key industry is the industry having one or more of the following characteristics, i.e. processes the strategic raw material for other industries, and employs sophisticated technology. These key industries are mostly capital and energy intensive and their production scale are huge (6 ) . The Duri oil field, being in the Riau Province, is just about in the junction of the above mentioned three regions for industrial growing centres. E-10
6. Nuclear Fuel Research The capability of fabricating the fuel elements for power reactor is important for Indonesia because it rises the following opportunities : a. securing the continuation of the fuel element supply at possibly more stable and reasonable price, b. growing other industrial activities since various products of other industries, e.g. chemicals and base metals as well as alloys, may be consumed giving rise to higher local content leading to the development of industrialisation. Considering above mentioned points, Indonesia has done various research activities in the fuel cycle although there has not any firm decision yet on the construction of nuclear power plants. The emphasis has been on Open cycle without enrichment since reprocessing1 is considered not necessary for at least couple decades after the operation of the future Indonesian nuclear power plants due to economics as well as technical reasons (7). The enrichment has also not been considered due to almost the same reasons. Indonesia will rely on international services which is secure enough for couple decades in the future (8). Meanwhile the demand for enrichment is in such as away that from economic point of view it will be better to have it done by someone alse's service in the international market. .In accord with this, Indonesia is now beginning to produce fuel element as well as control elements for her research reactor of 30 MW in Serpong (RSG-G.A. Siwabessy). At this nuclear research complex, experimental fuel element for power plant is soon be manufactured followed by various tests, both cold and hot using out of and inpile loops, rig, capsule as well as ramp test facility. Although research on HTR fuel element has not been carried out, it is expected that the capahility of producing this fuel element can be realized domestically if the production technology is transferred on time during the construction of the HTRs. This optimistic believe is supported by the facts that some of similar processes have been familiar to Indonesian scientists and engineers, there has also been some laboratory investigation on the process of nuclear grade graphite production. Furthermore, research activities on the HTR fuel element covering processes of fuel as well as its fabrication, to be carried out if the introduction of HTRs becomes more pronounced, can be done using the Serpong nuclear complex and other BATAN facilities in Yogyakarta, Bandung, as well as Jakarta. some
If the HTR fuel element is to be produced domestically, chemicals being the products of Indonesian industries, E-10
wilL be consumed giving rise to the more growing of industrial trees planted in surrounding regions for industrial growing centres. This kind of interdependency is actually one of important goals aimed by the Indonesian Government in the transformation toward industrialization (9,10,11). Those, chemicals are graphite, methyl trichlorosilane, C2H3/C3H6, and C2H2. For 30 years operation of 4 HTR 200 MWt modules, 4470 tons graphite, 4470 tons mehtyl trichlorosilane, 232 tons C3H6, and 163 tons C2H2 will be consumed. U-enriched (7.9%) consumption for first loading to operate the 4 HTR 200 MWt is 10.08 ton /year can be made available by having SWJJ service from international market using domestic natural U, or by purchasing enriched U from international market. The make up is about 3.6 ton enriched U (7.9%) per year resulting total requirement of 114 ton enriched U (7.9%) for 30 'year operation which is equal to about 1800 ton of natura'l U. Further more the nuclear grade ThO2, may also be domestically produced if it is feasible from economic point of view to reduce the U demand. Other chemicals, e.g. NH3, NH4OH, H2, PVC and additives needed for U02 kernel formation can also be supplied by domestic market. Moreover, other chemicals, to be used for water treatment, are also to be consumed. This, once again, will, lead to expected interdependency in the surrounding regions for industrial growing centres (6). For 30 years operation of 4 HTR 200 MWt modules, the consumption of these chemicals are 226100 tons HC1 , 118247 tons NaOH, 27086 tons FeCl, 25920 tons MgO, 1037 tons C12 and 752 tons polyelectroiyte. 7• Future Plans At present a more detailed study on coal and gas alternatives is being performed. It was suggested that coal from the Cerenti area (2 billion tons reserve, about 400 kM from Duri) can be opted, along side with the more detailed study of the Natuna gas pipeline system. The prefeasibility report (1) indicated an availability of appropriate site for the HTRs on the east bank of the Duri field. This assumption shall be the first to be verified in the next activity. There are efforts to encourage private companies to build and operate the HTR for producing the steam to be sold to Pertamina with the maxiumum price the same as the conventional one. Although the present study shows discouraging results, there.is still possibility that the development of the HTR as well as the future oil price change will lead to a favourable condition.
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Another alternative is to seek a sponsorship of a demonstration HTR unit to be constructed in the Duri oil field, subsidized appropriately by the Government in the beginning, and to sell the steam and electricity to the oil operator company. This second alternative may still be attractive if Indonesia is given the opportunity to gain and absorp the technology so that stronger industrial structure as well as more interdependency of the domestic industries, especially within and among regions for industrial growing centres in Sumatra island can be resulted. The fuel research leading to production of the fuel element of HTR will be intensified if the introduction of HTR is confirmed.
8• Concluding Remarks The nuclear solutic/n in the Markal study concludes that the cost of HTRs introduction is considerably high, but it equals to the cost of the reduced oil use policy itself. It confirms the previous thought that when oil price rises the solution appears. Then it depends on us, on how the Government will fare her own natural resources: shall a resources depletion cost be apllied and how high. Or shall the conservation consideration be weighted into the determination of lower discount rate? Sumatra is endowed with natural energy sources, but it has also to provide ample energy to the nation as a whole. The HTRs introduction for Duri oil field can favour the development of the region for industrial growing centers in Sumatra island, including the prolongation of source lifetime as well as the increase of electricity supply. Meanwhile the technological aspects of HTRs shall be thoroughly studied, in the harmony with other national and regional projects.
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1. "Tertiary Oil Recovery Using Steam and Electricity from HTR-Module Steam Generating Plant for the Duri Oil Field, Sumatra". Joint Prefeasibility Study (KWU, Interatom, BATAN, BPPT, MIGAS, PERTAMINA, LEMIGAS). Technical Part & Economical Part. June 1987. 2. "Economic Feasibility Investigation of First HTR-4 Module Cogeneration Plant for Tertiary Oil Recovery from Duri Oil Field, Sumatra". Kraftwerk Union AG. August 1987. 3. M. D.iokole 1 ono, [.Subki - "Study on Application of HTRs in Indonesia". IAEA Technical Committee Meeting on Criteria for Introduction of Advanced Nuclear Power Technologies for Specific Applications in Developing Countries. Vienna, 27-30 June 1988. 4. M. Djokolelono, R. Soedibjo, S. Padmosoebroto - "Economic Evaluation of HTRs as Applied to an Oil Industry". IAEA Technical Committee Meeting on Design Requirements, Operation and Maintenance for Gas-Cooled Reactors. San Diego, USA, September 21-23, 1988. 5. "Energy Strategies, Energy R+D Strategies, Technology Assessment for Indonesia". Final Report. Agency for the Assessment and Application of Technology (Indonesia), Nuclear Research Centre Juelich Ltd. (FRG). May 1988. 6. Anonym, "Wilayah Pusat Pertumbuhan Industri", bungan Masyarakat Departemen Perindustrian, 1985, Lampiran 4,5.
Biro HuIU/11/85,
7. S. Soentono et al, "BATAN's Activities in Fuel Development". Joint German - Indonesia Seminar on R&D Activities Using the MPR-30, Jakarta, Indonesia. August 19-21, 1985. 8. "Nuclear Power : Status and Trends", IAEA, Vienna, 1987. 9. S. Soepadi, S. Soentono, M. Djokolelono, "Contribution of BATAN's Multi Purpose Reactor and Its Supporting Laboratoria to the Nuclear Programme in Indonesia". International Symposium on the Significance and Impact of Nuclear Research in Developing Countries, Athens. 1986. IAEA. 10. S. Soentono, et al, "Possible Local Participation and Industrial Interdependency in the Construction and Operation of HTR in Indonesia "BATAN-KWU-INTERATOM Joint Seminar on HTR Module Application, Jakarta, Indonesia, June 29-30,, 1987. 11. "Rencana Pembangunan Lima Tahun Kelima", Departemen Penerangan Republik Indonesia, 1989. 10
E
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FIGURE %
RfGjON FOR• INDUSTRIAL QROWING CENTRES "•• . (REPELJTA IV ONWARQ). SOURCE INDONESIAN MINISTRY OF INDUSTRY
PA5E CHEMICAL INDUSTRIAL ZONES
o
BASE METAL JNDUSTR1AL ZONES PplONS FOR INDUSTRIAL G.ROWING CENTRES POSSIBLE REOIONS FOR INDUSTRIAL GROWING CENTRE 3ASEO ON KEY INDUSTRY'
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XA0101514 REACTOR TESTS AND POST-REACTOR EXAMINATION OP HTGR FUEL ELEMENTS AND COATED PARTICLES Yu.G.Degaltsev, A.A.Khrulev, I.S.Mosevitskii, N.N.PonomarevStepnoi, N.I.Tikhonov, V.V.Yakovlev I.V.Kurchatov Institute of Atomic Energy, Moscow, USSR
Report presented at the Technical Committee IAEA on HTGR, June 21-23, 1989, Dimitrovgrad, USSR ABSTRACT
Tests and examinations of HTGR fuel elements performed recently at I.V.Kurchatov AEI have been directed to
solve the
following problems: - determination of correspondence of the operability characteristics of fuel elements and coated particles from different lots to requirements to be imposed on VGM and VG-4OO designs being developed; - comparison of the characteristics of fuel elements manufactured using two different technologies of matrix graphite (KPD and GSP) ' as well as coated particles of various designs and material composition; - investigation of fuel element operability under severe accident conditions; - investigation of processes occuring in fuel element and coated particle, resulting in formation of defects and release of GPP; construction of physical models of fuel element behaviour in the reactor on this basis.
'KPD -carbonization under pressure GSP -graphite bounded with pyrocarbon E-ll
2. Tests and investigations were performed for fuel elements manufactured using the KPD and GSP technologies. For solving the above problems the experimental basis and methods described mainly in /1,2/ were used. The present paper is a short review of works covering these problems, with main emphasis being made on description and analysis of fuel element and coated particle tests under non-nominal accident and severe accident conditions. Such tests are of specific interest. While tests for confirming operability under nominal conditions are necessary for justification of the design characteristics and must be supported by sufficient statistics, test for severe conditions have at least two objectives: 1) estimation of the limits of fuel element operability depending on the basing affecting factors (temperature, burnup,etc.); 2) getting the data for studying processes resulting in loss of operability; on the basis of these data physical models of fuel element operability and accident sequence in the reactor unit may be constructed and verified. The experiments presented herein are the first stage of such investigations on determination of influence of the following effects on fuel elements: elevated temperature, increased burnup, neutron pulse and high oxidation. I. Reactor tests and estimation of fuel element and coated particle operability under routine conditions. Table 1 and 2 present the main parameters characterizing conditions of fuel element operation in VGM and VG-400 reactors and conditions under which the reactor tests were carried out. It is seen that the test conditions are essentially close to the routine conditions and in part of some parameters in a number of tests
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3. they were more severe; these tests are described in detail in next sections. For providing the required power levels in fuel elements and coated particles, enrichment (and loading) of uranium in specimen tested was much higher than in normal ones. In the process of 2?eactor tests samples of gas from the channels were taken constantly to control GPP release during the tests and determine the moment of coated particle depressurization. After the normal burnup (6-10% Pima for VG-400and VGM reactors) was reached in tests of KVG ' 1,2,4, and 5 channels of the PG-100 loop and Kashtan 2,3,4 channels no significant rise in GFP release was observed as compared with the initial level which had been determined by weak irradiation of each fuel element installed into the channels. In Udar channels groups of fuel elements (10 pcs) were tested in thermal cycles simulating transition regimes in reactor startup and shutdown. Measurements of GPP releases in the course of the tests and post-irradiation material-testing examinations of fuel elements and coated particles revealed that thermal fatigue processes in coated particle coatings and in matrix graphite do not cause damages affecting the fuel element operability. The test results obtained show that fuel elements tested which were manufactured by different technologies and subject to preliminary rejection using the weak irradiation method meet the requirements imposed on the fuel elements for VGM and VG-400 reactors. 2. Tests of fuel elements up to high burnups and their examination Tests of fuel elements up to burnups exceeding the design values were carried out in the ampoule channels Kashtan 2,3, and 4. The conditions of fuel element tests and fission product releases
E-11
4. measured separatly for each fuel element are shown in Table 3. The values of relative leakage listed in the table corresponding to the range of its variations during the tests, some constancy (increase in time dependence) was observed for fuel elements with high burnups ( £* 15%) in the Kashtan-3 channel. Measurements of fission product releases showed that even at burnups exceeding by 2 to 4 times the design values fission product releases, though increasing comparing with the initial level, do not indicate destruction of a significant number of coated particles (CP). Even at a bumup of 41%» for example, the leakage level is equivalent to destruction of 1-2 coated particles. It should be pointed out that the Udar and Kashtan tests were accompanied by a significant number of thermal cycle
so that their total number
exceeds that of the normal thermal cycles for VGM reactor. The post-reactor examinations included determination of distribution of caesium activity over the fuel element. For most fuel elements the test results revealed no accumulation of caesium in the fuel element graphite cladding which is indicative of the absence of significant damages of the SiC layer in coated particles. Only for some fuel elements, manufactured at the early development stages and tested in Kashtan-2 channel and having GFP leakage exceeding > 5.10
, accumulation ofcaesium in the fuel element graphite clad-
ding was observed which evidences destruction of microfuel protective coatings. Therefore, as far as the burnup factor is concerned, the fuel elements of experimental lots can be considered to possess a significant margin of operability.
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5.
3. Test and investigations of fuel elements and coated particles at high thermal effects. 3.1. During the long-term reactor tests of fuel elements in KVG-2 channel of PG loop, when a burnup of up to 4% Pima was reached, an accidental situation occured with depressurization of the helium loop, helium pressure drop and failure forced circulation with the heat density remaining at the same level. Change in the main parameters of this experiment is shown in Pig.1. The temperature of the fuel elements in the center of the fuel assembly increased up to 175O°C for 20 min. after reduction caused by helium escape from the loop. Then by the emergency protection signal the reactor was shutdown, the loop was filled with helium and the gas flow was restored. After the test conditions were restored GPP release did not increase and remained at the same level for 7500 hrs more when a burnup of 17% Pima was reached. 3*2. Prior to performance of the tests in Kashtan-4 channel one fuel element (KPD technology) was heated up to 25OO°C in the furnace (fuel element No.1). Subsequent tests under irradiation up to a burnup of 17% at 115O-75O°C showed that inspite of a high thermal effect and, as a result, an increased initial release of GPP («1.10"*5) the fuel element remained leak-tight. 3.3. The neutron pulse was applied to the fuel element tested subsequently in Kashtan-4 (fuel element No.2) and coated particles subsequently annealed in the Osa facility at 1000-1050<>C, with GPP release measured. The pulse for the 2
fuel element was 4*10 ^ 13 2 '
n/cm , and for coated particles it was (2-10).10 (
n/cm
&V =(1-2).10*""b, reactor GMra). The number of pulses for diffe-
rent group of coated particles was different. The energies in the pulses were such that in the adiabatic approximation in accordance E-ll
6. with the calculation the kernels of coated particles were melted. Measurement of GPP releases from the fuel element tested in the Kashtan channel up to a burnup of 20% at 115O-7OO°C showed that their value is at the level of release from fuel elements which had been not subject to the pulse effect. Test for leak-tightness of coated particles in the Osa facility revealed that coated particles subjected to the effect of single pulses remained leak-tight while with the number of pulses 5-10 GFP release began to increase which is indicative of appearance of defects. It should be pointed out that the pulse effect was applied to unirradiated fuel; a similar effect to irradiated fuel may have a more significant result. 4« Tests of fuel elements in high exidation These tests were carried out in KVG-3 channel of PG-100 loop. The coolant had increased content of COp and the nottest fuel elements were subjected to significant corrosion»of great interest was comparison of the experimental data on losses of the fuel element mass with the corrosion estimates as well as determination of the effect of increased corrosion on the fuel element state and increase of the activity in the helium loop. 4.1• Calculated estimates of fuel element corrosion. The initial data for the corrosion analysis were temperature regimes of fuel element performance in the channel, obtained by calculations of measured temperature and flow of helium at the channel outlet as well as constantly measured impurity concentrations in the coolant (Table 4 ) . The C0 2 concentration in the coolant was measured after termination of the experiment in the cold loop and was found to be equal to ^lOOCKrpm. E-ll
7. The U n A
concentrations in the course of tests can be estima-
ted by calculations as equilibrium by the CO +H 2 0sC0 2 +H 2 reaction; because of a weak dependence of the equilibrium constant on the temperature the loop
nonisothermality will weakly affect these es-
timates of U~ n . The U n n
estimates obtained amount to hundreds
of vpm in most regimes, i.e. they are close to the value obtained in measurements after completion of the experiment. The corrosion rate of the graphite fuel element resulting from its interaction with oxidizing components of C0 2 and H 2 0 by the reactions C + C0 2 = 2C0 and C + HgO « CO + 3HU is characterized by specific (per unit surface area) mol flows j c o and $„
0
absorbed
by the fuel element surface. For calculation of these flows an expression was used obtained for corrosion of a porous body in the flow of gas with a reacting gas component with allowance for massexchange both in the body itself and in the boundary layer, the typical dependences of the graphite corrosion kinetics in C 0 2 and Hr>0 were used with
allowance for the effect of inhibiting compo-
nents . Corrosion losses of the mass of each fuel element were calculated as:
?
In addition to the equilibrium estimate of UL n
another estimate
can be obtained from assumption that C 0 2 entered the loop from the purification system as a result of partial escape through the zeolite filter. In this case the balance equation can be written for C 0 2 and CO for the helium loop with allowance for reacting graphite and the purification system where CO is oxidized to C 0 2 in the copper oxide block and should be absorbed in the zeolite filter.
E-11
8.
The graphite reactivities relative to HgO and COp were calculated taking into account temperature distribution over the fuel elements. Solving the set of equations we found V c o » 1400vpm. For calculated estimates of fuel element mass losses it was assumed that Ynn =1000 vpm, which corresponds to measurements at the end of the experiment and lies between the estimates of in terms of equilibrium and the model with esacape through the filter. The results are listed in Table 5 along with the measured mass losses. Taking into account that the error in the calculation estimates of the fuel element temperatures and concentrations are high during the experiment ihe correspondence of the calculated and experimental data can be considered reasonable, except for fuel elements No. 13, 14 where the measured mass losses included significant spalling of graphite. 4.2. Reactor material testing studies. Corrosion occured in a wide range from appearance of roughness on the fuel elements subjected to low temperatures to change in the fuel element shape with destruction of not only the fuel element graphite cladding but also partial destruction of the fuel element kernel (for fuel elements at the maximum temperatures, Fig»2)» In the last case coated particles from the destructed part of the fuel element entered the helium flow and then were partially destroyed in the loop, which resulted in increase in its activity. Essentially non-uniform destruction of the fuel element over its surface should be pointed out, that was likely to be caused not only by corrosion. An essential factor which resulted in appearance of craters on the opposite sides of the fuel element and in a large loss of the mass (fuel elements No. 13,14), seemed to be mechanical interactions with the fuel elements over and under the destroyed fuel element.
ifOO
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9. It should be pointed out that open porosity of the GSP-type fuel elements did not practically change with a mass loss as low as 2%, while at higher mass losses it increased by 2-5% equally in the outer and inner fuel element cladding layers. For the KPD-type fuel elements a loss in the weight exceeding 5% at a surface temperature higher than 800°C increase in open porosity was mainly observed in the outer layer and reached 7%; in the inner layer the open porosity did not practically change i.eo corrosion was of the surface character. 4,3. Changes in the loop activity during the tests. Pig.4 shows change in the activity at the final stage of tests on the intermediate cooler at the outlet of hot gas, on the charcoal filter of the purification system and in the hall of the loop equipment. It should be noted that reduction in the maximum gas temperature by 14O°C by increasing the flow rate (after 10.11) did not result in any decrease of the activity, which supports the supposition on predominantly mechanical destruction of coated particles. Conclusion JPurther reactor tests and examinations of coated particles and fuel elements should be
carried out taking into account the fol-
lowing: 1. In view of some adopted changes in the technology of fuel element and coated particle manufacturing their operability under nominal conditions has to be checked and confirmed. 2. Tests under severe conditions have to be directed to systematic studies of fuel element and coated particle performance under accidental conditions in the reactor units and to obtaining the data permitting physical and mathematical models of their behavior to be constructed.
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10,
3. A detailed analysis of the methods for measurements of GPP releases, employed in different reactor units and comparison of the results obtained have to be carried out.
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REFERENCES 1. Yu.G.Degal'tsev, I.S.Mosevitakiy, N.N.Ponomarev-Stepnoy, N.I.Tikhonov, A.A.Khrulev: "Reactor tests of HTGR fuel elements. Paper presented at IAEA International Conf. on Gas-Cooled Reactors, Minsk, 1979. 2. V.A.Gurin, Yu.G.Degal'tsev, N.N.Ponomarev-Stepnoy, N.N.Tikhonov, Yu.M.Utkin, A.A.Khrulev, A.S.Chernikov, Ya.I.Strombach: "Reactor tests of spherical fuel elements and coated particles of HTGR and their post-irradiation examination". Paper presented at International Meeting on Fuel Elements of GasCooled Reactors". Moscow, 1983.
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'3
TABLE I Reactor Tests of Fuel Elements Parameter
Reactor VG-4OO VGM
Loop Channels (PG-1OO) KVG-1 KVG-2 KVG-3 KVG-4
KVG-5
Ampoule channels Kashtan-3 Kashtan-4 Udar-2
Udar-3
Pluence (E 0.18 jg_jg 30 8+23 6+14 7+17 2+6 i) 4,2+5,7 2,5+4,1*) 0,1 1+2,2 0,1 MeV) 1o2On/cm2 8+10 5,5+13 6,6+18 1,5+5,4 4,2+10 3,2+8,4 22+41 8-8 I I 14+20 Burnup,% Pima Max.temp erature 500-1240 700+1400 440+930 600+1080 640+1350 500+900 800+1270 600+1200 700+1340 400+800 700+1200 of fuel element on \>
Helium tempera- 350+950 ture (inlet-outlet),*^ Fuel power, KW 0,1+4 Total number thermal cycles ^U
300+950 300+600 400+800
400+920
400+850 400+900
0,5+2,1 0,5+2,4 1,3+2,6 0,9+5,2 0,5+1,4 1,0+4,8 0,4+1,3 0,4+1 50
120
100
10
90
30
120
70
1200
730
Test are being continued Range of thermal oycles TABLE 2 Reactor Test of Coated Particles Parameter Pluonce (E 0.18 MeV) i020n/cm2
Ampoule channels without removal of GPP Karat-2 7.5 - 10
Burnup (% Plma) 2.4 - 14 Temperature,°C 1300 -1700
Karat-3 20
Karat-4 10-20
7-17 7-H.5 800-1600 1130-1380
Karat~5 7.5-17 0.14 -24 880 -1600
Ampoule channel with GPP removed Mikrat-2 2.7-7.6* 5 -11 860-1100
'Test are being continued
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TABLE 3 Reactor Tests of Fuel-Element in Kashtan-3,4 Channels t
Chann e l
K-3
K-4
i
£ U-235 of fuel element
Temperature of kernel*) °C
Burnup,
Relative leakage of GFP
I
GSP
10004-680
21,7
I . I 0 % 3 . IO" 5
2
GSP
11004-660
28,4
3.I0" ? 4-I. IO" 5
3
KPD
10004-600
39,4
I.I0~64-8. I0~ 5
4
GSP
12004-600
33,0
7.I0"74-7. IO" 5
5
KPD
11004-570
41,0
5.I0~74-4. I 0 " 4
6
KPD
11504-600
18,9
2.I0~64-9. IO" 5
I
KPD
11504-750
17
5.I0"64-6. IO" 5
2
2
GSP
11504-700
20
4.I0"64-5. IO" 5
3
3
KPD
12004-800
14
3.I0~64-3. I0~ 5
4
KPD
13404-750
15,6
3.I0~64-2. io-
5
KPD
12004-700
15,2
4.I0" 6 44. IO" 5
6
KPD
13404-750
15 .
4.I0" 6 ^3. IO" 5
•
I
5
Range of temperature variations during the tests is shown
Note 1:Depressurization of coated particle 1-2 Note 2: Prior to test fuel element was subheated up to 25OO°C Note 3: Prior to test fuel element received an impulse of 4.1014 ^ cm
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105-
Note
TABLE 4 KVG-3 Channel Regimes
TBe °C
Date 11.10.84 17.10.84 22.10.84 29.10.84 10.11.84 15.11.84 16.11.84 19.11.84
G
o g/s
G
He g/s
800 810 870 880 890
58 53 49 42 35
750 750 750
58
C0 vpm
4.4 6.8 6.9
7.0 5.2 5.7
6.8
15.2 34.8
7.0 6.9 6,8
57 58
U
6.7
5.6 5.7 4.3
H
2 vpm 0.27 0.03 0.04 0.02 0.01 0.21 0.03 0.14
v vpm 0.4 2.2 2.0
2.5
U
C 0 2 (equal)
vpm 10
358 228 1480
6.0
1.6 1.5 10.1
342 372
TABLE 5 Losses of Fuel Element Masses in KVG-3 Channel Fuel ele-Type of Temperature ' Calculated loss ment fuel °C of mass,g number element
1.
GSP
2.
GSP'
3. 4. 5. 8.
GSP GSP GSP GSP GSP GSP
9.
KPD.
10.
KPD KPD GSP
6.
7.
11. 12.
13. 14. 15. 16. 17.
Measured loss of mass,g
GSP
GSP KPD KPD KPD
624 608 719 779 833 810 846 870 850 890 890 950 920 910 910 920
•
1,9-10"^ 8,5-10~4 I,MO"1 7,0-I0"1 3,7 2,0 5,2 9,3 5,8 . 14,5 . 14,3 41,6 25,4 21,7 21,7 25,4
3,3-IcrJ S^-ICT1 8,0-KT1 2,05 4,22 3,12 17,19 17,11 12,27 32,22 12,76 61,32 79,36 25,67 20,59 24,5
*)'For maximum regime
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FIGURE CAPTIONS Pig.1.
Change in the parameters of KVG-2 loop channel during accidental depressurization of the helium loop: P kg/cm
- pressure
C, g/s - gas flow rate N, MW - reactor power T shield, °C - shield temperature T fuel elem, °C - maximum temperature of the fuel elements. T gas, °C - temperature of gas at the channel outlet Is + - beginning of gas removal from the loop T g - operation of the emergency protection and make-up of the loop with gas Ifo - beginning of reactor power rise. Pig.2.
Appearance of fuel elements after removing them form KVG-3 channel* 1 - fuel element No. 5 2. fuel element No, 14.
Pig,3.
Section of a coated particle protruding from the matrix and interacting with gas flow.
Pig.4.
Radiation situation in equipment elements and in the PG-100 loop room at the last stage of experiment with KVG-3 loop. 1 - in room accomodating the technological equipment 2 - on carbon filter 3 - on intermediate cooler.
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T.;C /TOO
//OO
£00 f
300 -
/OO
faf
f. 3
E-ll
^ZJ
T~ 1 ' ft / !\ \ / • »\
CM
I 9Ao
i
1
XA0101515 STRUCTURAL
GRAPHITE GAS-COOLED
FOR
HIGH-TEMPERATURE
REACTORS.
Virgiliev Yu.S # , Avramenko P,Ya., Grebennik V Kalyagina I,P,, Lebedev I.G#, Filimonov V,A, , Shurshakova T # N,
Internal reflector blocks for high-temperature gas-cooled reactor ( HTGR ) are used in a wide temperature range from 620 to 1500 K and at neutron fluences up to 4 2: 10 nvt. To ensure the operability of graphite under the above mentioned conditions, previously non-practised,alrradiation-resistant grapite was required capable of working for the whole 30-year service life* And its strength characteristics are to exceed 2 to 2,5 times those of a standart nuclear graphite, PEM-K graphite of PP-280 grade used for the reactor core does not meet the strength requirements , Its irradiation stability cannot ensure a long-term serviceability of HTGR; at temperatures over 1070-1200 K after accumulating the fluence of 22 1 x 10 nvt, a secondary swelling of the material intensively develops ( Fig, 1 ) , The modification of nuclear graphite by means of pitch infiltrations followed by heat treatment ( FPII-2 graphite ) resulted in increasing the strength and thermal properties. However they were insufficient to meet the requirements put to the internal reflector blocks ( Table 1 ) , And thefrradiation stability has not actually changed, since the material base-calcihed coke , remained unchanged. Therefore it was necessary to produce a material of a new quality. The new class materials was created on the basis of non-calcined petroleum coke and a coal-tar pitch as a binder, . Here and below the neutron fluence is given with energy over 0 # 18 MeV, E-12
2-
J
320 -3WK
2~
0
jO70 -1200 *
-2
-2
' I I In i l
10
1
1 t M llll
I
t i i ••! i• I 1
1Q
il
i
i
i
i
•
i
i
i|
p di£SP I >
I'M*
Fig.1.Relationship between the neutron fluences and the change of a specimen length of nuclear graphit* irradiated at temperatures indicated on the graph. The specimens are cut parallel with(a) and normal to (b) the acticle extrusion. E-12
- 3Table 1 Properties of materials for internal reflector
Tech Requii ement
IT-280
rPII2-280
MITT
1,7
1,71
1,74-1,80
1,78
1,73
65
34/24
50/45
100
70/60
25
12,5/9,0
20/10
.54/53
Modulus of elasticity, GPa
—
6,5/5,0
10/9
10/11
G.T.E. (300-400K), m.K"1
4-6
3,2/4,9
3,8/5,1
6,6/6,0
Parameters
Density,g/cm compression Strength,MPa Bending 8trength,MPa
Indies
neat conduction coeff, (300K),W/mK
—
103/89
Resistivity, Ohm m
-
10/13
—
0,61
of
Materials
130/100
10/12
-
Sxperimen
95
Ji
30/25
11-13
5,1/5,3
65-83
10-14
15-17
1,34
0,76
Note: In the numerator - indiced for specimens oriented parallel to the blocr height and in the denominator for those oriented normal
E-12
- 4 For such materials: coarse-grain KJIT material and fine-grain MOT— material, high strength characteristics have been obtained by forming stronger structural bonds between the filler grains due to higher adhesion between the binder and the filler, Their strength properties attain values needed for the reflector blocks ( Table 1 ) , and theirradiation resistance is higher than that of nuclear graphite / 1 /• Theiradiation change of dimensions of these graphites is isotropical^ the initial growth of speciments, attaining 5,5% at 320 to 340 K / 2 /, abruptly decreases with theirradiation temperature increase, and in the range of from 700 to 900 K transforms to a slight shrinkage ( 0 # 2 to 0 # 4% ) # At a higher temperature grapite swelling occurs the rate of which increases with the&radiation temperature ( Fig* 2 ) # Thus, in comparison with nuclear graphite and its modifications^the deformation of graphite of this class actually lies in the swelling area in the whole temperature range investigated ( up to 1600-1800 K )0 The secondary swelling of these materials reveals itself to a lesser extent than in nuclear graphite, since a high initial strength of materials with a non-calcined coke-filler prevents their damage in the range of high fliences* However, graphites of this class ( MUT- type ) do not fully meet the requirements oniradiation stability put to the internal reflector blocks, either since at 1100 K and fluop
ence up to 2 x 10 nvt an intensive increase of the secondary swelling rate takes place. For the purpose of increasing thetrradiation resistance of the materials of this class, correction of a number of structural characteristics was required: perfection of the crystalline structure, porosity, anisotropy of properties. The quantitative meanings of these characteristics are to be optimal: the perfection should be sufficiently high; the isotropy ( both within the crystallite aggregates and in the macrovolume ) should be maximum; the pore distribution and volume should be close to those of M U T -material* It is known that the decrease of crystalline lattice perfection of materials produced by the conventional electrode technology, favouring the radiation shrinkage, shifts the secondary swelling to the area of less neutron fluences. Porosity, intrinsic of the M jjp -material, accomodates the crystalE-12
. 5 -
5
£00-1OO0X .7?O-87O><
-/
w
_J
I
I
I
t
I I I I
10to
I
I
I
I
I I I l I
10*
1
I t I 11
tO**
I
t
F
I
t
I i l I
Hear
P-
Pig.2.Relationship between the neutron fiuences and the change or a specimen length of MTtt1 grat'ite irradiated at temperatures indicated on the graph.
E-12
[
AS
- 6 lite c-expansion slightly during the radiation process in contrast to the rP-280 nuclear grapite,. When solving this problem, there were used such technological steps as introduction of an isotropic constituent into the material structure, the use of hydraulic pressing,the change of heat treatment conditions, etc,, which essentially effect the material serviceability. The development of graphite having the desired properties was based on the technology for the obtaining the MUF -graphite. The raw materials used in this case - petroleum coke and coaltar pitch - ensure the material high graphitizability. Therefore the developed experimental graphite passessed a sufficient perfection of the crystalline lattice ( Table 2 ), The pore size distribution was close to that in the M UT -graphite ( Fig. 3 )• The level and rate of specimens size changing can be controlled by changing the content of structural elements in graphite / 3 /, Thus with increasing the concentration of isotropic particles ( spherolites, globules ) in the materials of various classes, the size changing rate in the range of low temperatures ( 320 to 360 K ) decreases to valies typical of pyrographite ( 0% particles ) and then becomes negative - glassy carbon ( 100% isotropic parteiles ) , Nuclear graphites and their modifications occupy an intermediate position on the relationship under consideration / 4 /• With theirradiation temperature increase, the pyrographite swelling decreased, the glassy carbon shrinnage increased - the curve has shifted to the shrinkage area, ( Fig, 4 ) In view of the above-said, when preparing the experimental material, into the filler there has been introduced an additional ( atf ©Glared to MlIP) amount (by 30%) of isotropic particles exerting no essential influence upon the material graphitizability ( Table 2 ),.As a result the experimental graphite has undergone shrinkage ( Fig, 5 ) , Fig, 5 and fig, 6 compare for two irradiation temperatures ( 750 to 850 K and 1100 K ) the dependence of the specimen volumes and the dynamic elastic modulus of graphite: nuclear (EP-28Q) and its modification (IPn 2 ) ; M I I I \ / 2 / , / 5 / , / 6 /, the experimental one as well as ATR-2E ( FRG ) graphite is fluence of newtrons ATR-2E as a candidate for the HTGR interE-12
- 7Table
2
Characteristics of crystalline and porous struture of graphites
Indices Parameters C, run a, nm
TP-280
Materials MB?
Bxperimenta ]
0,6742
0,6760
0,6740
0,2461
0,2466
0,2460
Lc,nm
14
'13
II
La,nm
85
57
50
Graphitization factor (I112/ I110 ) Density
of
, g/cm-^ Volume Picknometric
Porosity , %
0,40
0,42
0,34
1,65-1,70
1,78
1,73
2,153
2,138
2,119
20-22
16,8
15,8
Volume of pores, cm-yg
0,131
0,126
0,114
Through p o r e s - t o t a l pore volume r a t i o ( v/V)
0,23
0,08
0,14
5-6
Total pore area, m / @ Maximum pore radius,micro-m
6,5
11,9
13,9
2,4
2,4
E-12
-8-
Fig.3-Integral curves of the equivalent radius size distribution of specific volume(v) of accessible pores (r) for graphites: nuclear(3); (2); experimental (1).
E-12
- 9-
If % lid
HO Fig.4.Relatioship between the change rate of a specimen volume of different graphites and the content (£) of spherical particles in their structure. Irradiation at a tempersture from 340 to 36OK (1) and from 720 to 780K (2) .
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- 10 -
$
/ 10
/
5
/
0
-5 ^
^
10
Fig.p» Neutron fluence dependence of the change of volume (a) and dynamic elastic modulus (b) for specimens of graphite : nuclear(1); MnT (2); experimental(3); kTR-ZE (4) /?/ • Irradiation at a temperature from 750 to 850K.
E-12
- II nal reflector blocks has been tested up to high fluences / 7 /, Ho moderation of shinkage of the experimental graphite speciments was observed at 850 K ( Fig, 5 ) and the accumulated fluence ( 1 x 10 nvt ), It suggest the appearance of the se22 condary swelling at fluences over ( 2 - 2,5 ) x 10 nvt. At the same time in nuclear graphite shrinkage is replaced by an 22 intensive secondary swelling at fluences over 1,3 x 10 nvt. The irradiation at a temperature about 1100 K ( Fig, 6 ) results in a secondary swelling of TP andfP/7-2 graphites 21
21
at the accumulated fluences 5 x 10 ' nvt and 3 x 10 nvt, respectively. The growth rate of MELT specimens increased several times. Apparently, its abrupt increase may be expected at about 22 2 x 10 nvt. For the experimental graphite at this irradiation temperature shrinkage takes place too, the secondary swelling 22 developing at fluences over 0,9 x 10 nvt. However no abrupt swelling rate increase was observed at the accumulated fluence of 1,3 x 1 0 2 2 nvt. The development of secondary swelling process is accompanied by the dynamic elastic modulus decrease^i,e, by the material loss of strength. And the shrinkage maximum and onset of the elastic modulus drop are inconsistent with each other as regards the neutron fluence. The elastic modulus drop proceeds more intensively, since the secondary swelling rate is higher. And speciments of the more strong FH1 2 graphite better remain the strength and integrity characteristics as compared to the less strong nuclear graphite. As concerns the low-porous high-strength MET -graphite, the modulus drop begins after the specimens has undergone swelling to about 1%, The modulus drop proceeds more slowly than in the nuclear graphite, since its strength is 2,5 to 3 times as higfc. This is a defining factor in retaining the integrity and strength of the blocks. The modulus drop in the experimental material proceeds at the same low rate as in the MTIF-graphite ( Fig, 5 ), The irradiation of the nuclear graphite speciments, its modifications (PHI -2 ) , as well as MHT at a still higher tempera ture ( 140Qc ) resulted in a further increase of the secondary swelling rate ( Fig, 1 , 2 ) E-12
- 12 At the same time a sligth ( about - 0,1% ) shinkage was observed in the experimental graphite at the resulting fluence ( 3 x 10 ' nvt ) , No rate slowing took place,.Hence the shrinkage development of the experimental graphite can be expected up toO, 22
at least, 0,7 to 0,9 x 10 nvt. It means that at the above mentioned temperature the experimental material ensures the serviceability of the blocks,. At the present there have been obtained articles of the experimental graphite of 700 mm, in diameter and 1000 mm, in length on the commercially available equipment. Conclusion An experimental-industrial technology of producing graphite having strength properties exceeding 2,0 to 2,5 times those of the conventional nuclear grapite has been developed. The irradiation-tests of the experimental graphite at SDo-to moo°K to the fluences of about 1.3.1022n/cm2 the usefulness of that rsrjshitG for IICGR.
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- 13 -
CM*
Fig.6. Neutron fluence dependence of tne change of volume (a) and dynamic elastic modulus (b) for specimens of graphite: nucclaar (3); IPII-2 (2); M T (3); experi. mental (4); ATR-2E (5) /7/. • . Irradiation at a temperature of 1100K.
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- 14 REFERENCES 1.I.P.Kalyagina, Yu.S.Virgiliev, V.R.Zolotukhin et.al. High-Temperature irradiation of Structural Graphite. Atomnaya Energiya,1974,v.36,3,p.212- 215. 2.I.P.Kalyagina,Yu.S.Virgiliev, Irradiation Dimensional Stability of Graphite with a Non-Calcinated Coke-Filler. Atomnay Energiya, 1983,v.55,3,P-157-159. 3.I.P.Kalyagina, Yu.S.Virgiliev.Effect of Tehhnological Factors on the Irradiation Dimensional stability of Structural Graphite.Atomnaya Energiya,1977,v.43,2,p.1o6111. 4.I.P.Kalaygina,Yu.S.Virgiliev. Investigation of Raw Material for Nuclear Graphite. Tsvetnye Metally, 1989,p.73-^5 5.P.A.Platonov, Ya.I.Strombach, V.I.Karpukhin et.al. Effect of Irradiation on High-Temperature Gas-Cooled Reactor Graph!te.In:Atomno-Vodorodnaya Energetilca i Tekhnolo&ty.a, Part 6,M. "Energoatomizdatn,'59a4,p.77-128. 6.Yu.S.Virgilier, I.P.Kalyagina,V.G.Makarchenko.Irradi a ation Changes of Graphite Properties in a Broad Range of Irradiation Temperature and Neutron Flueneas. IntKonstruktsionnjre Materially na Osnove Ygleroda,Part XV, M.,"Metallurgy",1980, p.60-70. 7. HflflG G . J DELLE 14. > NICEL H.5 ET BL DEVELOPMENT flND TESTING OF HUCLERR GRflPHITES FOR GERMftN PEBBLE - BED HIGH TEMPERflTURE REfiCTOR PROC. MftTER. IftEfl SPECIALISTS MEETING ON GRftPHITE COMPONENT STRUCTURflL DESIGN* JflERIj TOKVO* JFiPBHf SEPT. 3 - i l > 1986* P. 123-132.
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