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| !. REPORT ON THE FIRST MEETING OF THE WORKING GROUPS
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| {l OF THE U.S./U.S.S.R. JOINT COORDINATING COMMITTEE FOR CIVILIAN )
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| l NUCLEAR REACTOR SAFETY, MOSC0W, DECEMBER 5-9, 1988 j
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| I Compiled by Stephen N. Salomon International Programs Office of Governmental Affairs U.S. Nuclear Regulatory Commission Washington, D.C. 20555 May 1989 89072404BS 890707 PDR REVCP NRCUSUSR 8 PDR 1
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| 1
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| .I REPORT ON THE FIRST MEETING OF THE WORKING GROUPS OF THE U.S./U.S.S.R. JOINT COORDINATING COMMITTEE FOR CIVILIAN NUCLEAR REACTOR SAFETY, MOSC0W, DECEMBER 5-9, 1988 Compiled by Stephen N. Salomon International Programs Office of Governmental Affairs U.S. Nuclear Regulatory Commission Washington, D.C. 20555 May 1989
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| ABSTRACT The first working groups of the U.S./U.S.S.R. Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS) met in Moscow from December 5 through 9, 1988, in accordance with the Protocol of the First Meeting of the JCCCNRS signed by the Soviets and Americans on August 31, 1988 (hereafter called the protocol). Three working groups met at the Kurchatov Institute of Atomic Energy and discussed three of the ten areas cf civilian nuclear reactor safety that will be the focus of U.S./U.S.S.R. cooperative efforts as described in the protocol:
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| analysis of the safety of nuclear power plants in the U.S.S.R. and U.S., fire
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| . safety, and erosion / corrosion destruction of piping and components. Pressure vessel integrity, probabilistic risk assessment methods, and severe-accident code verification, all of which have features related to the protocol, were also addressed. The fire safety working group agreed on topics to be covered at the next working group meeting now scheduled to take place from June 5 through 9, 1989 in the United States. The erosion / corrosion working group adopted a draft plan for cooperation in the period of 1989 througn 1991, and prepared detailed plans for the next working group meeting. The U.S. Department of Energy inde-pendently conducted a seminar on its analysis of the Soviet VVER reactor. Some U.S. delegates visited the All-Union Researcn Institute for Nuclear Power Plant Operations in Moscow. The Soviets and Americans discussed additional subjects (not covered specifically in the protocol) including thermal-hydraulic research, cost sharing program on Chernobyl, plant aging, and safety concepts for the next generation of nuclear power plants, a description of the Scientific Technical Center of the State Committee for the Supervision of Nuclear Power Safety, pub-lic participation in the licensing process, the Soviet Academy of Sciences' Nuclear Safety Institute, forecasts of Soviet nuclear capacity and foreign sales, and the Armenian earthquake.
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| O
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| CONTENTS Page l ABSTRACT ................................................................ iii EXECUTIVE
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| ==SUMMARY==
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| ............... ....................................... ix
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| , 1 INTRODUCTION . ........ .. ........................................ 1-1 2 ~ PRELIMINARIES ...... ..................... ......................... 2-1 Thursday, December 1, 1988 ......................................... 2-1 Friday, December 2, 1988 ................. .......... ............. 2-1
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| . Saturday, December 3, 1988 .... .................................... 2-1 Sunday, December 4, 1988 ........................................... '2-1 3 PROTOCOL TOPIC 2: ANALYSIS OF THE SAFETY OF NUCLEAR POWER PLANTS IN THE U.S.S.R. AND THE U.S .............. ......................... 3-1 Protocol Topic 2.1: Exchange of Safety Analyses ............. ... . 3-1 Protocol Topic 2.1.1: DOE Analysis of Soviet VVERs ................ 3-2 Overview Session .. ... . .. . . ................. .......... 3-2 Working Group Session .. ... .......... .......................... 3-3 The DOE Report .. ............. ............ ................... . 3-3 Protocol Topic 2.2: Safety Research .......................... . . 3-4 Monday, December 5, 1988 ......................................... 3-4 The Next Generation of Nuclear Power Plants ........... ........ 3-5 Nuclear Power Safety in the Soviet Union . ...... .............. 3-5 Pressure Vessel Integrity and Plant Aging ..................... 3-6 Radiation Burden ... . ....................... ................ 3-7 Tuesday, December 6, 1988 . ... ............. ............ ....... 3-8 Soviet Studies on Severe Accidents With Core Melting .. ........ 3-8 American Studies on Severe Accidents ........................... 3-9 Soviet Studies on Severe Accidents Without Core Melting .. ..... 3-11
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| - Thermal-Hydraulic Research in the United States ..... . . ...... 3-11 Thermal-Hydraulic Research in the Soviet Union .. . ... ... .... 3-12 The Soviet Scientific Technical Center .. . ........... . ...... 3-13
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| . Probabilistic Risk Assessment Methods .. . ..... ... .. .... 3-14 Wednesday, December 7, 1988 ...... ... .. . .. . .. . .. ... 3-15 Public Participation in the Licensing Process . ... .. .... . 3-15 5 : $0 RRbSiONDISTRUCTIbN0hPIPINbANb f COMPONENTS . ... . . . .. . . . . ... 5-1 I
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| v
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| _ - _ _ , _ _ _ . - _ - - _ , - , - - - - - -,--,,--.ri-:2,-----m .
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| CONTENTS (Continued)
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| Page Wednesday, December 7, 1988 ............. ........................ 5-1 Thursday, December 8, 1988 ....................... ............... 5-2 Future Meetings .............. ................................... 5-4 6 OTHER TOPICS ....................................................... 6-1 Forecasts of Soviet Nuclear Capacity and Foreign Sales ........... 6-1 Cost-Sharing Program on Chernobyl ........................-........ 5-1 -
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| Soviet Academy of Sciences Nuclear Safety Institute .............. 6-2 Chairman Zech's Trip to the Soviet Union .... ..... .............. 6-2 Soviet Sale of Annealing Technology ........... .................. 6-2 .
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| Armenian Earthquake .............................................. 6-3 7 PREPARATION OF THE MEMORANDUM 0F THE FIRST MEETING OF THE WORKING GROUPS OF THE V.S./U.S.S.R. JOINT C0ORDINATING COMMITTEE FOR CIVILIAN NUCLEAR REACTOR SAFETY, DECEMBER 5-9, 1988 ..... .......... 7-1 APPENDICES 1 U.S. Participants in the First Meeting of Working Groups of the Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS),
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| December 5-9, 1988, Moscow 2 U.S.S.R. Participants in the First Meeting of Working Groups of the Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCNRS),
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| December 5-9, 1988, Moscow 3 Background Documents Sent to Soviets by NRC 4 Agenda of the Meetings in the Kurchatov Institute of Atomic Energy (KIAE),
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| December 5-9, 1988 5 " Nuclear Power Reactors of the New Generation," N. Ponomarov-Stepnoi and I. Slesarev 6 Three papers by V. G. Asmolov et al.:
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| (1) " Investigations on Justification of Nuclear Power Engineering in the U.S.S.R.," V. G. Asmolov et al.
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| (2) "The U.S.S.R. Approach to Safety Studies," V. G. Asmolov et al.
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| (3) " Development of Nuclear Power Plant Safety Research in the U.S.S.R.,"
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| V. G. Asmolov et al.
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| 7 " Theoretical Analysis and Numerir.al Simulation of Heat Transfer and Fuel Migration After Severe Accidents at Nuclear Power Plants, R. V. Arutyunyan et al.
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| vi
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| CONTENTS (Continued)
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| '8 Figure 1: Measurement of the heat output of the VVER-1000 reactor from the flow rate of the coolant in the primary loop by the radiation method.
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| Figure 2: Azimuthal flux distribution 9 Areas of Interest for Each Participating Organization in Erosion / Corrosion 10 Abstract: " Hydrodynamics and Heat and Mass Transfer When Adding
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| . Surfactant," G. A. Filippov, G. A. Saltanov, and A. N. Kukushkin 11 Three Fire Safety Documents Submitted by the Soviets to the Americans:
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| (1) "Providing for the Fire and Explosion Safety of Nuclear Power Plants in Connection With the Formulation and Accumulation of Hydrogen" (2) " Fire Danger of Electrical Cables (3) " Occupational Safety Standards" 12 Memorandum of the First Meeting of the Working Groups of the U.S./U.S.S.R.
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| Joint Coordinating Committee for Civilian huclear Reactor Safety, Deceraber 5-9, 1988 13 Memorandum on DOE Seminar in Moscow vii L ___.____._ - - . _ . _ . - . _ _ - E
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| l F.VECUTIVE
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| ==SUMMARY==
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| Working groups of the U.S./U.S.S.<'. Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS) met in Moscow from December 5 through 9, 1988.
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| The Protocol of the First Meeting of the U.S./U.S.S.R. JCCCNRS signed by the Soviets and Americans on August 31, 1988 (hereafter called the protocol) had called for a November 1988 meeting.
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| The United States and the Soviet Union agreed in the protocol to meet several times in both countries to discuss ten topics.
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| The American team was headed by Dr. Denwood Ross, Deputy Director, Office of Nuclear Regulatory Research of the Nuclear Regulatory Commission. Dr. Sol Rosen, Director, International Nuclear Program Division, Office of Nuclear Energy, Department of Energy (DOE), led the DOE seminar. DOE representatives partici-pated only in discussions of the VVER, the Soviet pressurized-water reactor (Topic 2.1.1 of the protocol). The Soviet team was headed by Dr. Nikolai N.
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| Ponomarev-Stepnoi, First Deputy Director of the Kurchatov Institute of Atomic Energy. A memorandum was signed that summarizes the meeting.
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| The working groups met at the Kurchatov Institute of Atomic Energy and at the All-Union Research Institute for Nuclear Power Plant Operations and discussed three of the ten areas of civilian nuclear reactor safety cooperation as described in the protocol. They are:
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| * Analysis of the Safety of Nuclear Power Plants in the U.S.S.R. and the U.S.
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| (protocol Topic 2)
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| * Fire Safety (protocol Topic 4T
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| * Erosion / Corrosion Destruction of Piping and Components (protocol Topic 10) ,
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| The Americans presented the Soviets with the 20-volume set of the South Texas Project Final Safety Analysis Report. Instead of presenting an analysis of the Zaporozhe plant (as called for in the protocol), the Soviets presented " Technical Justification Report on Construction and Operational Safety of Rovno Nuclear Power Plant Unit 3 With VVER-1000 Reactor," a 1200 page summary document. The Soviets, if requested, will provide detailed appendices to the Rovno report and will also detail the differences between the Rovno and Zapurozhe plants.
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| Delegates from DOE conducted a seminar on the results of DOE's analysis of Soviet-designed VVuRs (protocol Topic 2.1.1).
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| The U.S. team gave the Soviets a copy of NRC report NUREG-1319 which is a summary description of NRC research. The Soviets presented a paper or, their safety research program and gave a copy to the U.S. group. In advance of the meeting, the NRC had sent the Soviets material on a number of NRC research projects and would find it highly useful if the Soviet team would respond with documents of a similar nature.
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| Both teams developed a number of research proposals. Topics on pressure vessel integrity, probabilistic risk assessment methods, and severe-accident code ix
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| verification have features related to the protocol. Topics that are not men-tioned in the protocol, but that were proposed for study, are thermal-hydraulic I research, cost sharing program on Chernobyl, plant aging, and safety concepts for the next generation of nuclear power plants.
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| A program on pressure vessel integrity will be developed in accordance with protocol Topic 3, Radiation Embrittlement of the Housing and Support Structures and Annealing of the Housing.
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| The Soviets provided some methodology on the calculation of reactor pressure vessel rupture probability, and the Americans presented some results of calcu-lations performed. Both sides agreed that the exchange of the details of -
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| these methods and calculations is important.
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| The NRC could offer the Soviets its experimental data (8 year effort that is half .
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| completed) that evaluates: the effects of material chemistry (Ni, Cu, P); radia-tion flux effects in a test reactor; the effects of temperature at 500 F-600 F; ef fects of radiation, annealing and re-irradiation; and mechanisms of embrittlement.
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| In addition, the NRC has test results of Gundremmengen and Shippingport reactor vessel materials test (Charpy-V) of archive material and vessel material. How-ever, the Shippingport results have not been published yet. Also, there is a comparison of test reactor irradiation results with the results from operating power reactor irradiations on these materials.
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| A possible cooperative effort is to participate in the recently organized inter-national study group on radiation damage mechanisms which is sponsored by NRC.
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| The NRC requested from the Soviets the details of their reactor vessel annealing experience and research efforts. If the Soviets cut up their NV-1 reactor (as they plan to do), they could give NRC toughness values (Charpy-V) as a functirn of vessel thickness, and NRC could offer to perform fracture mechanics testing (compact tensile specimens).
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| NRC would like to get data showing effect of various flux (including low flux) levels as they affect the shift in nil ductility temperature (NDT) for mate-rials similar to American materials. NRC would also like to get data that show the effect of temperature (down to room temperature if possible) on the shift in NDT.
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| The Soviets could supply data on the embrittlement effects of various concentra- -
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| tions of phosphorus, and information on mechanisms of radiation damage.
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| The Soviets will consider the NRC proposals. .
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| There will be a meeting in late 1989 in the Soviet Union on the uses of prob-abilistic risk assessment (PRA) (protocol Topic 2.3). The two groups agreed that this meeting should cover both the results and also methodologies used in l each country. In particular, the Soviets noted a desire' to expand and improve l its methods for containment response (Level 2 assessment).
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| Both the NRC and the Soviets presented information on severe accident scenarios and on the interaction of molten core debris with concrete. The NRC suggested, and the Soviets agreed, that it would be useful for the U.S.S.R. to analyze one of the NRC experiments known as SURC-4. The NRC would provide initial con-ditions and geometry, and the Russians would predict the results with their code RASPLAV (MELT in English). Af terwards, the NRC would provide the measureti data for comparison with the Soviet predictions. This proposal is consistent with protocol Topic 6.2(b).
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| The NRC presented details on its thermal-hydraulic research program. The Soviets noted that they needed additional analytic capability for small breaks and transients, and thought that the NRC code RELAP-5 might be useful. The Soviets suggested that in return for the U.S. code, they could provide results from their integral f acility at Electrogorsk of 1:3000 scale and later the 1:500
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| . scale. To optimize the usefulness of this experiment, the NRC would like early involvement in the test matrix. The Soviets will consider this possibility.
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| The NRC stated that this exchange was not specifically provided for in the protocol, and that transmission of this code, as for any computer code, would have to conform to the export requireinents of the United States. Because of these factors, the Americans could not make a firm commitment.
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| The Soviets informed the NRC of the preparation of a cost-sharing program for exploration of the cause and consequences of the Chernobyl accident. Both groups agreed that such a program is important, although not specified in the protocol. The Soviets will provide details of this proposal.
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| The NRC presented research related to aging effects on nuclear power plant structures, systems, and components. The Soviets consider it useful to use a probabilistic approach for the assessment of the lifetime of the core struc-tural elements, plant equipment, ana buildings (fuel elements, graphite, piping, metal structures, etc.) given the condition of incomplete knowledge and high uncertainty. Such an approach is also being cor,sidered by the NRC. Although not specifically part of the protocol, it was agreed that research on plant aging was important to both nations and should be the subject of cooperation in the future, in accordance with direction from the JCCCNRS.
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| The Soviets presented a paper on nuclear power plants of the next generation.
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| Both teams agreed that it is imoortant to establish a working group to develop safety concepts for the new generation of nuclear power plants. At present, the protocol does not provide for such a working group, and this topic is recommended for consideration at the next meeting of the JCCCNRS.
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| The Soviets are conducting research on erosion / corrosion; the NRC, however, believes that f urther confirmatory safety research is not necessary. Both
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| _ groups recognize that although the Electric Power Research Institute (EPRI) is primarily conducting additional erosion / corrosion researct. for economic reasons, the topic also 5as safety implications.
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| The discussion of the fire safety working group covered exchange of information on the following:
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| * Capabilities of electrical cable to withstand fire, and design of cable penetrative, through bulkheads (protocol Topic 4.1.3) l l
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| * Methods for fighting large fires under high radiation conditions for extended periods of time (protocol Topic 4.1.4)
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| * Fire suppression systems and ventilation systems to protect the control room environment from external fire (protocol Topic 4.1.5)
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| Fire protection coating for structural steel (protocol Topic 4.1.6)
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| Use of probabilistic techniques in fire risk assessment at nuclear power plants (not in the protocol)
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| Hydrogen accumulation as it relates to fire safety at nuclear power -
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| plants (part of protocol Topic 4.1.1)
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| The Americans and the Soviets agreed that the following subtopics of protocol .
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| Topic 4, Fire Safety, would be discussed in the June 1989 working sessions:
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| Research related to, and methods for determining causes and effects of fires involving electrical equipment
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| * Research related to the modeling of temperatures within fires and within compartments Fire risk associated with computers utilized in nuclear power plants A. Computer equipment as a fire hazard B. The effect of fire and fire suppressants on computer equipment and the control of safety systems The categorization of components within power plants as regards the risks from fires and explosions. This will inciude a comprehensive discussion on U.S. and Soviet regulatory ef forts in fire protection including recent fire experiences. Subtopics will include:
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| A. Protection of structural steel B. Fire barrier penetration seals C. Fire suppression systems D. Protective enclosures for electric cable
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| * Modeling of hydrogen burn within containment The two teams would discuss a visit to a representative American nuclear power plant.
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| The erosion / corrosion working group adopted a draft plan for cooperation in the period of 1989 through 1991 and agreed on topics for discussion at the first workshop to be held in June 1989. This plan provides for holding annual seminars and exchanges of information and specialists. The working group will submit this plan to the JCCCNRS for approval. Realizing, the inportance of the two items on the plan that provide for the exchange of specialists and mutual visits,, the working aroup considers it essential to obtain additional agreement on these xii
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| items from governmental organizations, private firms, and the NRC. The working group noted that the Soviet Union has presented a list of the participants in this cooperative effort and that the United States will provide a similar list in the first quarter of 1989.
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| Dr. Ross and various U.S. working group members visited the All-Union Research Institute for Nuclear Power Plant Operations in Moscow. The role of this institute was outlined and the U.S. visitors discussed severe-accident modeling and several erosion-corrosion topics, including experimental facilities and results.
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| Conclusion Leaders of both teams believe that the exchange of information that took place and the seminers were worthwhile and very informative. They look forward to continued progress at the June 1989 meeting to be held in the United States.
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| Additionai subjects covered in this report, but not reflected in the signed memorandum of the meeting, include a description of the Scientific Technical Center, State Com.%ittee for the Supervision of Nuclear Power Safety; public participation in the licensing process; the Soviet Academy of Sciences' Nuclear Safety Institute with laboratories at Minsk; observations on the DOE analysis of the Soviet VVERs; forecasts of Soviet nuclear capacity and foreign sales, and statements regarding the Armenian earthquake.
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| l xiii !
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| l REPORT ON THE FIRST MEETING OF THE WORKING GROUPS OF THE U.5./U.S.S.R. JOINT C0ORDINATING COMMITTEE FOR CIVILIAN NUCLEAR REACTOR SAFETY, MOSCOW, DECEMBER 5-9, 1988 1 INTRODUCTION The first working groups of the U.S./U.S.S.R. Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS) met in Moscow from December 5 through 9, 1988, in accordance with the Protocol of the First Meeting of the JCCCNRS,
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| , signed by the Soviets and Americans on August 31, 1988 (hereafter called the protocol).
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| The United States and the Soviet Union agreed in the protocol to meet several times in both countries to discuss the following 10 topics:
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| (1) Safety Approaches and Regulatory Practices (2) Analysis of the Safety of Nuclear Power Plants in U.S.S.R. and U.S.
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| l (3) Radiation Embrittlernent of the Housing and Support Structures end Annealing cf the Housings (4) Fire Safety l
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| :(5) Modernization /Backfitting (6) Severe Accidents (7) Health Effects and Environmental Protection Considerations (8) Exchange of Operational Experience (9). Diagnostics, Analysis Equipment, and Systems for Supporting Operators (10) Erosion / Corrosion Destruction of Piping and Components
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| . The protocol provided a meeting schedule that gave approximate dates and places for the meetings and determined at which meeting each tor : would be studied. 4 The U.S. delegation (Appendix 1) was headed by Dr. Denwood Ross, Deputy Direc-tor, Of fice of Nuclear Regulatory Research of the Nuclear Regulatory Commission.
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| Dr. Sol Rosen, Director, International Nuclear Program Divisian, Office of Nu-clear Energy., Department of Energy (DOE), led the DOE seminar. DOE represent- !
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| atives participated only in discussions of Topic 2.1.1 of the protocol, which deals with a safety analysis of Soviet VVER reactors. The Soviet delegation (Appendix 2) was headed by Dr. Nikolai N. Ponomarev-Stepnoi_, First Deputy i Director of tha Kurchatov Institute of Atomic Energy.
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| In preparation for the meeting, the Americans sent background documents (see !
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| Appendix 3) to the Soviet Union 60 days in advance of the meeting (e mutual exchange of information preceding the meeting was provided for in the protocol).
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| The Soviets, however, did not send any documents to the Americans.
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| This report was compiled by Dr. Stephen Salomon of NRC's Office of Governmental and Public Affairs, based to a large extent on notes that he took while in Moscow.
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| Dr. Ross was responsible for oversight. Messrs. James Richardson (Office of Nuclear Reactor Regulation) and Alfred TabLada (Office of Nuclear Regulatory Research) wrote the sections describing the meetings on erosion / corrosion de-struction of piping and components and the visit to the All-Union Research Institute for Nuclear Sower Plant Operations. The section on fire safety was ~
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| written by Mr. Dennis Kubicki (Office of Nuclear Reactor Regulation). Mr. Wayne Hodges (Office of Nucle 0r Reactor Regulation) summarized the Department of Energy's seminar on VVERs (Snviet reactors). NRC's contract interpreter, - l Mr. Joseph Lewin, contributed substantially to the accuracy of this report as did the Editorial Section and Electronic Composition Services of NRC's Office of Administration.
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| 1-2
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| 2 PRELIMINARIES Thursday, December 1, 1988 l
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| In preparation for the first meeting of the working group 9, Dr. Stephen Salomon arrived in Moscow on Thursday evening, December 1, two days before the rest of the U.S. team was scheduled to arrive. Mr. Michael Nikitin, Protocol Officer, State Committee for the Utilization of Atomic Erergy (GKAE) provided transporta-tion for Dr. Salomon from the Sheremetyevo airport, Dr. Salomon went directly to ;
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| the U.S. Embassy to secure the South Texas Project Final Safety Analysis Report
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| , (FSAR), which was slated to be exchanged for the Soviet safety analysis of the Zaporozhe nuclear power plant according to protocol Topic 2.1. The reasons for safekeeping were twofold. First, NRC did not want the very bulky 20-volume set of the FSAR to be stolen or otherwise taken by unauthorized persons. Second, in the event the Zaporozhe safety study was not ready to be exchanged, a safe place was ne.eded to keep the FSAR until it could be delivered to the Soviets.
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| Friday, December 2, 1988 On December 2, Dr. Salomon cet with Mr. Phillip Otts, Acting Science Counselor, U.S. Embassy, to review safekeeping procedures for the South Texas Project FSAR and to report these procedures and details about accommodations for the U.S.
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| delegation to NRC's International Programs group by telex.
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| Saturday, December 3, 1988 On December 3, Dr. Salomon and Mr. Nikitin met the rest of the NRC delegation and several members of the DOE team when they arrived in Moscow. All members were able to stay at the Hotel Rossiya, a 1000-room hotel situated next to the Kremlin on Red Square.
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| Sunday, December 4, 1988 The NRC team met on Sunday evening, December 4, to plan how the Americans would I cover the topics during the five days of tne meeting. Safety research would be l discussed on Monday and Tuesday, erosion / corrosion on Wednesday, and fire safety on Thursday. Messrs. Ross, Richardson, and Taboada planned to visit the All-
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| . Union Research Institute for Nuclear Power Plant Operations in Moscow. Mr.
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| Hodges would monitor the Department of Energy (DOE) sessions on the safety of VVERS (Soviet pressurized-water reactors). Dr. Salomon would prepare the of fi-
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| , cial record of the meeting, " Memorandum of the First Meeting of the Working Groups of the U.S./U.S.S.R. Joint Coordinating Committee for Civilian Nuclear Reactor Safety, December 5-9, 1988." Mr. Lewin would interpret for all three topics: safety research, erosion / corrosion, and fire safety. Also scheduled was a meeting with the State Committee for the Supervision of Nuclear Power Safety (GAEN) to discuss public participation in the licensing process and the transmission of dv.uments from NRC's Executive Director for Operations, Victor Stello, to his Soviet counterpart, Committee Deputy Chairman Sidorenko, in re- !
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| sponse to some questions that were posed. Finally, during the week, the South Texas Proiect FSAR would be exchanged for the Zaporozhe safety analysis pursuant to the protocol.
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| 2-1 l
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| 3 PROTOCOL TOPIC 2: ANALYSIS OF THE SAFETY OF NUCLEAR POWER PLANTS IN THE U.S.S.R. AND THE U.5.
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| The joint safety working group met in 14cscow at the Kurchatov Institute of Atomic Energy (KIAE) to discuss the following sections of the protocol:
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| 2.1 Exchange of completed analyses of the level of safety of design of power units at Zaporozhe nuclear rower plant and the South Texas nuclear power plant. Following study of these analyses, questions and comments would be exchanged. Working group to meet to explain
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| , selected safety important differences. The co-leaders (U.S. and U.S.S.R.) of the working group would develop and recommend further safety assessment work to the JCCCNRS for approval.,
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| Group meeting - twice each year; the first meeting to be in April-May 1989 in U.S.*
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| 2.1.1 A seminar on the results of the US00E's analysis of features of Soviet desir#*ned VVERs will be held in the U.S.S.R. during November 1988.
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| 2.2 Topics and directions for safety research. Seminars once each year. The first seminar in November 1988t in U.S.S.R. concurrently with 2.1.1 seminar.
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| PROTOCOL TOPIC 2.1: EXCHANGE OF SAFETY ANALYSES Until Wednesday, December 7, the American component of the joint safety work-ing group anticipated that the South Texas Project Final Safety Analysis Report (FSAR) would be exchanged for the Zaporozhe safety report. Indeed, on December 7, Dr. G. L. Lunin, Head of the PWR Section, KIAE, said that the Soviets would be ready to exchange the documents on Friday, December 9. He said that although the Zaporozhe safety report, an 800 page summary document, had appendices that required explanations to be understandable, he could supply the appendices and necessary explanations later if the U.S. team wanted them.
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| . Also on December 7, Dr. Asmoltv said that all safety analyses dealing with risk assessment will be conducted according to protocol Topic 2, using the South Texas Project and Zaporozhe nuclear power plant safety reports. Dr. Ross said that the South Texas Project FS/R prepared by the utility could be used. How-ever, he believed that the NRC prepares a better safety analysis than the util-ity prepares because the NRC considers the concept of uncertainty, but the util-ity usually does not. Therefore, it is also important to exchange information on methodologies. Dr. Salomon was advised that the memorandum should say that
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| * Meeting scheduled for June 5-9, 1989 in Rockville, Maryland.
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| ** Meeting actually held December 6-9, 1988 in Moscow.
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| fMeeting actually held December 5-9, 1988 in Moscow.
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| 3-1 E_ _ _ - _ _ . I
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| the working groups will discuss the methodologies used for the South Texas and Zaporozhe safety analyses during the meeting scheduled for late 1989 in the Soviet Union.
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| On the morning of December 8, Dr. Salomon was informed by Drs. Sukhoruchkin and Gavrishin that a Rovno safety document, instead of the anticipated Zaporozhe document would be c.xchanged for the South Texas Project FSAR because Rovno has been better dacumented as a result af the International Atomic Energy Agency's Operational Safety Review Team (05 ART) inspection, Rnvno has been better docu-mented. In addition, Dr. Ross had a discussion with Dr. Ponomarev-Stepnoi on the subject. Dr. Ponomarev-Stepnai said that except for geological and some site-specific problems, the two plants, Zaporozhe and Rovno, are similar. He -
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| said that the Rovno plant will form the basis for the joint anaiysis. There seems to be some disagreement between toe Kurchatov Institute and the Ministry l
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| of Nuclear Power about which plant to study. The Minister of Nuclear Power wad .
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| not in Moscow at the time so Dr. Ponomarev-Stecnoi could not discuss this prob-lem with him. In fact, Dr. Ponoh;arev-$tepnoi did not know why the Ministry of Nuclear Power had initially chosen Zaporozhe for the study plant. In any event, wher, he visits the United States on January 6 [1989), Dr. Ponomarev-Stepnoi will cl6rify whether the working group is to study Rovno or Zaporozhe.
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| l On Friday, December 7, the Soviets presented the Americans with the Rovno safety summary report which is comprised of two volumes on Unit 3 conristing of about 1200 pages. Drs. Asmolov and Lunin took the South Texas FSAR to the KIAE library.
| |
| l PROTOCOL TOPIC 2.1.1: DOE ANALYSIS OF SOVIET VVERs The U.S. Department of Energy (DOE) seminar on Soviet VVERs was scheduled to L take place over four days (see agenda, Appendix 4). The seminar was based on the study prepared by DOE on the VVER (DOE /NE-0086, "Departh.cnt of Energy's Team's Analyses of Soviet Designed VVEP.5," October 1988).
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| A summary of the DOE seminar is provided by M. Wayne Hodges in the two sections j that follow.
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| Overview Session The DOE team met initially with the Soviet contingent on Tuesday, December 6, 1988. Discussions consisted of introductions and schedule. Because some of the Soviet delegation had come from laboratories outside of Moscow, Dr. Ponomarev- ,
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| Stepnoi requested that the general discussion be brief and the smaller working group meetings begin as soon as possible.
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| The DDE team then gave an abbreviated overview which condensed two days of presentation into approximately two hours. In this brief overview, Mr. Edward Purvis emphasized the draft nature of the report and solicited comments and criticisms. Dr. Ponomarev-Stepnoi indicated that the Soviets were still review-ing the supplements; therefore, the Soviet comments would not be complete.
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| Dr. Ponomarev-Stepnoi stated that the exchange was a first step and he wants 3-2 I
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| | |
| \ l l
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| to continue cooperation in the future. The Soviets appeared to have a strong )
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| interest in the DOE report on VVERs. !
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| l During the introductions, one of the VVER designers expressed an interest in !
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| commenting on the " Red Book" (DOE /NE-0084, Rev. 1, October 1987, Overall Plant Design Description VVER Water-Cooled, Water-Moderated Energy Reactor") in addi-tion to commenting on the analysis report Mr. Hodges' impression was that some of the assumptions made in the Red Book were in error and needed to be changed. The Red Book is descriptive only and presents no analytical results.
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| The overview pre:,entation was completed before lunch and it was agreed to begin s , the smaller working group sessions after lunch.
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| Working Group Session fir. Hodges was informed by Mr. Purvis (DOE) that DOE did not want an NRC presence at the meeting. Mr. Purvis insisted that although DOE had agreed to NRC presence for the overview session it had not agreed that an NRC member attend the working group sessions. Mr. Purvis stated that an NRC presence would inhibit the pre-senters. Dr. Rosen supported Mr. Purvis in the decision when questioned by Dr. Ross.
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| Although Mr. Hodges was unable to participate to the ex'ent planned, he learned enough from his brief perusal of the DOE report to recognize that there are a number cf problems with the report.
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| A comment by Dr. Rosen during a break for one of the sessions indicated that the Soviets questioned the use of a number of DOE assumptions. The Soviets will supply more detail so that DOE can do a re-analysis.
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| The DOE Report The following material is based on some conversation with Dr. Rosen on the DOE analysis of the Scviet VVERs.
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| Dr. Ro>en (DOE) informed Dr. Salomon on Thursday, December 8, ttat the DOE has some negative comments in its analysis of the VVER (Soviet pressurized-water reactor). He wanted Dr. Salomon to attend a discussion with Dr. Ponomarev-Stepnoi because of the political matters and the sensitivity involved. Because NRC had been previously excluded from participating in the working group discus-sions, Dr. Ross decided not to attend the DOE discussion.
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| The DOE analysis pointed out strengths and weaknesses of the VVERs. There are also some factual errors that Dr. Ponomarev-Stepnoi wants corrected before any limited publication. Therefore, the revised version will be more accurate.
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| Dr. Ponomarev-Stepnoi is afraid that environmentalists will use the DOE study to eliminate nuclear power plants in the Soviet Union, or slow the construction of installations. Consequently, he wants a good study to be published.
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| Dr. Rosen emphasized to Dr. Ponomarev-Stepnoi that if the interaction between DOE and the Soviets is too great, the credibility of the study may be jeopardized.
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| Too great an interaction would look like the Soviet Uniun and the United States are not giving a totally unbiased evaluation. An arms-length relationship needs to be maintained. The DOE does not have any more funds to do additional work beyond the final revision.
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| 3-3
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| l Dr. Ponomarev-Stepnoi wants to use the study to support some research in certain l
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| areas and to strengthen the system. According to Mr. Purvis and others, the l analysis shows that if there is an accident the system is very dependent upon how the operator performs. For comparison, according to Mr. Purvis, "the American reactors have automated features and the operator comes into play after .
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| a couple of heurs rather than at once." (Note: This is a misinterpretation on ,
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| the part of Mr. Purvis. ) DDE plans to correct some mainly factual errors and '
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| the distribution will be limited to a restricted number of persons.
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| Also, Dr. Ponomarev-Stepnoi asked Mr. Purvis if he could use the DOE report to support some safety analysis that he would submit to India. (Note: The Soviet i Union had sold two 1000-MWe VVERs as a turnkey (fixed price) project. Nobody -
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| l seems to know the cost or other details of the deal. Apparently, it will be sited in the south of India and will be cooled by sea water.) Dr. Rosen , aid that Dr. Ponomarev-Stepnoi may use the DOE study to help support marketing the .
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| VVERs, but that he (Dr. Rosen) is not at all involved in this use of the study.
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| PROTOCOL TOPIC 2.2: SAFETY RESEARCH Monday, December 5, 1988 The NRC safety working group met with its Soviet counterpart at the Kurchatov Inttitute of Atomic Energy (KIAE) located in northwest Moscow. The meeting was held in a cottage on the grounds of the Kurchatov Institute. KIAE has been characterized as the Los Alamos of the U.S.S.R., where wtapons and reactors are designed. KIAE currently employs about 10,000 people.
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| After welcoming the Americans, Dr. PonoEarev-Stepnoi brought the first session to order at about 10 a.m. in the central conference room. Dr. Ponomarev-Stepnoi is First Deputy Director of the KIAE, and is also Co-Chairman of the JCCCNRS.
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| The Soviets and Americans review t and discussed the agenda (Appendix 4).
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| Dr. Ross said that the agenda looked satisfactory.
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| Dr. Ross noted that a meeting relate 1 to a request of Dr. Victor sidorenko regarding public participation in the licensing process was not on the schedule.
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| Dr. Ponomarev-Stepnoi said thr; ur. Sidorenko was in the hospital and would be represented by Dr. Gutsaiov. That meeting would take place either on Wednesday or Thursday. The exchange of safety documents was discussed. Dr. Ponomarev-Stepnoi said that the Soviets were not ready on that day to exchange the Zaporozhe safety study but would later discuss the procedures on exchange of documents. .
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| Both delegations were introduced (see Appendices 1 and 2). The working group agreed that the presentations would be alternated between the Soviets and the Americans. After each delegation expressed its ideas about further procedures, general proposals would be made. These would be summarized in a memorandum.
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| (The ward " protocol" would be reserved for the August 31, 1988, document that was signec ir, Moscow.)
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| Dr. Ross presented an overview of NRC safety research (protocol Topic 2.2).
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| Soviet questions dealt with other U.S. groups engaged in safety research; the 3-4
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| | |
| relative costs of NR", Electric Power Research Institute, and DOE safety re-search; and the materials research program related to nuclear safety analysis.
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| (Before the meeting began, the U.S. delegation gave the Soviets a copy of NRC report NUREG-1319, which summarizes NRC research and also gives several research plans for a number of NRC projects.)
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| The Next Generation of Nuclear Power Plants The first Soviet research paper was presented by Dr. I. S. Slesarev on nuclear power reactors of the next generation. His presentation summarized some of the conclusions on nuclear power plant safety that Dr. Ponomarev-Stepnoi had
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| . presented at Jackson Hole, Wyoming last September [1988] at a conference co-sponsored by the American Nuclear Society, the Department of Energy, and the Atomic Energy Society of Japan (see Appendix 5). Among the reactors discussed were the advanced light-water reactor (ALWR), the high-temperature gas-cooled 3 reactor (HTGR), and the liquid metal reactor (LMR) thought to have a probability of <10 5 per reactor year for a Chernobyl-type accident and a probability of 1
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| < 10'J7 per reactor year for a TMI-type accident. The ultimately safe reactor l would have a probability for a Chernobyl-type severe accident of <10 7 per reactor year. Dr. Salomon asked Dr. Slesarov to what degree NRC safety goals were incorporated into the analysis. Dr. Slesarev replied that he is not sure he has all the necessary information to make such a determination. Dr. Ross y asked Dr. Slesarov how he had used the International Safety Advisory Group (INSAG)
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| * sport on reactor safety. Dr. Slesarev stated that the INSAG goals were useful in his analysis, but that he had found them oriented to past and present reactors rather than to future reactors. For these reasons, Dr. Slesarev 4 recommended that a working group be established to develop safety concepts for the new generation of nuclear power plants (see Appendix 12, Item 1.3.7). ,
| |
| 1 Nuclear Power Safety in the Soviet Union Dr. Asmolov discussed investigations of nuclear power safety in the U.S.S.R (see l translations of three of his papers in Appendix 6). He stated that his goal is nuclear power safety and reliability assurance. He hoped the two nations would not only exchange information but also would engage in commercial programs such ,
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| as sales of equipment and services. He was anticipating that this joint meeting '
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| would produce a final document, or memorandum, before it ended. In response to a question from Mr. Kubicki, Dr. Asmolov noted that in risk assessments fire protection was included in relation to both external and internal events; risk was to be defined in its broadest term as in risk analysis. With regard to uncertainty, Dr. Asmolov said that the Soviets have an entire organizational
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| , division devoted to the study of that subject, and another division devoted to the study of reliabilty. When questioned about the emphasis on prevention rather than mitigation, he answered that the Soviets stress prevention, but also pursue mitigation.
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| At this point, Dr. Ponomarev-Stepnoi interjected that at new power stations using the WER-88 m,dels, no firm decision has been made regarding three items:
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| the system used for discharging containment gas through filters (venting),
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| hydrogen detection systems, and the use of core catchers. However, he stated that the Soviets agree that they need to decrease the probability of a severe accident and that decreasing the frequency of accidents is a less expensive approach, both in terms of the economy and in terms of human consequences, than 3-5
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| | |
| implementing the mitigation technologies. Although Dr. Ponomarev-Stepnoi wants to work toward decreasing the possibility of accidents occurring, oe neverthe-less, wants to continue working toward decreasing the consequences of a :iuclear accident. He is having trouble coming up with a logical economic model that tells him how to proceed.
| |
| Pressure Vessel Integrity and Plant Aging Dr. Ross discussed the topic of the integrity of reactor components from the NRC perspective. Subjects included NRC safety issues and research on reactor vessels, piping, seismicity, and aging. Two questions were raised: Is the idea of extend-l ing the life of a reactor to 60 years based on technology or economics? Is such .
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| l an extension a real problem or merely a topic for discussion? Dr. Ross answered that according to the electric power utilities, the problem is economic and that it is now a real problem since Yankee Rowe, one of the oldest U.S. reactors, ,
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| licensed 28 years ago, is being considered for a license extension.
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| Professor Amaev discussed radiation effects on VVER reactor vessels, including embrittlement of the vessel material and the recovery of these properties by thermally annealing the vessel. Radia;;on embrittlement has been considered a majnr concern in establishing design safety factors for reactor vessels. Other factors include cyclic loadings and aging effects. The Soviets have done exten-sive research on radiation embrittlement including the effects of vessel material chemistry, fluence, and flux, and have developed an empirical correlation of embrittlement effects with these variables similar to the correlation given in NRC Regulatory Guide 1.99 (except that phosphorous and copper content are used for the chemistry factor). They have studied weld metal as well as base metal and also have a program to examine decommissioned reactors, including Novovoronezh Unit 1.
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| Thirteen of the VVERs ate operating w,'th unclad reactor vessels and some of these have no surveillance specimens. Early reactors had surveillance specimens that were removed because of operational problems. Later plants will have surveil-lance specimens installed in the reactor vessels.
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| The Soviets have performed extensive studies on thermal arnealing of reactor vessels to eliminate the embrittlement effects of neutron irradiation. They have studied the effects of metal chemistry, coolant, fluenr.e, irradiation temperature, annealing temperature, and length of time at annealing temperature.
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| As a result of these studies, three reactors have been annealed, including Novovoronezh Unit 3, Armenia Unit 2, and Lubmin Nord Unit 1 (in the German Democratic Republic). The process involves electrically heating the reactor . i vessel beltline weld region (above and below the circumferential region near the core) from the inside at a temperature of between 420 C and 460 C. No detri-mental effects of annealing were observed. For the VVER-1000, 180 C is the
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| " critical embrittlement temperature" or what in the United States is called nil '
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| ductility temperature (NDT).
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| Dr. A. A. Tutnov discussed a quantitative analysis of the probability of reactor failure as a result of leakage or rupture. The analysis relates flaw distribu-tion, flaw size, anc' possible mechanisms of failure, and determines critical values for flaws and the probability of their distribution. {
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| (Note: The NRC 3-6 i
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| | |
| g staff perceived that this mathematical approach had been validated only to a small degree with actual data, and plans to discuss this further at the June 1989 working group meeting.)
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| Two Soviet reactors, the RBMK and the VVER-1000 were evaluated for failure probability. The probabilities for leakage and rupture for the two reactors are as follows:
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| Reactor Leakage probability Rupture probability RBMK 6.3 x 10 4 8 x 10 5
| |
| . VVER 3.8 x 10 4 2.5 x 10 5 It was noted that, for the VVER-1000, 180 C is the estimated " critical embrittle-ment temperature" for the reactor vessel steel, or what in the United States is approximately Charpy-V toughness level. Because of the 180 C critical embrittle-ment temperature, the Soviets have instituted ultrasonic testing of the vessel every 4 years and have raised the inlet temperature to 270 C compared with the 250 C temperature for the early VVERs.
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| N ' alculation for fuel pin failures in the RBMK design-batis loss-of-coolant MimM (DFLOCA) is as follows:
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| Power (MWe) Probability (percent) 1000 2 x 10 5 1100 3 x 10 5 1200 8 x 10 1 1300 3.9 1500 18.0 This is a particular example for a particular accident. (Note: The NRC staff noted the sharp rise in probability at 1200 MWe. The reason for this was not explained. The NRC staff will ask the Soviets to explain this phenomenon at the June 1989 meeting.) The codes have been developed over a 20 year span. The Soviets gave Dr. Ross two recent articles from the journal Soviet Atomic Energy.
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| Although the articles do not give probabilities, the method of determining prob-abilities is described. The Soviets will present the latest research results at the June 1989 meeting.
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| Radiation Burden The next topic was related to radiation burden on reactor vessels and was pre-sented by Dr. A. V. Khrustalev. The steel used is 15KH2M4A (0.15% Cr, 2% Cr, 1% Mo, 1% Vn, A = special melt). Dr. Khrustalev reminded the group that the chi.tistry f actor had already been discussed by Professor Amaev and that there are other factors, namely fluence, to be considered. Dr. Khrustalev described a study of flux determinations for reactor vessels at the inner and outer wall using 60 symmetrical segments of the core. The parameters are T = 150,000 hours s 2 years, and $ = 2 x 1023 for energy greater than 1/2 MeV. For this study, the l Russians developed two codes, P03-6 (R0Z-6 in English) and RADUGA (RAINBOW in English). They also used two U.S. codes, DOT and ANISN. Calculations run for the reactors were compared with measurements taken on VVER-440 mockups in i
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| 3-7
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| __ _ _ _ _ _ _ l
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| | |
| Czechoslovakia. The latter (the calculations or the measurements) was done with Czech colleagues. (Note: Much of what Professor Amaev talked about can be found in the Russian paper by I. Shults, " Czech SSR," Atomnaya Energiya, Vol. 64, No. 6, June 1988, pp. 410-445.) The models represent realistic fluence values at the power plants. In general, diameters in Soviet vessels are much smaller than those of U.S. vessels because the Soviets must transport reactor vessels by rail; U.S. reactor vessels can usually be transported by barge, allowing larger vessels.
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| Tuesday, December 6, 1988 i On Tuesday, December 6, the NRC group met briefly before proceeding to the -
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| Kurchatov Institute to continue the safety discussion with the Soviets. The team decided to prepare a list of documents exchanged and to ask for the anno-tated description of the RASPLAV code (MELT in English). Dr. Ross had noted to .
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| the Soviets that NRC had transmitted documents to the Soviet Union before the meeting (see Appendix 3), but that the Soviets had sent the NRC none of their research planning documents.
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| Among the observations made by the group at this meeting, Mr. Richardson said that the Soviets seemed to be targeting mitigation toward older reactors, whereas prevention was targeted toward the newer reactors. He also observed that Dr. Ponomarev-Stepnoi is interested in cooperative research.
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| The U.S. team decided that it needed more detailed information before the discussions could proceed. One of the Soviet participants suggested joint calculations on a standard problem such as may arise from the South Texas /
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| !.aporozhe safety analysis project.
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| The delegations began discussions again at the KIAE at about 10 a.m. Dr. Asmolov introduced two more Soviet representatives--A. M. Kovalevich, Deputy Director, Scientific Technical Center, GAEN, and A. Mysenkov, Senior Researcher, KIAE.
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| The teams decided to continue the discussions on severe accidents and on thermal hydraulics, and in the afternoon, to prepare the joint memorandum which would cover joint interests and areas of future cooperation. At the Soviet request, the topic of waste management was deleted.
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| Soviet Studies on Severe Accidents With Core Melting Dr. R. V. Arutyunyan, KIAE, discussed theoretical analysis and numerical simula-tion of heat transfer and fuel migration after severe accidents at nuclear power plants. Dr. Arutyunyan's paper on this subject is translated in Appendix 7. .
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| After the Chernobyl accident, the Soviets began to work on such subjects and studied fuel-concrete interaction. At Chernobyl, the Soviets dropped sand on the reactor to extinguish the fire. The fuel fragment melted the sand and ,
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| started to move down 15-20 cm at a speed of about 1-3 cm per day. There is a two- and three-dimensional (2-D and 3-D) heat transfer and thermal hydraulics.
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| The physics is complicated.
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| To illustrate the action of fuel fragments on the sand, a slide presentation was given. In addition, a temperature distribution map was shown as well as a com-puter simulation as a function of time. Two fragments were studies with spacing, and 2-0 and 3-D geometries were investigated.
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| 3-8
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| | |
| r_.
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| A flat very unstable layer of fuel also was studied. At a later stage, drops form together to make a solid layer. Later, the layer goes through 11 tubes that are cooled by water. Then the fuel fuses into a layer. Instabilities form and then go to 40 cm. Then the fuel divides into several drops. There are lots of layers of different thicknesses.
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| The passage of the flat layer through the system of cooled tubes was studied.
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| After the layer flows around the tubes, the movement continues. A thin layer becomes solid in contact with the tubes. Under RBMK, Unit 4 Chernobyl, a flat heat exchanger was used. It takes 10 days for flow over the flat heat exchanger.
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| , Other studies of a more general nature than Chernobyl have been conducted. The results are similar to those predicted by studies done in the United States, the Federal Republic of Germany, and the Soviet Union. The CORCON computer code, which is a 2-D analytical model of fuel-concrete, is used. It can be run in about 10 minutes on a personal computer. The influence of different factors on the speed results in times from 1 to I times the normal time.
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| Dr. Arutyunyan asked the Americans to verify the Soviet codes, to check their reliability and the constants used in the codes. All these influence the final results. The fuel movement has to be studied in materials other than concrete.
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| There have been proposals to study ceramic materials at both KIAE and at the V. G. Khlooin Radium Institute near Leningrad. During 1989 and 1990, the Soviets hope to verify the models applied to fuel movements in these types of materials.
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| When Dr. Ross asked h,w many kilograms of fuel were used in the experiments already conducted at KIAE and the Khloplin Institute, he was told that several grams of irradiated fuel and 20 to 30 kg of non-irradiated fuel, with some zir-conium added, were used. The large number of uncertainties in U.S. experiment versus theory was noted, such as the heat coefficient of overlying water pool, the oxidic fraction, and the void fraction. Dr. Ross suggested that it would be interesting to compare the theoretical work done in the Soviet Union with the experimental work done in the United States, such as the SURC-4 (sustained uranium concrete) experiment. The work at WECHSEL in the Federal Republic of Germany was also discussed. Dr. Ross then suggested a blind experiment to the Soviets: The NRC would provide the Soviets with initial conditions and geo-metry, and the U.S.S.R. would predict the results; then the NRC would give the Soviets the experimental results so they could compare the results of the experi-ment with the predicted results.
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| Dr. Asmolov recounted how the experiments got started in May 1986. Different
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| , heating approaches were used, such as laser and RF induction. Their results formed the basis for the facility built under the damaged unit at Chernobyl. By means of the ACE computer program, fuel-concrete interaction is studied. These experiments were conducted at the Khlopin Institute using 20 to 30 kg of simu-lated fuel and dioxide fuel. Part of these results had just been presented by Dr. Arutyunyan. These experiments will be conducted at KIAE and at the Khlopin Institute in 1989 and 1990 to verify the models.
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| American Studies on Severe Accidents Dr. Ross presented NRC's severe-accident research. This covered topics on core melt, reactor system failure, containment structure failure, reactor risk analysis, severe-accident policy, the eight areas of uncertainty, and a 30 year 3-9
| |
| | |
| l overview of regulatory policy. A number of questions were asked. The first was whether the steam explosion was an optimistic or final view. The answer was given that it was a final view. The United Kingdom is doing more in this area, l and the NRC is watching this work. The second question dealt with the observa-l tion.that there are a number of publications that describe very rapid steam explosions. In one, the entire installation could be destroyed. An NRC re-spondent said that Sandia National Laboratory has experimented with destructive consequences. It is believed that the probability of complete destruction is sufficiently low that it is not worth pursuing. The core will be fractionalized.
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| Core water interaction that will take place is not that rapid.
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| The Soviets wanted to know to what degree the NRC was engaged in experimental -
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| work. In reply, Mr. Taboada described his work as an NRC project manager and told the Soviet group that NRC owned some equipment at national laboratories operated by NRC contractors. .
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| As to uncertainty areas, the question was asked: "Where does the vessel fail?"
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| The answer given was that vessel failure could be melt-through of a penetration or gross bottom head collapse.
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| Someone asked if iodine is elemental or combined. The reply was given that the answer is not completely known, but it may not make any difference. For most scenarios, given enough time, elemental iodine recombines before the con-tainment fails.
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| The question of the main conclusion of fuel-concrete interaction was raised.
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| In answer, it was stated that there are several uncertainties involved, such as corium-steel surface; the question of whether a crust forms; water-heat transfer coefficient; core-concrete interaction is well defined, but in present models, homogeneous. The working group should spend more time on corium-oxide-metallic interactions.
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| The group was told that experiments are being performed on the effects of boron on organic iodine at NRU (a Canadian reactor).
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| The group was also told that George Parker is doing experiments at Oak Ridge National Laboratory on the eutectic effect, but the results have not yet been published. The main interest-of these experiments is with the boiling-water reactors (BWRs) that have kilogram quantities of B 4C.
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| In corium-concrete U0 2 , the layer at the bottom mixes with iron. The layer forms afterwards. The production of gases mixes things up. .
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| Dr. Ross told the group that because the Soviet long-range plans are not under-stood, it is difficult to say in what fields NRC will produce work for this ex- ,
| |
| change program. During the next two months, NRC will be redoing a risk analysis on risk contributors. Dr. Ross then discussed the topic of containment integrity including the actual structural capability and the test of a reinforced-connete containment model in which steel failed catastrophically. The finite element l
| |
| analysis that was perforraed was too gross to pick up the abnormality where it failed. Both the concrete and steel containment tests confirmed in part the l
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| 3-10 1
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| t_____________m______ _ _ _ - - _ _ - -
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| | |
| validity of codes and gave the confidence that margins of codes are indeed true.
| |
| It is not prudent to assume that steel containments will leak before failura; this will affect the ultimate consequences of a severe accident.
| |
| Core damage frequency distributions will be published shortly. Last year's calculations for Surry, Grand Gulf, Sequoyah, and Peach Bottom nuclear power plants were given (see NUREG-1150). A detailed uncertainty analysis using expert opinion will appear in the revised NUREG-1150.
| |
| Soviet Studies on Severe Accidents Without Core Melting Dr. Kramerov then talked about Soviet work on severe accidents in which the ,
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| reactor core does not melt. The accidents were described specifically for i channel reactors for mass fuel melting which are several orders of magnitude less than light-water reactors (LWRs). There is partial failure of cooling, dryout of core, and fission fragment release. There is a two phase mixture, rupture of tube plate on loss of integrity in the second circuit. Therefore, the Soviets are planning to study (1) using rigs or real reactors, (2) iodine and cesium, and (3) a number of other problems. They are studying release ki-netics with increased overpressure. Studies of probability include short-term heating 600 C to 1000 C and with increased outer pressure. Then the working group should study the distribution of released products between liquid and steam phases. The kinetics of the approach and the distribution of phases should also be studied.
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| When the loss'of cooling is studied, pulses in water flow are noted. Then the study addresses changes in iodine forms and distributions between phases and leakage into the coolant. The Soviets are investigating how much of the iodine is left in the steam phase to move into the containment and possibly be released through different penetrations. All these studies are at the initial stage and would be explained if the Americans wish. Some of these should be based on U.S.
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| data already obtained. What is important from the U.S. point of view should be known for severe accidents without core melting.
| |
| Dr. Ross said that the Soviet studies sound very interesting. He described one sequence that the United States is required to evaluate, namely, the rupture of one steam generator tube. NRC does not require any other analysis when there is no core melting. On the contrary, the Federal Republic of Germany has such design requirements. Something might be put into the memorandum on this subject.
| |
| Thermal-Hydraulic Researcn in the United States Dr. Ross discussed the topic of thermal hydraulics.
| |
| His presentation covered modeling activities of current codes (RELAP and ldAC) and the associated separate effects and integral experiments. Special activ-ities related to the Babcock & Wilcox design were noted. The overall budget i for this effort was given. The new Appendix K was discussed. This topic was included under the research category " prevention of core damage."
| |
| In answer to the question: "How well have the codes been verified especially when the processes are slow?" Dr. Ross replied that the codes are essentially 3-11 l
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| | |
| l l
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| completed and support experimental data. RELAP and TRAC are believed to be good enough. The codes, however, are still being modified because such work is a commercial enterprise.
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| Vendors improve TRAC and have a product to sell. Only mistakes are corrected.
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| In addition, the vendors try to find ways to run codes cheaper and faster. The NRC does this too. For large-break and small-break loss of-coolant accidents (LOCAs), calculation show that the results are in agreement wit.h experiment.
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| The vendors incorporate proprietary infor nation.
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| If the Soviets want to procure current codes, RELAP is recommended as a basic code. Sometimes TRAC is used. TRAC would take 100 hours of computer time to -
| |
| run. NRC has furnished RELAP to a number of countries but always for exchange of information. These codes are not for sale Dr. Asmolov wanted to include this point in the next day's discussions. Dr. Ross said that it could certainly ,
| |
| be included ia the memorandum as an item for next year's meeting. Dr. Asmolov said that the Soviets could use RELAP to verify some experiments.
| |
| The Soviets questioned if the results can be reliable because of the smaller pipes, velocities, and phase. They also asked about scaling. The answer was given that the break area to volume ratio is generally scaled. That is why the term semiscale was invented.
| |
| Dr. Asmolov wanted to discuss flows at low velocity and mixed and various phases.
| |
| Dr. Ross said that the Dubrovni conference on severe accidents in May 1989 would be the right place for such a discussion if the appropriate people attend.
| |
| In addition, Dr. Ross noted that the protecol does not include thermal hydrau-lics as a topic. On the contrary, Dr. Asmolov believed that it did under protocol Topic 6 , Severe Accidents. (Note: Thermal hydraulics is not men-tioned specifically.)
| |
| Thermal-Hydraulic Research in the Soviet Union Dr. Asmolov then presented the Soviet view on thermal hydraulics. Research began in 1986 to understand severe accidents in the primary circuit. Thermal-hydraulic codes were based on conservative assumptions. The be t-estimate codes were used for a more raalistic description of parameters. Codes that describe core rupture in a large volume are new developments. The Soviets tried to use a code developed in the United States, such as MARCH-2, after it was received from the International Atomic Energy Agency (IAEA) package of codes. They de-cided to combine the Soviet code with the Source Term Code Package (STCP) (IAEA package of codes). There was sufficient detail before core rupture to use the ,
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| STCP and then come developments. It could be used for the design-basis accident but not for severe accidents. The main codes in this set include MOCT-11 (BRIDGE-11 in English). Previously. M0CT-7, 9, and 10 were useo. The codes describe the VVER in different modes excluding large leaks. The code uses 3-D ~
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| neutron kiretics. There is a detailed description of the fuel elements. The DRIFT model is used in the core. The pressurizer, steam cenerator, and all the factors that affect reactor parameters are included. It considers process, such as small-break LOCA, and secondary rupture.
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| Codes such as RELAP are used for the VVER and for other reactors. MAST is another computer code used for heating reactors. This is the same class of 3-12
| |
| | |
| codes-as M0CT. Some modules are the same. In order to support these codes, KIAE has developed a code for fuel elements, MYLA-2. 3-D calculations are based on a specific key which take into account gap conductance. It is called GAP-THERMAL.
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| For large leaks, a code developed at the Hydropress Institute (located at Podolsk south of Moscow) is called TECH-M4 (LEAK-M4 in English). This program interfaces with the STCP in MARCH-2, but only under conditions for large-scale accidents. Total station blackout accident is one of the accident conditions examined. The important parameter is how fast this develops. There is a hori-zontal steam generator which takes a long time to evaporate. It takes twice
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| ~
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| u long to dry out. Under Soviet conditions using horizontal generators, the iate of development of an accident may be one-half the rate of vertical genera-tors. Total generator blackout would occur under total station blackout. Such problems exist. This is a brief description of the Soviet codes. They are in-terfaced with U.S. codes. There is the same problem of verification. Two of the facilities described by Dr. Asmolov to develop thermal codes and verify.
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| them are 1:3000 and 1:500 scale test stands. It has several loops and enabled the model to describe high pressures beyond the critical pressures. This facil-ity will enable us in the future to make designs, as well as to model without SCRAM.
| |
| The MOCT can model processes with pressures of 250 atmospheres and higher.
| |
| Because the Soviets are unable to construct such a facility within Moscow city limits; the Electrogorsk facility (72 km [s45 miles] from Moscow) is used.
| |
| Dr. Nigmatulin works at Electrc,gorsk which is part of the All-Union Research Institute for Nuclear Power Plant Operations. Also, the Soviets are aware that there is SPES in Italy which is an international effort. In addition, there is a Hungarian facility and a Czech loop called the large water loop. The first test stand will be a 1-MWt rig at the Hydropress Institute. The KIAE is devel-oping at KIAE a stationary facility of 6 MWt boiling phase with a loop.
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| During the team's visit to the All-Union Research Institute for Nuclear Power Plant Operations on Thursday, December 8, the Soviets presented calculations of thermal hydraulics for core melt based on large pipe rupture and a station blackout. The sequence is as follows: 850-mm pipe rupture, lack of coolant dries core (144 seconds), fails core (960 seconds) which raises in temperature to 1200-1400 C, core melts through reactor vessel (12,300 seconds), reacts with concrete and builds pressure in containment to beyond capacity, and fails in-stantaneously at the weakest link (10 hours). The Soviets checked this analysis with the American MARCH program. Dr. Ross commented that the sequence is con-unlikely in the United States. The probability of a large-break sidered very" per reactor year, and the mode of failure of containment expected LOCA is 10-is a minor tearing of the liner for a concrete containment.
| |
| The Soviet Scientific Technical Center l Dr. O. M. Kovalevich, Deputy Director, Scientific Technical Center, State Com-m'ttee for the Supervision of Nuclear Power Safety (GAEN), described the new center. It was created within the last year and employs atout 150 people, most of whom are scientists. It is an independent scientific body, not just part of 3-13 l f l L__________ i
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| | |
| the centralized organization. The budget now is several millions of rubles.
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| The center can contract but can't spend as much money as the NRC can spend.
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| The center has two groups. The first one, which takes up to 30-40 percent of the working time, is a scientific and technical support staff of GAEN. About 50
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| . percent of its staff performs research. They make conclusions on the documenta-tion and norms developed in the center. Conclusions are made on specific de-signs. GAEW issues what really amounts to a license. Accident investigations are conducted. The center also performs research in a manner similar to the way NRC parforms research. Designers and builders analyze designs indepen-dently. Because of this, they can supply expert opinion. The seven divisions of the organization cover all subject areas which include:
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| (1) Safety requirements and criteria, both general and specific ones (2) Physics (3) Strength of materials and equipment adequacy _ i (4) Instrumentation and controls for reactors and human factors (displays, layouts,etc.)
| |
| (5) Personnel training (6) . Radiation safety on site (7) Experimental research GAEN has just begun to issue recommendations on regulations. Research programs coter most of the problems, but most activities have just started. Most re-searchers have continued their own topics. Naturally, many of the staff come from industry so the direction of their research had to be changed.
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| Probabilistic Risk Assessment Methods Dr. A. A. Tutnov then discussed probabilistic risk assessment (PRA). These acti-vities are carried out at the center by some of the U.S.S.R's leading scientists.
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| At present, some estimates for core rupture have been prepared, but because there are large uncertainties these estimates are not presently useful. If some PRA is introduced into the regulations, this will be a good step. In the future, there will be more frequent meetings and the center hopes to increase cooperative '
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| activities with other research groups.
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| l The center has established several requirements for design, manufacture, control !
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| of defects (quality control), and requirements for operation. These are taken seriously.
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| The center is monitoring fluence at the reactor vessels at operating reactors.
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| There is a non-uniformity at the azimuthal direction. There is a deviation .
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| from the median up to a factor of 2. The non-uniformity varies with time.
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| They could correctly monitor oower by measuring N M downstream from the core. -
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| Whether this should be a standard system or an experimental one has not been determined because the evidence so far is fragmer.tary. Other people are inter-ested in the results too. Two diagrams were distributed (see Appendix 8) that illustrate measurement of the heat output of the VVER-1000 reactor from the flow rate of the coolant in the primary loop by the radiation method.
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| With regard to " bubbling condensers" (called supression pools in the United States), there is one system at the bottom of the Cuban reactor containment.
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| 3-14
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| ___-____-.-.___---___._,m_ - _ _ _ -
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| | |
| There are no oscillations or pressure changes. .This is the center's own devel-opment. A core catcher concept is being considered to be used to prevent core-concrete interaction. The core should be cooled by natural convection. The center has paid a lot of attention to PRA.
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| Dr-. Asmolov concluded the day's working group meeting by saying that the U.S.S.R.
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| has wasted 7 years between the accident at Tnree Mile Island and the Chernobyl accident and now looks forward to working with the United States. He said that the Soviets would have suggestio7s for topics to be included in the joint memorandum.
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| _ Wednesday, December 7, 1988 Public Participation in the Licensing Process The NRC safety working group met with Dr. A. T. Gutsalov, State Committee for the Supervision of Nuclear Power Safety (GAEN). Dr. Gutsalov represented Dr. Victor Sidorenko who was hospitalized. The subject, public participation in the nuclear power plant licensing process, was being discussed in response to a request made by the Soviets to Victor Stello, NRC, for some material on the sub-ject. The transmittal telex and material were given to Dr. Gutsalov. He stated that the public has to be more involved in the licensing process. Furthermore, the public should be involved at various stages. Right now, the Soviets are preparing a law on nuclear power licensing and they want to know what has happened in other countries when all branches of society that are not normally involved in the licensing process become involved in that process.
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| Dr. Ross described the safety 'and' environmental review processes as they take place in the United States and told how private citizens as well as local and State governmer.ts participate. He also described the roles played by the NRC staff and the licensing boards. Public critiques of both the environmental report prepared by the electric utility applicant and the NRC staff's environ-mental impact statement were described. Dr. Ross emphasized the fact that NRC maintains uniform radiological safety standards throughout the country.
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| The Soviets asked many questions on how NRC paid the licensing board salaries and the relationship between (nard members and NRC staff.
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| e 3-15 i
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| | |
| 4 PROTOCOL TOPIC 4: FIRE SAFETY The joint fire safety working group met in Moscow on Thursday, December 8,1988, to discuss the following sections of the protocol:
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| 4.1.1 Exchange of information on the effects of hydrogen concentra-tion levels on the propagation of flames, maximum expected pressure following an explosion, and the rate of hydrogen release in contain-ment following a severe core-damage accident. x
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| , 4.1.2 Exchange of information on the combustion processes of hydroger-containing vapor gas mixtures in large volumes (e.g.,
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| containments) 4.1.3 Exchange of information on capabilities of electrical cable to withstand fire, and design of cable penetration through bulkheads.
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| 4.1.4 Exchange of information on methods for fighting large fires under high radiation conditions for extended periods of time.
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| 4.1.5 Exchange of information on fire suppression systems, venti-lation systems to protect the control room environment from external fire.
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| 4.1.6 Exchange of information on fire protection coating for structural steel.
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| . Working group to meet and discuss Topics 4.1.1 through 4.1.6 in April-May [1989] in U.S. ,* with information exchange beforehand.
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| Mr. Dennis Kubicki, NRC, met with four engineers / researchers from the Soviet Ministry of the Interior at the Kurchatov Institute to discuss various topics within the overall category of nuclear fire safety. In the Soviet Union, the Ministry of the Interior is responsible for fire safety at all industrial and other facilities, including nuclear power plants. Because of delays associated with meeting site logistics, the need to agree upon a tentative schedule of events for the June 1989 meeting, and the time taken by sequential translations,
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| . the actual exchange of technical information included about 2 hours of dialogue.
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| Mr. Kubicki presented the Soviets with copies of NRC's fire protection cri-teria, including Branch Technical Position (BTP) CMEB 9.5-1 and Appendix R to 10 CFR Part 50, and supplemented these documents with additional papers related to. fire brigade composition and training and electrical cable combustibility.
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| rir. Kubicki briefly summarized the American nuclear fire experience and regula-tory efforts and the philosophy for protecting safe shutdown systems. The Soviets expr s sed much interest in NRC's utilization of probabilistic risk I assessment (PRA) techniques in the decisionmaking process associated witF the
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| * Meeting scheduled for June 5-9, 1989, in Rockville, Maryland.
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| 4-i 1
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| | |
| provision of fire protection features at nuclear power plants. They, in turn, described briefly that in the design of all types of structures (including power plants) in the Soviet Union, PRA-type analyses must be conducted to deter-mine the level of fire protection required. Their goal is to design the build-ing in such a way that the probability of a f're fatality is less than 10 6 In addition, cost-benefit analytical techni @es are applied as well as strict technical standards. Af ter Mr. Kubicki explained NRC's process of granting approval for deviations from NRC's fire protection criteria, the Soviets stated that their standards are rigidly applied and only under rare instances are variances approved.
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| Mr. Kubicki discussed in some detail technical issues related to the protection -
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| of structural steel, cable penetrations through fire bar*.iers and partitions, fire fighting strategies and systems, and protection of cables. The Soviets responded to a number of Mr. Kubicki's questions regarding the protection of -
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| safety related systems by explaining that all safety-related cables are pro-tected by an "in house"-developed cable wrap system consisting of alternate layers of a fiberglass tape and a "flammastic-like" product referred to as "OnK" (OPK in English). This configuration has been fire tested and is re-portedly capable of withstanding a 3-hour fire without cable damage. In addi-tion, the cables are protected by automatic fire suppression systems. Contrary.
| |
| to Mr. Kubicki's expectations, the Soviets said that this protection scheme exists for all plants, including theh oldest plants. Mr. Kubicki inquired whether non-safety related circuits are powered off the same bus as safety-related circuits. The Soviets said that they were. The Soviets indicated that non-safety-related circuits are not protected as are safety ci rcuits. In this configuration Mr. Kubicki believes that there is a potential for loss of power to safety-related systems in a fire. Safety-related components are protected in the United States by divisional separation, relying principally upon spatial separation in lieu of complete fire barriers.
| |
| Mr. Herman (G. I. ) Smelkov presented a surnmary of a paper delineating a proba-bilistic approach to the analysis of cable fires. He indicated that half of all fires in $'viet plants involve cables. Most of these fires are started exter-nally, such as during hot work operations. Mr. Smelkov did not disclose how many of these fires had occurred.
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| Mr. Anatoly Shevchuk presented a paper on the use of halogenated hydrocarbons to mitigate the risks of hydrogen generation inside the containment. This approach, if implemented, would appear to violate the recently signed Montreal protocol on the production and use of halogenated hydrocarbons.
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| From Mr. Kubicki's perspective, there appears to be a degree of fragmentation in the regulatory authority for fire protection at Soviet power plants. Spe-cifically, separate organizations are responsible for fire safety during the .
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| design, construction, and operational phases of the life of a power plant.
| |
| There does not appear to be nuch communication and cooperation among these entities. This topic will be discussed at the next working group meeting in June 1989.
| |
| The working group agreed on a tentative agendum for the planned Soviet visit to the United States in June 1989. The Soviets' initial proposals for an agendum did not coincide with the protocol. In addition, they did not originally request 4-2 E _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - - - - - - - - - - - - - -
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| I
| |
| | |
| l to visit an American power plant. This led Mr. Kubicki to conclude that the Soviets are more interested in the research aspect of fire safety at nuclear power plants than in gaining insight into the practical aspects of fire protec-tion. Mr. Kubicki asked the Soviets to prepare a comprehensive presentation on their nuclear fire protection program, including recent fire experience.
| |
| For the U.S. reciprocal visit to the Soviet Union in October 1989, Mr. Kubicki proposes that a two person team from NRC conduct a typical Appendix R-type
| |
| - inspection of an older Soviet plant.
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| Appendix 11 comprises Soviet publications submitted to Mr. Kubicki.
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| 6 4
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| 9 4-3
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| | |
| t f
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| L f 5 PROTOCOL TOPIC 10: EROSION / CORROSION DESTRUCTION OF PIPING AND COMPONENTS i
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| The joint erosion / corrosion working group met in Moscow on Wednesday and Thurs-day, December 7 and 8, 1988, to review the working plan for future meetings in the following areas:
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| (a) Exchange of information on prediction methods for determining
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| .. . piping degradation through erosion and/or corrosion, including ,
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| primary causes, mathematical modeling, experimental verification, '
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| and prevention.
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| ^
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| (b) Exchange of information on methods for predicting the .ocation of erosion / corrosion degradation and its rate in piping sy:,tems and criteria for controlling continued degradation.
| |
| (c) Exchange of information in developing new corrosion-resistant steel for nuclear power plant piping and eouipment.
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| (d) Basic research on the mechanism of nodule corrosion and stress corresion cracking of zirconium alloys.
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| (e) Developing future water-chemistry processes and automated pro-grams for calculating corrosion, erosior products output, and radiol-ysis, taking into account physical and structural properties of reactors.
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| (f) Further studies of the mechanism and kinetics of electro-chemical and structural processes at the apex of maximum permissible defects which determine the corrosion and mechanical properties of the material.
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| (g) Devaloping automated systems of control of the corrosion con-dition of nuclear power plant equipment (corrosion monitoring).
| |
| Wednesday, December 7, 1988 Mr. Richardson and Mr. lat,oada met at KIAE with the Soviet members of the working
| |
| . group on erosion / corrosion. The meeting was co-chaired by Mr. I. A. Stepanov, Institute of Power Technology, and Mr. James Richardson, NRC. The Soviet par-ticipants are listed in Appendix 2. The area of interest for each participating organization is described in Appendix 9.
| |
| The Soviets made a large contribution to the subject of erosion / corrosion.
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| Dr. Tutnov spoke on microcracks and the conditions that cause such cracks to grow to material f ailure. Factors, such as length of time of stress, stress intensity, carbon inclusions as well as other inclusions, and morphology were discussed. He also gave other failure mechanisms for zirconium.
| |
| Mr. Stepanov summarized the Soviet work related to erosion / corrosion. This includes erosion / corrosion of piping, fittings and components, material selection 5-1
| |
| | |
| l to avoid environmental deterioration, inhibitor development, water treatment, and corrosion of zirconium alloys. The Soviets have also worked on development and application of low and high alloy steels for piping to mitigate stress cor-rosion cracking. They have experienced stress corrosion cracking in stainless steel as has the rest of the world. The Soviets are concerned about nodular corrosion of zirconium alloys for pressure tubes and would like to exchange information on fundamental studies, kinetics of nodular corrosion, methods of accelerating testing, and zirconium-to-steel joints. They have studied kinetics of crack development in reactor environment and corrosion at crack tips. Their approach to estiraating allowable defects is similar to the approach taken by the American Society of Mechanica~1 Engineers (ASME).
| |
| D . Nigmatulin described work at the All-Union Research Institute of Nuclear Power Plant Operations on erosion / corrosion of pearlitic steels in secondary-side coolant of VVERs, water droplets, and pure steam. They have also studied -
| |
| deposition of radiolytic products in the primary circuit. They have developed a computer model to predict erosion / corrosion, but need engineering test results to validate the model. The program is written in BASIC and has impressive graphics.
| |
| Dr. G. A. Saltanov described Soviet work on the use of auines films to inhibit erosion / corrosion. He presented the U.S. team with a Tussion book titled
| |
| " Hydrodynamics and Heat and Mass Transfer When Adding Surf.actant" by G. A.
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| Filipov, G. A. Saltanov, and A. N. Kukushkin which was published in 1988. An English abstract of the book appears in Appendix 10.
| |
| Dr. Chabak spoke on water chemistry studies related to VVER reactors. He de-scribed additiv'es, controls, corrosion products, and reactor experiences in piping and steam generators. The Soviets use hydrazine initially to scavenge oxygen from their reactor systems and maintain an oxygen content of less than 5 micrograms per kilogram of water. They maintain a high pH with ammonium addi-tives. Cladding is protected by water chemistry control to less than 0.02 per-cent fuel failure. The condensate has 100 percent polishing with electromag-netic filters and parallel mixed-bed filtei.5. The Soviets use stainless steel tubes in their horizontal steam generators instead of nickel-based alloys, such as are used in the United States. The Soviet experience with these tubes exceeds their specification of less than 0.08 percent plugged tubes. Failures are not a function of time but result from chloride in leakage. In August 1983, 20 tubes failed in the Novovoronezh Unit 3 steam generator at spacer locations.
| |
| Mr. Richardson updated the status of erosion / corrosion experience in the United States; he presented results of the latest inspection of thinned pipes at the -
| |
| Surry nuclear power plant.
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| Thursday, December 8, 1988 .
| |
| On Thursday, December 8, the teams met at the All-Union'Research Institute for Nuclear Power Plant Operations with Soviet scientists from that organization.
| |
| Mr. V. Koslov, head of the Electriral Engineering Department at the All-Union Research Institute, and Dr. Ross co-chaired the meeting.' The Soviets presented the following topics:
| |
| I -
| |
| Calculation of Thermal Hydraulics for Core Melt (see page 3-13)
| |
| Erosion / Corrosion of Piping 5-2 w_-__-_____________. . - . __
| |
| | |
| Problems of Erw ion-Corrosion Operat'an-wise Operational Reliability of Safety Sy tems Mr. J. Richardson summarized the erosion / corrosion problem in U.S. plants.
| |
| Other Americans at this meeting were Mr. W. Hodges, Office of Nuclear Reactor Regulation, and Mr. A. Taboada, Office of Nuclear Regulatory 'esearch.
| |
| Mr. Koslov,_Mr. N. Gareshkin, and Mr. M. Saltanov presented information on erosion / corrosion in the Soviet Union. These problems occur in the secondary circuit of Soviet PWR plants in steel pipe described as "pearlitic" :;teels.
| |
| . These steels have approximately 0.16 percent chromium and 0.5 percent mangan-ese. Erosion problems in austenitic steels have been purely mechanical. In the 1000-MW VVER reactors, erosion / corrosion occurred under such conditions as 225*C temperature, water chemistry with less than 10 ppb of oxygen, and pH of 8 to 8.5. Hydrazine is added to the coolant.
| |
| The Soviets collect wall-thickness data and analyze them. They divide these data into three categories: >3 mm, I to 3 mm, and <1 mm. Most material crodes in the 1-to-3-mm range. Erosion / corrosion has been found in pipes with wet steam, single phase water, and boiling coolant. The maximum operating time of~
| |
| reactors examined is 8 years.
| |
| The institute staff has determined the principal locations or flaws where erosion / corrosion rates have been exceeded in walls that are 3 mm thick. The staff has recommended permanent monitoring in the areas where these rates were observed. The program developed includes the use of permanently installed (on the pipe wall) ultrasonic testing transducers, check in the steady-state mode and during transients, and every two weeks. The results of these procedures have not yet been validated.
| |
| Test facilities have been designed at the institute to study erosion / corrosion using simulated water conditions, pipe materials, and thermodynamics. All the work is empirical. One test rig uses electrochemical measurements to establish erosion / corrosion rates on small flat specimens of materials tested. Institute staff members offered to test American materials on their rig. The institute also has rigs for lifetime testing. The Soviets have tested seven different grades of steel, including pearlitic, low alloy (Ni), and 18-10 stainless steels.
| |
| They have also tested elbows and straight pipe, both in and out of headers.
| |
| They have accelerated erosion / corrosion rates by increasing flow rates by a factor of 4.5. At present these rigs use pipes 36 mm 0.D. and 28 mm I.D. Insti-tute staff members predict long-time erosion / corrosion from 200-500-hour tests using a model that has been computerized.
| |
| The Soviets discussed improving reliability of valve systems. Except for the main isolation valves, this subject was not given any special attention until the 1970s. Several valves have failed because they were manufactured improperly.
| |
| From 10 percent to 40 percent of reactor shutdowns have been caused by valve failures. More than 30 percent of all repairs done are repairs on valves. The Soviets have developed regulatory documents that now place requirements on all manufacturers. These include general specifications, environment specifications, I
| |
| reliability requirements, construction requirements, and quality assurance 1
| |
| l 5-3 L
| |
| | |
| requirements. All valves are inspected by rs iresentatives of the institute.
| |
| The institute also has test facilities at tht mal power plants outside of Moscow where safety / relief valves-and other large valves are tested. Power plant valve data have been evaluated and compared with French and German valve data. Major weak points of valves include insufficient sealing of valve glands, reliability of control valves (drives are insufficient), erosion / corrosion wear in regula-tors controlling steam level, low pressure heater dump regulators to dump tank, and derators. A program to avoid failures has been developed that uses contin-uous monitoring of parameters. The Soviets have particular problems with check valve jamming.
| |
| Dr. Ross indicated that the NRC is not happy with the high failure rates of -
| |
| U.S. valves and that NRC issued instructions to reduce unreliability of valves.
| |
| Research in this area includes dynamic tasting, signature analysis of motor-operated valves, transient and electrical current measurements during operat- .
| |
| ing and closing, cooperative valve testing with the Federal Republic of Germany, and surveillance testing of most critical items.
| |
| Future Meetings !
| |
| The two groups agreed that three workshops would be held (June 1989 in the U.S.,
| |
| May'1990 in the U.S.S.R., and May 1991 in the U.S.). Topics for the 1989 work-shop will cover power plant experience, engineering considerations, and research activities. The 1990 workshop will cover the control of water chemistry. The 1991 workshop will cover corrosion-resistant materials. In addition, the group agrees to exchange information in the areas of:
| |
| (1) Methods for predicting steam generator degradation (2) Development of new, alternative materials for piping and steam generator tubes (3) Mechanisms of corrosion at the crack tip of material defects (4) Methods for determining piping degradation due to stress corrosion cracking (5) Erosion / corrosion and corrosion cracking of zirconium alloys (6) Water chemistry (7) Development of on-line methods and systems for monitoring stress corrosion cracking .
| |
| The two groups also agreed to exchange visits, if appropriate and approved, to research facilities to gain further insights and information relating to erosion / corrosion of piping.
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| l 5-4
| |
| | |
| I 6 OTHER TOPICS Forecasts of Soviet Nuclear Capacity and Foreign Sales i On Tuesday, December 6, the Kurchatov Institute hosted a luncheon banquet at
| |
| ! the Sovietskaya Hotel for the NRC and DOE teams. Some interesting discussions took place during lunch. When Dr. Salomon questioned him on the forecast of
| |
| . nuclear power in the year 2000, Dr. Pnnomarov-Stepnoi said that his best esti-mate is 100 GWe, but that there is no agreement among knowledgeable people in the Soviet Union. (Note: The official Soviet 5 year plan low projection is
| |
| . 150 GWe and current installed capacity is 38 GWe. R. Caron Cooper and A. T.
| |
| Birman presented estimates at the Slavic Studies Conference in Honolulu in Novem-ber 1988 of 110 GWe and 100 GWe, respectively. These forecasts incorporate proposed conservation policies by the Soviet government.) Foreign sales after the year 2000 were estimated to be 50 GWe. Two 1000-MWe VVERs were jud sold on a turnkey (fixed price) basis to India using ocean cooling. Dr. Ponomarev-Stepnoi had accompanied General Secretary Gorbachev on the trip to India to conclude the sale. In addition, other foreign sales are contemplated to COMECON countries.
| |
| Cost-Sharing Procram on Chernobyl After lunch, Dr. Gagarinskyi discussed Chernobyl Unit 4--the accident, evacua-tion, fire fighting, and the consequences of mitigation and cleanup. He had spoken on the same subject in November 1988 at the American Nuclear Society meeting in Washington, D.C. He was especially interested in the long-term consequence stuM es that included radioactive contamination, monitoring, trans-port of radionu,. ides, and radioecological consequences. Also, fuel behavior,
| |
| " entombment" design and severe-accident modeling were important. In this regard, the long-term integrity of the entombment is important.
| |
| He noted that no potential dangers have been observed. Forty ports were drilled into the reactor entombment. Some holes went into the reactor core chamber.
| |
| Diagnostic investigation showed that the temperature there is 200 C and that the fuel has been dispersed. Neutron flux measurements have been conducted. In 2 to 3 years, all the necessary data should be obtained on the location of the fuel.
| |
| He would like cooperative investigations, especially for the radioecological analyses and contamination of the biosphere.
| |
| The group was told that the program at Chernnbyl will be discussed at a major meeting in the Soviet Union in 1989.
| |
| In the United States, the NRC uses the CRAC environmental transport code in i the Soviet Union, a modification of CRAC called CREDIT is used; the names gre similar but the codes are not.
| |
| To the question: "Can this code be validated at Chernobyl?" the answer was given that there are other codes, such as the French code called RESAC. A lot 6-1
| |
| | |
| i' 1
| |
| o depends on the source term. There is the code LACE which was used before ACE.
| |
| In any event, this is an area to be worked on in the future.
| |
| Soviet Academy of Sciences Nuclear Safety Institute On Wednesday, December 7, Dr. Salomon asked Drs. Lunin and Gutsalov about the new Nuclear Safety Institute that reportedly was to be created at Minsk, as the i
| |
| Americans had heard from Dr. Ponamarev-Stepnoi at the Araerican Nuclear Society (ANS) meeting in November 1988. The Soviets replied that the institute would be moved to Moscow within the Soviet Academy of Sciences. The laboratories i would stay at Minsk. Currently, the activity is only in the planning sta2e. l Dr. Bolshov of KIAE will be the First Deputy Director. The organizing director -
| |
| is Dr. Velikhov of KIAE and also of the Soviet Academy of Sciences. Its basic task would be to act as an independent expert organization to investigate all fundamental safety questions. It is impossible to say now what the institute .
| |
| will become. There are presently only ten people assigned to the organization; they have been given two old buildings that need repair and remodeling.
| |
| The budget for the new institute is being determined. By 1989, 150 people should be employed at the institute, and a few million rubles a year 9 Juld be budgeted for its activities.
| |
| The Soviet Academy of Sciences, like the State Committee on the Supervision of Nuclear Power Safety (GAEN), is suffering f' t.:n a lack of public credibility.
| |
| The creation of the institute is an attempt to improve the Sov et Academy's 4
| |
| credibility. The institute is not like NY 's Advisory Committee on Reactor Safeguards. The institute will work clo< ,1y with KIAE.
| |
| For many years, A. K. Krasnan headed the Minsk laboratory, which is part of the Byelorussian Academy of Sciences. The Minsk laboratory was responsible for determining cooling with regard to Chernobyl. That work is no longer useful.
| |
| Dr. Bolshov is trying to get the Minsk laboratory to cooperate in new studies.
| |
| The Minsk laboratory is a source of good mathematicians and theoreticians who can work with KIAE. The specific mission of the new Nuclear Safety Institute will not be defined until the new director is appointed.
| |
| Chairman Zech's Trip to the Soviet Union On Wednesday, December 7, Dr. Salomon presented to Dr. Asmolov and Dr. Gutsalov copies of the NRC publication, " News, Reviews and Comment," September 1988, Vol. 5, No. 3, which featured Chairman Zech's trip to the U.S.S.R. in August -
| |
| 1988. The Soviets were keenly interested in the publication.
| |
| Soviet Sale of Annealing Technology 5
| |
| On December 9, Dr. Ponomarev-Stepnoi spoke with Dr. Ross about the Soviet sale of annealing technology to the United States. He was very enthusiastic about this. Dr. Ross responded that this was not an NRC responsibility since NRC does not import such technologies. Dr. Ross suggested that Mr. Byron Lee, Jr. ,
| |
| President of the Nuclear Management and Resources Council (NUMARC), who repre-sents the nuclear industry, was a better prospect. Dr. Ross promised to tell Mr. Lee about Dr. Ponomarev-Stepnoi's interest in selling annealing technology.
| |
| 6-2
| |
| | |
| Armenian Earthquake In addition, on December 9, Mr. Nikitin, GAEN, reported to Dr. Salomon that the intensity of the earthquake that occurred on December 7,1988 at 11:15 a.m. at the Armenian nuclear power plant was 5.6 on the Soviet scale. (Note: The top of the scale is 12.) The reactor is set to trip at 6, so it operated through-out the earthquake. It was reported that 30,000 people died and that 100,000 people were homeless. Dr. Salomon was amazed at Mr. Nikitin's candor.
| |
| 9 l
| |
| l -
| |
| 1.
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| 6-3 f 1
| |
| | |
| 7 PREPARATION OF THE MEMORANDUM 0F THE FIRST MEETING 0F THE WORKING GROUPS OF THE U.S./U.S.S.R. JOINT COORDINATING COMMITTEE FOR CIVILIAN NUCLEAR REACTOR SAFETY, DECEMBER 5-9, 1988*
| |
| Wdnesday, December 7 was taken up with discussions on the memorandum. Alexander Gavrishin, a research associate in the Nuclear Safety Division, KIAE, served as interpreter for these discussions. Both groups agreed that the main ideas would be first written down in an English text, and that the English version would be signed. After the session, the text could be translated into Russian. Dr.
| |
| Salomon was to prepare the memorandum, using the August 31, 1988 protocol as a
| |
| . model.
| |
| Some interesting points were brought out in the discussions on the preparation of the memorandum. The participants wandered from strict adherence to a discus-sion on the contents of the memorandum as they sparked each others interests.
| |
| Dr. Asmolov, head of the Soviet group, stated that toe memorandum should ncte that the JCCCNRS should become a vehicle for the exchange of information. Infor-mation about commercial and government programs related to safety should be ex-changed. For example, the Electric Power Research Institute (EPRI) has proposed a joint U.S./U.S.S.R. enterprise for radiation embrittlement and annealing.
| |
| Dr. Asmolov's view is that each nation can play a valuable role in the JCCCNRS, and that both nations can contribute important information to this joint effort.
| |
| l Many suggestions were made for areas of future cooperation:
| |
| l l Dr. Slesarev talked about new safety concepts. He suggested that a new kind of working group be created to study concepts and share information on a new gen-eration of nuclear power reactors, including the so-called " inherently safe" reactors.
| |
| Erosion and corrosion are important topics. Dr. Asmolov noted that American companies are interested in erosion / corrosion. Another concern is the steel and zirconium dissimilar metal weld. The Soviets have the RBMK pressure tube l experience. Dr. Asmolov expressed his view that the JCCCNRS should embody all l these subjects.
| |
| 1 i . Dr. Gagarinskyi spoke about joint programs regarding Chernobyl. A commercial l approach should be noted similar to the ACE progrm. One-nalf of the cost would be borne by the Soviets and the rest by the pz,rticipating countries.
| |
| Dr. Asmolov suggested that models and codes need to be verified by experiment.
| |
| The blind experiment idea suggested by Dr. Ross should be pursued: The Americans would describe the fuel / concrete interaction in general terms and the Soviets would calculate the interaction using the computer codes. Then, the United States would give the Soviets the experimental results so the Soviets could com-pare their calculations with experimental results.
| |
| *The memorandum is reproduced in Appendix 12.
| |
| 7-1 I
| |
| | |
| The2-D/3-DprojectinvolvingjointparticipationoftheUnitedStates, Japan, and Germany at a cost of $450 million ($150 million each), was discussed. -The Soviets said that they do not have the $150 million to contribute and would wait until the information was released; one to two years of waiting is estimated.
| |
| l Dr. Asmolov told the group that the Soviets have a thermal-hydraulic facility at Electrogorsk. He felt that the latest version of the RELAP code could be used for verifying experiments for transients and small breaks. Dr. Ross said that this was possible since about a dozen countries already use the code. Howe'.er, military and commercial authorities in the United States must review the use of the HELAP code before it can be exported. -
| |
| Dr. Asmolov said that for calculating meltdown the Soviets use CORCON and RASPLAV whereas the United States uses SURC-4 experiments. Dr. Asmolov said that it .
| |
| would be beneficial to compare the use of the codes.
| |
| Dr. Kovalevich asked if the NRC can provide information about the probability of vessel rupture. Dr. Ross answered that the United States would suggest in-formation that NRC could share and would tell the Soviets what information NRC would like to receive.
| |
| Both groups agreed that the memorandum need not list the documents that were exchanged. Documents that the NRC sent to the Soviet Union pursuant to the pro-tocol are listed in Appendix 3. The Soviet Union did not send documents to the NRC before the meeting.
| |
| The NRC group met at 9:30 p.m. on Wednesday, December 7, to review progress on the memorandum. The outstanding issues that remained to be incorporated in-cluded the exchange of safety documents regarding the South Texas Project and the Zaporozhe nuclear power plants and adding the Soviet input regarding a research planning document. Otherwise, the team was satisfied with the points made in the memorandum.
| |
| On Friday morning, December 9, Drs. Ross and Salomon went to KIAE to continue work on the memorandum. The original plan was for both of them to go to the American Embassy to report their progress to Mr. Phillip Otts, Acting Science Counselor, telex the unsigned final draf t of the memorandum to NRC headquar-ters, and pick up the South Texas Project FSAR to be exchanged with the Soviets.
| |
| However, because the GAEN microbus never arrived for them and it was impossible to hail a taxi, Drs. Ross and Salomon proceeded (without the FSAR) to KIAE on the bus GAEN provided for the DOE team. -
| |
| At KIAE, Drs. Ross and Salomon continued final editing of the memorandum. They submitted it to Ms. Olga Proshaina who began to type it. She types in English .
| |
| and uses an American Research Corporation personal computer that has a Canon printer. While this was being done, Drs. Ross and Salomon went by GAEN bus to the American Embassy. There, they transmitted to Mr. Phillip Otts, the unsigned final draft of the memorandum for telexing to NRC headquarters, and also gave him the answers to questions posed by a Soviet freelance journalist. They picked up the South Texas Project FSAR.
| |
| 7-2
| |
| | |
| l' l
| |
| l-L In the afternoon,.Dr. Ponomarev-Stepnoi' signed the memorandum of the.first meet-l' ing of the working groups for the Soviets. Dr. Ross signed for the Americans (see Appendix 12).
| |
| l l Some important points were made at the signing ceremony. This first' meeting was a good sign of improved cooperation between the United States and the Soviet Union. Such exchanges of information will help both countries. Finally, Dr. Ponomarov-Stepnoi said that maybe in the future environmentalists will participate in such meetings because environmentalists are very active in the Soviet Union.
| |
| 9 l
| |
| O e
| |
| i i
| |
| I I
| |
| 7-3 ,
| |
| i
| |
| _ _ _ _ = _ _ _ _ . _ _ _ . _ _ _ _ _ . _ _ ._ . _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ __ _ . _ _ _ _ _ _ _ _
| |
| _ . _ _____._j
| |
| | |
| b APPENDIX 1 U.S. Participants in the First Meeting of Working Groups of the Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS) i December 5-9, 1988
| |
| . Moscow Acting JCCCNRS Co-Chairman:
| |
| Derwood Ross Deputy Director, Office of Nuclear Regulatory Research, NRC Members M. Wayne Hodges Chief, Reactor Systems Branch, Division of Engineering and System Technology, Office of Nuclear Reactor Regulation, NRC Dennis Kubicki Fire Protection Engineer, Office of Nuclear Reactor Regulation, NRC Joseph Lewin Contract Interpreter for NRC James Richardson Assistant Director for Engineering, Division of Engineering and System Technology, Office of Nuclear Regulatory Research, NRC Stephen Salomc, Technical Analyst, Internatior.al Programs, Office of Governmental and Public Affairs, NRC Alfred Taboada Materials Engineering Branch, Office of Nuclear Regulctory Research, NRC e
| |
| Note: The Department of Energy participants are listed in Appendix 13.
| |
| 1 Appendix 1
| |
| _ _ _ _ _ - - _ _ _ l
| |
| | |
| i 1
| |
| APPENDIX 2 U.S.S.R. Participants in the First Meeting of Working Groups of the Joint Coordinating Committee for Civilian. Nuclear Reactor Safety (JCCCNRS)
| |
| December 5-9, 1988 Moscow JCCCNRS Co-Chairman:
| |
| ~
| |
| N. N. Ponomarev-Stepnoi2 First Deputy Director, I. V. Kurchatov Institute of Atomic Energy Members:
| |
| : 1. V. Kurchatov Institute of Atomic Energy (KIAE)
| |
| A. D. Amaev2 Laboratory Head R. V. Arutyunyan) Senior Researcher V. G. Asmolev2 Department Head A. F. Chabak2,3 Laboratory Head A. Yu. Gagarinskyi2 Deputy Director A. N Gavrishin2 Nuclear Safety Division Research Associate A. V. Khrustalev2 Senior Researcher A. Ya. Kramerovl Laboratory Head G. L. Lunin2 Department Head L. M. Luzanova2 Laboratory Head D.. A. Proshina2 Nuclear Safety Division Engineer E. P. Ryazantsev s Deputy Director L. V. Sergeeva 2 ,3 Senior Scientific Associate
| |
| : 1. S. Slesarev2 Department Deputy Director V. K. Sukhoruchkin2 Laboratory Head A. A. Tutnov 2 ,4 Laboratory Head Other Organizations
| |
| . V. N. Belouc 2 ,3 Head of Research and Development Laboratory, Institute of Power Engineering
| |
| ^
| |
| D. L. Greenevich2,3 Research and Development Engineer, Institute of Power Engineering L. P. Kabanov2 Professor, Moscow Institute of Energy L. N. Karahanyan 2 ,3 Head of Research and Development Group, Institute of Power Engineering B. A. Kuvshinnikov2 Deputy Director, Department of Scientific and Technical Cooperation, State Committee for the Utilization of Atomic Energy (GKAE) 1 Appendix 2
| |
| | |
| Other Organizations (Continued)
| |
| V. B. Mayorov 2 ,3 Engineer in Research and Development Group, Institute of Power Engineering E. V. Kuznetsov2 ,3 Department Head, Central Research Division, Institute of Heavy Engineering A. A. Nazarov 2 ,3 Laboratory Head, Central Research Institute of Struc-tural Materials (Prometheus) .
| |
| B. I. Nigmatulin 2
| |
| ,4 Department Head, All-Union Research Institute for Nuclear Power Plant Operations G. A. Saltanov2 Deputy Director, A'.1-Union Research Institute of f Nuclear Engineering M. G. Soltanov 2 ,3 Scientific Associate, All-Union Research Institute for Nuclear Power Plant Operations I. A. Stepanov 1 ,4 Head of Research and Design Department, Institute of Power Technology All-Union Research Institute for Fire Protection (Working Group for Fire Safety)
| |
| : 1. F. Poedintzev2 Division Head Yu. N. Shebeko2 Division Head A. P. Shevchuk2 Laboratory Head G. I. Smelkov2 Division Head l
| |
| l l
| |
| 1 Note: See Appendix 13 for a list of the Soviets who participated in the DOE seminar on the VVER reactor.
| |
| 2 Participated in December 5, 1988 meeting and thereafter.
| |
| 2 Participated subsequent to December 5, 1988 meeting.
| |
| 3 Expert in Working Group 10 (Erosion / Corrosion).
| |
| 4 Member of Working Group 10 (Erosion / Corrosion).
| |
| 2 Appendix 2
| |
| | |
| APPENDIX 3 Background Documents Sent to Soviets by NRC in November 1988 Safety Research Seminar Documents:
| |
| : 1. Nuclear Safety Research Mission Area
| |
| . 2. NUREG-1144, Rev. 1, " Nuclear Plant Aging Research (NPAR) Program Plan"
| |
| : 3. NUREG-1147, Rev.1, " Seismic Safety Research Program Plan"
| |
| : 4. NUREG-1222, " Piping Research Program Plan
| |
| : 5. NUREG-1245, Vol. 1, " Radioactive Waste Manages.ent Research Program Plan l
| |
| for High-Level Waste - 1987"
| |
| : 6. NUREG-1252, " Nuclear Power Plant Performance Research Program Plan"
| |
| : 7. NUREG-1264, " Containment Integrity Research Program Plan" Erosion / Corrosion Worki,,nq Group Documents:
| |
| : 1. Proposed Items for Discussions (Preliminary)
| |
| : 2. NRC Bulletin No. 87-01: " Thinning of Pipe Walls in Nuclear Power Plants"
| |
| : 3. NRC Information Notice No. 88-17: " Summary of Responses to NRC Bulletin 87-01"
| |
| : 4. A 2 page explanation prepared by NRC staff of contents of Electric Power Research Institute Users Manual NSAC-112, "CHEC" (Chexal-Horiwitz Erosion-Corrosion)
| |
| : 5. NUREG/CR-5131, " Preliminary Review of Mass Transfer an<. Flcw Visualization Studies and Techniques Relevant to the Study of Erosion / Corrosion of Reactor Piping Systems"
| |
| : 6. NUREG/CR-5149, " Erosion / Corrosion of PWR feedwater Piping; Survey of Experience, De:;ign, Water Chemistry and Materials"
| |
| : 7. NUREG/CR-5156, " Review of Erosion / Corrosion in Single-Phase flows" 1 Appendix 3
| |
| | |
| Lf APPENDIX 4 Agenda of the Meetings in the Kurchatov Institute of Atomic Energy (KIAE)
| |
| - December 5., Monday 10.00-17.00 Safety Research Overview Room A a
| |
| . 13.00-14.30 Lunch Break (Canteen)
| |
| December 6, Tuesday 10.00-17.00 Safety Research Overview Room B 10.00-17.00 DOE Seminar on VVER Rooms A, C, 315 13.00-14.30 Lunch Break (Sovietskaya Hotel) _
| |
| December 7, Wednesday 10.00-17.00 DOE Seminar on VVER Rooms A, C, 315 10.00-17.00 Erosion / Corrosion WG Room B 10.00-17.00 Visits to KIAE Facilitics:
| |
| "ASTRA", F-1, MP, T-10, T-15, Kurchatov Museum 13.00-14.30 Lunch Break (Canteen)
| |
| December 8. Thursday ,',
| |
| 10.00-17.00 DOE Seminar on VVER Rooms A, C, 315 '
| |
| 10.00-17.00 Fire Protection Meeting Room B
| |
| ~
| |
| 10.00-17.00 Visit to All-Union Research Institute for Nuclear Power Plant Operations 13.00-14.30 Lunch Break (Canteen) i l
| |
| December 9, Friday 10.00-13.00 Final Meeting of DOE Seminar on VVER Rooms A, C, 315
| |
| ^
| |
| 10.00-13.00 Discussian of the Memorendum Room B 13.00-14.30 Lunch Bfeak (Canteen) 14 30-17.00 Signing of the Memorandum Rooms A, B Note: Rooms A, B, and C are located in the cottage; Room 315 is located in ,.
| |
| ti.a main building of the institu+.e.
| |
| 1 Appendix 4 i
| |
| r
| |
| | |
| APPENDIX 5
| |
| " Nuclear Power Reactors of the New Generation" N. Ponomarov-Stepnoi and I. Slesarev (Kurchatov Institute of Atomic Energy)
| |
| From Proceedings of t.he International Conference on Reactor Physics, September 18-22, 1938, Jackson Hole, Wyoming Appendix 5
| |
| | |
| It is ad:dited t ut tht. benc f act ors which c a termine the sta-tus of nuclear power as an important link in the national econoqy currehtly and in the future a.e the econceic competitiveness in cot.
| |
| parison dth the alternative energy sources, expansion of the nucle-ar energy sou:eq application (application sphere) and possibility to organize fuel scif-supply (fuel supply). The above factors can be con-sidered as I W ng the idea of the efficiency of the nuclear energy plants. The most important objectiva in dert. loping the new 6enerati-
| |
| ~
| |
| on nuclear reactors f.s the radical increase of the efficiency i.e.
| |
| rc,atization cf the principle step in solving even if one of the abo-ve problems (nat.trally not at the cost of retreat in any of the re-lating probler.s). These reactors of the new ceneration could be con-ventionally called effective cuss.
| |
| It would seem that usin6 the economical criteria these dif fe-ring factors could be reduced to a common deno=inator; however at present no reliable quantitative reintions can be established bet-ween these factors because of the lack of experience. The relative importance of the components of this triad has been subject to sig-nificant changes within the short history of nuclear power (NP) formation and development.
| |
| At the initial stage when NF was so small-scele that there was no doubt that it could be unlimitedly supplied with fuel and the safety for the design operatien modes was considered to be a solved problem, attainment of the eceaomic co=petitiveness was the permit for NP to be introduced to the natienal econocy. The most suitable
| |
| ~
| |
| final product fer NP has proved to be electricity. Use of the con-ventional energy conversion methods, the possibility of NTP siting at large distances from densely populated c.reas have made it suc-cessful to apply the nuclear source et the power plant.
| |
| S-2 1 Appendix 5 E__
| |
| | |
| Search f or other fields of El applici. tier., in addition to elec-tricity generation, was stimulated not only by intellectual ; cur $ o.=1-ty of scientists but also necessitated by rapidly changing situation in the oil and gas market. Auong these fields the moet power consu-ming are the industrial and residential distr $ct heating and encrgy supply of industrial technologies including, in perspective, the te-chnology of synfuel production, pealication of these trend , elens with solut$ on of the competitiveness problems, necessitates search for new designs of the nuclear source permitting its optical comp 11- '
| |
| ance with the technological processes and increase of the safety re-quired by the necessity to site the energy source as close to the censumer as possible.
| |
| Extensica of the scales of NP introduction required a more atten-tive attitude to the problem of fuel supply, necessity has arisen to caster all ec ponents of the fuel cycle, to develop more effecti-ve fuel utilization in the current thermal reactors and to design the breeders with high fuel reprodu,ction characteristics. Possible deter-f oration of the economic characteristics of the reactors resulting from increased fuel reproduction must be compensated by improvement of the fuel cycle characteristics and the total positive effect may manifest itself with increased rates of NP development on the back-ground of continuous rise of the conventional ft el cost.
| |
| Currently the N? safety problem has become the focus of atten-tion.The recent severe accidents have demonstrated the real NP dan-cer which in earlier probabilistic assessments see:ed to be remote
| |
| ~
| |
| and impossible. They have made the public attitude to the need of intensive N? develop =ent sceptical and even hateful. It has become clear that solutien of the safety problem exceeds the possibili-t$ es of a sincie senber of the werid ce=nunity and the necessity has matured to este.blish the internati< nal regice of safe NP de-5-3 t
| |
| 2 Appendix 5
| |
| | |
| velopment. "he do ;nant of the current st'aSe is chreful considera: ice and realicction of s revitalized concept of . mate h7 6evelopment. Ho-wever when workinE over new br'.vanced nucacar systems it would be un-wise to concentrate only on the stfety probles. It is evident that in the limiting case the e.bsolutely safe r.uclear syste= is the kivi, a non-flying tird.
| |
| - The new nuclear reacters thrule meet most eff ectively demands of the mankind for energy, while being zir.imum hazardous for its well-being. ,
| |
| Therefore before proceecin5 to discussion of the specific featt-res of the current approach to the h7 saf.ety problem, let us consider in more cethil the relating problems, those to whose soludon, as :t appears, must be directed the efforts of tbc cesigners of the advan-ced reactor plants. .
| |
| 7uel supp1v ef nuclear power At the present stage the econo =ic co=petitiveness of nuclear po-wer is deter =ined by a relatively low cost of uranium. However scar-city of inexpensive uranium resources may lead, within a few decades, to the rise in the cost of energy ' produced for nuclear power based only on ther:a1 reactors (TR). It is also'known that h7 has a unique ability of fuel reproduction. Therefore it :ls appropriate to raise .l the question of its cc p;ete self-fuelling. The presence of the bre-eders in the N? structure may slow down or even stop the above-men-tiened rise of egend:Liures for energy production. It follows'from the f oregoing that the funda ental problem is dete.~ination of the possibility and conditions for construction of economically-justifi-ed se.t ?-fuel".ing nuclear power. Solution of this problem cay be of importance bete f or chcice of particult. pcwer plants rund fer all the h7 strue: re.
| |
| S.4 3 Appendix 5 j l
| |
| | |
| At the Iresect stage cf }& development are being solved such s
| |
| problema as increces of the co= petit 2veness potential by lowering the coPte of TR end their fuel cycles, reduction of fuel consu=ption cha-racterintics for preventing increase in the fuel cost.
| |
| There are cany ways for realization of this goal:
| |
| - increase of the fuel burnup in TR operating in the open fuel cycle; ,
| |
| - closing the TR fuel cycle, which leads to appreciable reducti-on cf the fuel rike-up q (e.g., for LWR from qcit 235 U/cw(e). year .
| |
| to O 6./GW(e).yeart
| |
| - increase of the breeding ratio of the thermal reactors and, au a conse:;nence, further reduction of q in tu. closed fuel cycle.
| |
| This may be reached, for cr.c=ple, on account of the " shift" of the neutron spectrum in TR. For the light-water TR this effect is ob-served in transition to the closely-spaced fuel element arrays (RCVS reactors), with the breeding ratio increased up to 0.7-0 9 Even more radical seems to be the method based on significant reduc-tion in the coolant density (for exa=ple, transition to the steam or steam-water cooling) with reaching BR7 1 in the fast-resonance neu-tron spectrum /1-2/.
| |
| An icportant characteristic of the breeder is the breeding po-tential by which is =eant the level of excess fuel production 2or BR with allowance for the breed.w fuel consumption for the initial lon- ,
| |
| ding. If cue proceeds from th prerequisite that the breeders will .
| |
| ]
| |
| I re=ain core mpensive than TR by more than 20-30% then they need to !
| |
| I have, as sisted above, a high level of excess fuel production, par- -
| |
| j ticularly, under the 'cenditions of a multicc=ponent IG structure, i
| |
| At the p. resent the =ethod for essential i=prevement of r and ]
| |
| BR in the breeders have been developed. The f ollowing =ethods see=
| |
| to be realistic: .
| |
| j S5 )
| |
| l 4 Appendix 5 l
| |
| 1
| |
| | |
| L'
| |
| - transition. co the dense fuels (metal alloys, carbide, nitride
| |
| .. i fuels etc.). This trend may recuire-prolonged experimental verifica-tions of the operability of the new fuel elements at high therms 1 los.de, fauer.ces S.nd burnups which are characteristic of the power breeders;
| |
| - use of heterogeneous oxide-metallic and other breeder core.
| |
| compositions ensuring " sparing" Soed and burnup conditions for the depleted uranium-containing compon:nt. Such compositions little yielding to the." homogeneous" dense-fuel breeders in the breeding le-vel and even in tbc thermobydraulic characteristics can per=1t the BR values to be increesed up to 1 5-1.6 /3/.
| |
| It follows from the above discussion about the role of the bre-eders that the generalized 2ndex of their ."qualityn is their abili-ty to stop the growth of expenditures for' energy production in NP and keep them at as low level as possible. Choice of the criterion for . comparison of the breeder qualities is discussed in /4/.
| |
| The analysis /4/ shows that:
| |
| - increased BR makes the breeder more profitable at the earli-er stage of NP development;
| |
| - sensitivity of expenditures to relative changes in production of excess plutonium r by the breeder and the thermal reactor make-up q is nearly the sa=e, which indicates that the measures directed to improvement of the characteristics of the TR and breeder fuc1 cyc-
| |
| ~ les are equally importari t.
| |
| The opinien'that the ER value is not as important for the bre-efers ca the level of non-fuel ec=ponent expenditures is only correct for low fuel ec penent expenditures when the breeders themselves are still net i:pertant for the nuclear power.
| |
| The solution of the proble of eceplete fuel self-supply depends !
| |
| 5-6 5 Appendix 5 1
| |
| _m._.-_-.___m.______ __m
| |
| | |
| i
| |
| . no; only on the quality of the thermal and fr.n reactors but also on the NP stnicture and possible abarc cf breeders there.
| |
| Important seems to be The problem of c.pansion of the breeder application sphere, study of the possibility of their application not l
| |
| only for electricity generation but also for thermal energy producti-on, which, in turn, may require radical measures for enhancement of their safety. Otherwise in the conditions of the multicomponent po- ,
| |
| wer engineering the requirement to the breeding level and need of a d.rastic reduction of the fuel make-up of the rest of the reactors wo- ,
| |
| uld be too high.
| |
| Improve =ent of the breeder designs is one of the i=portant prob-lems in reactor building. De same cen-be said about the reactors of other types: their improvement will permit the economic efficiency of all NP to be 1.ncreased. The most urgent NP tasks are the followings
| |
| - dev'lopnent e and com=ercial production of advanced 7T. with the improved fuel characteristics and fuel recycle, for the purpose of maximum elongation of the period of relative inexpensive uranium fuel and low expenditures for energy production, " inexpensive a breeders based on the TR . technology 1
| |
| - development of the breeders with a high fuel conversion (E3-1 5) and. the lowest pessible cost of their construction;
| |
| . development of the " safe" breeder concepts, which will per-mit the sphere of their application in NP to be expanded.
| |
| With the fuel problem solved, NP, using the efficient breeders, ,
| |
| will make it possible to use a wide spectrum of other reactors ha-ving specific qualities, such as the high temperature reactors, .
| |
| without imposing stringent requirements on fuel consu=ption.
| |
| S7 6 Appendix 5
| |
| | |
| yxonnet cc of the enhere o! nuc5 ear nower ann 16::ution.
| |
| The possibility of increasing the share of nuclear electrici-ty up to 50-60% of the whole ciectricity generation is of no doubt for a number of reasons such as gained experience of L7P construction and operation, availability of high-energy systeca, absence of direct contact d th the concurer etc. However, such a scale' of development recuires to take into acccunt so=e additional conditions inclucint; use of the nuclear power plents for ecvcring the alternatir.g part of the load curve and the possibility of heat and electrhity co-genera-tion for a part of the power plants. The estimtes show that af ter the year 200030-40% of hT?s are expected to be operated .in the ma-n0euvring regime. Use of energy-accu =ulatinC systems would per.di this fraction to be reduced to 20-25%. In the perspective structure of the nuclear electricity the base fraction must be covered by the fast creeders and the alternating part of the load curve could be covered by the ther=al reactors, which provides favourable conditions for
| |
| ' fuel supply. Such an approach will require improvement of the exis-ting reactor plants in order to ensure their reliability and, prima-hily, the fuel element stability at alternating powers. The need n!-
| |
| so ariees in the additional operative reactivity margin, which will lead to increased fuel consumption. These factors as well as reduc-tion of the installed power utilization factor will result in the kiee of the specific expenses for generated electricity. All this
| |
| ~
| |
| bakes it necessary to improve the reactor plants for preration in the semipeak regice of daily and weekly load changes.
| |
| Another approach previding better conditions for the nuclear source is develcynent of the accum:lators for energy produced by h??
| |
| in the period of reduction in the energy oystem loads.
| |
| S-8 7 Appendix 5
| |
| | |
| l Today and dr the nearest futtu o a significant place in the con-1 sumption part of the fuel-cnergy balence (FEh> of this country is occu '
| |
| . J pied by energy cer.su=ption of industries and t esidential district he-ating. The direct (in the form of fuel) and secondary (in the form of conversed energy cerr$ crs: stea=, e) .ctricity, hot water) consump, tion of the fuel e.nd energy resources by the industrial technology requires more than 60% of their production in the country. .
| |
| In determination of possible development scales of various types of nuclear heat sources (IMS) the etructure of heat nreduction must ,
| |
| be taken into account both by the type of tne ener;;y carrier (steam,,
| |
| hat water, high-temperature coolant) ant. by its temperature poten -
| |
| tial.
| |
| The analysis reveals that about 50% of the therrtc1 energy is co-nsumed at a te=perature below. 400-45000 and about 50% - at higher temperatures. Therefore, IMS =ust ensure the possibility for heat generation over a very wide range of te=peratures, up to 100000 (abo-ut 80% of all heat consumption). An important condition is the ne-cessity of developing IGS of various purposes for production of energy carriers of various types (steam, hot water etc.). Tht: ta-ble presents the prognoeis of the change in the USSR district hea-ting structure.
| |
| s TABLE Predicted changes in the USSR structure of heat supply, %
| |
| Index 1980 Perspective
| |
| , l' Bot water output 28 30 Sten: generation 28 40 H16h-temperature procces heat production 44 30 T ot al 100 100
| |
| . 5-9 8 Appendix 5 i
| |
| | |
| At present the following NHS are considered as promir:. E ,o;cre for hot water and steam production: NPP with uncontrolled; Ites.m ex-tractions, itueles.r co-gencration plants (NOGP). nuclear re.zidential district beating plants (N DHP), nuclear industry d.istrict heating plants (NIDHP), nuclear long-distance heating plants CIDIP).
| |
| Heating of the NPP itself, site, personnel settleren etc. can be provided by iteans cf uncontrolled steam extractions frcn the plant turbiue unit. The thermal load for these purposes from, fr example.
| |
| SP/ with VVER-1000 may amount to 500 p(th) (per one unit). Under ,
| |
| favou able conditions (intence heat consumers nearby, sufficiently snooth load curves of RTP etc.) these reasures can be significantly effective if the scale c.f NPP development is taken into account. On the other hand, this means' that the site for.NPP must be closen with ellowance for the possibility of heat supply for the nearby consu-ters. In the USSR rather extensive experience of utilizati:n of he-at extractions from the condensation NPPs has been gained. Por exam-ple, at the Beloyarsk, Novo-Voronezh, Kursk NPPs and othersone -
| |
| knd'two-circuit NPPs steam extractions from the plant turbines are used for heating the NPPs, sites and personnel settlements.
| |
| The nuclear co-generation ple.nts (NCGP) have the thernedynami-cal advantages over the plants 6enerating only beat or only electri-eity. The technoeconomical analyses have shown that it is ;rofita-ble to have the NCGP reactors maximum loaded during the yen at the
| |
| .~ cost of additional electricity generation in the condensing modt and reduce the NCGP thermal load during the varm seasons.
| |
| In some cases the problem of using the nuclear reactc s for industrial heating can be successfully solved by development of the $iCG? on the basis of the reactors with high initial pa s.=eters 5 10 9 Appendix 5
| |
| | |
| f of generated stett, such as tha fart,sodiuo (PS) and high-temperatu-re gee-cooled resciors (HTGF). In the caee of using an HTGR-based nu-ClePJ/ powcr unit and the Close g4r-turbine cycle at NGGE some additi-onal increase in the efficiency of the heat and electricity co-gene-ration (by 25-30%) becones possible. As the heat potential of recoved heat at the cooling systems of such NOGp 18 high enough, there can be used dry cooling towers. . .
| |
| The nuclear residen'.ial district heatins; plants (NF.DHP) and nu-clear industrial district heatirs plants (NIDHP) cea be considered as a sufficiently high-power (300-500 Ges1/hr) district heating sou-ree, when sited in highly-populated settlements. The capacity of the reactor units at such plants is 300-500 LT(th).
| |
| In principle, the heat-only plants (NIDHP) can also be used for cteam heating. Of course, the steam production at NIDHP will requi-re to increase the. pressure in the reactor vessel, which will make the reactor design more complicated and core expet.sive and at the stece tine will reduce the NIDHP safety. Since in practice the steam transportation for distances exceedin6 5-7 km de impossible, NIDEP must be sited near the steam consumers, which makes the safety re-quire =ents core stringent.
| |
| Complex b'est-sten: supply can be provided by t).e industry-he-ating co-generation plaats of enhanced safety with HTGR cf various unit powers as well as by the nuclear long-distance heating plants (NLDHF). The nuclear energotechselogical pir.nts (hT!?) are designed .
| |
| for co= plex best-and-e::ergy supply (high-temperature heat up to 1000*C) of industrial technological productions of various purposes. -
| |
| In some cases they can serve (particily) as the long-distance bes-ting plante. Under consideration are NETP for: nitrogen industry (pro-5-11 l
| |
| 10 Appendix 5
| |
| | |
| duction cf conversed fac, sient), 17. particular, for a- otia or ,
| |
| methano; production; ferroue metallurgy (production tnd heating of reducing gaees); pttrolent chetistry and petroleus'intustry (hect and steas supply for craching and petroleum refining); oil producti-on (co: plex stent, hot water and electricity supply of oil ficids);
| |
| synthette fuel production (hydrogen production from water, gaseous and liquid hydrocarten production from coal). Also NETF could be used in non-ferroue metallurgy, mining industry, production of con-struction cattrials etc. /5/.
| |
| Depending on the NETF purpose the characteristic capacities of the nuclear power units for such p3 ants are within the range '-3 GW(th). 4. Significant part of the industrial consu+ ca demand for capt. cities lying near the lower boundary of this rance. Theref ore the possibility is being considered of developing the multiunit NE7P and NCGP with the low-power (200-300 ET(th))A=odular reactors.
| |
| Siting NETP near large industrial heat consumers poses heighte-ned saf ety requirements, similar* to those of NDHP, upon these plants.
| |
| One of the reactor types, meeting the NETF requirements for temperatures and appropriate sa(ety, is HTGR.
| |
| The HTCR introduction requires use of an effective energy car-rier which would.per=1t the ther=al power produced by the power unit to be transformed and accu =ulated in a form suitable for the con-sumer. Along with the electricity and steam as well as the most ra-1 -
| |
| dical, ai ed at long-term objectives, processes of 5 ydrogen produc-tion frc the wster and further use of hydrogen e universal effec-
| |
| ~
| |
| tive energy carrier, attractive are the available proven technolo-l gies based en application of hydrogen-containing gases obtained from l the cata;ytic ccnversict of hydrocarbons (= ethane) using the heat l
| |
| l fro HTGR.
| |
| l l S-12 1
| |
| l 11 Appendix 5 l
| |
| l
| |
| )
| |
| | |
| l The ant;yc .s cf the strt.ctre cf -he the::-C. enerc* mnrkets of some mos developed in:urtrial cer.tries hLr 2 eveLled that in addi- )
| |
| tion te the conventional high-power nuclear plants it is nEcessary ]
| |
| to develop a new concept of a nucleer small end medr.s power enerc .
| |
| source of modular type, whose coer,ficient of installed po e utili-iation and safety could meet the requirements of continuous processes.
| |
| The power required for large industrial works can be prov$ded ,
| |
| by use of several (2-6) one-type modulcr unite idth e unit power of 200-300 W(th), which also permits the required thermal power margin ,
| |
| 'o be created.
| |
| Increasing interest of the pcwer-generating and industribi firms in low-power ree.cte,rs is mainly dictated by the desire to reduct: the significant financibi risk connected with construction of ihrge !;TPs.
| |
| The modular reactors are not only evoking ever growing inte-rest at the inland mark'ets of' the developed countries but, in the opinion of the IAEA specialists, this trend stimulates development of the nuclear power in ens 11 and developing countries.
| |
| Concarative estimates of power reactor efficiencies.
| |
| The future nuclear power is expected' to be multicomponent, i.e.
| |
| having a number of the reactor application e7heres such as electric 1-ty generation and various te=perature heat production. Of particular importance in solution of the fuel supply problem will be breeders primarily developed for the electricity generation sphere.
| |
| Therefore actual renain the questions on the criteria for compa- -
| |
| rison of different conceptual propositions.
| |
| As noted, the eccnonic indices can constitute the basis for de-velop=ent of the universal criterion if the confident qualitative estimaten of these indices can be obtained. The currently existing difficulties in cbtaining the uns:biguous answer must not prevent j development of this trend. A useful example of the qualitative 5-13 I
| |
| l 12 Appendix 5
| |
| | |
| coalyrie of the econo:ic ef ficiencies of dif ferent-type reactors is given below. _
| |
| Return to the multico=ponent NP structure model adopted ehrlier.
| |
| pt a part of it be occupied by the base electricity production ( A)
| |
| ,rith the thermal reactors, the mort inexpensive ones, the rest of it ye occupied by heat production with the reactors for energotechnology
| |
| . '(3), low-te=perature heat production (C), industrial district hea-Sing (D) etc. -
| |
| - ' The reduced annual expenditures for energy production of the i- ,
| |
| type reactor (i n A,B,C,...) can be written as 4 = Ki " jn: Cv ,
| |
| 2 were C, is the expenditures f or production' of the uranium mass unit; gg is the annual make-up of the reactor with uranium; kg is the re-Auced capital expenditures including those for organization of the fuel (open or closed) cycle and initial fuel load.
| |
| Nuclear power development began from the thermal reactors (TR) of group A, because they were the most inexpensive ones and more prepa-red for realization. As noted the breeders proved to be much more expensive than it had been expected and could not bis used for in:me-tiate replacement of TRg . However with increase of the uranium cost
| |
| ! bey can be introduced into nuclear power. Appearance of the breeder (ts) la economically justified if the expenditures for energy production l
| |
| by the breeder and TR are equal
| |
| *t,C,u y
| |
| ,- EA + $O
| |
| * dna there t is production of excess second%7 nuclear fuel (here the ef-ticiencies of urandum and secondary fuel are equalized)
| |
| Eence l
| |
| gm g _ E M Es M N f U' Gr+& l J
| |
| S-14 13 Appendix 5 j 1
| |
| | |
| I g 1
| |
| got. then if the requirement to the "nen-fuela co=ponent of the e7-J penditures for energe productic:
| |
| *L' 1
| |
| Qg__ a~i --
| |
| A -.
| |
| ~"
| |
| pt ta met the. breeder is econo 1cally eure profitable than 'fRg .
| |
| Introduction cf sufficient number of the breeders to nuclear .
| |
| power can stop the, rise o! the fuel cost (complete breeding). The condition of such breedin6.(without anowance for fuel dectw& for ,
| |
| further development of nuclear power) is E_ g =. 7 $5d'c, s u ....
| |
| where 1 is the shares of the reactors of various types in UP.
| |
| It follows from these considerations that economically the NP future is fully deter:ined by the breeder econo =y and, particular-ly, by the "non-fuel" expenditures Kbr. This is why the important proble= to. be solved is forantion of the 21 century breeder concept.
| |
| Now let us proceed to the question of the co=petitiveness of the nuclear way of energy production. If the level of the reduced l expenditures of the energy production alternative to the nuclear one is Lg (the coefficient of .the competitiveness margin in NPP can L
| |
| be then /x: p; ) then the foll' wing c NPP competitiveness condi-tions can be obtained Kg + [ -v 'O b[
| |
| for TR j Kn - s C s 4. L A or
| |
| ! [# #_
| |
| N44* D for breeders k RB = -
| |
| A_
| |
| t _ _pu_b Transition to the breeders ee.rlier then the energy production by the A-type reccters will becone sufficiently expensive is un-5 15 14 Appendix 5
| |
| | |
| prefitable eince vith tht :urt.niu:. cor; beicg reit.t:,vely lo4 czpen-ditures w1.11 be L&.tr. In this case it is more profitable 3,o
| |
| * continue operation c' .he the mh1 rehetors, it:peding the process of uranium cest increase en account of reduction of its consumption q, (in the corglete treecing the fuel expenditure will increnee at k ca- G 1 east to C. J.
| |
| * AjA-- t ,
| |
| Ehen the coat of uranium production increases the E,0 - type re-actors may become competitive to the niternative energy source. The
| |
| . conditipn of eccnotic justification of these reactors in nuclear po-wer ensi neering in realf 5,ation of breeding will ,be s
| |
| ^
| |
| th + Sc4 b(. 3 C B, C, . . . ;
| |
| Lt*-C
| |
| ~
| |
| or = const. .
| |
| . s depending on the breeder characteristics.
| |
| As (2cphasized earlier, the economic attractiveness of nuclear f power will increase when the sphere of breeder utilization (inclu-ding their use for ther:a1 energy production) is extended on account of economically profitable realizat' ion of their high safety poten-tini (so-called enhanced and ultimate safety breeder designs). Use of these will ease the require =ents to the amount of excess produc-tion (due to reductice of the total uranium consumption), t'creover l the value of the fuci expenditures of the high-safety breeders, which is the econo:dc " permit" to nuclear power a could be K,*,
| |
| -( 4 Kc 4 %'( W h ,
| |
| For the case of the current low fuel cost (q,0f 0.2 Ye ) and excess production corresponding to the breeding ratio ~ 1 5 (re 0 4 q,) we dbtain the rer.uirenent to E as *
| |
| ]
| |
| Rd < Ke. + (0,c+1)Cu 9 Ke (L+ c.2 - ob 4,3 4 )
| |
| which seems to be atts. int.ble in the ner. rest future taking into i
| |
| account the high enough current level of non-fuel expenditures cf 5-16 15 Appendix 5 l
| |
| | |
| the ' low-tenpercture and andUs rial district her.";ing. Such t. fast reactor will no: loose its econocical attractobility for a lone tire-puelear ria :t safe:e concent ,
| |
| Ef fecti've research in nuclear saf ety 0.ni implementation of a revitalized nuclear saf ety concep; is the du.r.inant at the curr :nt sta-ge of nuclear power developmen't. This necessity is dictated by requ-1rement of improving both the safecy of indiv$due.] power plants and
| |
| * that of nuclear power as a whole.' Safety of any, power-intens e plant, fr.cluding the nuclear power one, cannot be absolute - cuch is the
| |
| " technical" nature of safety. Attainment of safety implies a conti-nuous fight for its improvement where concrete technically feasible advancea.ent een te foteseen.
| |
| In assessing hacardous consequences of human industrial acti-vity and, particularly,'of accidents having a stochastic nature, the tools of " probability-riek" assessment have been playing a Gro- l wing' role.
| |
| It is known that the probability is the numer'ical characteris-tit of the degree of possibility for occurence of a certain event (for example, accident) under certain repeatable conditions. j In contrast, to the probability we shall use the notion " risk" 1
| |
| as the nuserical characteristic of the degree of possibility for occurence of an event for a certain reason in the presence of some competing reasons under certain repeatable conditions.
| |
| In 1987-88 the International body of experts in IAEA developed
| |
| *Pasic NTP Safety Principles T where the world's experience of the
| |
| . j current idea cf the " picture of good state." of nuclear power instal-4 q
| |
| 1ation safety has been generalized.
| |
| In the above decunent the safety assuring activities in regard to the optre.tf =g N??s have been systematized c.nd the most 1 pertant safety principles and engineering means specified. Nevertheless it 5-17 16 Appendix 5 l
| |
| L_____. _ __ __ _ ____ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ . _ . _ _ _ _ _ _
| |
| | |
| it ap;11ei ra:r.er ;c. the pa.9 than to the future and did not prac :1-cally contain anclyris of new tendencies tc assurasee e :-I safety, E
| |
| particularly with respect to reduction of severe a:ciden; risk, which has been dictated by the recent accidents at the Three Ylle Ie3and (Tril-USA) and Chernebyl (CNPP-USSR) nuclear power plants. In view of particular danger of severe accidents 11 ITP?s, highly threatening human health and even life, such safety concept of advanced power unita is needed, that is aimed at averting just severe accidente ear-ly considered as hypothetic. .
| |
| The safety concept is intended to develop the general requdra -
| |
| ments to advancee K??s with respect to severe accidents, */aich is the main subject of dissatisfaction felt by today specialists and she public. The main current approach to assessment of the severe accident risk and danL6e should be, presumably, considered as an optimal combination of the probabilistic and deterministic approa- -
| |
| ches comprising a consistent and as much as possible detailed consi-deration of various accident s'equences with exception of those of them whose probability under particular accident conditions can be admitted as negli51bly small one.
| |
| Saf ety chilosethy of advanced nuclear tower units.
| |
| Current Requirements of the nuclear plant safety follow from the established ideas about unper:dssible exposure levels for the public, operating personnel and environmental contamination, speci-fic features of typical accident processes taking into account real damages of the severe accidents having occured. Social da age fol- j
| |
| )
| |
| lows from the danger of the El obal character of severe accident i I
| |
| consequences, high uncertainty in the scale of their real har=,
| |
| p'essible loss of tunan values which are beycnd ecenende evalue: don, social unacceptatility of significant risk to public exposure to S-)B 17 Appendix 5 i l
| |
| 1 I
| |
| l
| |
| | |
| 'l railose:1vity and environmental conta= tta ion. ,
| |
| I.cono=ic and ecciorical derage is determined by injud of the expencive resetor, nuclear power unit (plant) and, hence. cause sig-l nificant damage even if the released radioactivity hac been locali-l set.
| |
| Health damare co relates ritb the individual and d'ollecti'e ex-l posures (apprenr.ately with exceedin6 Permissible radiation level). -
| |
| The above types of damaEes depend on possible frequency of ac-cidental events, scale of the accidents themselves (radioactive re- ,
| |
| l 1 ease levels), NFP site, population density in nearby areas, direc-l tion c')d cechants..: of radioactivity Transfer and me.ny oth*r facicts.
| |
| Div:rcity (typical vectoriality) of the damages entails difficulti--
| |
| es in determination of a single reduced damaSe; in practice this can be done (then only with a g,reat uncertainty) solely for two ac-cidents, such as occured at TEI and CNPP. To provide singularity of the requiremen'ts to advanced ppwer units (NPP) with respect to se-1
| |
| ~ vere accidents let us co: sider that risk of severe accidents at the power unit is determined by the frequency of accidents and their scales (level of released radioactivity) taking into account the ATP site and population density. As " standard" scales of the real severe q accidents can be chosen those I
| |
| - with degradation of the power. unit and si6nificant release
| |
| }
| |
| of radioactivity to the environment (Chernobyl, A-type); .
| |
| t l - with'ecre celting and localization of radioactivity release -
| |
| (TMI, B-t37e).
| |
| The values of the boundary probabilities of severe accidents
| |
| , I (unpermissible with respect to the anticipated risk) can be obta-ined on the basis of the follev.ing criteria:
| |
| l l
| |
| l l
| |
| 5-19 I
| |
| 18 Appec. dix 5 I i
| |
| l
| |
| | |
| 1:-
| |
| l
| |
| - s ocin'_ tea:inc :.n.c acc ou .; ne c:.rcuncicnce thr.: eve:. n sre_11 nzber of severe accicents, threate:1.nc with the public overe-exposurc, nay shake faith in nuclear power t.nd cas- doubt in exps-dience of ite f.:rther development;
| |
| - ecomor.de and ecolorical, ascuning that the reduced economic risk from accidents at the nuclear power unit or the nuclear reacter
| |
| - must be noteceably lower than the profit from the energy output.
| |
| From the socini criterics follows the require =ent of the follo-wing type: du-ing the forecasted period of nucles.r power development .
| |
| (or,'which is epfroximately equivalent, within the lifetime of one humui cent.ation, f.e. cbout 50 years) and independent 3y of the number of the constructed and orcr. sting NF?s, no more than one me-vere accident with a si6nificant radioactivity release (of the Che-rnobyl-type) r.2st occur within the confidence interval of probabi-lities. Such a requirement correlates also with the psycholo61 cal barrier of unsecepterace of repeating dangerous events by the can and mitication of this barrier with change in the human generation.
| |
| D30tribution of the density of the probability, P(x),of the accidents resulting from the N?P nechanisms and ecuipment failures obeys the discrete distributions of binominal type. The Poisson dis-tribution is one suitable for description of the frequency of rare events: .h 3 2C7)= 6
| |
| - vihere x is the number of severe accidents at the total energy out-put of the reactor unite equal to n reactor-years.
| |
| The r_athenr.tical expectation L'y of the number of much accidents 2 l and the squared dispersion yD = D are I w I M .t = TGN=1 '
| |
| h =. $ = .t4 @- 1[ = x o(.x- d2N c.$
| |
| S 20 19 Appendix 5 1
| |
| 1
| |
| | |
| 1r. icrna cf the kr.cvtr. Fernou11 'r thecren ice near. frequene: ci tht- events (cceiden;s) under censictratior., eeut.: sc b, -
| |
| r tenir, e:
| |
| an : nfinite y proving n, to the even* probability f, i.e.Nn.
| |
| ~
| |
| IM ~/
| |
| Irom ine socini criterion one can obtain the resriction on the probability [, of severe accidents with radios.ctivity releases, if one requires that at the given enercy output n the teta: probabi-lity of occu-ence of at least twe accidents would not exceed an accep- ,
| |
| tablevalu*) , say, A = 0.01 (i.e.1%) . Otherwise L
| |
| 22(1)@('-h
| |
| ~
| |
| L In this case 7gg 1._ g ( 1. N and no D, L h = } .c % 0, i [a - g
| |
| - s The ecenetic and ecologier.? s: iterien icposes restriction on the number of accidents a f or the energy output n by a relation ha-sing the ior: .
| |
| 3v ,
| |
| c/
| |
| u = pm. S cl j where N is the reduced risk caused by the severe accidents of the given type (Y is the scale of a single damage front one accident).
| |
| d is the reduced (e.nnual) profit from the energy output; 3 is the coefficient showing excess of the profit over the risk (f = 1-10).
| |
| Two types of the accidents having occured per::itted the esti-mate of the reduced damages to be obtained: for the accidents si-tilar to that of Chernobyl (A-type) the de.= age is co= parable with .
| |
| docens of capital investments C for construction of a power unit (Y e 106), while for the TEI-type accident (B-type) it is equal to about C (Y e c).
| |
| As energy censumptien is universt.1, the relative annual pro-fit frc: the encrcy output is cc parable with increase in the natie 5 21 20 Appendix 5 I
| |
| l i
| |
| .m_-__-_ _________._ _
| |
| | |
| si
| |
| ~
| |
| (; ,
| |
| Q V. Q V "
| |
| enal nec nt '. (i.ei nce'.:nts to severs'. . ptr cent !. The:. a the *2rn2a reduced er;enditurer for enercy produc. :c:. by one reactor, e:;ual'to t, we obtain the estitz. ate f or d:-
| |
| .o";*c
| |
| [ ~ y*L ~ =..?C_
| |
| 3c ~ - ,
| |
| i where f is the discount factor (f = 0.10).
| |
| . ihen' Y tr.6- 10'3 nc
| |
| . at 6t =1C In assessi.:s the permissible number of accidents a we shall. keep to a prudent (conservative) strategy, having in rind a considerable I exten", of the sente accident damage ev under the conditions of the n
| |
| safe 5uard- :und save up, whic'h $ s speci:2c uf nucher power. She prudent strategy implier meeting the eccnomic and ecological requi-recents not only for the mean number of accidents a deter =ined by l .
| |
| the mathematical expectation, but even up to a so high a for which the probability density is low enough.
| |
| It follows from the Chebyshev's inequality for the probability density e 7.
| |
| P ( @- 1} }.) G b b r 2 h w a .) * ., m
| |
| .(where a is the positive nu=ber) '
| |
| .: -that with I deviating from the mathematical expectation !!, by a va-lue hisher than dispersion f , the probability of the event reduces sharply. ihen the upper estimate of m may be i tvt = Mit +N'Dx.== h*D and, therefore, the economic and ecolo61 cal restrictions to the se- 4 vere accident probability Pg ( b r.fc ) can be written as (ye-rMn{e.)-54N -
| |
| 5 22 21 Appendix 5 l
| |
| l I
| |
| | |
| . gwc types of necidents end, th :, two Y ve. luer impose appropriate A
| |
| economic ' t.nd ecciorical rectiremen s en the probe.tilines Pc and 5
| |
| pc . The need fer meeting both, the socini and the econctic and ecolo-Si cal criterib 61ves rise to the following safety requirements (to probabillites of A - and E - type acci6ents):
| |
| A f 6 min (fs, fc )
| |
| It should be emphasized enee more that the above iwqualities are the reference points in denlopment of the ter 6eneration reac-tors and i= ply fulfillment of the following guarantees (within the '
| |
| prctability crnfidence interval): ,
| |
| : 1. The anticipated number of severe acci?ents will not exceed
| |
| ,o_ne_dttrin6 the ferecasted time. - \
| |
| i
| |
| : 2. The economic and ecological risk will not exceed the permis-sible level determined by the KPP efficiency.
| |
| Dependence' of the limitin6 values of f3 and fc determining the boundary of thekermitted severe accident. probabilities on energy output n (i.e. nuclear power development) is presented in Fig.1 sho-wing change of the require =ents to the severe accident probabilities with ti=e. Determination of the concrete region of the permitted probabilities must be carried out taking into account both the anti- '
| |
| cipated nucieny power develop =ent rates and assumed energy output and the circumstance that the operation time of the existing NPP po-wer units is len6 enough and, according to the current assess =ent, cr.tst reach 50-60 years. The latter means that determination of the
| |
| " boundary" valuer must be done not by the enerEy output n expected to be obtained by the given year, but with allowance for the whole lifeti=e of the power units built (there seems to be little hope, if any, to che.nSe their safety level. ) Tor example, ,for the world o
| |
| 5-23 22 Appendix 5
| |
| | |
| - /. -- he.r.s a::: v'..! Caernonvl type ICort. distbotion - rac. reIcase)
| |
| E, - heavy a::.ct nts TMI type teore d:sruotton
| |
| -l; P A ( 1/yei!!
| |
| ._.__...... - . - . ....w... ... ...........
| |
| cuton; 7 :::::::::::::::::::::::::::::::::::::::::p?
| |
| '::*::::::::::::::::::::::::::::::::: ** lk.
| |
| =U
| |
| * P -
| |
| A'/"::::'::::::::':"
| |
| e AFre ::::
| |
| AREA:.........::::i I U/.f.-..:;.c#
| |
| i 6 _-
| |
| " ,:::U.:.::::::::-
| |
| p,O #
| |
| 3 Ay p$ ,
| |
| .
| |
| * ECONWICAL AND 2 (COLOGIC AL CRIT E Rt A
| |
| ~ (PRUDENT STRATEGY) 3 -._
| |
| a i m 2 4 "
| |
| 6 1n 9
| |
| ENERGY OUT PUT(REACTOR YEARS)
| |
| ADMISSIBLE PROBABILITIES '
| |
| P A < (10'' + 10-6 ) 1/ year = min (P A , Ap )
| |
| ^ d e (PRUDENT ,
| |
| P B < (10.s + 10'') 1/ year = P B STRATEGY)
| |
| * ENHANCED SAFETY NPP:
| |
| PA <10-7 1/y. ear, PB < 10-5 -l/ year MEANS: INHERENT PROPERTY + PAS $1V E SAFE SYSTEMS +
| |
| , + ACTIV E DEFENCE + DIAGNOSTIO , LOCAll2 ATION 1 M E A N S + ... .
| |
| SU RETY: PRA NPP-TYPE: ALWR. HTGR. LMR
| |
| * ULTIM ATE SAFETY NPP: PB _, pA = 10-7 1/ year MEANS: MAX (INHERENT + PASSIV E) SAFETY SU RETY: TESTS CN SELF DFFENCE PRCFE.: TIES PRA NPP TYPE: MHTGR, MLMR, SA LT.R E ACTOR. . .
| |
| S-24 23 Appendix 5 1
| |
| 1
| |
| _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _____________.___.___________w
| |
| | |
| l g b.e va*.ue cf corres; oni:.nt . tr.: beginning cf the 7.I cen:q-
| |
| =3y cpprqact ic 10'. , ; ,
| |
| /.t seen fro: Fig.' 'he f ellevring r.rzimv; probabilities i FA di 10~I 1/ reactor. year I
| |
| P mE 10~5 1/ reactor. year '
| |
| can be taken as the reference ones for a long enough time. These vs-lues are Lot teo stringe:t if ont takes into censiderati $n signifi- -
| |
| cant errors which can arise in severe accident probability assessment for particular designs of the odvanced power units. ,
| |
| The following requirements 3
| |
| JA, 7 5E 10~I 1/reacter year can be considered as sufficient.
| |
| As the nuclear power installations can be effectively used in various power productic: industries (other' than electricity genera-tion) a number of various-purpose nuclear plants have appeared, which differ, in principle, from the electrical N7Ps in the reactor design, plant' arrangement, supposed siting and, hence, in safety. Specific requirements to the safety of corresponding power units can be taken into account using the traditional method determining the permissi-ble limits of public energency exposure and environmental contamina-tion as well as the permissible siting regions, characteristic of the particular power unit. This can provide a correct transfer (no-realization) of the damage caused by the real accident to the condi-tions of the power units considered.
| |
| The above requirements =ust be references in designing the nue-lear reactors and nuclear power units of new generation. When choosing the directicn in moving to these require =ents the following two ap-preaches are possible'.
| |
| a) reaching the beundary value of the [A severe accident prcba-5-25 24 Appendix 5
| |
| | |
| tr
| |
| . bility c: . e ;.ower un:. with the "two-f unp" way of rr.dd oscid vity ;e.
| |
| lease i.e. by: - -
| |
| - reduciac the reactor tes-acation probability {3 to a level lying vathin the range 10'E - 10~7 1/ reactor year;
| |
| - developing (strengthening) means assuring loccl$zetien of the consequences of possible reactor (core) degradation'for both the .
| |
| . in-reactor reasons and cyternal effects so that the probability of the public overexposure would not exceed 10~7 1/ reactor year. This -
| |
| approach is more applicable for the conventional types cf units, ,
| |
| wherc it. crease of the accident stability of the reactor itself is a very complicat ed task. Reduction of the reector degradation pro-bability to a value not exceeding 10-5 1/ reactor yecr eheuld be only expected if the current nuclear reactor and power unit desi6ns be radi~cally improved and . the reliability of the equipment and dia-gnostic means essentially increased. In the USSR such reactor units are conventionally called the enhanced safety nuclear power units; b) meetin6 the sufficient "one-jump" requirements on account '
| |
| of re, duction of reactor (core) degradation down to the "liniting" value (10'7 1/ reactor year). Such an approach requiring development of quite new design solutions or, even possibly, creation of new reactor technologies may prove to be justified economically and =cre reliable with respect to final results. Such reactor unite are cal-led the ultimate safety power units.
| |
| + The analysis of the reactor safety means shows that the most reliable ones should be censidered the means which are able to act on the beris of the physical and other laws of nature i= edie.-
| |
| J tely in the reactor materini microstructure or the core e,tructural )
| |
| elenents and unlike the traditional active safety means require 5-26 25 Appendix 5
| |
| _ _ _ -- - __ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ . _ _ . . _ . _ _ 1
| |
| | |
| ne power sup;.17 reo:. any externa'. 6,uret. These are primarily the so-called inherent saf ety trover tes of the reector components, which g.anifest themselves in cot.biustion of thermohydraulic, nuclear and physical processes, chemical, strene ;th and ether properties as well as appropriate qualities of real etrnctural elements (facilities).
| |
| The h paesive protection means, including passive heat removal and reactivity centrol ones can r.lso be considered as the reliebic .
| |
| means if accident situations cannot aff act their perforson:e and they require no external power,are always operable, require no ope- -
| |
| l rating personnel interference.
| |
| The combination of the i'nherent safety properties and passive protective systems forms the system of internal safety means c! tne corresponding " barrier zone" (the reactor section limited by the protection barrier) and the reactor plant as a whole.
| |
| An important factor in increasing the reactor safety is reduc-
| |
| ' tion of the stored radioactivity (fission fras=ents) which can be l attained in a number of appropriate reactor concepts by, for example, a regular re= oval of fragments [6-7/ ,
| |
| The failure-controlled active means can also be considered as l a way to reach enhanced safety if their sufficient reliability is proved in concrete eituations.
| |
| The a::ti-accident stability of barrier tones (BZ) depends on both their own cafety properties relative to" internal" processes and environmental eff ects and external BZ. The environ = ental im- ,
| |
| pacts (ee.rthquakes, plane crashes, hazardous = meteorological condi-tions, etc) are blocked by special engineering means depending on -
| |
| the character of the KI? siting e.nd identifies additional require-sents to the prettetive bn rier cf the external 3Z.
| |
| l 5 27 !
| |
| i 26 Appendix 5 1
| |
| l 1 l
| |
| | |
| p the orciectivenece of each M arcinst the considered accident or
| |
| .. 2 a Group of accident situations :Is dete;t. mined by its ability not to exceed' the technically per=icef bic. - (from the dw-point of degrada-on) level of influence on its own protective barrier (with the tech-nically possible effect fren the external BZ), taking into account gli p2 otective means available of the reactor plant. The protective-ness against technically real (P) > 10"I I/ year) internal accident initiat* ors of the given type cnly at the et,et of application of the
| |
| , internal safety means of the reactor and the power unit will be cal-1ed the self-protectiveness. ,
| |
| If inherent safety properties and passive prot 0ctive treans are enough to prevent all severe accidents caused by the $n-plant isdti-ators -(ruling out the events whose probability, taking into account j the above criteria,is practically negligible and therefore, can be acceptable: 10-7 1/ year) without using the active protection means, control syste=s and operating personnel, then this protectiveness may be called as the tot al self-protectiveness.
| |
| The present-day concept of ensuring the NPP high safety is ba-sed on the principle of " maximum reliable protectiveness, attain-ment of self-protectiveness of the barrier zoneis and the reactor plant as a whole and, within the limit, on the total self-protecti-senese.
| |
| To es,ticate quantitatively the 3tvel cf stability to severe accidents for,"nen-self-protected" reactor plants it is reasonable to use the probabilistic analysis apparatus and t:0 cc= pare the pro- ;
| |
| . Ptb111 ties of development of anon-self-protective" accidentc1 pro-tessta of the above type (cr thti;P continuations) with their pern.is-I sible values. .
| |
| 1 1
| |
| . )
| |
| S 28 27 Appendix 5
| |
| - - ___-_ _ l
| |
| | |
| The self-protectiveness (i.e. the t+.f ety potent $ e.1 of various reactor concepts) can be assessed basing on the consideration of reactor's asymptotic states as resulting from development and attenu-ation of the above cecidental processes.
| |
| The advantage of such an approach /8/ da its relative simplici-ty, the possibility of obtaining a concrete result, taking into ac-count only the physical properties, general arrangement and reac- -
| |
| tor cooling scheme within the scope of the concept under considara-tion. ,
| |
| 3 In assessment of tht bazard of any accident processes a number l ;
| |
| of important contr$tuting factors must be taken into account. Among them is the po:,sibility of release of the energy " stored" in the re-actor when exceedin6 the operatin6 conditions of the reactor compo- ,
| |
| . l nent's. An exa=ple of such situations is exceeding of the tempera- '
| |
| i ture " threshold" of active zirconium - cladding-coolant interaction in the light water reactors, possible burning of the moderator and coolants in oxygen ingress into' the reactor plant (graphite and sodium reactors). Account of such reactors narrows vitality areas of the reactor components, reduces their stability, margins, etc.
| |
| These effects can also be consequences of some specific design fe-atures (such as. the local everheats ef fect, etc. ).
| |
| below are given the formulas for nuclear power units of the enhanced e.nt ultinste safety, covering the states bneic requirements to the safety. -
| |
| Euclear ,pewer unit _of enhanced saf ety for the new <
| |
| generation Y??:
| |
| The nuclear power unit (NPJ) which taking into accout site-specific er.tcr.;.nl acciden; initictors, ensures: .
| |
| 5-29 28 Appendix 5
| |
| | |
| l es
| |
| - n:n-czeeeding the ;c=issible lir_ite'" r,f radioac tive relea-
| |
| ,,e f er all r;ccident evente 1:.clud:.ng f aul:3 ac: ions of il.e' nie '
| |
| personnel, with a probability higher than 10''I I/ year:
| |
| - nont degradation of the reactor core for all accidtnt events including faulty actions of the site personnel, with a probability hiGh er than to-SI/ year; p;c3 ear tmwer unit of ultimate saf ety f.o.r the .
| |
| gw ceneranen .i?P ;
| |
| e '
| |
| W nuclear power unit (NPU) which, taking into r ecount site -
| |
| :;ecific external accident initiators, ensures non-excee(ing the permissible limits of r'adioactive releases as well as non-degrada-tion of tut. resctor ccrc in all accident events, including farity actions of the- NFP personnel, with a probability higher than 10~7 I/yearmnly due to the self-protectiveness of the reactor plent.
| |
| S"- dng up the discussion of the problems it should be enpha-sized that designers of the new generation reactors face a co=pli-cated but technically solvable task developing highly, safe, effici-ent ,and econo:ic nuclear power sources having a wide sphere of their applications.
| |
| The process of recenprehension of requirene:.ts to power plants
| |
| - tt. kin:; place u.ntil recently, teterni:Atien of touchening the ehfety criteria hae resulted in f;rev.h cf ;repi.cale c:. the pcrt of scien-tiste and desic.ners to crea ,a the ralically ain. .ced new generation
| |
| ~
| |
| i.eaeters. A 1rdef resiew, net pret end1rg to ec=pietences but riving an idea of trends in ees.rch for solutions of this problen is pre-it:- below.
| |
| * /Is i +:en.:.ned de;ending u he 57 pu ;cce tnd citing.
| |
| 5 30 29 Appendix 5
| |
| | |
| 1 l
| |
| The light **.ter thermal reerters (of the VTER, ???R and Bt3 ty-
| |
| ... J L ' pes) have been predominantly developed in the world NP by the present
| |
| - tine. The machine building b se created and technology mastered can also appraise the do=1: ant role of the light eter thermal reactors (Lta) in $P to for the nearest perspective /12/ .
| |
| At t'ae same tine the modern L%3 designs are characterized by a number of essential disadvantages, includir.g: ,
| |
| .. relatively high reactivity margins in the core;
| |
| - weak protectiveness agednst LOCl-type acciden.s, sensitivi-d
| |
| ;y loss-of-flowacc[ dents; ' ' ' "" *
| |
| - low-efficient fuel utilization; o .
| |
| . low parameters of the ther=ocyramic cycle;
| |
| - low heat potential 11mitir4 the sphatre of 1.%3 applicability.
| |
| A scant interest is expressed in stuhing and realization of the pro =ising energy production and safety potential of the high-temperature gas-cooled reactor (HTGR) designed for nuclear power planta of different purposes - NP ?, NIHp, NPP. l The reactor (Fig.3) having no netal structures in the core, with the graphite with sealed ceramic fuel, as the main co=ponent and with the inert belin: Eas as the coolant can provide high tem-peratures and high effectiveness of nuclear fuel utilization..The concept assures the erhanced saf ety of the reactoreby self-pro-tectiveness, as well as the satisfactory ecology of the NP? (see 4
| |
| the table). .
| |
| In the USSR the R and 3 works are in progress on some pilot _,
| |
| 1 units of various c.apab131 tics as well as on HTGR f*ppl:! cation for '
| |
| power *,echnol'c;ical ;Lpcees, !
| |
| I S 31 30 Appendix 5 l
| |
| | |
| The integrated arrangement of the main equ'ipment of, th,e primary cooling circuit in the reinforced concrete ve.esel, low volume power density, low resetivity margin reat. bed due to continuous refuelling, one-phese reure of the coolant result in the extended "surviabilitya of the installation even in accidents with complete loss of forced cooling and insertion of exceso reactivity. Additional it.eaptres himed
| |
| . at prevention of possible chemical activity of the graphite in the depressurization of the primkry circuit, removs1 of r'er,idual heat by the passive means make it possiolw ic satisfy the highest safety ,
| |
| requirements. ,-
| |
| Tor a high power NPTP a pilot 4Lcrgotechnological plant VG-4CA is being designed, that is intended for co-generation o.t hic 4 tempe- .
| |
| rature heat (up to 950cc), steam and electricity, which will per-r.it it to be used in various intense-energy branches of industry (Fig.4).
| |
| HTGR Features Safety High temperature Ecology potential 1 2 3
| |
| .Eigh inherent charac- Efficient substitution Minimum heat I IO8811 I"*1 teristics pollution Tassive ren: val ef Multi-purpose energy Efficient utili-residual heat source; dnduetriel ration of nuclear
| |
| , technological proces- fuel
| |
| , ses
| |
| ,. High raiintion- Long-distance energy Saving of water thermal strength supply resources with use of fuel of dry cooling tostra 5-32 31 Appendix 5 i
| |
| | |
| i
| |
| /E .
| |
| I W :'{
| |
| A'% I f y ,
| |
| &( m l h h.
| |
| / \ >
| |
| i 12 t -
| |
| q
| |
| /
| |
| NI N4
| |
| 'N "N l l z ) ,e 9
| |
| 0 O
| |
| *) 00 o
| |
| ' O 00 O
| |
| O mmU
| |
| 'o p "I o O '
| |
| O
| |
| 'O O*
| |
| O ../
| |
| 5 1 - Reattor Vessel
| |
| - 2 - Reador Cover 3 - Easket 3f
| |
| , 4 - Contr6 Rod j# $ -IJis Assembly 6 - Graphite Bricks ,
| |
| s P 7 - Upper Shield YFu (Ik 8 ~~ Annular Sh:cid
| |
| {j lf 9 - Upper Refie: tor
| |
| '#,/p u-b r 4, g
| |
| 4MeL '0 i - Control Rod Drive 11 - Main Gas Blower 12 - Steam Generator 33 32 Appendix 5
| |
| | |
| i ' y_ . _ _ , _
| |
| w n l\ e--
| |
| 8 t -
| |
| .c.
| |
| .J .-
| |
| 5y 2.
| |
| as 5
| |
| i
| |
| "-l l L -- a: r :,
| |
| : i. s . , .. , 'lO .,
| |
| J
| |
| .e = - -
| |
| h 5 :.. *)
| |
| - - . . .. . _ . ._ ~ , .f_
| |
| [
| |
| t !
| |
| e i. p '
| |
| l .--- -- ::. . ,,; :.m.J . i F l. ; !
| |
| s ~~
| |
| e l' I N . ; .
| |
| it o
| |
| :. $ ,'l <
| |
| 3, IF .I !7g = %; l'..A.re;v.;,r.w.13._,1._...
| |
| /
| |
| F m L =J ll T..m eo '-
| |
| 'V 1); . w. ..b.ga . .li %_.a-O t e e ,s 3
| |
| _ s 11 I t- 5
| |
| ,9 O.
| |
| c =
| |
| w.~ . =\3"{
| |
| : c.une g!_..f. i 1 2
| |
| =..q..2.g
| |
| . . . b.
| |
| .- . . , i V.*,~
| |
| qu.R- y 0- ~~.~..$
| |
| 2-a w
| |
| 5 r7
| |
| \
| |
| \ _lV m*g.yt.(;7-
| |
| .. 2 6
| |
| e- 4 <
| |
| E __j . . '?g l3.L,.
| |
| e .. 1,_ ... n. . .
| |
| E
| |
| . g y M-~ei.-;.gh _____g " _ -
| |
| y 1."-
| |
| .., . g y-- g
| |
| , t t . i .. . . . s, d, _d;2^
| |
| i . i
| |
| ..__..__r m e ,
| |
| o _ . ,E y
| |
| - i* .''' -l .. . ~ft
| |
| -. rJ c
| |
| .5 e-C ' . * .' - .: . ; <Q
| |
| '\ d o .3 o . l 1e&.c v0 u, -
| |
| y) og e ,,
| |
| < t _ - + - .
| |
| o H o
| |
| , n s _I__ .-
| |
| ;: I 7.r,s 3: .~,d. 53
| |
| < C- l
| |
| * g g -
| |
| p . ;dh
| |
| ~~
| |
| w w
| |
| \ ] =j . eO. -=:. nv.''7 c i4
| |
| - . '"3 F.~" it g g te
| |
| * E a.
| |
| w
| |
| . '* f
| |
| * l' yt ~1e !I*ijQ.
| |
| k_ _r v lf .,_b J%,&'=$_ ._
| |
| su \ _-9. : -
| |
| J I
| |
| 1 5-34 33 Appendix 5
| |
| | |
| -)
| |
| I 1 2- -
| |
| 3
| |
| ^
| |
| Impossibility of Steam production and Lot releases of re-
| |
| ..;'re me? ting residential district dioactisc urnducts heating durin6 P l ant opera-tion lack of phase tran- Elect;-ic power in- Small volume of ra-citions of helium dustry dioactive waste s coolant -
| |
| - High safety characteristics can be expected fgom use of the ,
| |
| modular HTGR providing the effective heat removal 'through the ther-mal conductivity L3 natural 4 convect $ on in the most seveie accident situations. These properties of the modular type HTGR are ensured by use of a small core arranged in a metal vessel and by limiting the mean power density to 3-4 MW/m3 .
| |
| The preliminary investigations have shown / 3,4,8,11/ that the fast reactors exhibit a high , potential for safety increase by pro-viding a co=plete self-protectiveness (i.e. only at the cost of the inherent safety means). Below is presented the concept of NP with the modular liquid satal-cooled fast reactor (LEFR), which is really feasible at the current level of development of the reactor engineering and technology.
| |
| The final purpose of the proposed 6cncept is development of a poner reactae with a high saf ety level ensured by its 'self-pro-tectivenees and hsving at the st=e time a high fuel breeding Poten- .
| |
| tikl. .
| |
| To solve this problem the following approaches are proposed / 3 /: -
| |
| - uee of the rodium coolent enabling to perform a reliable he-at rete,cval at the atmes-heric pressure and the sufficient boiling cr.-
| |
| Igin; ,
| |
| 5-35 i
| |
| i 34 Appendix 5
| |
| | |
| ds
| |
| - the use of an unconventionally " wide % array of the fuel ele--
| |
| ments and abare" fuel ancemblies making it possible to reduce the foid reactivity effect', to decrease sharply the probability of vio-
| |
| -lation of heat removal (VPLCD) becau.* cf the local clogging of the coolant flow section, to provide a high level of the natural'circu-lation of,the, coolant; ,
| |
| - use of intra'-nesembly oxid'e metallic (carbide oxide) hetero-genaity with a regular array of the fuel oxide (UO2 + P902 ) and ura-nium-containing metallic (carbide) fuel elements with the ratio 2:1 in order to ensurei nu inie'rnal breeding ratio of about 1 (BRA +1) t.nd "uel reactivity diec1.s needed for celf-protectiveness;
| |
| - use of modular arrangement of the reactor with its relative-ly low unit power of 500-1500 p /th permitting the residual ener-gy release to be removed by natural convection of air outside the reactor vessel and/or through the built-in emergency passive heat-
| |
| .exchangers of the primary circuit.
| |
| The level of the coolant natural circulation is expected to be in the range of- (30-100)f. of the rated one, the version with the 100% natural circulation having the extremely high resistance to LO yrs-type accidents and a simple design.
| |
| Later on, at a due level of technological and design assessments, l
| |
| 'the fire-safe heavy liquid-cetal coolant 11 , for examp3e, Pb, Pb-1!.s and others, having great temperature toilfug rargins can be considered, together with the sodium enes, This can mske it por,aible to avoid such a neEative property of the sodf um coolant as fire-hazard, to reduce radically the a=ount of the stored energy produ-ced by the possible che=ical interaction between nodium and air and water, to obtain te:.:perature, power, density and void reactivity effects even more favourable.for safety, to increase the fuel bre- l 5 36 35 Appendix 5
| |
| | |
| {
| |
| l i
| |
| 1 l
| |
| eding ratio. - .
| |
| j i
| |
| The high IN.TR safety assured by a[b protectiveness,is; also ;
| |
| J associated with reduc tion of the reactor rated power. The power se- 1 l
| |
| 1ection is deter--d.ned by the condition of fulfilling the following i requirencms :
| |
| - the density reactivity coefficient and void reactivity ef-feet cust be minimum and 3RA must be close to 1; .
| |
| - it is necessary to ensure the papsive removal of tha residu-al energy through the reactor vessel and/or chrough the built-in ,
| |
| energency heat exchangers of the primary circui: by the natural air convectien
| |
| - selection of the power must not reduce significantly 'ba IRF2 competitiveness.
| |
| .ae preliminary assessments of the charac'aristics have shown the.t the acceptabic pressure drop at the mean coolant heat-up of about 150cc,which cculd completely assure sodium flow by the natu-ral circulation, shonid be erpected at the fuel element diameter of about 9 mn and a relative pitch of the triangular array of 1.5. In this case the draught section height (the difference in the heights of the coolant er.it from the core and its entrance into the heat-exchanger) will be 20-25 e if the natural circulation (NC) is 100% and 10-15 m if NC is 30-40%.
| |
| The practically constant reactivity level during the core -
| |
| lifetice and the long reactor operation at the rcted power (as ,
| |
| long as 1 year) without refuellini;s and a high breeding ratio of
| |
| '1 5 is reached using the core with intra-assembly heterogeneity. -
| |
| Tlie ttble gives the maxi ==a coolant temperatures in the post-accident asynptotic m:ste for the IFR reactor (as one out of the peseitle desigem of highly se.f e fest reactor e_th the retal fuel U-ha-Er /4 2' e.nd IG.. As known these chars.cteris tics are very in-S-37 36 Appendix 5 l
| |
| | |
| portant in.aseessing the consequences of, thc accidental transi-
| |
| .. I ent process ATWS. Let us consider the follovrinS situations and their combinations: ~ 3oss .of the primary circuit coolant flow; - LOFW; loss of the heat' sink - LOHSYS; the uncontrolled overcooling (up to 3.,e ,reezing te=perature) of the primary circuit sodium - OVCWS; the uncor. trolled excess reactivity intiertion - TOFWS.
| |
| The calculations results show a high stability of IFR ar.d MTR reactors even to the most nevere, extremel r improbe.ble accident -- )
| |
| sequencies. Note that in contrast to the concepts with the traditi- ,
| |
| onal array of.the fuel elements and a relatively low NC level.
| |
| Um! has no undecired " surge" of the aodium temperature at a sudden shutdown of the pumps.
| |
| TABLE 5 The assessment of the r.aricum sodium te=peratures T at the ATWS accident sequences T"E" (*C)
| |
| ATES-SPI IER LMFR LOFWS 800 660 s 600 LOHSWS 850 600 550 TOP WS 900' 700 670 DVCWS -
| |
| 540 540
| |
| . (LOF+70P)WS
| |
| - 810 710 1
| |
| .. (LOES+TCP)WS -
| |
| 720 660 (LOF+ TOP +0VC)WS ,
| |
| 870 690.
| |
| ~
| |
| S 38 l
| |
| 37 Appendix 5 i
| |
| l l
| |
| l o _ --_ !
| |
| | |
| s The investiga*. ions carried out have shown the tdvantages of the concept discussed for development o$ a reliable 2T assaring not-exceeding the permissible limits of radioactive releases as well as E aiot-degradation of the core in all accident events having the pro-bability h$gher than 10~7 I/ year due. to the reactor self-protecti-veness. ..
| |
| An' important advantage of IJUR 10 that in its development use .
| |
| can be made of any available technical approaches, structural mate-rials and technological processes widely employed in reactor buil- .
| |
| ding.
| |
| .The advantages provided by use of a incombustible cecient ariae the interest in development of" the celten-salt reactors with the optimal, fuel balance (BR es 1). In this case it may prove to be pos-sible to minimize even more the factors determining the potential risk of degradation of any nuclear reactor /6,7/ .
| |
| The high-temperature molten-salt reactor concept is based on use of the high-resistance of the graphite materials in molten flu-orides and saod thermal and neutron physical properties of the melts, though their disadvantage is the high melting temperature.
| |
| At least, two promising physical and technical schemes (Fig.5) can be mentioned. One of them uses the forced circulation of the molten-salt coolant and spherical graphite fuel elements. The other is supposed to have the natural convection of the molten-salt coolant in the primary circuit. The fuel is eith. ? loccted in the indepen- -
| |
| dent salt fuel circuit, or is inser ed into the, coated particles and circulates over the primary circuit together with the coolant. Here one can trpect attainment of a high working temperature of the co-t o'lant at the natural convection.
| |
| l 5-39 !i 38 Appendix 5 )
| |
| 3
| |
| | |
| ; ,r ,. .
| |
| c
| |
| , m m mA -
| |
| *. e e s see i
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| M,
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| ,j
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| ,, !A r
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| m.
| |
| r.
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| -y-
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| . #< M nt.. ,J -G,F C... i
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| /, p - ,n sp_'i!;;"u:
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| . p'ti .
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| , ...c
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| .,> ., , , .Wp
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| *.- ~.
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| i
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| , .. twiv , L ,. .,' . .. .:e :ma ..y.
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| I.3a.
| |
| m,
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| %, ;6 3./.a%s.
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| ., t ...4 o ', m,4. te..
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| s
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| ,..., : l ir ".p . ,
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| o
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| . , . o .;.u . . .
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| 3 ,,i. .
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| . : ,,: : .5 y,,a c +..a v.6 i. . . ,
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| ii i
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| ~1..y lf.p.p y. ,,, , .. w. ;. y m w w .
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| . .i .
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| i .
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| . ,q c.r p; ;. ..,;,j-
| |
| -i i rf u n3.e M.H. N...,:" e u.. *. ..9
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| . - . ,;. ; *P. S'.y g'.*,.'j:
| |
| lili:.! .i!!
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| f
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| + "*g
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| <9 sj .t-Q t.
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| yys-
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| . at.u
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| . . . .s . ' . . !u
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| *o
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| ?,.'..i e
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| s
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| ' f f .&.:t n y:
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| 5-- .
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| s%Ec ..':iE. y' ., a r,y,.j,;,w@;s..<..:1,{'j, . 11... q e - .-), (y '
| |
| % : % . ,.u 4 , p f w .,. u .i . . :;, ,....,
| |
| i -
| |
| ':Q s.siv.if tt 8.>!yp.4 (p. g. ->1 r, !p 1 -
| |
| *y 14y ep ,s.
| |
| . el4-9 3
| |
| . 3$ . ,ty .ssy .. fil. *
| |
| ' ( A's'Q *' 2 _f._ r (*! M.,..'./
| |
| * i s -- .
| |
| ( [
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| .t
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| ~
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| , %e :. , m . n : ,. .
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| .- .r - w .
| |
| ^ i
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| ((Q \ ''
| |
| v < _p -
| |
| s k .. 'g *
| |
| ,3 p
| |
| , e l 3 s .- -
| |
| ', ti
| |
| . ~ .. -
| |
| 1 - t r e.'.
| |
| y . r.t " titCAL P UEl. ELEut N18.
| |
| . 3 = LOOLAtil
| |
| , d . E tt * ' L t * **t:
| |
| 3, . #tt ibi.g v e' es e *JTg:
| |
| so .
| |
| * ts v. n .t- rn$s t,.
| |
| ~
| |
| t . .'s I t 'Lt A
| |
| ** . e/(tt enit sa tes, ti . .n. e > ep ..ec.e te:
| |
| 10 = 'C8"C4 A 8 4 *".
| |
| .i .....,...
| |
| IJ .
| |
| * I f*i'itl* f CATICN f YS 9t. % .
| |
| . :. . .. . . . ,, , , . . . r . i, , , y .
| |
| 1 l
| |
| 39 Appendix 5 l
| |
| l l
| |
| L
| |
| | |
| l p The'most in uresting seen.: to be the ET!JSR modular concept be-
| |
| : sed on the combination' of the molten-sal't coolant and spherical fu-el elementa developed for the helium-cooled reactors.
| |
| ll The possibility of using the graphite epherical fuel elements is provided by the co=patibility of the molten-salt coolant with graphita within a wide range of temperatures (up to 1500'C).
| |
| 1
| |
| ! The disadvantage of the molten-salt coolants is pr.rtly a corpn- ,
| |
| ratively high melting temperature ( ) 35000). This circumstunge is p
| |
| ' thermodynamic 11y related to a low partial pressure over the melt ,
| |
| ' mirror,(helow the atomosphere pressure up to treperatures of 1500-l n 1700*C ).Tf the s;on1 of the nractical enetering th( high-temperature l l range of nuclear district neating and the hiE h efficient conver ion :
| |
| of heat to electricity is set, realizing that its solution is essen-tially simplified at low pressures, then th'e solution of technical t .
| |
| problems associated with the high melting point of the molten-salt
| |
| .j l.' .
| |
| i coolant is the natural and inevitable pay for simplification of the l approach to'high te=peratures. The positive feature of the increased melting te=perature of the chemically inert salt coolant is the improvement of the conditions of its' retention in the core in some accident situations and simpli[ication of the' conditions of lag li-ning on the vessel wall. If is worthwhile to emphasize also the high radiation resistance and the low p,u=p power typical of the mol-ten-salt coolants, which opens the way for high specific loads of the core (> 100 U/m3 ) at low ( d 100*C) te=persture drops of - .j the coolant in the heat consumption zone. For the HTL'5R modular j
| |
| . version with a low specific power Izvel of the core (~20 W/m3) _
| |
| : l. at the te=perature drops of the coolant in the heat censu=ption zone (* 25CL300eC) the heat re=evel e.::d her.t transfer in the pri-U ca y circuit een be ob',ained due to the coolant nat' ural circulation, which creates prerec,uisites for a cachine-free high-te=perature
| |
| . 5 41 40 Appendix 5 1
| |
| | |
| 3 eat transfer to 'he cons serr..
| |
| The low pump power typical of the m lten-salt' coolant is pro- #
| |
| videt by the f avourable combinet.4.on of thermal and physical proper-ties of the fluoride melts. If the chemical inertness of the salt coolant is taken into account, the riqaired low working pressure in the ;;1t*.ry circuit, its rediation stability, the availability of graphite - and metal-based structural materials as well as 'its cys-11 ability and low-prices, the attracti.ven-er of this coolant for the reactors of the next generation charactorired by t'.ne enhanced safety, efficiency, ecology becomes evident.
| |
| Of particular interest is the ET' 233 concept with a gas-turbine plant (OTP) operating in the open air cycle. Due to the high e cc.-
| |
| nomy along with schematic gimplicity HTMSR w:tth GTP can be a pro
| |
| * mising energy source even for desert, permafrost and remote regions.
| |
| Among the inherent safety means of this reactor are the fo3-lowing
| |
| ! - the ability of the coolant and structural materials to effec-tive heat accu =ulation;
| |
| - a low pressure of coolant saturated vapors;
| |
| - the large boiling margin (80000); i
| |
| - the good strength of structural materials in the coolant;
| |
| - the mutual chemical passivity of the coolant cogonents as well as passivity in interaction with foreign substances from the core surrounding;
| |
| - the possibility of self-healing of the vessel microcracks;,
| |
| ~
| |
| - the natural circulation of the molten-salt coolant;
| |
| - the natural circulation of cooling-down coolant-airt
| |
| - the low pressure of the reactor coolant.
| |
| 1 I
| |
| S-42 4
| |
| l 41 Appendix 5 u________.______.________. 1
| |
| | |
| e Continuing the works on improving the current reketors; search ,
| |
| for and' development of new reactor' technology in the direction of increasing-the reactor self-protectiveness, improving lot,alization means,- broadening the sphere of utilization and improvinC the !
| |
| fuel breeding, the mankind'will have a reliable and effective energy source.for years. ,
| |
| R 9
| |
| l
| |
| [. -
| |
| l S-43 42 Appendix 5 O -. _ _ _ _ . ..
| |
| | |
| r :-
| |
| D - t
| |
| 'J,
| |
| '['
| |
| PREFERENCES.
| |
| : 1. M.A. Schultz, M.C.Edlund. A New Steam-Cooled Reactor. Nucl.
| |
| j .-
| |
| Sci. Eng. 90, 391-399, 1985
| |
| : 2. D.D.Es2sfati. I.S.Slesarev, V.A.Stukalov, T.D.Shebepetina.
| |
| The analysis of breeding characteristics of- the steer-cooled fast neutron reactor. Preprint I AE, V 3996/4, Moscow,1984.
| |
| 13 Orlov V.V., Slesarev I.S. , Saritsky S.M. e.a. The Theoretical Possibility of' Reducing the Doubling Time in Fast Reactors by Using Heterogeneous Configurations of'Various Types of Puel. ? cst React or Physics,1979, Vienna, I AEA-5E-244/76,1980,
| |
| : v. 2, p. 409-480~ . ,
| |
| 4.- Slesarev I.S. On criteria of the breeding efficiency In Nuclear Power. In Proc. 'ICENES-4, Madrid,1986,' p. 330-333.
| |
| : 5. -E.K.Nazarov A.Ja. Stolyarevski1. The power technologic . use of' high-temperature nuclear reactors. - Voprosy atoanoi nauki i tekhniki. Ser.: Atomno-Vodorodnaya energetika i tekhnologlya.
| |
| M. , ' Atomizdat,' 1980, vyp. 3, p. 58-128.
| |
| : 6. Novikov V.M., Molten Salt reactors end molted salt carriers for industrial heat supply. In Proc. ICENES-4, Madrid 19,86,
| |
| : p. 199-20 -
| |
| 7 Ost Y., Trauger P., White' J. Nuclear power opticus viability.
| |
| Ock-Ridge National Laboratory's . Study Ibid, p. 206-210.
| |
| : 8. D.C. Wade, Y.I. Chang. The IRp-reactor concept. Proc. of Int.
| |
| . Topical Meeting on Advances in Reactor Physics, Mathematics and Computation, Paris, 27-30 April 1987.
| |
| ~
| |
| 9 V.V.Orlov, E.I.Grisbanin, V.M.Murogov. Certificate of author-ship N 776334 (USSR). The method cooling of the fast reac-tor core. - Bull. izobretenii, 1966, N 36.
| |
| 5 44 43 Appendix 5
| |
| | |
| o .
| |
| I
| |
| ~
| |
| L ..f 10. v.v.Bulygin et a]. The atomic power, technological plant with
| |
| .. I the VG-400 reactor. The report at TC IAEA on gaa-cooled re-l actor and their utilization, October. 20-?3, 1986, Jul.t e$,
| |
| PRG. ;
| |
| : 11. V.T.Orlov. The nuclear breeding and reactors safety. The re- l port'at the World scientif1: forum on energy prob] ems el III century. M., October, 1987, 1-17. '
| |
| o
| |
| : 12. Technical Committee Meeting on Advances in Li@t Water Technology.
| |
| Washinston, USA, November 24-25, 1986. Summary IAEA.
| |
| * l l
| |
| i 5 45 44 Appendix 5 I
| |
| | |
| APPENDIX 6 Three Papers by V. G. Asmolov et al.:
| |
| . (1) " Investigations on Justification of Nuclear Power Engineering in the U.S.S.R."
| |
| (2) "The U.S.S.R. Approach 13 Safety. Studies" (3) " Development of Nuclear Power Plant Safety Research in the U.S.S.R."
| |
| I t
| |
| Appendix 6
| |
| | |
| INVESTIGATIONS ON JUSTIFICATION OP NUCLEAR POWER ENGINEERING IN THE USSR ALMOLOV' Y.G. .ERMAKOV N.I. ,PONOMARJEV-STEPNOJ R.N. ,
| |
| PDTZENK0 A.N.,SUKCHORUCHKIN V.K, (USSR)
| |
| Strategic importance of a wide-scale and accelerated de-velopm et of nuclear power engineering in the USSR is deter-mined by ce necessity to solve a series, of specific problems pertained to economy and power. supply including territorial ut and consumption in power resources, disproportions the great queta'of s in outp/
| |
| oil gas .fgel in electric power engineering, ecological problems et cetera. Growing demands in electric and thermal energy cah~ nit be satisfied in the meanst future
| |
| ~
| |
| only on account of resources of organic fuel especially in the European part of the USSR even with rega2d for realisati-on of power saving policy in growing but real scales. , The pri-ority development of nuclear power engineering in Eumpean part of nearest. the andUSSM is laid long-te2n down in the economic USSR Energy Programme, plans.
| |
| Peculiarity of nuclear power engineering development in the USSR implies except application in traditional,electrio power area also its introduction into the heat power enginee-ring ( district heating and industrial . heat generation of dif-forent potentis1). Building-up of "small" nuclear power en-gineering (nuclesr power plants (NPP), nuclear power-and-heat-ing plants, nuclear heating-only plants of maall power) for remote areas of the country is also being planned.
| |
| The necessary ters of realization of piens related to nuclear power engineering development in the USSR is provision of its social acceptability that in the first place, is oca-nected with maximum decrease of risk of severe accidents. At-tention to these questions especially have aggravated after the accident at Chernobyl NPP. .
| |
| It is completely obvious that development and implemen-totion of measures to improve nuclear power safet performed at rates exceeding its power increase. y are to be As all over the world we meet with the difficulties to er, plain the. population methods for determination of acceptabi-lity of. this or that safety level of nuclear power as well as of comparison of safety levels of nuclear power and other ourw rent industrial technologies proportions of risks created by nuclear power with the genera,l background of technogenic risk 1 Appendix 6
| |
| ___a.m.-.__---- - - - - - - - - - - - - - - - -
| |
| | |
| (
| |
| L in the modern society and also of necessity of nuclear power engineering development to decnase the very background. ,
| |
| However, at present inside the nuclear community in the USSR a strong conviction has been fomed that quantitative-probabilistic approach offers a powerful means in detemina-tion of necessary safety-related characteristics with respect to probability of severe accidents that meet the expected rate of the power engineering development.
| |
| In th6 USSR two ways of reaching the necessary characte-ristics of safety are considered in creation of the reactor -
| |
| units of new generation.
| |
| The first way - creation of reactors with increased sa-Inj;I. Reactor of such type, with regara ror possible outsice -
| |
| accident events shall secure the following safety indices for its siting:
| |
| - nonexceeding of allowable limits of souret terms for all acc{/1reactoryear; 10- dent events taking place with probability grea.ter than nondestruction of the reactor core in all accident events taking place with probability greater than 10-2 1/ reactor year.
| |
| Reaching thp probability of the reactor core destructi-on less than 10-21/ reactor year ( against 10-3 + 10-4 ob-tained at present) shall occur on the basis of drastic design improvements, increase in equipment n11 ability, improvement of diagnostic means and so on within the framework of tradi-tional types of the nactor plants. In this case reaching the required value of probetility of the accidents with con-siderable source tem shall be provided on account of strenth-ening of means for localizing the consequences of the core destruction on account on containments also.
| |
| The second way - development and creation of the mae-nors with untimete safetr. Here reaching the requinments for prk bability of accidentF Fith considerable source tems of 10- 1/ reactor year shall happen due to further decrease 3
| |
| of probabil of the core destruction up to the " ultimate" value - 10 yty/1 nactor year. This result may be reached on-d ly by way of increasing the self-protection of the zleacter plant, i.e. due. to application of inherent safety and passi-ve safety means. For such reactors the necessity in locali-sing means will be possibly detemined only by urposes of protection against the outside accident initiat events.
| |
| Such an approach, evidently will demand developmen. of princi-pally new designs and may he even creation of new reactor technologies. Revetheless, it may turn out to be mon econo-mic and more reliable as to final results from the point of view of safety assurance.
| |
| 2 Appendix 6 I
| |
| | |
| o At present in research and development organizations *of the USSR, the development of WER-1000 reactor design with increased safety comes to an end. This is the.so-called
| |
| (: WER-88 design. Possibility or filling the reactor shaft with .
| |
| boric acid solution, building-up of the increased basement (or creation of dry trep) to hold the melted core in case of
| |
| . severe accident and also a system of filters for pressure and activity dumping - all this is provided in this design in addition to safety censures reelized in the current de-signs of cotr.ercial WEE-1000 units. .
| |
| . Eext important issue is creation of WIR-92 reactor de-sign wherein should be used at most home and world scienti-fic and technical achievements in nuclear power engineering, gathered experience in design, construction and operation of
| |
| - the reactors and ensured the required safety and economy-n-lated indicas basically on account of radical simplification .
| |
| of the design, standardization, perfection of diagnostics and expansion of use of massive safety means. Wide-scale in-troduction of NPP with WER-92 reactors shall start at the end of.1990, not later.
| |
| .. In solving the tasis directed to safety improvement of nuclear sources of energy, measures to improve nuclear and radiation safety of all the stages of nuclear fuel cycle are to be simultaneously developed and applied due to existance of potential hazard of radioactivity n isase at the stages of storage, transportation, reprocessing and disposal of ra-dioactive waste formed in the course of nuclear reactors ope-ration. .
| |
| Attainment of the set aims is impossible without the de-ve*1opment of nuclear power engineering safety-related scien -
| |
| tific base without cardinal raise of level of computation and ding-upexperimental of new objects studies,ofreconstruction,d experimental an computation base, expansion and buil-resources and personnel provision of work.
| |
| Solution of p'oblems r related to n11 ability assuarance of nuclear stations and other utilities of nuclear power an-gineering requires application of complex af measures which structurally may be aivided into the following elements:
| |
| - setting of aims and safety norms and criteria corresponding to them
| |
| - - developm:ent, substantiation and applicatica'of technical safeguards which guarantee achievement of accepted safety norms;
| |
| - development of wide spectrum of administrative and organi-
| |
| . zational requirements and measures which guarantee accorp-lishment of accepted safety norms and criteria and provide functioning of engineered safeguards.
| |
| I-3 Appendix 6
| |
| | |
| l O
| |
| To werP out measurea for safety control it is necessary to possess the effective means for prevention accidents, in-formation on initial events both of inside (including equip-ment failures and personnel errors) and outside character, be '
| |
| able to describe the course of accident, evaluate its conse-quences, work'out means of protection and localization to the optimum. It is possible only by way of simulation of the who-le complex of physical-and-technical processes and control ectienu-by automatic process control system (APCS) and person-nel with evaluation of initial events probability and buil-dine-up of schemes of possible ways of the. accident develop-
| |
| * ment. Anthev+4Mtv of safety analysis results with applica-tion va ni.cname samdlation means shell be secured. by reliable experimental and operational infomation added with inferina-tion from producers of equipment and nuclear station systems. ,
| |
| The developed Complex:cf programs of resea'ch-and-deve-r lopment work directed to increase reliability and safety of nuclear power installations is subject to the solution of enu-merated problems.
| |
| This Complex embraces main directions of R&D in the area of nuclear power engineering safety and involves:
| |
| : 1) a special-purpose progrowne of R&D to secure a reliab-le and safe operation of nucles ;ower installations;.
| |
| : 2) program of work in the saa of probabilistic safety
| |
| . analysis;
| |
| .. 3) safety concept bases of perspective nuclear power
| |
| * units;)4 program of development of YVER-1000 reactor with in-proved safety. I A special-purpose progrensne of R&D work to provide a re-liable and safe operation of nuclear power installations re-lies upon the experimental base being improved and newly-built up. Development of this base is provided by way of no-dernization and construction of research and experimental com-plexes, facilities and stands within the period till 1995. -
| |
| T In the special-purpose programme are formalated'bacio L direetions of experimental investigations. Structurtl and i
| |
| . functional disgrams of the special-purpose progzunene are pre- "
| |
| sented in Figs.1 and 2, The work is provided on studies of heat transfer hydro- -
| |
| dynamics in the reactor plants during accidente, behaviour of fuel elements and fuel assemblies under accident conditi-ons, fuel behaviour and radioactive fission yield in esse of severe accidents with fuel element melting and core destrus- -
| |
| tien, investigation of hydrogen safety-related problem as well as of problem pertained to vapour (" physical") explosions, creation of scientific bases of safe control of the processes 4 Appendix 6
| |
| | |
| et the nuclear station, designing 'of engineered means and systems for equipment diagnostics, study of problems of nue-leer safety during fuel transportation. The studies are pro- .
| |
| vided also on the complex of processes and hydrodynamic ef-facts in localizing systems under the accidents with loss of the primary circuit sealing, on the rating of radiation-sanitary and ecological factors of the nuclear station in-fluence on the environment and population health, on the problems of strength and destructive dynamics of structures and the reactor plant equipment, on investigation of safety affected by outside factors such as force majeure, aircraft crash, terrorism, etc.
| |
| A part of above-mentioned work is provided as the de-velopment of investigations carried out earlier. A part of work is set for the first time.
| |
| Importance of experimental verification of modem and perspective safety concepts dictates urgent necessity of build $mg-up of t -
| |
| - large-scale universal safety stands of great power that meet the requirements of themohydraulic processes st-muistion and oriented not only to modern but also the per-spective reactor plants equipped by modern means of measure-ment, control, collection,and processing of experimental da-ta;
| |
| - special experimental complex to conduct large-scale physical-and-chemical investigations usdag unirradiated fu-el;
| |
| - special loop research reactor to study fission yield in simulation of accident situations;
| |
| - middle and full-scale models to study explosure phe-nomena accompanying out-of-design accidents; ,
| |
| - multiple computer simulation complex and special si-
| |
| - mulators of nuclear station to study processes in complica-ted systems : Including studies in the accelerated time scale with connection to actumi equipment and studies at the nuo-lear station; I
| |
| - special stand which pemits to conduct experimental I investigations of physical-and-chemical foms ana compounds I I
| |
| of fission products in the interaction with reactor materi-als.
| |
| Obtaining authentic experimental information will ori-
| |
| . tically affect basic technical and nomative-organizational safety-related approaches and , finally, economic nuclear station indices as well as . development of nuclear power engineering as a whole. ,
| |
| Probabilistic safety analysis and risk evaluation I sented in the program of work in the area of probabilistkre- e safety analysis of nuclear station are assigned to be one of
| |
| * j I
| |
| 1 1
| |
| '- 5 Appendix 6
| |
| | |
| the central places in the program.-
| |
| Improvement of nuclear station safety is predetermined
| |
| - to a large extent by use of probabilistic approach to analy-sis and control of safety.
| |
| Its preference, in comparison with traditional detemi- '
| |
| nistic approach, consists in the fact that it permits to pro-
| |
| , vide the balanced process of safety-related decisions on the i basis of complex and successive approach curing the design development of nuclear stations, their operation and,also their regulatien. Application of this analyais permits to -
| |
| use quite other qualitative level of: analysis. of safety of nuclear stations reliability of which depends on the great
| |
| . number of compone,nts; substantiation of requirements for nuclear station siting; substantiation of technical and or-
| |
| * ganizational safety-related measures; determination of the requirements for reliability of' systems and equipment; deve-lopment of optimum schedule for their maintenance; prepara-tion of operators; determination of activities in out-of-de-sign accidents as well as their preventive maintenance;arran-gement of problems and effective use of resources for safety assurance; determination of inspection activities; ;lustifice-tion of regulatory decisions; generalized quantitative esti-mation of safety and risk of nuclear station usage.
| |
| ' Simulation means used for probabilistic safety analysis (PSA) and risk estimation involves
| |
| : 1) data base with experimental and operational informa-tion on characteristics of systems, equipment and processes important for safety used for analysis and des 43 m4M of nuc-leer stations, check of authenticity of computation-and-ana-lytical models of accident processes by way of comparison with date of representative experiments; .
| |
| : 2) complexes of computer codes including'*fasta compu-ter codes for operational personnel of nuclear stations and central' control bodies to perform probabilietio safety ana-lysis and expert information related to reliability of the equipment and systems of nuclear station;
| |
| : 3) complexes of modern powerful computer codes f.. ana-lysis of accident course at the nuclear station based on the results of fundamental investigations of physical and-}hysi-cel-and-chemical processes.
| |
| * Program of development of nuclear station with VVER-130 reactor with the improved safety is based upon the results of operation referred-to the complex program of safety and directed to building-up of the nuclear station that mesta-the modern regulations and safety ofiteria for realization -
| |
| till the end of the current century.
| |
| We attach a great im3nrtance to international cooperati-L l -
| |
| f l- ,
| |
| 6 Appendix 6
| |
| =- - _- - - _ - _ _ . _ _ - - .
| |
| | |
| o which, besides information, technical and 3
| |
| _.onal economiccooperation aspects, ,took at present and for the nearest futu- !
| |
| re a special significance due to the following circumstances:
| |
| - to increase public attention both in the country and abroad to the questions perte$ned to. nuclear power engine-ering development j
| |
| - significenc.e of time consumption to perform necesse-ry R&D with support only by our own forces. J With the help of effectively organized. international
| |
| - cooperation it.is possible to reach economy in time and re-sources for the program of nuclear power engineering asfety-related work to be perform.
| |
| We provide for various foms of intemational coopera-tion that involve realization of joint studies, mutual exa-minations equipmentofofdesignslear the nuc station ,etc up to formation ofjoint development of computer mixed utilities.'
| |
| In necessary cases, when accelerated introduction of !
| |
| measures, directed to improve safety and home developments on specific measures are absent, there is envisaged purcha-se of licences or equipment abroad. !
| |
| l Realization of programs directed to improve NPP safetv i in the USSR requires considerable investments both for buil- I ding-up the required experimental base arrd for conducting scientific studies. The real tem of its accomplishment is 10-15 years. 'We believe completely that realization of this program pemits nuclear power engineering to enter a new century as safe and economic means for provision of our ener-gy needs.
| |
| t j
| |
| l i
| |
| 7 Appendix 6 l
| |
| l *
| |
| \ - -- - __
| |
| a
| |
| | |
| 't O
| |
| ~
| |
| NUCLEAR POWER PLANT SAFETY RESEARCH PROGRAMME .
| |
| PROCESS & FUEL INTEGRAL FULL-SCALE ELEMENTS ELEMENTS OPERATION RIG EX?ERIMENTS STUDIES STUDIES STUDIES 4 4 4
| |
| -EXPERIMENTAL DATA BANK -
| |
| 4 4 4 i
| |
| PROGRAMME PROGRAMME SUBUNITS PHYSICAL & COMPUTER MODELS TESTINGS SET OF PROGRAMMES : i I
| |
| 4 NUCLEAR POWER PLANT ,
| |
| , SAFETY ANALYSIS j l
| |
| 4 NUCLEAF POWER PLANT TM IMPROVEMENT SAFETY SYSTEMS OPTIMIZATION i
| |
| FIG 1. -
| |
| 8 Appendix 6 L.
| |
| | |
| L I
| |
| PROGRAMME STRUCTURE HYDRODYNAMICS DNakCH& DEVELOPMENT ANALYSIS OF AND PROGRAMME BEYOND THE TO ENSURE : DESIGN-BASIS HEAT TRANSFER :
| |
| IN REACTOR NUCLEAR SAFETY ACCIDENTS FACILITIES 4 4 STRENGTH FUEL ELEMENTS BEHAVIOUR EXTERNAL ENGINEERING AND EVENTS STRUCTURES -.-+ RUPTURE UNDER e-- -
| |
| DYNAMICS ACCIDENT ,
| |
| CONDITIONS I
| |
| QUANTITATIVE PROBABILISTIC SAFETY ANALYSIS ,
| |
| AND NutirAp ACCIDENT NUCLEAR POWER PLANT SAFETY MITIGATION DURING
| |
| : RISK .a N N WT ----+
| |
| C CONTAINING)
| |
| ETORAGE &
| |
| SYSTEMS
| |
| * TRANSPORT ENVIRONMENTAL FUEL .
| |
| M -+ EF'/ECTS MELT-DOWN OF NUCLEAR PROBLEMS POWER PLANTS
| |
| ~
| |
| umuts '
| |
| 0 l HYDFLOFH AUTOMATIC SYSTEMS &
| |
| + -- + COMPONENTS PROWLEM + -- + CONTROL DIAGNOSTICS SYSTEMS RELIABILITY FI G. 2.
| |
| 9 Appendix 6 w_w. --.__ _. --____--- . . . . _ . . _ . _ - _ _ - - - _ _ - - . - - .- -_ -_ _ _
| |
| * THE USSE APPROACE TO SAFETY STUDIES {
| |
| ', Y.0. ASMOIOY '
| |
| I.Y.Kurchatov Atomic Energy Institute, Mor, cow, USSR In the lighrcI the- seWWecidents at the t'MI and the Cherncbyl nuclear power plants (NPP) in the USA and the USSR, respectively, the problem of ensuring safety became
| |
| ~
| |
| of cardial 1:aportance for furtherstable nuclear power develop- <
| |
| ment ari necessisted the development and implementation of ra-dical measures to improve safety p( the existing nuclear power plants and of those deing constructed or designed. It is obvio that safety improvement snould be more intensive than the de-I velopment rate of nuclear power.
| |
| The analysis carried out in our country af ter the ,Cherno-byl accident has resulted in substa$tial changes in our plans on the introduction of nuclear power sources. The development of nuclear power is a necessity both for the USSR econosy 'and for the world economy as c, whole. The analysis of the antici-pated development rate of the world and the national nuclear power as"well as economic and social damage assessments of the severe accidents which have happened show that the probability
| |
| ~
| |
| of the core damage which does,not lead to the population radia tion exposure above the al'lowable emergency level, should not be scre than one event per 10 5 reactor-years and the probabili ty of the core damage, that results in zudioactive releases above the prescribed limits.,should not be greater than one event par 107 reactor-years. ,
| |
| To solve the problem of nuclear power reliability and i
| |
| 11 Appendix 6 )
| |
| | |
| 2 safety it is.necessary to implement s whole range of measures that can be structurally divided into the following elements:
| |
| - establishments of safety goals and corresponding stan-dards and criteria,
| |
| - development, , justification and introduction of enginee .
| |
| safety features and measures to ensure the established safety
| |
| ' standards implementations .
| |
| - development of a wide range of noministrative and orga:
| |
| zational requirements and measures to ensure the implementatie of the established safety standards and criteria as well as t1 functioning of engineering safety syratems.
| |
| After the Chernobyl accident the Soviet Union adopted e, spectrum of organizational and technical measures to substan-tially increase the nuclear power safety. ,
| |
| First priority technical measures were developed and imp:
| |
| mented' to exclude the possibility of reoccurrence of Chernobyl-
| |
| ~
| |
| type ace (dents at ItBE reactors /1/. The implemented measures dealt mainly with two problems: decrease the void reactivity coefficient p,. off and to improve the reactor emergency prote tion system. Further measures at the existing nuclear power plants with RBE reactors to increase the . fuel enrichment from 2 to 2,4% and to decrease the graphite inventory in the reacte cell will practically lead to zero influence of the void react vity coefficient. Reactor scram system is being develcped. .
| |
| that will ensure the negative reactivity insertion up to 3 &
| |
| J. .S I .
| |
| during 2 - 2,5 sec.
| |
| Besides the development measures for RBE rea6 tors, activ ties to increase safety are being carried out at nuclear power plants of all the types.
| |
| 12 Appendix 6
| |
| _ _ . _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _-- m
| |
| | |
| 3 Measures to improve safety are being implemented also at nuclear power plants with VVER reactors, which are the leading and promicing N?ps in the USSR energy programne and the safety level of whhn is similar to the one of PWRs 127
| |
| ~
| |
| .. Based on the results of investigations jdesign and techno-logical changes were proposed and implemented,''that will enabl to realize the design life of pressure vessels.
| |
| At present the design of VVER-1000 reactor with improved safety is being developed. It is called VVER-88 desigh. The i
| |
| design provides for the possibility of filling the reactor vault with boric acid 1; wtion as well as construction of thiel basement ,(or dry trap) to antain the melt-down cors in case o:
| |
| a severe accident and installation of filtors the pressure relief line to capture the activity.
| |
| The diversity of requirements to improve safety /3/ calls for different approaches to the development of the so called improved safety reactors as well as ultimately safe reactors.,
| |
| The first approach to reach the goal is to improve the ,
| |
| proven and tested reactor facility design and the severe accidt consequences mitigation systems.t,The second one is the implemet tation of requirements *to decrease the probability of severe accidents due to inherently safe features of the reactor facil ty. This approach requires the development of principally ner design solutions and, probably, of new reactor technologies.
| |
| The complementary nature of the two approcches is evident and justified at the present stage of nuclear power development
| |
| / L An picreased resistence of nuclear power sources to severe accidente is a precondition to reach the above goal.
| |
| 13 Appendix 6
| |
| | |
| 4 Simultaneously with the nuclear power sources safety improvement it'is necessary to deve3op and introduce measures to improve the nuclear and radiation safety ' of all the facilit of the nuclear fuel cycle since there is a potential danger of radioactive releases during storage, transportati'on, repro-cessing and disposal of radioactive materials genevated'by nuclear reactors.
| |
| To achieve the established goals the USSR has developed a research programme for improvement of nuclear power facili-ties safety /4/.-
| |
| In accordance with the programme) activisies are planned and being. implemented to study hydrodynamics and heat transfers pr'oblems during severe accidents, behaviour of fuel elements and assemblies under accident conditions, fuel behaviour and radioactive fission product releases during severe accidents with fuel elements melting and reactor core disruption, hydrogt safety problems, steam
| |
| * physical" explosions, durability and rupture dynamics of reactor facility structures and shipment, well as external safety related factors such as astural disaa-ters, airplane crushes, terrorism etc.
| |
| 1 Some of the above studies are follow-up activities, others are newly formulated research..
| |
| ~
| |
| On the basis of the research rsaults computer codes for NPP safety analysis are improved for all possible transients and accident conditions. including beyond the design-basis conditions, and simulation systems and complexes are developed.
| |
| Studies are carried out on reactor design with passive safety 1 systems - t'ho so called inherently safe reactor design. Activit are intensified on probabilistic safety analysis, nuclear power 14 Appendix 6
| |
| | |
| c risk analysis, development of conceptual and methodological g
| |
| basis for radiation safety optimis.ation. Cont 4=nine efforts to improve the system of regulatory and technical documents are made within the framework of the generalized list of regu.
| |
| lations and standards development in the nuclear power area.
| |
| -- In order to accelerate the necessary research and to impri their quality the USSR develops the international cooperation -
| |
| within the frameAork of the general scientific and technical progress programme of CEA member-countries and IAEA programes and on the basis of bilateral agreements ud,th other countries.
| |
| The Soviet' Union har proposed the idea of the internationt
| |
| > regime of safe nuclear power development. It opens very large opportunities,which have not yet been fully used and which can form the basis of new relations in the international nuclear pcwor cooperation.
| |
| ~
| |
| REFERENCES I. AAauos E,0. , Acuenos B.P. , BacunescanR B.H. m Ap. HosmeHNe 6esonacHocTu ABC e PSMK.ATouHER sHeprNs,1987, T.62, sn.4,
| |
| : c. 2I9-226.
| |
| : 2. Acuosos B.P., Boposet A.A. , AeuwH B.G. x Ap. Asapus MS 9epHQ duabercR A3C: ron cnycts. ArowHaR sHeprus,1988, T.64, sun.I, c.3-23.
| |
| ~
| |
| : 3. HoHouapes-CrennoR H.H., Caecapes HX,. Be3onaCHoCTh N 3@@eRTH:
| |
| nocTb mepHon sHepreTHEN - OCHosa a p060 tax HSA peaMTopaMK '
| |
| HosO-
| |
| ~
| |
| re nonomeHus. ATousas sHeprxx,19' W , T.64, sun.I, c.28-29.
| |
| 4.~ Acwonos B.P. , Epuaron H.H. , Cngopetao B.A., Eax 0.H. "O pas-
| |
| ~
| |
| iiMTun Hay 4HuX McCneAosaHMR Ho 06ecneveHMC 683onaCHoCTN AC s
| |
| 'CCT Report of the 6th Meeting of 'the TK of Therpal Reactor Safety Ress-a 's:A, Vier.na, 9-12 June.1987 15 Apperidix 6
| |
| . .i
| |
| | |
| 2)IY3LOPCIT CF NUCI2A7. P077't PIJLNT SAFITY RESEARCH IN THE USSR Asmolov V.G., Irmakov N.I., Sidorenko V.A., Shakh 0.Ia.
| |
| l
| |
| * The prog ess in the development of any industry, including nuclear power, depends substantially on its se.fety for man and his environment. The most serious potential danger of nuclear power develop =ent is radioactive contamination of the biosphere.
| |
| Themanisoneoftheelementsofthebiosphere,wkchisthe *
| |
| +
| |
| most sensitive to such contamination. Radioecological data de-monstrate that the protection of man against radiation effects
| |
| ~
| |
| ensures usually a corresponding protection of the whole com= unity of flora and f auna. In the context of USSR nuclear power the cafe-ty usually means the quality of nuclear power plants which as a result of the following: \
| |
| - inherent design features:
| |
| . - special engineering means, and -
| |
| - organizational and technica} measures excludes the exceeding of the established internal and external radiation dose limits for'the plant personnel and ths population at a nuclear power plant,on its site and beyond the site at a distance prescribed by the regulations, as well as of norms and stand.ards fo$ radioactive substances concentrations in the envirors ment both during. normal operation and in the cese of any ace'ident.
| |
| Prom the very beginning of the construction of nuclear power plants and other nuclear facilities special attention was paid to the problen of safety. To solve the problem it was gensrally
| |
| - s neccesary to inplement a whole set of measures, wich may be cate- .
| |
| gorized as follows:
| |
| - establishment of goals and corresponding >.riteria (no=s) ofse.fetyi
| |
| - develop =ent of engineering safety means ensuring imple en-17 Appendix 6 1
| |
| | |
| n .
| |
| 2 tation of the adopted safety edteria; .
| |
| . . . - development of a wide range of administrative and organiza-tional: requirements and measures ensuring'i=plementation of the adopted safety criteria and functioning of engineering safety means. * ' '
| |
| To develop measures for the control of safety it is necessary to create effective accident prevention means, to be able to des-cribe accident sequences and to asses". its consequences, and to
| |
| ~
| |
| find optimum means of protection and mitigation (safety means).
| |
| It is necessary to model the whole range of physical processes; the modelling includert 14 development of computer codes
| |
| : 2) checking their credibility Af representative experimental devices; I
| |
| : 3) probabilistic safety analysis and risk assessment.
| |
| These principles form the basis of the USSR research program to ensure nuclear power plant'safet'y, that covers priority areas where further research and technical data are needed.
| |
| : 1. A range of hy'drodynsmic and heat exchange studies at 4 reactor facilities in the case of accidents. .
| |
| A comparative analysis of Soviet and foreign computer prog-rams for safety assessment showed that the levels of computer .
| |
| codes development are rather close as far as their coverage of safety problems is concerned. However the main Soviet codes use -
| |
| a homogeneous equilibrium model of two-phase flow (some of them take into account slip ratio and partial inequilibrium), and -
| |
| l modern foreign codes of ? improved assessment" are based on complete two-fluid inequilibrium models. Eodern se.fety approach requires codes of " improved assessment *.
| |
| 1 18 Appendix 6 (
| |
| | |
| 3
| |
| * Studies of thermophysic and hydrodynamic processes using integral . experimental rigs, which.are similar structurally, 1 hydrodynamically and thermophysically to real circulation loops l of nuclear power plants with different reactor types .are a 1' ~
| |
| necessary element of the modelling. ,
| |
| .The-USSR carries out such studies &f integral safety rigs w
| |
| thatmodel circulation aircuits of Wu and RBE reactors.
| |
| M Now new large-scale universal safety rigs of large espacity l are being erested, that correspond to modelling requirements of
| |
| - thermohydrolics in nuclear power plants circuits with WZR and RBE reactors, and that are oriented both at existing and future reactor facilities, and that are equipped with modern systems of . instrumentation and control, and of experimental data .collec- l l
| |
| tion and processing. * '
| |
| The rigs will be used for the following studies: -
| |
| thermohydrolic processes during transients and accidental
| |
| . e conditions, including accidents with a "small" leak from a circu-lation loop;
| |
| - effectiveness of emergency core cooling systemag
| |
| - departure from nucleate boiling (DNB) during transients that night occur at a nuclear power plant in case of a failure of one or more main circulation pumps or an ejection of control rods from the reactor core, etc.3
| |
| - heat exchange in fuel channels models with failed fuel elements in stationary conditions with DNB and post DEB;
| |
| - conditions of post-accident heat exchanges'
| |
| - diagnostics of processes that occur during accident con-ditions and transients:
| |
| - set of test experiments to confirm and improve computer programs.
| |
| 19 ,
| |
| Appendix 6
| |
| --------__._---___.l.-_--------- - - - - - - - - ^ - - -
| |
| | |
| l
| |
| : 2. Research in the aret of fusi elements and assenblies behaviour in accident conditions with the following main purposes
| |
| - tests of fuel elenents and assentlier behaviour in rig and reactor conditions to find the main factors influencing the heat transfer in accident conditions:
| |
| - detailed studies of physical and mechanical properties ,
| |
| of fuel element components.
| |
| To achive these, aims the following studies are suggested: .
| |
| : 1) reactor tests of .WER fuel assemblies and individual
| |
| ~
| |
| fuel elements in the conditions imitating accidents using renearch reactor loops to study fuel elements behavioul during loss of l I
| |
| coolant accidents ("large", " medium" and"small" LOCAs) and during 1 power surge accidents: .
| |
| : 2) rig tests of fuel bundis and individual fuel element imi- j tators of VVER and RBMK types during Lf)CA to study processes of
| |
| * f deformation, heating-up, oxidation, and influence of specific pa-rameters:
| |
| : 3) studies pf properties of fuel cladding structural materi-als and fuel matrix.
| |
| : 3. Aesearch of fuel behaviour and radioactive fissin products ;
| |
| release it ring severe eccidents with fuel melting and core damage.
| |
| One of th'e main problems related to severe nuclear power plant accidents is the behaviour of fuel elements and radioacti-vity release in case of fuel element failures including fuel melting. The quantity of radioactive products released from a fuel element is a central factor for the assessment of radiation consequences of such accidents.
| |
| Mathematical modeiling of the process is difficult due to complex physical phenomena occuring in such case. Suen phenomena 20 .
| |
| Appendix 6
| |
| | |
| . .- 5 include heat transfer conditions, fuel-cladding and cladding-coolant. interactions, kineties of cladding oxidation and rupture on of materials structure changes, interaction of molten fuel with stntetural materials, shieldings materials and the coolant. Direct experimental measurements of fission product releases in case of
| |
| , fuel melting are practically the only sourse to get credible information. - *
| |
| . In small-scale experiments using samples it is practically impossible to reproduce some processes, such as the reaction between steam and zirconium 'or interaction of molten fuel with structural materials, shielding materials and the ecolant, that are possible during a real-accident. These processes determine the total quantity of generated non-condensable gases that influ- I ences the pressure in the containment or in some other mitigation systen and the probability of their failure. ,. ,
| |
| Credible information based on studies of melting processes e
| |
| using a special experimental facility for large scale experiments
| |
| - will 'significantly influence the technical decisions and regula-t tions in the area of safety and, finally, the nuclear power plant economics and nuclear power development as a whole.
| |
| These include: -
| |
| t Regtiirements to emergency cooling systems effectiveness.
| |
| : 2. Requirements to mitigation syste u effectiveness.
| |
| : 3. Siting of different-purpose nuclear power plants around towns and. generally on the territory of the country. ,
| |
| i
| |
| : 4. Emergency planning to protect the public.
| |
| 1 Moreover it is planned to earry out experiments using 1& ops '
| |
| of recearch reactors to study fission produett. releases in simu-lated accide'nt conditions,to construct a special rig fare experi-mental research of physical and ehemical forms and compounds of 21 Appendix 6
| |
| | |
| u -
| |
| L h
| |
| e fission . products as a result of their interaction with reactor materials and for testing the effectiveness of emergency loca-lization and hitigation systems.
| |
| : 4. Problem of hydrogen at nuclear power plants.
| |
| L The presence of hydrogen in water cooled nuclear reactors l'
| |
| ig a reality that should be taken into, account. ,
| |
| In design-basis normal and accident conditions without i~
| |
| reactor core superheating (melting-down), that is necessary for -
| |
| L . chemical interaction between steca and reactor structural mate-rials, the presence of hydrogen (or explossve mixtures) is not usually dengerous and does not cause sequences. leading to a severe radiation accident.
| |
| . Hydrogen-related safety in desigh-basis events is ensured by design. An exemple of this approach is the design of RU AST-500 reactor facility for central heating. -
| |
| In a hypothetical accident, the generation of a large quan-tity of hydrogen may lead to very severe consequences that will determine-the nature of the accident sequences and the level of the dancge. In such cases even unter the assumption that it is impossible to mitigate hydrogen effects, the post-accident measures should take into account the possible consequences of. {
| |
| hydrogen behaviour. For this purpose in order to assess the acci-dent consequences it is necessary to study ,the whole range of ,
| |
| the hydrogen problems including hydrogen generation and ignition (explosion). ,
| |
| The following questions are ple.nned to be studied within the framework of the hydrogen problem research programa l
| |
| - hydrogen generation in radiolytic and chemical processes, ,
| |
| - hydrogen distribution in the reactor circuit and plant rooms, 22 Appendix 6
| |
| | |
| '7 ,
| |
| i
| |
| - characteristics of hydrogen burning,
| |
| - methods of hydrogen detection,
| |
| - methods of mitigation of hydrogen influene'e on accident development,
| |
| . - viability of safety systems sM equipment, I
| |
| - legislation and regulations development.
| |
| : 5. Reliability analysis of nuclear power plant systems and I
| |
| _ components, research on probabilistic safety analysis and risk assessment.
| |
| At present the improvement of nuclear power plant safety is connected with application ofprobabilistic safety analysis.
| |
| The analysis ensures a balanced approach to cafety decision making process during nuclear power plant design, operation and regulation. Its application permits: to analyze nuclear power plant safety; to identify yeak eldments in nuclear power plant systems, the reliability of which depends on a large number of components; to justify technical and organizational safety sea-sures; to determine reliability requirements for syster.s and equipment; to develop optimum procedures of their maintenance to train operators to determine preventive measures for hypo-thetical accidents: to set priorities and effectively use resour-ces to ensure safetyg to determine inspection activities; to justify regulations: to present a general quantitative assess-ment of nuclear power plant safety.
| |
| Disadvantages of probabilistic approach utiliza. tion are determined mainly by: insufficient knowledge of the plant per-sonnelbehaviourreliabilityinnormaloperationandsecjdent conditions; existing uncertainty in analysis results and diffi-culties in'their processing; incomplete modelling of accident processer incomplete data base on reliability of nuclear power
| |
| . 23 Appendix 6
| |
| | |
| - e ,
| |
| plant components; political and psychological particularities of probabilistic safety criteria application.
| |
| The proposed research is connected with the following studies:
| |
| a] determination of primary failure probabilities (probabi-lities of accident initiators) -
| |
| ~
| |
| b) development of possible accident sequences t.aking into account the operation of service safety systems c) determination of safety systems failure probabilities for given accident sequences:
| |
| * d) analysis of physical, hydrodynamic, mechanical, chemical and other processes during accidents e) determination of composition and quantities of radionuc-lides escaping the nuclear power plant; f) determination of radionuclides distribution beyound the nuclear power plent l ,
| |
| g) assessment of consequences for the population taking -
| |
| into account meteorological conditions and population distribution h) development of probabilistic indexes of nuclear power i
| |
| plant safety and choice of criteria values.
| |
| The in11owing studies are planned as well '
| |
| j
| |
| - develo'pment of scientific basis for safe control of technological processes at nuclear power plants: -
| |
| - development of technical means and equipment diagnosfic systemsg ,
| |
| I
| |
| - nuclear safety studies for fuel trnasportation
| |
| - studies'of processes and hydrodynamic effects in locali-zation systems during a:cidents with a rupture of primary circuit:
| |
| - development of norms and standarde for radiation, medical and ecological effects of nuclear power plants:
| |
| 24 Appendix 6
| |
| | |
| 9
| |
| -stredthandrupturedynamicsstudiesofreactorfacilities l structures and equipment.
| |
| Considerable attention will be paid to studies of hypothe-tical accidents with a possible destruction of protective barriers and fuel melting in order to assess thsir probabiMties and radiation consequences. The studies should contribute' to the de- '
| |
| ~
| |
| velopment of proposals on'the destgn of stand-by technical means
| |
| - for nuclear power plants aimed at mitigation of severe accident ,
| |
| 1 consequences and on m,rtmum possible release of radioactive substancec to be taken into account in nuclear power plant siting '
| |
| and development of population protection measures.
| |
| The implementation of the above program should permits- .
| |
| - to carry cut a comprehensive safety analysis of existing and future nuclear power facilities and to dustify noms and stan-
| |
| , dards in this area at a qualitatively new level;
| |
| - t'o create an experinantal basis with modern computers and instrumentation to solve problems of improved reactor facilities
| |
| ~ ~
| |
| with increased safetyg -
| |
| - to choose scientifically justified criteria for nuclear.
| |
| power plant., safety and to fomulate a national consept pn the issue. . . -
| |
| 25 Appendix 6
| |
| | |
| 1 APPENDIX 7 !
| |
| " Theoretical Analysis and. Numerical Simulation of Heat Transfer and Fuel Migration After Severe Accidents at Nuclear Power Plants," R. V. Arutyunyan et al. !
| |
| i 1
| |
| 1 i
| |
| 1 Appendix 7 1
| |
| | |
| Theoretical Analysis and Numerical 'Simulat. ion of Heat Transfer and Fuel Migration After Severe Accidents at Nuclear Power Plant.s Arutyunyan R. V. , Bolshov L. A. , Vityukov V. V. , Goloviznin B. M. , Dyikhne A. M. ,
| |
| Kiselev B. P. , Klementova C. B. , Krayushkin I. E. , Moskovchenko A. B. , Pismennyj B. D., Popkov A. G., Khoruzhij 0. V., Chernov C. Yu., Chudanov V. V., Yudin A. I.
| |
| To keep in step with.the latest American fashion, we will preface this pre- 1 sentation with a Russian proverb:
| |
| "P0KA GR0M NYE GRYANYET, MYZGIK NYE PEREKRESTITSYA," which translates to:
| |
| "A peasant won't cross himself until the thunder crashes."
| |
| American and Russian scientists engaged in stndying reactor safety problems
| |
| .have behaved like the peasant in the proverb: They were impelled to increase their studies by the Three Mile Island accident in the United States and the Chernobyl accident in the Sovfet Union.
| |
| In our theoretical and mathematical research we consider the following phy- 2 sical phenomena:
| |
| : 1. Nonsteady three-dimensional heat transfer with temperature heat capacities and conductivities:
| |
| : 2. Thermal effects of melting and evaporation.
| |
| : 3. Motion of dry substances in molten ones.
| |
| : 4. Dissolving of fuel in core, diffusion in liquid.
| |
| : 5. Thermal convection in liquid core.
| |
| : 6. Evaporation of volatile fission products due to fuel heating, and their adsorption upon the fuel cooling. .
| |
| : 7. Heat transfer by means of black-body radiation.
| |
| In testing our conclusions exp'erimentally, we simulated radioactive heat- 3 ing by cw-lasers of appropriate strength.
| |
| In the material that follows, we will illustrate the following physical 4 effects:
| |
| : 1. " Meltdown" or " China syndrome." the penetration of molten core through the basemat (sand, concrete, clay, and other materials);
| |
| : 2. Instability of the meltdown phenomenon; l
| |
| : 3. Adsorption-desorption wave, which leads to volatile substances rising through dry solids;
| |
| ~
| |
| : 4. Heat-exchanger that v prevent molten core from penetrating the ground-water and the design oY such a heat-exchanger.
| |
| Note: Numbers in the right margin indicate page number in Russian text.
| |
| 1 Appendix 7 i
| |
| | |
| p . . -
| |
| ^
| |
| p; h
| |
| f ' Meltdown 5 Meltdown occurs only if the fuel heats up beyond.a certain critical value. In the first several frames of the movie on the TV screen we see a small black rectangle. This is a piece of fuel that is not moving because its temperature is not high enough to melt the surrounding material.
| |
| Next you see on the screen what happens when the temperature is high enough 6 to melt the surround-ing material. Here the fuel begins to sink in the molten material. After several hours, a steady picture of the meltdown wave has been established. How fast this wave moves depends on how much fuel is involved. -
| |
| The yellow color represents liquid fuel, the red represents molten sand.
| |
| t the bottom of the screen you can see the temperature scale in degrees Kelvin.
| |
| On one axis, numbers of days, hours, and minutes are represented; the other axis shows distance the fuel has moved (in centimeters).
| |
| Next on the screen is a demonstration of the case in which the fuel melts in 7 a liquid environment. ' Our experiments illustrate that the fuel does not dis-solve in a pure molten sand, but that it dissolves very quickly in sand that contains 5 percent clay. In this case the meltdown velocity slows.
| |
| Nextweseeillustratedhowtwodropsoffueljointogetherasaresultof 8 thermal effects. The space between the drops is hotter than the outer regions.
| |
| This leads to individual drops of fuel joining together. This phenomenon is demonstrated next on the screen.
| |
| If the drops are of different. sizes, such confluence is accompanied by stream-ing of molten fuel from one drop to another. This phenomenon implies excita-tion of rapid motion of fuel resulting from thermal effects.
| |
| Instability of Meltdown Phenomenon 9 Melting may occur in a'n unstable way with respect to space inhomogeneities.
| |
| Small differences in depth of the molten layer lead to some displacement of fuel; this, in turn, causes increases in the heat source inhomogeneity and greater modulations in the depth of the molten layer.
| |
| Such instability may lead to drops of fael breaking away from the fuel layer, andinsomecases,joininglateron.
| |
| Both these possibilities are illustrated on the screen.
| |
| Volatile Substances Rising Through Dry Substances 10 In addition to fuel, fission products can be present in the ground or sand.
| |
| Their atoms may be adsorbed by sand particles. When the temperature rises, .
| |
| they may be desorbed and form a volatile fraction. If there is a hot area in a medium, an upward flow of air occurs (Archimedes' principle). The fuel ,
| |
| layer may also move up because:
| |
| : 1. A volatile fraction rises up when the medium is heated. !
| |
| : 2. The volatile component moves to a cooler area because of filtration flow.
| |
| : 3. Volatile components are adsorbed and accumulate in this area.
| |
| l 2- Appendix 7
| |
| | |
| j,
| |
| : 4. As a result of consequent self-heating, the precedihg three steps keep occurring.
| |
| Next.on the screen we can see what happens when meltdowr, and volatile rising occur simultaneously.
| |
| Designing a Heat Exchanger That can Prevent Molten' Core From Penetrating 11-Groundwatr; The next part of- the presentation relates to the design of a heat exchanger that prevents the fuel from moving down.
| |
| On the screen you will see the interaction between the melting stream ar.d the
| |
| , periodically arranged water-cooled tubes. If the distance between the tubes i is equal to (or more than) the tube diameter, the fuel penetrates into the space between the tubes and finally flows together as a united whole.
| |
| 12 If the distance between the tubes is less than the tube diameter, the melting wave may be stopped before the cooling tubes.
| |
| Next on the screen we see the fuel stream stop. After several hours the steady-state configuration of fuel is established.
| |
| The next frames demonstrate the fuel sinking onto the surface of the 13 water-cooled heat exchanger.
| |
| If all the fuel is failing simultaneously, it takes about two months for the fuel to lose all its heat. We tested the final design of such a heat exchanger by laser simulation. In the design, we took into account the possibility of
| |
| " thermal blow" should a large mass of fuel fall suddenly on the device. The simulation indicated that such an accident could be avoided by covering the heat exchanger with graphite.
| |
| t 3 Appendix 7 a
| |
| _ _ _ _ _ _ _ _ . _ _ . _ _ . _ _ . _ _ . m _
| |
| | |
| Figures Fig.1 - 3 The numerical modeling of the fuel flat layer motion instability.
| |
| The zone'of the solid and melted fuel is on the left side.
| |
| The temperature isotherm is on the right.
| |
| Tmin = 0'C, Tmin = 1600*C, X
| |
| max
| |
| = 28,cm, Y max
| |
| = 35 cm, time = 800 min Fig. 4 - 6 The confluence modeling of the two cubical fuel fragments with -
| |
| 10 on edge, the calculations have been made for 3 dimensions.
| |
| The' capacity of energy liberated is q = 0,5 wt/cm 3 The total confluence time is 25 hours. ,
| |
| Fig. 7 - 8 The promotion of. the fuel flat layer through the periodically arranged watercooled tubes. -The rise of the region X x Y is 1mx1m.
| |
| 1' 4 ,
| |
| Appendix '
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| 10 Appendix 7
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| TPA6WK AWWWR @ ,
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| TEMnEPSTYPR =CONST K0AW4ECTBD YPD8HER- 1 l FMIN= 3.199821 wE+ 2 FMAX= 3.160931 NE+ 3
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| FPA9WX AHHHR b TEMDEPATYPR =CONST K0AH4ECT80 YP08HER- 1 FNIN= 3.199910 wE+ 2
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| l-
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| | |
| I l-APPENDIX 8 l
| |
| l Figure 1: Measurement of the Heat Output of the VVER-1000 Reactor From the Flow Rate of the Coolant in the Primary Loop by the Rediation Method Figure 2: Azimuthal Flux Distribution I
| |
| 4 Appendix 8
| |
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| i APPENDIX 9
| |
| -1 4
| |
| Proposed Topics for Discussion in the Erosion / Corrosion Area l
| |
| Appendix 9
| |
| | |
| Proposed' Items for Discussion at the First Meeting of the Erosion-Corrosion Forking Group on November 15, 1986 '
| |
| : 1. A plan shall be developed for a workshop on the suhject of erosion-corrosion to be held in the U.S. A. in April /May 1989. Enclosure-1 proposed topics for-this workshop that fall into two broad catagories; (1) power plant experiences cnd engineering considerations and (2) research and development activities. This workshop is expected to be of two or three days duration.
| |
| : 2. A plan shall be developed for further cooperation for the per'3d of 1989-1991.- This plan shall include exchange of information, joint workshops, joint studies and plant / laboratory visitations.
| |
| Enclosure 2 proposed topics for consideration for this plan.
| |
| l l
| |
| 1 1 Appendix 9
| |
| -a-_-___________
| |
| | |
| )
| |
| {
| |
| Enclosure 1 i
| |
| (
| |
| Proposed Topics for US/ USSR Workshop on Erosion-Corrosion to be Held April /May 1989 ,
| |
| I. Power Plant Experiences and Engineering Considerations -
| |
| t 1
| |
| a) Codes, standards and general practices for design, materials, q water chemistry and inspection applied to the construction of piping
| |
| ~
| |
| j systems for nuclear power plants. !
| |
| b) History of pipe wall thinning in nuclear power pl:,nt piping including experiences of pipe thinning in both liquid phase and liquid vapor phase lines.
| |
| c) Application of computer code to identify inspection locations.
| |
| d) Verification and prevention of erosion-corrosion in nuclear power plants.
| |
| e) Remedial action to address pipe wall thinning issue including control of degradation, repairs and replacements, criteria for structural evaluation of deteriorated pipe.
| |
| f) Regulatory efforts to address pipe wall thinning.
| |
| II. Research and Development Activities
| |
| ~
| |
| a) Review of major variables that control erosion-corrosion.
| |
| b) Phenomenological aspects of erosion-corrosion.
| |
| ~
| |
| c) Predictive models for erosion-corrosion in piping systems and computer code development.
| |
| d) Feedwater system design and operating characteristics that relate to potential erosion-corrosion induced wall thinning, e) Wate'r chemistry practices in nuclear power plants and their effects on pipe wall thinning.
| |
| 2 Appendix 9
| |
| | |
| f) Inspection Procedures for locating, characterizing 6nd verifying. existence of erosion-corrosion in piping of nuclear power _ plants.
| |
| g) Methods of preventing erosion corrosion.'
| |
| h) Analytical methods for establishing the acceptable-level of wall thinning in piping of nuclear power plants.
| |
| ^
| |
| : 1) ~ Flow Considerations in erosion-corrosion of pipe.
| |
| I e-3 Appendix 9 i
| |
| | |
| v _ ._. -_ ._- _ _ - _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ - - _ _ _ _ _ _ - _ _ _ _ _ - _ _ - _ _ _ _ _ . . - _ _ _ . _ - - _ _ _
| |
| 4 Enclosure 2 .
| |
| 4 Proposed Topics for US-USSR Work Plan for 1989-1991 a) Exchange of information on prediction methods for detemining piping degradation through stress corrosion cracking including primary causes, analytical modeling, experimental verification, nondestructive examination,'and prevention.
| |
| b) Exchange of infomation on prediction methods for determining steam generator degradation including primary causes, analytical modeling, experimental verification, nondestructive examination
| |
| ~
| |
| procedures, and prevention.
| |
| c)- Exchange of information on the development of new, alternate materials for power plant piping and steam generator tubing.
| |
| d) Exchange of information on corrosion and stress corrosion cracking of zirconium alloy fuel cladding.
| |
| e) Exchange of information on water chemistry controls including -
| |
| on-line monitoring for mitigating stress corrosion cracking, erosion-cortosion, and transport of radioactive products.
| |
| f) Exchange of information on the mechanisms of corrosion at the crack tip of material defects and effects on corrosion rates and mechanical properties. -
| |
| g) Exchange of infor1 nation on the development of on-line procedures -
| |
| and systems for control of stress corrosion cracking in nuclear power plants.
| |
| 4 Appendix 9
| |
| | |
| n .
| |
| Topics for cooperation within the WG 10 J
| |
| - Exchange of information on: Rzchan6e of information and specialists ons a) Prediction methods for a) Methods for determining piping degradation due to determining steam genera-'
| |
| tor degradation including sirens corrosion cracking primary causes, analytical including primary causes, modeling, experimental ve- analytical modeling, experi-rification, nondestructive mental verification, nondes' examination procedures, ructive e?=4 nation, and and prevention, prevention.
| |
| b) The development of new, b) Corrosion and stress corro-alternate materials for sion cracking of zirconium power plant piping and alloy fuel cladding.
| |
| steam generator tubing. c) Water chemistry control c) The mechanis:na of corrosion including on-line monitoring at the crack tip of mate- for mitigating stress corro-rial defects and effects sion cracking, erosion-corr (
| |
| on corrosion rates'and sion, and transport of
| |
| ' mechanical properties. radioactive products.
| |
| d) The development of on-line methods and systems for control of stress corrosion cracking in n:1 clear power plants.
| |
| 9 5 Appendix 9
| |
| | |
| ['-
| |
| l i
| |
| List of participants on behalf of the USSR l.
| |
| i l'
| |
| l.
| |
| Organizations tress for cooperation g Research and Development 2 a,b,o des,f,g Institute of Poirer Engineerlag Institute of Ateedo Energy 2 a,bedef 1
| |
| l' I
| |
| All-Union Research Institute 2 a, bed e of Nuclear Fourer Plants ,
| |
| All-Union Research and F-4y 2 be sg e
| |
| Institute of Power foobaology 1
| |
| Central Remoarch Institute of 2 a,becg structural Materials 'Prametheus" l
| |
| l Research and Psoduction Associa- 2 becf ;
| |
| tion Central Researth Institute of Heavy Engineering Research Institute of Euclear 2 a,e Engineering h t2W b Se.hM k YY t -
| |
| K L;f- <^me a'& f" 7afw/FET~/S$l-g ,
| |
| 1 I 6 Appendix 9 j
| |
| | |
| ~
| |
| 1 .
| |
| l l
| |
| l.
| |
| APPENDIX 10 Abstract: " Hydrodynamics and Heat and Mass Transfer When
| |
| * Adding Surfactant," G. A. Filippov et al.
| |
| I Appendix 10
| |
| | |
| Hydrodynamic and Heat and !! ass Transfor When Adding Surfactant Filippov 0 4., Saltanov G.A.,
| |
| Kukushkin A.N.
| |
| Abstract The investigation results of the influence of surfactant microadditives on single- and two-phaco medium hydrodynantics, boiling and condenantion heat and masa trtinsfer procesces, i
| |
| Physico-technical processos and crosion-corrocion wear obtained by a number of the scientific groups in the USSR (VHIIAM,MEI, NPO CKTJ, the Kol'skaya nucient poirer plant, etc.) and in GDR (Institut fur Enerced u, Institut fur lierkatofforschung' der Akademie der Wiscenschaften der DDR. Institut fur Encreetik don VE Kombinates KKlf "Druno Leuscluier", etc.) are procented in
| |
| 'the study. The inves tigntioita nre directed to solving the importruit practical problema colatenteel with nn incrence in rollability, lifetime and efficiency of the power-cencratin6 equipment, the nuclear power plant stenm-water equipment, in the first place.
| |
| The inventications were carried out in the largo and diverse test unit complex (steam dynamic tunnel, test steam generaters, model conder.sers, test turbines, e to.). Larco-scale field testing data on a number of the installations beinc in use (the Dimitrei s tean utetion, the Kol'skaya nuclear poner plant, KKW "Bruno Leuschnc/)
| |
| . are given.
| |
| The effects, such as
| |
| - steam-liquid flow diaparsion and increase in wet atoam turbine efficiency
| |
| - sharp decrease in the pipe-line and power-cenerating equipment erosion-corrosion wear intensit,y (droplet-impincoment 1 Appendix 10 l
| |
| | |
| 2'.
| |
| nroolon, cavitation eroolon, e tc. );
| |
| - the heat transfer ourface noter-rape 11ency treatment and protection due to monomnieculer film formation;
| |
| .Intenalve doorecee in che corrnsion procesn of different types (oxy 6en and carbonic ecid corroaton, stresa corrcaton cracking, etc.):
| |
| - intensive heat transfer surface decontamination of vnricus .
| |
| impuritica, including corrosion-bacardnble ones;
| |
| - the ability of effective controlling ::he nonntatinunry 1
| |
| procennes in the spontaneously condensible and wet nteam flows ,
| |
| haveheen rovenled and confirmed convincingly.
| |
| Thorough resulto in n now field of hydrodynamic, heat and mans trnnnfer, erneion-corrosion near', a number of which was used in developing the new effectiva technolo61 cucf incroase in the nuclear power ninnt and sterua ntntion equipment ro11 ability and efficiency on the basin of film-forming surfactant (octadecylamino) microadditives of very email amounts (concentration ~10-6), wore obtelved.
| |
| Theco data may be used in the other brnnches of traditionni and new oncineering, as well.
| |
| Publishing Of fice "Energoatomizdat",1980.
| |
| 2 Appendix 10
| |
| | |
| l 4
| |
| l I
| |
| APPENDIX 11 Three Fire Safety Documents Submitted by the Soviets to the Americans:
| |
| (1) "Providing for the Fire and Explosion Safety of Nuclear Power Plants in Connection with the Formation and Accumulation of Hydrogen" (proprietary)
| |
| (2) " Fire Danger of Electrical Cables" (proprietary) ,
| |
| (3) " Occupational Safety Standards System. Fire and Explosion Hazard of Substances and Materials. Nomenclature of Indices and Methods of Their Determination" Appendix 11
| |
| | |
| PROVIDING FOR THE FIRE AND EXPLOSION SAFETY OF NUCLEAR POWER PLANTS IN CONNECTION WITH THE FORMATION AND ACCUMULATION OF HYDROGEN
| |
| /1 The formation of hydrogen in reactor equipment of nuc1' ar e
| |
| power plants leads to the fire and explosion danger of the corresponding equipment.
| |
| Different features of this danger may be shown schematically as follows (Figure 1).
| |
| : 1. Generation of hydrogen in a normal mode, measures to suppress radiolysis.
| |
| : 2. Propagation of hydrogen in a normal mode.
| |
| : 3. Incidents leading to the discharge and escape of hydrogen and the heat carrier.
| |
| : 4. Formation of hydrogen in metal-vapor react. ions during eme'~3encies.
| |
| : 5. Escape of hydrogen within the containment.
| |
| : 6. Formation of combustion mixtures (concentration limits of the flame and detonation propagation)
| |
| ~
| |
| : 7. Af teref fects of the combustion processes.
| |
| ~
| |
| The direction of research in the field of the hydrogen safety of nuclear power plants may be seen schematically as follows (Figure 2):
| |
| : 1. Control of the ga's mode.
| |
| : 2. Control-of the hydrogen content.
| |
| Numbers in right margin indicate pagination in original text.
| |
| 1 Appendix 11
| |
| : 3. Methods making it possible to avoid the accumulation of hydrogen.
| |
| : 4. Propagation of hydrogen over the reactor circuits and within the protective shells.
| |
| I
| |
| ~
| |
| : 5. Removal of hydrogen. .
| |
| l
| |
| : 6. Methods making it possible to avoid detonation. .
| |
| : 7. Development of safety criteria and the corresponding standards.
| |
| /2 Certain results have been obtained in the USSR in these areas.
| |
| 'The concentration limits have been established for the propagation of flames in mi::tures of the hydrogen-oxygen (a ir ) --
| |
| diluent type at room temperature and atmospheric pressure. The results are shown in Figures 3 and 4. A comparison of the stabilization curve of the hydrogen-oxygen mixture with nitrogen with the literature data shows a good agreement in the results, which points to the relf. ability of the values obtained for the concentration limits of the flame propagation.
| |
| It was found that the fuel concentration at the lower branch of the stabilization curve C p in the case of dilution with chem-ically inert stabilizers is described by the relation
| |
| ' III j- # [
| |
| where C H --I wer concentration limit of the flame propagation of the combustion substance (hydrogen) in air, % (volumetric) ;
| |
| y--quantity characterizing the stabilization capacity of the diluent; C --concentration of the diluent in the mixture, % (volume-
| |
| ^4 tric).
| |
| 2 Appendix 11
| |
| | |
| r ]
| |
| i
| |
| .. l For nitrogen and carbon dioxide the value y may be calculated according to the formula
| |
| . y= .(H;-4}/(dwj) .- m .
| |
| where'H,, HB --absolute molar enthalpies of the stabilizer and air.
| |
| The indices and ' refer to the adiabatic combustion temperature
| |
| .and the initial temperature of the fuel mixture. i
| |
| )
| |
| 1 The concentration limits of the flame pr6pagation in a H +0 +C F Br mixture have certain characteristics (Figure 5) .
| |
| 2 2 24 2 It is known that vapors of 1,2--dibromtetrafluorethane burn in oxygen in a tube with a diameter of 0.05 meters and a height of 1.5 meters with the propagation of the flame from below upwards.
| |
| Th'us, the Le-Shatel' rule is satisfied
| |
| ~
| |
| [p; * /3
| |
| ,4 where Cg --concentration o,f the ith combustion component of the mixture; ,
| |
| -Ogj --its lower (upper) concentration limit of flame propa-gation.
| |
| Segments of the lines connecting the flame propagation limits must correspond to relation (3) on the plane of C p and C, (Figure 5). As follows from Figure 5, there is a great infringement of the Le-Shatel' law, which must be considered in practice.
| |
| As ,a rule, harmful mixtures (one exception is the case of stabilization with argon and helium, where the mixtures are stoichiometric) correspond to the stabilization points. This
| |
| ~
| |
| phenomenon is caused by the bubble nature of hydrogen combustion, which is caused by a much higher hydrogen diffusion coefficient in air as compared with ' gases such as oxygen, nit,rogen, carbon dioxide, chlorpentafluorethane and 1,2--dibromtetrafluoret,hane.
| |
| Thus, the flame front is not solid, but represents a set of small fock.
| |
| 3 Appendix 11
| |
| | |
| O The magnitude of the excess coefficient of the oxidant a for the stabilization points is given by the following '
| |
| a (DD4 ~Ci
| |
| * Co8- (4) c,. reap .
| |
| where B--stoichiometric coefficient of oxygen in the combustion reaction (for hydrogen 8 = 0.5); C --c neentration of oxygen in O
| |
| the oxidizing medium n (volumetric)2 ,
| |
| It is found that for the mixtures studied the value of a
| |
| ~
| |
| - may be described by the electric expression where
| |
| /s {00 # { $6 x (5) g,,
| |
| @, *b at b* d (6)
| |
| Q 40 1.g at 12 --hydrogen diffusion coef ficient in air; 13 "-diluent diffusion coefficient in air.
| |
| From (1) and (4) we obthin the expressions describing the /4 concentrations of the diluent and the fuel at tne stabilization point C, and C7 :
| |
| S VOO/t**>) .
| |
| (7) 4 " o(00*Ser a ((f l/gy4)cjeo](w,sroo/a,)
| |
| (8)
| |
| /OS-[e.
| |
| ['a rWp roo/[,,
| |
| According to formulas (7) and (8), the calculation accuracy is suitable for practice. ,
| |
| To provide fo.r the hydrogen safety of nuclear power plants, it is of great interest to study the concen'tration limits of
| |
| ; flame propagation in a mixture of an explosive gas with water vapor at high temperatures (up to 200*C) and pressures (up to l 4 Appendix 11 l
| |
| l-L___.__ _ _ _ _ . . __ _ - . - _ . - _ _ _ _ _ ~ . _ m - -.
| |
| | |
| . i I
| |
| The experiments were carried out in a reaction vessel j 2.0 MPa).
| |
| 3
| |
| . with an in?.ernal diameter of 200 m and a total volume of 16 dm .
| |
| The combustion of the mixture was done by burning a nichrome wire with a voltage of 24 volts and a combustion energy of about 2 J.
| |
| The flame propagation was recorded using pressure sensors DDI-2I with water cooling, and with subsequent amplification of the ID-2I amplifier signal and recording on a light beam oscillograph H-ll5.
| |
| The reaction mixture was prepared at partial pressure by pumping individual components of the mixture into a previously evacuated reaction vessel. The mixing quality was satisfactory.
| |
| Figure 6 shows the results of determining the lower concen-
| |
| . tration limit of the flame propagation (LCFP) of the explosion
| |
| ; gas (2H 2 +0 2 ) in a mixture with a heated water vapor at a temperature of 200*C. There was practically a rnonotonic decrease of LCFP with a pressure increase, although the absolute change in the limit was relatively small. Figure 7 shows the corresponding dependence for saturated water vapor. Attention should be called to the higher values of LCFP for saturated vapor as compared with heated. Thus, the value of LCFP for 100'C and a pressure of 0.15 MPa is 16.7 i /5 0.6% (ma s s) , and at 200*C and 20.5 MPa 12.7 1 0.5% (ma ss) . A change of LCFP with temperature at a fixed initial pressure 0.4 MPa was relatively small and was 0.2% (mass) for each 10*C.
| |
| In order to provide for explosion safety as a result of the formation and accumulation of hydrogen, it is of interest to study the critical conditions for the formation and development of detonation.
| |
| Numerical modeling for the propagation of a detonation wave was performed in gas volumes of a finite width in order to deter-
| |
| ~
| |
| mine the critical width of the gas layer. On the one hand, the gas layer bordered a solid non-heat conducting wall, and on the other hand--a chemically inert gas medium. The competition between the heating of the combustion nixture in the shock wave and its cooling due to' lateral loading Jed to the formation of a critical width of the gas layer with respect to the stationary or quasi-5 Appendix 11
| |
| | |
| stationary (pulsed mode) propagation of the detonation wave. The experimental determination of the critical width undoubtedly gives the most reliable values; however, such experiments are very difficult and take a long period of time. Computational methods are more accessible, among which the most reliable are the numerical modeling of the detonation propagation in a gas layer.
| |
| The critical widths of the gas layers were calculated for the mixtures 2H +0 , 29% H +71% air, 23% H +77% air, 4H 0*
| |
| 2 2 2 2 2 2 Two dimensional equations of gas dynamics in the LaGrange -
| |
| notation were used to model the detonation wave propagation.
| |
| These were supplemented by the equation of chemical kinetics in Arrhenius form. The parameters in the Arrhenius formula were selected so as to correctly describe experimental dependence of the induction period for the burning of the combustion mixture on temperature. The ef fective ' specific heat liberation was selected as compared with the calculated and experimental values of the detonation wave velocity for a large (much more than the critical) width of the gas layer.
| |
| /6 As a result of the calculations, the following values were obtained for the critical width of the gas layer.6:
| |
| 2H2+0 2 1.35 ( 6 kp 4 1.80 cm; 29%H 2
| |
| +71% air 9.0 < 6 kp ( 13.5 cm;
| |
| - 23%H2 +77% air 27.0 ( 6 kp ( 36.0 cm; 4H2 +0 2 4.5 ( 6 kp 4 6.8 cm.
| |
| Many studies attempt to connect the limiting characteristics of a detonation wave (critical initiation energy, critical -
| |
| diameter, and dimension of a rectangular channel) with the {
| |
| dimension of the detonation cell A . The question arises of l whether there is a linear dependence between A and the critical width 6 kp just as was observed for the critical diameter. It was found that within the accuracy limits of calculations in this study, this inter-relation holds with a proportionality coefficient
| |
| ~
| |
| 6 Appendix 11 l
| |
| h_mhm______.___.m__ m. ._.m._. _ _ _ m .n-.
| |
| | |
| from'6 to 12.
| |
| - A' comparison of the calculated. values of 6 kp with the experimental values for a 2H 2+02 shows the following. The experimental value of 6 kp for this mixture is no less than 2.5 cm, which greatly exceeds the calculated values '(1.35 4 6 kp 4 1.80 cm), i.e., the use of the calculated values provides a certain reliability reserve in problems of explosion safety.
| |
| /7 Reducing the fire and explosion safety of gas mixtures con-taining hydrogen is of great practical interest, using the stabili-zation methed, including haloid-containing compounds which have an inhibiting action. As noted alove, in the case of the combustion of a hydrogen-oxygen mixture, one of the most effective explosion substances for combustion in air (1,2--dibrometrafluorethane) does not give the necessary effect. It was found that in the case of stabilization with this cold mixture of hydrogen with air its consumption is much greater than in the case of stabilization of hydrocarbon-air mixtures. In order to determine how to increase the effectiveness of the stabilization agents several studies were carried out on the mechanism of oxidation and inhibition of hydrogen combustion with coolants. During these studies it was found that the mechanism of hydrogen oxidation at atmospheric pressure differs greatly from that which was previously assumed.
| |
| A mechanism was proposed for the hydrogen oxidation (Figure 8) in which a mode was assumed of the degenerate branches by H0 Considering these concepts, the partial 22 decomposition.
| |
| combustion of hydrogen with coolants may be explained by the reactions H02+HBr + H220 +Br and Br+H2 + HBr+H, as the result of which atomic hydrogen is regenerated and hydrogen percxide is formed, which readily decomposes into active hydroxyl radicals.
| |
| Apparently for, harmful mixtures of hydrogen with air, a mechanism is characteristic with straight chains, and this may explain the small influence of the coolant on their combustion (the lower branches of the stabilization curves). Therefore, to increase, 7 Appendix 11
| |
| | |
| the effectiveness of coolants, it was proposed to carry ouc the inhibition at the same time as there was a decrease in the oxygen
| |
| ' content in the combustion mixture (i.e., to form mixtures with a more abundant combustion component).
| |
| Figures 9 and 10 show the results of' experiments on the stabilization of mixtures of hydrogen with air with the simultan-
| |
| ~
| |
| eous introduction of an inert diluent (nitrogen, carbon dioxide).
| |
| It can be seen, that with a decrease in the oxygen content, the
| |
| " peak" concentration of the coolant sharply decreases. On the /8 -
| |
| basis of these studies, combined compositions for stabilization and volmetric quenching were proposed. The results of tests of one of them (95% nitrogen, 114B2-5% cooling agent) are shown in Figure 11, from which it dan be seen that the coolant consumption may be one order of magnitude lower with the introduction of a moderate amount of nitrogen.
| |
| In conclusion, we give data,on stabilization.and quenching of a diffusion fleme of hydrogen by supplying nitrogen to a tube with hydrogen. At flow rates up to 10 m s -1 (flame separation was observed at 140 m s~1) the flame quenching was achieved with a ten-fold dilution (theoretically an 18-fold dilution is necessary).
| |
| 1 8 Appendix 11
| |
| _ _ _ _ _ _ _ ____ _ _ _ _ i
| |
| | |
| FIRE DANGER OF ELECTRICAL CABLES In designing the AES cable system, the number of combustible / ly .
| |
| materials must be reduced-to the' minimum; in this case it is neces-sary to use fire-resistant electrical cables that do not spread combustion. . Cables are of great national economic value considering their high cost of production and laying. Fires in cable structures
| |
| , are a danger not only for the cables themselves and these struc-tures, but also for all' units connected to them and the AES as a whole.
| |
| The material damage from cable fires, in ad.dition to the direct
| |
| -losses related to the cost of the burnt cables and destroyed rooms and units, is determined a great deal by the indirect damage from underproduction of products due to the interruption in electricity, supply, as well as underdistribution of electricity by the power-plants.-
| |
| Prevention of fire, as well as limiting itr spread,are'imple-mented by affecting the combustible system and the heat sources to exclude conditions under which outbreak of combustion is possible.
| |
| In discussing the fire danger of cable lines, we have to keep in mind that they include the fuel system and the heat source.
| |
| Comprehensive solution to be problem of fire protection for cables of specific facilities is a complicated problem; its solution requires detailed analysis of the possjble dangerous situations.
| |
| Fires for electrical engineering reasons are prevented both in our country and abroad in different ways.
| |
| One of the primary trends is fire prevention work. /2 The VNIIPO { All-Union Scientific Rescarch Institute of Fire Danger] has developed a technique for determining the Numbers in right margin indicate pagination in original text.
| |
| 1 9 Appendix 11
| |
| | |
| probability of fire breaking'out from a cable
| |
| , and wires of electrical circuits. It covers electrical wires laid in pipes, in troughs, in boxes, on insulating supports and cable lines. The basis of the technique is determining the probable emergence of ignition and the probable spread of fire through the combustible material of the cables and wires.
| |
| The probable outbreak of ignition is determined on the basis -
| |
| of data on.the op5 rating reliability of the electrical equipment, electrical protection and cables (wires). The probable spread of .
| |
| fire over the route is determined depending on the combustion heat of the cables and wires, the volume occupied by them and the arrangement. The method of laying the cables and wires is con-sidered to have satisfied the fire safety requirements if the value of probable emergence of fire from the electrical lines (cable lines) per year does not exceed 1 x 10 -6 ,
| |
| The probable emergence of fire is determined by the following expression: ,
| |
| Gn 03. pod where O p is the probability of fire spreading, Q 3 is the probability of electrical network ignition, O is the probability of extinguishing cables and wires by T3 the automatic fire extinguisher, h3''hN*OB The probability of electrical network ignition is determined by defining the probable appearance of an ignition s,ource (Og) ,
| |
| and the probability of inflammation of the combustible material when an ignition source falls on it (Q B). /3 l l
| |
| hu = I - fl (l.- hyg)'
| |
| 2 10 Appendix 11 i
| |
| MNM%
| |
| | |
| l 1
| |
| 1 where Q HK is the probable appearance of an ignition source in a group
| |
| ' lay'ing made of n cables from the K-th cable.
| |
| The probability (Qg) depends on the group of combustibility of the material or is determined experimentally. -
| |
| Definition of (OHK) takes into consideration the probable emergence of emergency dangerous fire conditions in each cable (wire), the probable presence of a dangerous fire range for the characteristic dangerous fire factor, the probability of igniting particles falling into'the combustible medium of the facility, the probability of the electrical protection malfunctioning and the possible malfunctioning of the user (cable load).
| |
| Calculation of the probable spread of fire through the group of cables and wires (Qp ) starts from the combustion heat of the combustible material contained.in a unit of length of the laying (W). that is defined from the expression:
| |
| n-W = 1 Wi.
| |
| L=i where W g is the combustion heat of one running metor of cable of the i-th type size, n is the tota'l number of cables in the laying.
| |
| In addition, the numerical value of the volume (V) occupied by the group laying of length equal to 1 m is defined.
| |
| The presence or absence of fire spread is determined from the specific combustion heat of the cable laying (J) that is computed from the formula:
| |
| J=
| |
| The resulting value of combustion heat (J) is compared with the permissible for vertical or horizontal arrangement of' the cables.
| |
| 3 11 Appendix 11 ,
| |
| | |
| 1 l
| |
| If the calculated value of combustion heat is greater than the 44 l permissible, then additional fire prevention measures must be taken.
| |
| 1 Studies made in recent years in the Soviet Union and abroad to evaluate the fire danger of electrical cables have made it possible to develop general re'quirements for cables in fire-resistance. The i
| |
| main indicators characteri=ing cable behavior in a fire have been l established: l l
| |
| --fire resistance,
| |
| --nonspreading of fire, -
| |
| --optical density of smoke formation,
| |
| --corrosion activity of gas-release products,
| |
| --toxicity of gar-release products.
| |
| ]
| |
| Requirements are made for individual indicators or together for the cable depending on the application conditions. Higher fire safety requirements are made for cables used'in AES, including all the aforementioned indicators.
| |
| i i
| |
| We will discuss some of them. Nonspreading of combustion characterizes the capacity of the electrical cable to extinguish itself after there is no longer an open flame. The publication of the International Electrotechnical Commission 332-3/1962 is used in the USSR to test cables for nonspreading cf combustion.
| |
| Fire resistance characterizes preservation of the efficiency of an electrical cable under the influence of an open flame source for a set time. Fire-resistant cables are special purpose cables that are used in especially important electrical circuits, f or -
| |
| example, in the AES safety system circuits. In particular, the main parameters that can be controlled in a fire for fire resistance of power and control cables are permissible leakage current through [5 the insulation and preservation of electrical strength of the insulation. We use the method presented in IEC[ International Electrotechnical Commission) publication 331/1970 to test cables for fire resis,tance.
| |
| 4 12 Appendix 11, A
| |
| | |
| The characteristics of fire safety of electrical cables are mainly governed by the properties of the insulation materials ,
| |
| and to a lesser degree, by the design of the cables, although the-indicators of nonspreading of combustion and fire resistance are intimately linked both to the properties of the materialn and to ,
| |
| the design of the cable. The nomenclature of the fire danger indicators of materials and the methods of de'termining them are
| |
| - mainly similar both in the USSR and abroad, however there are some differences in terminology. In the USSR, the nomenclature of the fire da'nger indicators of substances and materials, and the methods of determining them are established in GOST 12.1.044-84. Such indicators as combustibility, oxygen index, smoke formation coef-ficient, as well as toxicity of the combustion products should be isolated for polymer materials used also for insulation and shells tof cables.
| |
| Acc,ording to GOST 12.1-044-84, combustibility characterizes the espacity of material for combustion. Materials in the USSR are divided by combustibility into three groups: ,
| |
| incombustible--materials incapable of burning id air; difficult-to-burn--materials capable of igniting in air from an ignition source, but incapable of burning inde-pendently after it is removed; combustible--materials capable of self-ignition, as well as ignition from an ignition source and independent burning after it is removed.
| |
| The method of evaluating combustibility of plastics by the OI /f (oxygen index) value has become the most popular in the USSR and in foreign practice. The OI of a material is characterized by the minimum volume of oxygen content in an oxygen-nitrogen mixture in which stable combustion of the material under test conditions is possible.
| |
| The smoke formation coefficient characterizes the optical density of smoke forming during combustion of material with specific 5
| |
| 12 Appendix 11
| |
| | |
| saturation-in the. room. Because of the increased concentration of cable in a' unit'of volume of cable' structures, questions of evaluas.ing smoke release during'comoustion of materials and cables
| |
| . and standardization ~ of - the smoke -formation indicator have become very important. According to GOST'12.1-044084, materials -in the USSR 4 are divided into three groups according to the degree of smoke formation:
| |
| ! Classification group of Coefficient of smoke materials with smoke-forming formation capacity
| |
| ~
| |
| Low' < 50 j Medium 7 50. (up to 500)
| |
| 'High > 500 J Toxicity of the combustion products according to GOST 12.1-044-84 is characterized by the' toxicity. indicator HCL 50, the ratio af the quantity of material to s unit of volume of closed space during whose combust, ion the released products cause death of 50% of the experi-mental animals. Polymer materials are classified into the following groups by value of the toxicity indicator: .
| |
| Classification group Toxicity igdicator HCL 50' 9/"
| |
| Extremely dangerous up to 13 Highly dangerous from 13 to 40 j Moderately dangerous from 40 to 120 Not very dangerous over 120 The- requirements for nn: spreading of combustion or fire /_7 j
| |
| resistance'in creation of electrical cables can be met by using .{
| |
| special elements in the cable ~ design or by using materails of l reduced combustibility with specific properties under the influence I of fire. Of the design elements that guarantee the required '
| |
| f resistance of the cable to the effect of fire, special protective j covers or metal shells (screens) jointly with protective cover -
| |
| are used most oft'en.
| |
| The results of the tuales were used to develop a classifier of electrical cables for fire resistance and nonspreading of com- !
| |
| l 'bustion which'are produced at plants. l l
| |
| 6 14 Appendix 11
| |
| __.__:-_.-____-__-----------------------------------------------^^-----^-^^^^^-- - - ^ ^ ^ ^ - ^ ^-' - ^ ^ ^ ^^
| |
| | |
| r
| |
| , Fire-resistant cables are mainly fabricated using incombustible materials, generally of inorganic origin.
| |
| When combustible materials are used in the design, elements are introduced which determine the fire resistance of the,latter.
| |
| Fire-protective coatings applied to the cables af ter they' are laid in bundles and give them the required' fire resistance have also become widespread. In order to reduce the fire danger of cables at the AES, the USSR uses the OPK coating which protects the combustible protective coatings and cable shells from ignition and spreadof combustion under the influence of a local fire source with temperature' B00'C for 30 min.
| |
| The OPK coating is a gray paste consisting of a mixture of thermally stable, gas-forming and fibrous fillers with water-emulsion binding agent. The mass percentage of nonvolatile substances must be (75,1 5)t.
| |
| After the coating is applied, the permissible le,ngthy current loads on the cable drop by (3 - 7)% depending on the thickness of the paste layer and the conditions of cooling the cable route.
| |
| It is not recommended that the OPK coating be used for fire pro- f8 tection of cables laid in rooms where oils, solvents, bitumen, alkalis, aggressive vapors and gaser could affect the coating.
| |
| The applied coating is 3 mm thick in the dry state. In order to gua'rantee this thickness, the coating is applied in 3 - 5 layers.
| |
| Each subsequent layer must not be applied earlier than in 12 hours of drying of the previous layer at temperature 60*C.The final drying of the last layer of coating occurs on the 4th - 5th day.
| |
| In order to improve the fire protection of the cable system in
| |
| . nuclear power plants, unify the technical solutions and use the most effective materials and methods of protecting the cables f at the stage of installation and during operation, the following order has been established for making the fire-blocking seals and partitions, and methods have been pinpointed for laying the cables at AES under construction, under reconstruction and active:
| |
| 7 15 Appendix 11
| |
| | |
| l I
| |
| : 1. Before the cables are installed in the cable structures and production rooms the following must be done:
| |
| 1.1. The planned amount of construction work and water insulation must be completed, including application of protective coating to the walls and metal structures. A certificate is drawn up to release the rooms for electrical installation work. j l
| |
| 1.2. Systems of fire extinguishing must be installed and ,
| |
| i adjusted according to section 5.11 of " Fire Standards of Designing AES." Before the cables are laid, there must be advance incorpora-tion of fire extinguishing units in die temporary mode.
| |
| 1.3. The temporary illumination and welding (for cable structures) lines are dismantled.
| |
| 1 1.4. Standard lighting must be fitted with power on a temporal scheme and releaded to the customer by certificate (for cable 1 structu'res). !
| |
| l.S. There must be primary fire extinguis'hing resources with 49 equipping of the posts in an amount required by the norms.
| |
| 1.6. Individuals responsible for the fire-prevention condition of specific cable structures and for the operation of the fire extinguishing units are appointed.
| |
| : 1. 7.. Instructions are developed for the actions of the operating personnel of the power plant, construction, installation i
| |
| and adjustment organizations in cases of outbreak of fire that must j be coordinated with the facility fire pro,tection. .
| |
| : 2. During installation of cables in the cable structures and production rooms: -
| |
| All places where cables pass through walls and floors, in rooms in which work is done to lay cables, regardless of their design <
| |
| (finished opening, modular or tubular passages, metal boxes) must be temporarily sealed with fire resistant r.iaterials.
| |
| 8
| |
| , 16 Appendix 11
| |
| | |
| All temporary seals disrupted during laying of the cables.
| |
| must be restored daily over the entire length of the route after the end of operations by filling the free space between the cables and the walls of the passage with material that ja allowed for temporary sealing. .
| |
| i If there are no combustible cables (without the "ng" index) ;
| |
| in the cable structure, requirements for daily temporary sealing of the cables go into effect by the time that voltage is supplied or when the structure is filled with cable with 71 volume of
| |
| . polymer materials per running meter.
| |
| : 3. Af ter the cables are laid in each passage (room, structure) ,
| |
| but no later than the moment when the cable system starts to operate fj (the cable system of the block is put into operat' ion on the whole
| |
| ~
| |
| as an object).
| |
| In-this case:
| |
| 3.1. All places of cable passage through walls and floors regardless of the design must be sealed.
| |
| 3 . 2 .. In the metal boxes, except for sealing of places passing through walls and floors, permanent fire-blocking zones must be made on vertical tracks every 20 m, on horizontal tracks every 30 m.
| |
| The type of cable-laying (open on cable structures, in boxes, in troughs, etc.) is determined by the planning organization.
| |
| 3.3. Distribution of a cable in production corridors and .
| |
| rooms must be done based on the condition that on each cable structure (shelf, trough, box, etc.), the volume of polymer materials
| |
| - is no more than 7 liters per running meter (condition of nonspreading of combustion for cables with "ng" index). In this case, OZS *
| |
| . coating of the cables is not required.
| |
| The greatest distance between individual structures in the rooms, corridors and cable structures must correspond tothe PUE.
| |
| fire-protective composite.
| |
| 9 17 Appendix 11
| |
| | |
| [
| |
| i If the volume of polymer materials on the cable structure is more than 7 liters per running meter, then the cables should be covered with OZS when,they are laid in corridors and rooms of
| |
| ' the plant not equipped with automatic fire-extinguishing units:
| |
| --the entire surface.of power and single control cables;
| |
| --the upper row of control cables laid in boxes in many layers; fj
| |
| --the outer layer of control cables laid in bundles or i
| |
| l troughs.
| |
| A similar requirement for OZS coating refers to any cable routes . .
| |
| if they include cables without the "ng" index.
| |
| 3.4. Welding on of' covers is forbidden in metal boxes of any type.
| |
| 3.5. Asbestos materials (asbestos bulk flexible cable, asbestos fabric, etc.) cannot be used for sealing cable passages.
| |
| 3.6. Rooms for control panels (central control panel, BShU.
| |
| [not fur:h,er identified), RShU [not further identified), etc.),
| |
| as well as rooms with electronic and electrical apparatus (UVS, UKT*, AKNP, SUZ, VRK, AKRB [not further identified), and so forth) should have a coating by fire-protective composite for the com-bustible cables placed between panels in boxes or within the lower part of panels. In this case, the fire-protective coating should cover each power cable and the upper row of control cables laid ,
| |
| in multiple layers.
| |
| 3.7. The power, control cables, and communications cables in the machine halls must be laid in metal boxes when they pass '
| |
| near oil tanks and oil stations (less than 10 m in distance) and at sites of possible mechanical damage. In this case, the control -
| |
| cables, communications cables, as well as the power cables passing through these sections are coated with fire-protective materials on the section of the route where an external fire could affect the cables (in -limits of the indicated equipment plus 10 m to each side). The upper layer of the cables is covered in the control 10 18 Appendix 11 1 i
| |
| | |
| point boxe's when there are multiple layers.
| |
| : 4. The composition and type of fire-protective coating materials ;
| |
| and fire-bloching seals, a s well as the inst;uctions for their use must be coordinated with the' Main Administration of Fire Protection of the USSR Ministry of Internal Affairs and for the permissibe current loads with the All-Union Scientific Research, j Planning-Design and Technological Institute of the Cable Industry of the USSR Ministry of the Electrical Engineering Industry.
| |
| : 5. Organizations and direct manufacturers of fire-blocking seals, partitions and zones as well as coatings of the cables with coating materials are responsible for strict observance of the chnological instructions and quality of work.
| |
| Methods of the International Electrotechnical Commission for Testing Cables /13 IEC 332-3/1982--Characteristics of electrical cables that do not spread combustion and are laid in bundles IEC 331/1972--Characteristics of fire-resistant electrical cables General Indicators of Fire Safety for Cables ,
| |
| /14
| |
| : 1. Nonspreading of combustion
| |
| : 2. Optical density of smoke formation
| |
| : 3. Corrosion activity of gas release products
| |
| : 4. ToK,icity of gas release products
| |
| : 5. Fire resistance
| |
| - Degree of Smoke Formation according to GOST 12.1.044-84 /j Classification group of materials coefficient of smoke with smoke-forming capacity formation Low > 50 Moderate > 50 (up to 500)
| |
| High > 500 Groups of Toxicity Indicators according to GOST 12.1.044-84 /1E 11 19 Appendix 11
| |
| | |
| . Group of Toxicity Indicators according to GOST 12.1.044'-84 h Classification group Toxicity indicator HCL 50' 9/"
| |
| Extremely dangerous -
| |
| up.to 13 Highly dangerous from 13 to 4')-
| |
| Moderately l dangerous. _ from 40 to 120 Not very' dangerous. over 120 Combustibility according to'GOST 12.1,044-84 /1 Incombustible -- materials incabable of burning in air .
| |
| . Difficult-to-burn --materials capable of igniting in air from an ignition source, but' incapable of burning
| |
| ' independently after it is removed combustible -- materials capable of self-ignition, as well as ignition from an ignition source and independent
| |
| -burning after it is removed Probabi itylof Outbreak of Fire /18
| |
| ~
| |
| Qn= Q - Q,. (1 - Q a)
| |
| Qa = 0.V Q a' .
| |
| n kH ?I~ g ( ~ kHK '
| |
| n g" v W''
| |
| where: -Q 3 --probability of ignition of electrical network , .)
| |
| Op --probability of spread of fire QT3- -Probability of extinguishing cables and wires by auto- .
| |
| matic-fire-extinguisher On --probability of ignition source appearing 03--probability of burning of combustible material from ignition source O Hk --Pro'bability of appearance of ignition source in group I
| |
| 12 j
| |
| . 20 Appendix 11
| |
| _ _. = _ _ _ - -
| |
| | |
| made of n-cables froin k-th cable -
| |
| Wg --combustion heat of one running meter of cable of i-th type-size V--volume occupied by group laying of cables 1 meter in length y--specific combustion heat of cable laying Suggestions for Agenda for Spring 1989 Seminar in the United States 9
| |
| : 1. Methods of determining causes of fires related to operation of electrical equipment.
| |
| : 2. Modeling of the temperature conditions of fires in AES rooms.
| |
| ~
| |
| : 3. Fire' danger of computers at AES. -
| |
| : 4. Classification of AES rooms for explosion-fire and fire danger.
| |
| : 5. Safety of people during AES fires.
| |
| : 6. Modeling hydrogen combustion in protective shells of nuclear power plants.
| |
| : 7. Discussion of agenda for fall 1989 seminar in the USSR.
| |
| 13 21 Appendix 11
| |
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