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NEA/CSNI/R(97)7

INTERNATIONAL SEMINAR ON

THE SAFETY RESEARCH NEEDS FOR RUSSIAN-DESIGNED REACTORS

KOUKUKAIKAN, TOKYO, JAPAN 8-9 July 1997

MATERIAL PRESENTED AT THE INTERNATIONAL SEMINAR

OECD

OCDE COMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATIONS

OECD NUCLEAR ENERGY AGENCY Le Seine Saint-Germain- 12, boulevard des îles

F-92130 Issy-les-Moulineaux (France) Tel. (33) 1 45 24 82 00 Fax (33) 1 45 24 11 10

ORGANISATION FOR ECONOMIC CO-OPERATION AND DEVELOPMENT

Pursuant to Article 1 of the Convention signed in Paris on 14th December 1960, and which came into force on 30th September 1961, the Organisation for Economic Co-operation and Development (OECD) shall promote policies designed:

— to achieve the highest sustainable economic growth and employment and a rising standard of living in Member countries, while maintaining financial stability, and thus to contribute to the development of the world economy;

— to contribute to sound economic expansion in Member as well as non-member countries in the process of economic development; and

— to contribute to the expansion of world trade on a multilateral, non-discriminatory basis in accordance with international obligations.

The original Member countries of the OECD are Austria, Belgium, Canada, Denmark, France, Germany, Greece, Iceland, Ireland, Italy, Luxembourg, the Netherlands, Norway, Portugal, Spain, Sweden, Switzerland, Turkey, the United Kingdom and the United States. The following countries became Members subsequently through accession at the dates indicated hereafter: Japan (28th April 1964), Finland (28th January 1969), Australia (7th June 1971), New Zealand (29th May 1973), Mexico (18th May 1994) the Czech Republic (21st December 1995), Hungary (7th May 1996), Poland (22nd November 1996) and the Republic of Korea (12th December 1996). The Commission of the European Communities takes part in the work of the OECD (Article 13 of the OECD Convention).

NUCLEAR ENERGY AGENCY

The OECD Nuclear Energy Agency (NEA) was established on 1st February 1958 under the name of the OEEC European Nuclear Energy Agency. It received its present designation on 20th April 1972, when Japan became its first non-European full Member. NEA membership today consists of all OECD Member countries, except New Zealand and Poland. The Commission of the European Communities takes part in the work of the Agency.

The primary objective of the NEA is to promote co-operation among the governments of its participating countries in furthering the development of nuclear power as a safe, environmentally acceptable and economic energy source.

This is achieved by: — encouraging harmonization of national regulatory policies and practices, with particular reference

to the safety of nuclear installations, protection of man against ionising radiation and preservation of the environment, radioactive waste management, and nuclear third party liability and insurance;

— assessing the contribution of nuclear power to the overall energy supply by keeping under review the technical and economic aspects of nuclear power growth and forecasting demand and supply for the different phases of the nuclear fuel cycle;

— developing exchanges of scientific and technical information particularly through participation in common services;

— setting up international research and development programmes and joint undertakings. In these and related tasks, the NEA works in close collaboration with the International Atomic Energy

Agency in Vienna, with which it has concluded a Co-operation Agreement, as well as with other international organisations in the nuclear field.

© OECD 1997 Permission to reproduce a portion of this work for non-commercial purposes or classroom use should be obtained through Centre français d'exploitation du droit de copie (CCF), 20, rue des Grands-Augustins, 75006 Paris, France, for every country except the United States. In the United States permission should be obtained through the Copyright Clearance Center, Inc. (CCC). All other applications for permission to reproduce or translate all or part of this book should be made to OECD Publications, 2, rue André-Pascal, 75775 PARIS CEDEX 16, France.

COMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATIONS

The NE A Committee on the Safety of Nuclear Installations (CSNI) is an international committee made up of scientists and engineers. It was set up in 1973 to develop and co-ordinate the activities of the Nuclear Energy Agency concerning the technical aspects of the design, construction and operation of nuclear installations insofar as they affect the safety of such installations. The Committee's purpose is to foster international co-operation in nuclear safety amongst the OECD Member countries.

CSNI constitutes a forum for the exchange of technical information and for collaboration between organisations which can contribute, from their respective backgrounds in research, development, engineering or regulation, to these activities and to the definition of its programme of work. It also reviews the state of knowledge on selected topics of nuclear safety technology and safety assessment, including operating experience. It initiates and conducts programmes identified by these reviews and assessments in order to overcome discrepancies, develop improvements and reach international consensus in different projects and International Standard Problems, and assists in the feedback of the results to participating organisations. Full use is also made of traditional methods of co-operation, such as information exchanges, establishment of working groups and organisation of conferences and specialist meeting.

The greater part of CSNI's current programme of work is concerned with safety technology of water reactors. The principal areas covered are operating experience and the human factor, reactor coolant system behaviour, various aspects of reactor component integrity, the phenomenology of radioactive releases in reactor accidents and their confinement, containment performance, risk assessment and severe accidents. The Committee also studies the safety of the fuel cycle, conducts periodic surveys of reactor safety research programmes and operates an international mechanism for exchanging reports on nuclear power plant incidents.

In implementing its programme, CSNI establishes co-operative mechanisms with NEA's Committee on Nuclear Regulatory Activities (CNRA), responsible for the activities of the Agency concerning the regulation, licensing and inspection of nuclear installations with regard to safety. It also co-operates with NEA's Committee on Radiation Protection and Public Health and NEA's Radioactive Waste Management Committee on matters of common interest.

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FOREWORD

The International Seminar on The Safety Research Needs for Russian-Designed Reactors was held in Tokyo, Japan, July 8th and 9th, 1997, This seminar on international, national and bi-Iateral cooperation programmes on the safety research needs for Russian-designed reactors was hosted by the OECD Nuclear Energy Agency and the Science and Technology Agency (STA) of Japan at the Kouku Kaikan Conference Centre in the center of Tokyo. More than 70 participants attended the seminar. Represented were experts from OECD/NEA member countries, Russia, the International Atomic Energy Agency (IAEA), the International Science and Technology Center (ISTC), the International Nuclear Safety Center (INSC) and the Russian International Nuclear Safety Center( RINSC). Eighteen papers were presented in five sessions.

The purpose of the International Seminar was to bring together experts involved in this work, to provide a hearing for progress in programmes sponsored by governments and OECD/NEA, and to review significant new technical information coming from these programmes.

The International Seminar was structured around four main areas of co-operation: co-operative programmes of the OECD/NEA, programmes of international organisations, bi-lateral programmes, and national programmes of OECD/NEA member countries having reactors of the VVER type. Each of these areas corresponded to a separate session.

This document contains the material presented, as well as the Summary and Conclusions from the International Seminar.

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TABLE OF CONTENTS

Page

PROGRAMME 11

SUMMARY AND CONCLUSIONS 17

LIST OF PARTICIPANTS 25

PRESENTATIONS 31

Overview of the OECD Nuclear Energy Agency 33 by Dr. Gianni M. Frescura, Head, Nuclear Safety Division, Nuclear Energy Agency

Overview of the NEA Co-operation and Assistance Activities with the CEEC/NIS in the Field of Nuclear Safety 49 by Dr. Gianni M. Frescura, Head, Nuclear Safety Division, Nuclear Energy Agency.

Safety Research Needs for Russian-designed Reactors 59 by Mr. Eric S. Beckjord (USA), former Director of Research, USNRC, and Chairman of the OECD Support Group on the Safety Research Needs for Russian-designed Reactors.

Status and Plans for the VVER-440/213 Bubble Condenser Containment Research • 75 by Prof Dr. Ing. Helmut Karwat (Germany), Chairman of the OECD Support Group of the WER-440/213 Bubbler Condenser Containment Research.

WWER Thermal-Hydraulic Code Validation Matrix for Reactors of the WWERType 93 by Mr. Klaus Liesch (GRS, Germany), Chairman of the OECD Support Group on the W E R Thermal-Hydraulic Code Validation Matrix, and Dr. Michel Reocreux (IPSN, France), Chairman of NEA's Principal Working Group No. 2 on Coolant System Behaviour,

Integrity of Equipment and Structures for RBMK: An Assessment of the Research Needs •.• 133 by Dr. Yoshitaka Hayamizu (PNC, Japan), member of the OECD Support Group on the Safety Research Needs for Russian-designed Reactors.

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Operational Safety Issues Identified by the OECD Support Group on the Safety Research Needs for Russian-designed Reactors 143 by Dr. John R. Honekamp (Consultant to the US International Safety Program Office), member of the OECD Support Group on the Safety Research Needs for Russian-designed Reactors, and Mr. Robert L. Moffit (PNNL,USA).

WER-1000 Large Scale Test Facility (PSB) 153 by Dr. S. Michael Modro (INEEL, USA), member of the OECD Support Group on the WER-1000 Large Scale Test Facility (PSB). Currently on attachment at the OECD Nuclear Energy Agency, Paris, France.

Latest Finding from the OECD/NEA RASPLAV Project 161 by Dr. Vladimir Asmolov (NSI RRC 'Kurchatov Institute', Moscow, Russian Federation), member of the OECD Support Group on the Safety Research Needs for Russian-designed Reactors.

The OECD SCORPIO - VVER Project 179 by. Mr. Atsuo Kohsaka and Mr. T. Suzudo (Japan Atomic Energy Research Institute, Japan)

An Overview of the IAEA Extrabudgetary Programme on the Safety of W E R and RBMK NPPs 195 by Mr. Kazuo Shimomura, Division of Nuclear Installation Safety, International Atomic Energy Agency, Vienna, Austria.

Co-operative Programmes of R&D Work on Nuclear Safety through the International Science and Technology Center (ISTC) 213 by Dr. Alain Gerard (ISTC Executive Director, Moscow, Russian Federation)

Bilateral Programmes of Work between Germany and the CEEC/NIS in Nuclear Safety 229 by Mr. Klaus Liesch (GRS, Germany), Chairman of the OECD Support Group on the W E R Thermal-Hydraulic Code Validation Matrix, and Dr. Klaus Wolfert (GRS, Germany), member of the NEA's Committee on the Safety of Nuclear Installations.

Bilateral Research Programmes of Work between France and the CEEC/NIS in Nuclear Safety 263 by Dr. Michel Reocreux (IPSN, France), Chairman of NEA's Principal Working Group No. 2 on Coolant System Behaviour.

Bilateral Research Programmes of Work between IPSN/GRS and the CEEC/NIS in Nuclear Safety Within EU Programmes 269 by Mr. Klaus Liesch (GRS, Germany), Chairman of the OECD Support Group on the W E R Thermal-Hydraulic Code Validation Matrix, and Dr. Michel Reocreux (IPSN, France), Chairman of NEA's Principal Working Group No. 2 on Coolant System Behaviour.

NEA/CSNI/R(97)7

Bilateral Programmes of Work between the United States and the CEEC/NIS in Nuclear Safety 277 by Dr. David J. Hill (Director, International Nuclear Safety Center (INSC), ANL, USA)

Co-operative Programmes of Work between the United States and Russia in Nuclear Safety through the Russian International Nuclear Safety Center (RINSC) 289 by Prof. Sergei Bougaenko (Director, RINSC, ENTEK, Moscow, Russian Federation), member of the OECD Support Group on the Safety Research Needs for Russian-designed Reactors.

Japan's Assistance and Cooperation for the Community of Independent States and the Countries of Eastern Europe • 315 by Mr. Yoshiharu Shigeiri (STA, Government of Japan) and Mr. Tatsuya Shinkawa (MITI, Government of Japan).

Safety Evaluation of WWER Type Reactors 329 by Turkish Atomic Energy Agency (TAEK). Not presented.

National Programme in the Czech Republic for VVER Reactors , 333 by Dr. Jiri Zdarek (Nuclear Research Institute, Rez, Czech Republic)

New Programme for the R&D Activity Sponsored by the Hungarian Nuclear Safety Authority 355 by Dr. Lajos Voross (Director, Nuclear Safety Inspectorate, Hungarian Atomic Energy Commission)

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PROGRAMME

INTERNATIONAL SEMINAR on

THE SAFETY RESEARCH NEEDS for RUSSIAN-DESIGNED REACTORS

KOUKU KAHCAN, TOKYO, JAPAN

8 - 9 July 1997

Sponsorship

This seminar on international, national, and bi-lateral co-operation programmes on the safety research needs for Russian-designed reactors is sponsored by the Nuclear Energy Agency (NEA) of the Organisation for Economic Co-Operation and Development (OECD) in collaboration with the Science and Technology Agency (ST A) of Japan.

Background And Purpose of the Seminar

The Nuclear Energy Agency of the OECD has a broad-based programme of co-operation and assistance to both the Central and Eastern European Countries (CEEC) and the Newly Independent States (NIS) of the former Soviet Union in planning and executing safety research programmes with a view to building up know-how and capabilities in safety technology pertaining to their nuclear power plants. This programme is intended to contribute to and improve the CEEC/NIS nuclear safety culture by concentrating on long-term objectives, as a complement to the near-term technical upgrades to the highest risk plants and improvements of operational safety.

The OECD Nuclear Energy Agency is carrying out this programme of co-operation under the auspices of the Centre for Co-operation with Economies in Transition (CCET) of the OECD with additional funding (grants) from Japan and the United States.

In carrying out its programme of work, the NEA actively co-operates with other national and international agencies, such as the European Commission (PHARE, TACIS), the International Atomic Energy Agency and the International Science and Technology Center. Representatives from these agencies are active participants in this International Seminar.

Participation

• CEEC/NIS safety experts • Experts who have participated in NEA programmes, • Representatives from international agencies, • Japanese experts who are active in Japanese assistance programs, and • Experts from national and bi-lateral programmes.

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THE SAFETY RESEARCH NEEDS for RUSSIAN-DESIGNED REACTORS

KOUKU KAIKAN, TOKYO, JAPAN

8 <- 9 July 1997

SEMINAR LOCATION

The International Seminar will be held at

KOUKU KAIKAN Conference Hall (6th floor) 1-18-1, Shinbashi, Minato-ku Tokyo, Japan

SCHEDULE OF EVENTS

I. Registration Tuesday, 8th July 1997, 8:30 - 9:00 At KOUKU KAIKAN, 6th floor where the seminar will be held

II. International Seminar Tuesday, 8th July 1997, 9:15-17:30 Wednesday, 9th July 1997, 9:00 -17:00

III. Reception Tuesday, 8th July, 1997, 18:30 -At SUEHIRO (KOUKU KAIKAN, 8th floor) Co-hosted by the OECD Nuclear Energy Agency and the Science and Technology Agency (STA) of Japan

IV. Technical Tour Thursday, 10th July 1997 Tokai Area (Eastern part of Japan) Visiting: Japan Atomic Energy Research Institute (JAERI)

The Japan Atomic Power Company (JAPCO)

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PROGRAMME OF SESSIONS

Tuesday. 8 July

9:15 Opening Remarks Dr. Kenji Sumita, Nuclear Safety Commisioner Mr. Makoto Takahashi, Deputy Director, OECD Nuclear Energy Agency Mr. Eric S. Beckjord and Dr. Masao Nozawa, Seminar Co-Chairmen

Overview of the OECD Nuclear Energy Agency by Dr. Gianni M. Frescura, Head, Nuclear Safety Division, Nuclear Energy Agency.

Session I

PROGRAMMES OF THE NUCLEAR ENERGY AGENCY Session Co-Chairmen: Dr. Michel Reocreux and Dr. Michio Ichikawa

10:00 Overview of the NEA Co-operation and Assistance Activities with the CEEC/NIS in the Field of Nuclear Safety by Dr. Gianni M. Frescura, Head, Nuclear Safety Division, Nuclear Energy Agency.

Safety Research Needs for Russian-designed Reactors by Mr. Eric S. Beckjord (USA), former Director of Research, USNRC, and Chairman of the OECD Support Group on tire Safety Research Needs for Russian-designed Reactors.

Status and Plans for the VVER-440/213 Bubble Condenser Containment Research by Prof. Dr. Helmut Karwat (Germany), Chairman of the OECD Support Group of the WER-440/213 Bubbler Condenser Containment Research.

VVER Thermal-Hydraulic Code Validation Matrix for Reactors of the VVER Type by Mr. Klaus Liesch (GRS, Germany), Chairman of the OECD Support Group on the W E R Thermal-Hydraulic Code Validation Matrix.

12:15-13:30 Lunch

Session II

PROGRAMMES OF THE NUCLEAR ENERGY AGENCY (Continued) Session Co-Chairmen: Prof. Tadatsune Okubo and Prof. Sergei Bougaenko

13:30 Integrity of Equipment and Structures for RBMKs: An Assessment of the Research Needs by Dr. Yoshitaka Hayamizu (PNC, Japan), member of the OECD Support Group on the Safety Research Needs for Russian-designed Reactors.

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Operational Safety Issues Identified by the OECD Support Group on the Safety Research Needs for Russian-designed Reactors by Dr. John R. Honekamp (Consultant to the US International Safety Program Office), member of the OECD Support Group on the Safety Research Needs for Russian-designed Reactors, and Mr. Robert L. Moffit (PNNL,USA).

The VVER-1000 Large Scale Test Facility (PSB) by Dr. S. Michael Modro (INEEL, USA), member of the OECD Support Group on the WER-1000 Large Scale Test Facility (PSB). Currently on attachment at the NEA.

Latest Finding from the OECD/NEA RASPLAV Project by Dr. Vladimir Asmolov (NSI RRC 'Kurchatov Institute', Moscow, Russia), member of the OECD Support Group on the Safety Research Needs for Russian-designed Reactors.

The OECD SCORPIO - VVER Project by. Mr. Atsuo Kohsaka (JAERI, Japan)

16:00-16:30 Coffee Break

Session III

PROGRAMMES OF INTERNATIONAL ORGANISATIONS Session Co-Chairmen: Dr. Yoshitaka Hayamizu and Mr. Tadakuni Hakata

16:30 An Overview of the IAEA Extrabudgetary Programme on the Safety of VVER and RBMK NPPs by Mr. Kazuo Shimomura, Division of Nuclear Installation Safety, International Atomic Energy Agency, Vienna, Austria.

Co-operative Programmes of R&D Work on Nuclear Safety through the International Science and Technology Center (ISTC) by Dr. A. Gerard (ISTC Executive Director, Moscow, Russian Federation)

Wednesday, 9th July

Session IV

BILATERAL PROGRAMMES Session Co-Chairmen: Mr. Atsuo Kohsaka and Dr. Vladimir Asmolov

9:00 Bilateral Programmes of Work between Germany and the CEEC/NIS in Nuclear Safety by Mr. Klaus Liesch (GRS, Germany), Chairman of the OECD Support Group on the W E R Thermal-Hydraulic Code Validation Matrix.

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Bilateral Programmes of Work between France and the CEEC/NIS in Nuclear Safety by Dr. Michel Reocreux (IPSN, France), Chairman of NEA's Principal Working Group No. 2 on Coolant System Behaviour.

Bilateral Programmes of Work between the United States and the CEEC/NIS in Nuclear Safety by Dr. David J. Hill (Director, International Nuclear Safety Center (INSC), ANL, USA)

Co-operative Programmes of Work between the United States and Russia in Nuclear Safety through the Russian International Nuclear Safety Center (RINSC) by Prof. Sergei Bougaenko (Director, RINSC, ENTEK, Moscow, Russian Federation), member of the OECD Support Group on the Safety Research Needs for Russian-designed Reactors.

Bilateral Programmes of Work between Japan and the CEEC/NIS in Nuclear Safety by Invited Speakers from STA and MITI, Japan

12:15-13:30 Lunch

Session V

NATIONAL PROGRAMMES Session Co-Chairmen: Dr. David J. Hill and Prof. Kenkichi Ishigure

13:30 National Programme in the Czech Republic for VVER Reactors by Dr. Jiri Zdarek (Nuclear Research Institute, Rez, Czech Republic)

New Programme for the R&D Activity Sponsored by the Hungarian Nuclear Safety Authority by Dr. Lajos Voross (Director, Nuclear Safety Inspectorate, Hungarian Atomic Energy Commission)

14:30 -15:00 Coffee Break

15:00 Panel Discussion, Co-Chaired by Mr. Eric S. Beckjord and Dr, Masao Nozawa "International Co-operation in Safety Research: The Need for Integration of CEEC/NIS Efforts with those of the OECD/NEA Member Countries"

16:30 Conclusions and Recommendations

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HOST ORGANISATIONS

Science and Technology Agency(STA), Government of Japan OECD Nuclear Energy Agency

represented by:

Mr. Yoichi ITO Director for International Cooperation Research and International Affairs Division Atomic Energy Bureau Science and Technology Agency(STA) Prime Minister's Office 2-2-1 Kasumigaseki Chiyoda-ku TOKYO 100 JAPAN

Mr. Gianni M. FRESCURA Head, Nuclear Safety Division OECD Nuclear Energy Agency Le Seine St-Germain 12, boulevard des Iles F-92130 ISSY-LES-MOULINEAUX FRANCE

SECRETARIAT

For enquiries regarding this International Seminar, please contact:

Mr. Hidetaka ISHIKAWA Tel: +81 3 5470 1981 International Affairs and Planning Department Fax: +81 3 5470 1989 Nuclear Safety Research Association(NSRA) Net: [email protected] 5-18-7, Shinbashi, Minato-ku TOKYO 105 JAPAN 105

or

Dr. Herbert E. ROSINGER OECD Nuclear Energy Agency Le Seine St-Gennain 12, boulevard des Iles F-92130 ISSY-LES-MOULINEAUX FRANCE

Tel: +81 3 3581 2597 Fax:+81 3 35815198

Tel: +33 1 4524 1050 Fax:+33 145241110 Net: [email protected]

Tel: +33 1 45 24 10 62 Fax:+33 1452411 10 Net: [email protected]

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mailto:[email protected]



NEA/CSNI/R(97)7

SUMMARY AND CONCLUSIONS


THE SAFETY RESEARCH NEEDS for RUSSIAN-DESIGNED REACTORS KOUKU KAIKAN, TOKYO, JAPAN

8 - 9 July 1997

The International Seminar on The Safety Research Needs for Russian-Designed Reactors was held in Tokyo, Japan, July 8th and 9th, 1997. This seminar on international, national and bi-lateral cooperation programmes on the safety research needs for Russian-designed reactors was hosted by the OECD Nuclear Energy Agency and the Science and Technology Agency (STA) of Japan at the Kouku Kaikan Conference Centre in the center of Tokyo. More than 70 participants attended the seminar. Represented were experts from OECD/NEA member countries, Russia, the International Atomic Energy Agency (IAEA), the International Science and Technology Center (ISTC), the International Nuclear Safety Center (INSC) and the Russian International Nuclear Safety Center( RINSC). Eighteen papers were presented in five sessions.

The purpose of the International Seminar was to bring together experts involved in this work, to provide a hearing for progress in programmes sponsored by governments and OECD/NEA, and to review significant new technical information coming from these programmes.

The International Seminar was structured around four main areas of co-operation: co-operative programmes of the OECD/NEA, programmes of international organisations, bi-lateral programmes, and national programmes of OECD/NEA member countries having reactors of the VVER type. Each of these areas corresponded to a separate session. The general conclusions, followed by the specific technical conclusions, from the International Seminar are included in the following points.

GENERAL CONCLUSIONS

1. By wide acclaim of the participants, the International Seminar on the Safety Research Needs for Russian-designed Reactors was very interesting, informative, timely, and useful. There were 5 Sessions for papers. Sessions I and II: Programmes of the Nuclear Energy Agency; Session III: Programmes of International Organisations; Session IV: Bilateral Programmes; and Session V: National Programmes. The emphasis of the presentations was on the research needs and not on specific safety issues that are design related. The papers presented at the seminar show the large breadth and depth of safety research underway applying to Russian-designed reactors. The seminar was apparently the first to include the entire range of work in this field.

2. The goal of nuclear safety assistance programs in the CEEC and NTS should be to transfer relevant safety knowledge and technology, and to develop indigenous capabilities in ways that will be self-sustaining when assistance eventually comes to an end.

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3. One of the important lessons of Chernobyl, and of Three Mile Island before it, was the necessity of understanding severe accidents in order to prevent or mitigate them. Severe accident research is difficult and expensive, but international co-operation has proved to be very beneficial through pooling of resources and bringing the best minds and facilities to work on problems, and through dissemination of research results. Work in the field has greatly expanded the common language of reactor safety. Specific NEA initiatives, such as the TMI Vessel Investigation Programme and RASPLAV gathered together teams of skilled people and achieved important results. NEA initiatives with the CEEC and NIS are assisting national programmes by disseminating current knowledge and helping to generate new knowledge in these countries.

4. The Overview of NEA Co-operation and Assistance, and reports of co-operative research between NEA country research centers and Russian research centers clearly indicated that work is progressing well, and is achieving NEA goals of building know-how and capability in safety technology. In particular, CSNI initiatives of RASPLAV and Safety Research Needs for Russian-designed Reactors are good models for future research and technical co-operation. Reports from the Czech Republic and Hungary were also very positive, and indicate keen interest in cooperative programmes from the CEEC point of view.

5. Information exchange between NEA and other international organisations interested in CEEC and NIS nuclear power plant safety is very beneficial, and is contributing to better understanding of safety technology and issues in these countries. Information exchanges show each organisation working in the field what other organisations are doing. Some co-ordination of activities would be useful for disseminating information to these organisations, and for eliminating unnecessary duplication. These are roles that NEA could perform well.

6. With regard to the Safety Research Needs of Russian-designed Reactors, development of a strategic plan for this research has top priority. The key players, i.e., design engineers, plant operators, regulators, and safety researchers in the CEEC and NIS must be involved in developing the plan.

7. It is clear that assistance programmes will eventually come to an end. Meanwhile the best thing to do is to replace assistance with integrated, co-operative safety research programmes. Technical co-operation programmes should lead to self-sustaining nuclear safety improvements and sustainable improvements in the safety culture. CSNI with its long experience can contribute to this development.

8. Regarding co-operation with Russia, we understand that Russian research centers are able to obtain funds from their government for international co-operation on specific projects. Ministerial approval is required, and there are two conditions: a) proposed research must have technical merit, and show reasonable promise for success and practical application in Russia; b) the funds that partners offer to provide, and thereby risk sharing, make the project more attractive than it would be with purely Russian funding.

9. The International Science and Technology Center (ISTC) is dedicated to non-proliferation purposes, but nevertheless, because of special capabilities of the scientific community that it serves, has the potential of contributing to nuclear power plant safety. The ISTC structure exists,

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and funding is available to support projects approved by sponsoring governments. The way to approach ISTC is to provide clear definition of nuclear safety research needs. If successful, the only cost to nuclear safety projects would be the cost of co-ordinating ISTC work with the project.

10. Although the Russian Regulatory Authority is responsible for certification of codes used in Russian nuclear power plants, the Authority does not have means of independently verifying codes, because it does not have a research budget. It depends on the Ministry of Atomic Energy for this purpose. This dependence is at variance with OECD/NEA member country practice. Further, there is the potential for conflict, as thermal-hydraulic code validation reaches completion. Conflict would be apparent if different decisions on code certification were made for similar applications in Russia and CEEC countries.

11. The NEA methodology of co-operation and collaboration on nuclear safety issues could serve as a model to be extended to other parts of the world.

12. Transfer of western codes is a very effective basis for technology transfer by providing up-to-date technology and by training on how to use the technology. The recipients of the codes can contribute further by developing new models to be included in the codes.

TECHNICAL CONCLUSIONS

1. The report of Status and Plans for the VVER-440/213 Bubble Condenser Containment revealed thorough technical evaluation and planning since inception five years ago. However, experiments and data are urgently needed to demonstrate the effectiveness of the bubbler condenser containment in accident conditions. Unfortunately, contracting and procurement problems are delaying the start of experiments. In view of the safety implications of the bubbler condenser to VVER confinement, strong support should be given to getting the experiments underway and completed.

2. Human performance continues to be very important to plant operational safety. There are three specific research needs: (a) a common methodology for performance evaluation; (b) a human error data base for PSA; and (c) expert systems to help prevent accidents caused by human error.

3. The VVER Thermal-Hydraulic Code Validation Matrix work began in 1992, and is progressing well. Additional tests are needed to complete the work. The absence of fully-validated codes is now slowing the implementation of symptom-based EOIs. The first draft of the report on the VVER thermal-hydraulic code validation matrix is expected early in 1998, and completion in two years.

4. The VVER-1000 Large Scale Test Facility (PSB) would be very useful for obtaining data for the VVER Validation Matrix, and the case for completion and operation of the thermal-hydraulic facility is very convincing. Completion of the PSB test facility would provide a full height, approximately 1/300 volumetric scale, for test of the VVER-1000 design. About $2M is needed

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to complete construction, and about $1M annually for three years is needed to operate the facility, for a total expenditure of about $5M.

5. The SCORPIO - VVER Project is going well. Its purpose is to adapt the SCORPIO core surveillance and prediction systems, which has been used at six OECD plants, beginning with Ringhals-2 in 1987, to VVER units. Dukovany is the current target plant. The SCORPIO core surveillance and prediction systems are to be extended to other VVER nuclear power plants in coordination with upgrades to the instrumentation and control (I &C) systems.

6. With regard to Integrity of Equipment and Structures for RBMK, the integrity of the fuel channel is the top priority. A major effort is needed on this issue, including NDE development, materials properties determination, ageing, seismic and dynamic loading, and full-scale testing facilities. The question of the possibility of multiple concurrent raptures of fuel channels should also be addressed.

7. Important points of view emerged from CEEC and NIS countries: • the bubble condenser greatly improves W E R confinement systems, and therefore research to

prove performance is very important and urgently needed; • most, if not all, thermal-hydraulic codes used in the CEEC come from other countries, and the

ability to understand and use them properly is necessary; • the nuclear safety research strategic plan should definitely be developed; • the end of assistance is anticipated, and intentions are to develop their own expertise; • nuclear safety information and research co-operation is very important and useful.

8. Representatives from the CEEC and NIS countries also pointed out that there are also many positives aspects of the VVERs. For example, credit is due to VVER conservative design, because of the large water inventory, which slows the evolution of accidents.

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I H T » SEMINAR on 1 MBIT SESËASCH UBS for HBSUHBKB M M KOUKU KAIKAN, TOKYO, JAPAN ON 8-9 JULY 1997

LIST OF PARTICIPANTS I. OVERSEAS CZECH REPUBLIC

Dr. Jiri ZDAREK - Director of Integrity & Materials Division, Nuclear Research Institute, Rez

France Dr. Michel REOCRBUX - Senior Expert, Safety Research Department, Institute of Protection and

Nuclear Safety (Chairman of NEA's Principal Working Group No. 2 on Coolant System Behaviour)

GERMANY Prof. Helmut KARWAT - Consultant to OECD, (Chairman of the OECD Support Group of the

VVER-440/213 Bubbler Condenser Containment Research)

Dr. Klaus LIESCH - Group Leader, Project Manager East Europe, Thermal Hydraulics and Process Engineering Division, Gesellschaft fuer Anlagen und Reaktorsicherheit, (GRS) mbH, (Chairman of the OECD Support Group on the VVER Thermal-Hydraulic Code Validation Matrix)

HUNGARY Dr. Lajos VOROSS- Director, Dept. Nuclear Safety Inspectorate, Hungarian Atomic Energy

Commission

RUSSIA Dr. Vladimir ASMOLOV - Director for R&D RRC, Kurchatov Institute

(Member of the OECD Support Group on the Safety Research Needs for Russian-Designed Reactors)

Dr. Sergei BOUGAENKO - Director, International Nuclear Safety Center of Russian Minalom, Research and Development Institute of Power Engineering (Member of the OECD Support Group on the Safety Research Needs for Russian-Designed Reactors)

Ms. Lioudmila BOUKCH - Principal Specialist, Ministry of Fuel and Energetics of RF

Dr. Alain GERARD - Executive Director, International Science and Technology Center (ISTC)

Dr. Lev V. TOTCHENYI - Senior Project Manager, Internatinal Science and Technology Center (ISTC)

Ms. Alevtina G, TOTCHENAIA - Vice-Professor, Nuclear Methods Dept., Moscow Physics & Engineering Institute

TURKEY Ms. Gulin DINC - Second Secretary, Turkish Embassy

USA Mr. Eric S. BECKJORD- Consultant in Nuclear Energy, Former Director of Research, USNRC

(Chairman of the OECD Support Group on the Safety Research Needs for Russian-Designed Reactors)

Dr. David J. HILL - Director, International Nuclear Safety Center, Argonne National Laboratory

25

Dr. John R. HONEKAMP - Consultant to the US International Safety Program Office (Member of the OECD Support Group on the Safety Research Needs for Russian-Designed Reactors)

Dr. S Michael MODRO - Consultant for Idaho National Engineering and Environment Laboratory (Member of the OECD Support Group on the VVÏÏR-1000 Large Scale Test Facility (PSB) Currently on attachment at the NEA)

IL INTERNATIONAL ORGANISATION EU

Mr. Maurice BOURENE - First Counsellor for Science & Technology, European Commission

OECD/NEA

Dr. Gianni FRESCURA - Head, Nuclear Safety Div., OECD/Nuclear Energy Agency

Dr. Herbert ROSINGER - OECD Nuclear. Energy Agency

Mr. Makoto TAKAHASHI - Deputy Director, OECD/Nuclear Energy Agency

IAEA Mr. Kazuo SHIMOMURA - Special Projects Unit, Div. of Nuclear Installation Safety, IAEA

m. mm 1. NUCLEAR SAFETY COMMISSION

Dr. Kazuo SATO - Vice Chairman Dr. Kenji SUMITA - Commissioner

2. UNIVERSITY

Dr. Kenkichi ISHIGURE - Professor, Graduate School of Engineering, the University of Tokyo Dr. Yutaka KUKITA - Professor, Nagoya University Dr. Tadatsune OKUBO - Professor, Sophia University

3. INSTITUTE 1AERI

Mr. Kiyoharu ABE - Deputy Director, Dept., of Reactor Safety Research

Mr. Yoshinari ANODA - Head, Thermal-hydraulic Safety Engineering Lab., Dept. of Reactor Safety Research, Tokai Research Establishment

Mr. Masashi HIRANO - Head Nuclear Safety Data Evaluation Lab., Dept of Reactor Safety Research, Tokai Research EstaMisment

Mr. Atsuo KOHSAKA - Deputy Director General, Dept., Tokai Research Establishment

Mr, Makoto SOBAJIMA - Head, Risk Assessment Analysis Lab., Dept. of Reactor Safety Research, Tokai Research Establishment

PNC Dr. Kiyoto AIZAWA - Deputy Senior Director, Reactor Development Project Head Office

Dr. Yoshitaka HAYAMIZU - Deputy Director, Monju Construction Office (Member of the OECD Support Group on the Safety Research Needs for Russian-Designed Reactors)

Mr. Hiroshi IKEDA - Deputy Director, Fugen Nuclear Power Station

26

Mr. Kenichiro KANEDA - General Manager, Material Monitoring Section, Fuels and Materials Division,

0-arai Engineering Center

Mr. Takuya KITABATA - General Manager, Periodic Safety Review Group, Fugen Nuclear Power Station

Dr. Michilaka KOIKE - Senior Staff, Technology Management Division Head Office

Dr. Tadashi MARUYAMA - Senior Staff, Material Monitoring Section, Fuels and Materials Division, 0-arai Engineering Center

Mr. Mitsuo MATSUMOTO - General Manager, Safety Assessment of Fugen Group Reactor Development Project Head Office

Mr. Yoshitsugu M0R1SHITA - Assistant Senior Engineer, Waste Management Section, Administration Division, 0-arai Engineering Center

Mr. Yosuke NAOI - Assistant Senior Chemist, Safety Assessment of Fugen Group Reactor Development Project Head Office

Dr. Sususni SHÏBUYA - Senior Staff, Safety and Chemical Section, Fugen Nuclear Power Station

Mr. Soju SUZUKI - General Manager, Reactor Technology Section, Experimental Reactor Division, 0-arai Engineering Center

Dr. Toshio WAKABAYASHI - General Manager, Corephysics Research Section, Advanced Technology Division, 0-arai Engineering Center

4. ORGANIZATION CONCERNING NUCLEAR ENERGY Mr. Hiroshi ABE - Manager, Cooperation Department, International Cooperation Center, Japan

Electric Power Information Center, INC.

Mr. Haruo FUJI I - Senior Research, Japan Electric Power Information Center

Mr. Masaharu KOBAYASHI - Deputy General Manager, Department of International Relations, Japan Atomic Industrial Forum, INC.

Ms. Haruko MITSUISHI - General Manager, Department of International Relations, Japan Atomic Industrial Forum, INC.

Mr. Katsuma NAKAYAMA - Research Director, Investigation and Research Department, Institute of Research and Innovation

Mr. Masaki 0HKUB0 - Manager, Office of International Affairs, NUPEC

Dr. Koreyuki SHIBA - Director, International Nuclear Technology Cooperation Center, Radiation Application Development Association

Mr. Shinnosuke YAGUCHI - Senior Engineer, Eaergency Response Technology Division, Safety Information Research Center, NUPEC

5. ELECTORIC POWER COMPANY

Dr. Michio ICHIKAWA - Technical Adviser, The Japan Atomic Power Company

Mr. Isao IMURA - Tokyo Electric Power Company

Mr. Hiroyuki HAMANO - Manager, R & D Department, The Japan Atomic Power Company

Mr. Hiroyuki KUSUNOKI - Senior Manager, Reactor Core Design & Safety Group, Nuclear Power Department, Electric Power Development Co., LTD.

27

Mr. Kunihiro TOMITA - Nuclear Power Programs Department, Tokyo Electric Power Company

Mr. Kouji USHIJIMA - General Office of Nuclear & Fossil Power Production, The Kansai Electric Power CO., INC.

Mr. Kouji YOKOTA - Tokyo Electric Power Company

Mr. Mitsuiu YONEYAMA - Nuclear Power Engineering Department, Tokyo Electric Power Company

6. PRIVATE COHPANY Mr. Masaki ANDO - Chief Specialist, Safety Design Engineering, Reactor Design Engineering

Department, Nuclear Energy Division, Toshiba Corporation

Mr. Tadakuni HAKATA - Acting General Manager, Nuclear Energy Systems Engineering Center, Nuclear Energy Systems Headquarters, Mitsubishi Heavy Industries, LTD.

Mr. Takanao HIGUCHI - Researcher, Energy & Environment Department, IEA of Japan CO., LTD.

Mr. Katsuyuki KUMASAKA - Senior Engineer, Nuclear Reactor Engineering Section, Hitachi Works,

Hitachi, Ltd.

Mr. Takaaki K0NN0 - Manager, Nuclear Structures Engineering Department, Kajima Corporation

Mr. Nobuo MAKI - Senior Researcher, Institute of Nuclear Safety System

Dr. Hiroyasu MOCHIZUKI - Director, International Technical Cooperation Division, Engineering Department, PNC Engineering Service CO., LTD.

Mr. Etsuzo MURAMATSU - Staff Consultant, Energy & Environment Department, IEA of Japan CO., LTD.

Mr. Takashi SAWADA - General Manager, Reactor Control S Safety Engineering Department, Nuclear Energy Systems Engineering Center, Nuclear Energy Systems Headquaters, Mitsubishi Heavy Industries, LTD.

Mr. Hiroshi TOCHIHARA - Deputy Chief Engineer, Core Engineering Department, Nuclear Energy Systems Headquaters, Mitsubishi Heavy Industries, LTD.

7. STA

Mr. Akio YUKI - Director General for Nuclear Safety Commission, Nuclear Energy Bureau

Mr. Kazumasa HIOKI - Director, Office of International Relations, Nuclear Safty Bureau

Mr. Kenji SEYAMA - Director, International Affiars and Safeguards Division, Atomic Energy Bureau

Mr. Yoshiharu SHIGEIRI - Deputy Director, International Affairs and Safeguards Division, Atomic Energy Bureau

8. HITI Mr. Tatsuya SHINKAWA - Deputy Director, Nuclear Power Safety Policy Division, Public Utilities

Department, Agency of Natural Resources and Energy

9. NSRA

Dr. Hideo UCHIDA - President

Dr. Tatsuji HAMADA - Executive Director

Dr. Masao NOZAWA - Director

28

10. STAFF Mr. Hidetaka ISHIKAWA - (NSRA) Mr. Takeshi 0SUGA - (NSRA) Ms. Tazuko MÏKI - (NSRA) Ms. Atsuko TAKANO - (NSRA) Ms. Chiaki INOKOSHI - (NSRA) Ms. Itsuko IIJIMA - (NSRA) Mr. Hideo KAWAI - (NSRA) Ms. Naomi HIDAKA - (NSRA)

29

NEA/CSNI/R(97)7

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KOUKU KAIKAN, TOKYO, JAPAN

8 - 9 July 1997

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Overview of the OECD NUCLEAR ENERGY AGENCY

Presentation at the INTERNATIONAL SEMINAR on THE SAFETY RESEARCH NEEDS for RUSSIAN-DESIGNED

REACTORS, Tokyo, Japan by Gianni M. Frescura

8 July 1997

OECD/NEA Nuclear Safety Division July 1997 SEMOECD.ppt

Introduction

What is the OECD ?

What is the NEA ? -Background - Objective -Main Programme Areas

Other than Reactor Safety and Regulation Reactor Safety and Regulation


34

What is the OECD

Organization for Economic Co-operation &Deuelopment

29 Member countries < 20% Worlds population > 3/5 worlds exports produce > 2/3 goods & services provides > 4/5 economic aid to developing countries generates > 4/5 nuclear power in world


What is the OECD

SCOPE OF WORK PERFORMED

General Economic Policies Trade policies Financial, Fiscal & Enterprise affairs Energy Environmental Policies Food, Agriculture & Fisheries Social Affairs, Manpower & Education Science, Technology & industry Relations with Non-Member Countries Public Management

OECD/NEA Nuclear Safety-Division July 1997 SEMOECD.ppt

35

WHAT IS THE NEA

• ONE OF 15 BODIES OF THE OECD

• ITS AIM IS TO PROMOTE THE DEVELOPMENT OF NUCLEAR ENERGY AS A SAFE, ENVIRONMENTALLY ACCEPTABLE ENERGY SOURCE


|>ai!X BASIC CHARACTERISTICS

Semi-autonomous directorate of OECD Two parts - Main Secretariat and Data Bank, with separate membership and budgets Total NEA budget: Aproxlmatety $16 M Staff- 38 professionals/38 support


36

OECD Member Countries

Australia Austria Belgium Canada Czech Republic Denmark Finland France Germany Greece Hungary Iceland Ireland Italy

Japan Korea Luxembourg Mexico The Netherlands New Zealand Norway Poland Portugal Spain Sweden Switzerland Turkey United Kingdom United States


AIMS OF THE OECD

TO PROMOTE POLICIES DESIGNED:

-TO ACHIEVE HIGHEST SUSTAINABLE ECONOMIC GROWTH

-TO CONTRIBUTE TO ECONOMIC EXPANSION

-TO CONTRIBUTE TO EXPANSION OF WORLD TRADE


37

Oi EVOLUTION OF NEA

Became NEA in 1970s when U.S., Japan, Canada, Australia joined Membership stable until mid-1990s Korea, Mexico, Czech Republic, Hungary joined Present membership: 27 countries Role has evolved - focused was joint projects - has become forum for co-ordination of

policies/programmes - outreach since early 1990's


•^™ NEA BASIC ROLE *%*! NEA

Provides forum for peaceful nuclear co-operation Provides Members access to info and experience of others Facilitates pursuits of R&D and studies in areas of mutual interest Pools expertise of Members Promotes development of common views/approaches


38

NEA STRENGHTS

• Homogeneity of membership small club like-minded approach

climate of mutual trust relatively non-political

• Provides added values and is cost-effective • Strong scientific/technical/Jegal work

narrow focus does not deal with non-proliferation, safeguards

• Work methods flexible and responsive to Member needs


J W N METHODS OF WORK 'SfeiSS N E A

Standing Technical Committee Structure Committees decide their programme of work under the guidance of the Steering Committee Working groups meet as required to integrate their efforts Expert Groups, Peer Reviews Secretariat consists of technical experts who co-ordinate the activities and assist the chairmen "Share" not "Assist"


% A E N

39

(OECD Nuclear Energy Agency^

OECD/NEA Nuclear Safely Division July 1997 SEMOECD.ppt

Radioactive Waste Management Committee (RWMC)

Develops common approaches for radioactive waste management strategies. Builds confidence in long-term safety assessment of waste disposal systems & site evaluation methods. Arranges International "peer-reviews" of national programmes. Promotes Internationally accepted technical solutions for decommissioning of obsolete nuclear facilities.

OECD/NEA Nuclear Safety Division Juty 1997 SEMOECD.ppt

40

Committee on Radiation Protection & Public Health

(CRPPH)

Provides advice to Member Countries in the implementation of International Radiation Protection Standards. Assesses radiation in the context of other risks and integrating public concerns into decision-making. Co-ordinating International Nuclear Emergency Exercises (INEX). Manages NEA's information exchange system on Occupational Radiation Exposure (ISOE)


Legal Affairs

• Nuclear Third Party Liability • Contributes to modernisation of

international nuclear liability regime (Paris, Brussels & Vienna Conventions).

• Provides advice on the establishment of national nuclear laws and regulations, and assists CEEC & NIS to develop domestic nuclear legislation.


AI » i

Nuclear Development and the Fuel Cycle Committee (NDC)

• Provides analyses of the role of nuclear power in meeting energy policy goals, including assessments of economic, resource & technology factors.

• Publishes statistics & projections on nuclear energy in the OECD area, and estimates supply & demand in the field of uranium and nuclear fuel cycle services.

• Assesses impact of environmental concerns on nuclear policies.


Nuclear Science Committee (NSC) and Data Bank

• Supports development of scientific knowledge & understanding in fields important for nuclear power.

• Encourages basic research aimed at maintaining & developing expertise in Member Countries.

• Data Bank - Supplies nuclear data & computer programmes for

nuclear technology applications. - Works towards internationally approved data base of

evaluated data


42

Committee on the Safety of Nuclear Installations (CSNI)

• Members are senior safety scientists and engineers

• Main objectives: - to exchange information on nuclear safety

research and operating experience - to develop common technical positions on current safety issues

- to assess status of research in member countries and sponsor international projects



• Main "products":

- Consensus on specific issues - network for information exchange - consensus on research needs and priorities - qualification of analytical tools - stimulus for research - sponsoring of international projects


43

•» TïUTrt;ri:iO«tc!ttorcga


Principal Working Groups • PWG1 - Operating Experience and Human Factors • PWG2- Coolant System Behaviour • PWG3 - Integrity of Components and Structures • PWG4 - Confinement of Accidental Radioactive

Releases • PWG5- Risk Assessment Other Groups • Working Group on Fuel Cycle Safety • NEA Co-operation & Assistance with CEEC & NIS

OECD/NEA Nuclear Safety Division July J997 SEMOECD.ppt

44


Typically:

-about 30-40 techncal reports per year

about 10 to 15 workshops and specialist meetings


Current Sponsored Projects

• HALDEN - New Three year project started in Jan. 1997

• SCORPIO-VVER - Core Surveillance for VVERs On-going

1996/97

• RASPLAV - PHASE I Ended June 1997

• RASPLAV - PHASE II 1997/00 -Agreement being signed

• CABRI - RIA 1998/2002 -Under Discussion


45

OECD/NEA Co-operation and Assistance Activities with CEEC/NIS

• Three Main Areas within Nuclear Safety and Regulation

-Enhancement of Nuclear Safety Research Capabilities

-Provision of Nuclear Safety Information

-Support to Regulatory Authorities


Committee on Nuclear Regulatory Activities (CNRA)

Created in 1989 to guide NEA's programme regarding regulation, licensing and inspection of nuclear installation with regard to safety. Main tasks include:

» exchange of information and experience » review of developments which could affect

regulatory requirements with the objective of providing better understanding of motivation for making changes and improvements and

» review current practices and operating experiences in the Member countries


46

Committee on Nuclear Regulatory Activities (CNRA)

Main "products":

- state of the art reports - compilation of regulatory practices - collective opinions - learning through discussion of

operating experience and research results


CNRA - "Special" Issues

- Regulatory Approaches to Human Factors in Operational Safety (1990) - Regulatory Requirements and Experience Related to Low-Power and

Shutdown Activities (1991 ) - Regulatory Requirements and Experience Related to Steam Generators

(1992) - Dealing with the Safety Case for Ageing Plants (1993) - Regulatory Approaches to Severe Accidents (1994) - Regulatory Approaches to PSA (1995) - Technical Support for Licensing of Computer-Based Systems

Important to Safety (1996) - Review Procedures and Criteria for Different Regulatory Applications

of PSA (1997) - Regulatory Problems of Ageing Reactors (1998)


47

CONCLUDING REMARKS

• The OECD/NEA provides a forum to exchange technical information and develop common positions

• Most of the information available to all countries


48

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Overview of the OECD/NEA Co-operation and Assistance Activities with CEEC/NIS in

the Field of Nuclear Safety

Presentation at the INTERNATIONAL SEMINAR on THE SAFETY RESEARCH NEEDS for RUSSIAN-DESIGNED

REACTORS, Tokyo, Japan by Gianni M. Frescura

8 July 1997

OECD/NEA Nuclear Safety Division July 1997 SEMCCET.ppt

General Features of the Programme

• Areas of traditional strength of the NEA

• Selected areas

• Long term improvements to safety

• Co-ordinated with that of other organisations


50


• Funding for the CEEC/NIS experts and NEA secretariat efforts is provided by the OECD's CCET and by grants from the Japanese and US governments

• Participation of experts from OECD Member countries provided cost-free

• Co-ordination/co-sponsorship with other organisations



• Three Main Areas within Nuclear Safety and Regulation

-Enhancement of Nuclear Safety Research Capabilities

-Provision of Nuclear Safety Information

-Support to Regulatory Authorities


51

Enhancement and Support of Safety Research

• Safety research is important basis for safe reactor operation - it provides the knowledge to:

- Assess safety margins - Develop an effective regulatory regime - Prevent and manage accidents

• Safety research comparable to that carried out in the OECD countries not performed in the former Soviet Union



NEA can play a role Concentrate on Nuclear Safety Research Safety • needed to ensure long-term safety of Russian-designed

plants • promote scientific and technical co-operation with

CEEC/NIS experts in solution of nuclear safety problems • promote common understanding of nuclear issues • involve CEEC/NIS experts in the solution of nuclear

safety problems • promote a nuclear safety culture • encourage international co-operation and collaboration

on nuclear safety issues


52


Support groups are set up to

• identify needs • encourage and evaluate R& D

proposals, • share experience accumulated in OECD

countries, • assist in identifying sources of funding, • arrange for training in Western labs, etc.

OECD/NEA Nuclear Safely Division July 1997 SEMCCET.ppt

ENHANCEMENT OF NUCLEAR SAFETY RESEARCH CAPABILITIES

Four OECD Support Groups • VVER-440/213 Bubbler Condenser

Containment Research • VVER TH Code Validation Matrix • Research on Thermal-hydraulic

Behaviour of VVER-1000 Coolant System (PSB), and

• Safety Research Needs for Russian-designed Reactors


53

BUBBLER CONDENSER RESEARCH WORK

SAFETY CONCERN BEING ADDRESSED Capability of the last physical barrier to withstand DBA and limit radioactive releases to the environment

RECOMMENDATION OF SUPPORT GROUP Perform additional experimental work to validate analytical tools and test dynamic response of structures and structural elements (trays)

Presentation by Prof. Dr. Helmut KARWAT

OECD/NEA Nuclear Safety Division July 1997 SEIWCCET.ppt

VVER-1000 THERMAL-HYDRAULIC RESEARCH LOOP (PSB)

SAFETY CONCERN BEING ADDRESSED

Lack of experimental data on the thermal-hydraulic behaviour of VVER-1000 in accident conditions

RECOMMENDATION FROM THE SUPPORT GROUP Complete construction of PSB facility, only integral facility addressing VVER-1000 thermal-hydraulic behaviour

Presentation by Dr. S. Michael MODRO


54

VVER TH CODE VALIDATION MATRIX

SAFETY CONCERN BEING ADDRESSED Lack of systematic assessment of codes and models used for safety analysis

RECOMMENDATION FROM THE SUPPORT GROUP Complement the CSNI Validation Matrix to address specific phenomena of

VVER reactors

Presentation by Mr. Klaus LIESCH


SAFETY RESEARCH NEEDS FOR RUSSIAN-DESIGNED REACTORS

BACKGROUND

• Review of Safety Research needs is a traditional function of the NEA and CSNI, e.g., SESAR

• A Support Group was set up with significant participation of Russian experts

• Complementary to EU work (USA, Japan, Canada)

• Group gave consideration to the applicability of safety research in OECD Member countries to limit additional experimental and analytical needs

OECD/NEA Nuclear Safety Division July 1S97 SEMCCET.ppt

55

SAFETY RESEARCH NEEDS FOR RUSSIAN-DESIGNED REACTORS

• Presentations by

• Mr. Eric S. BECKJORD,

• Dr. Yoshitaka HAYAM1ZU, and

• Dr. John R. HONEKAMP


Other Activities

SCORPIO/VVER - Core surveillance System for VVERs

- Work to be performed by Halden and by Czech organizations

- System will be made available to other VVER operators

- Funding from Japan

Presentation by Mr. Atsuo Kohsaka


56

Provision of Nuclear Safety Information

• GENERAL OBJECTIVE: - Contribute to make available to CEEC/NIS safety

experts the knowledge accumulated in the OECD countries

• Experts from CEEC and NIS countries are sponsored to participate in some NEA activities - Workshops - Specialist Meetings - International Standard Problems (ISPs)

• CEEC and NIS experts are sponsored at Western Laboratories


Provision of Nuclear Safety Information

Typically In a year:

-About 30 to 50 CEEC/NIS experts are invited to NEA workshops and specialist meetings

-CEEC/NIS experts contribute to two or three International Standard Problems (ISPs):

» In 1996 to: » (ISP 37) aerosol distribution, » (ISP 38) low power transients/BETHSY, » (ISP 39) quenching and fuel-coolant interaction/FARO

- 3 to 5 CEEC/NIS attachments » Have used: »> BETHSY (France), PKL (Germany), PACTEL (Finland)


37

Support To Regulatory Authorities

GENERAL OBJECTIVE: To strengthen and elevate the status of

regulatory bodies in the CEEC/NIS

• The CNRA has established a link with the Association of Regulatory Bodies of Countries Operating VVERs

• The Association is invited to attend the meetings of the CNRA. The NEA Secretariat is an observer at meetings of the Association


CONCLUDING REMARKS

Successful programme • encouraged the research needed for the safety

of Russian-designed reactors • involved CEEC/NIS experts in leading edge

nuclear safety issues, • encouraged international co-operation and

collaboration, and promoted a common understanding of nuclear safety issues, e.g., need for TH code validation matrix

Methodology utilised could be model for co-operation and collaboration on nuclear safety issues in other areas of the world


58

Safety Research Needs for Russian-designed Reactors

A Study Sponsored by the Nuclear Energy Agency

Eric S. Beckjord

International Seminar

July 8, 1997

Tokyo

59

Introduction

The Safety Research Needs of Russian-designed Reactors is an important study carried out by the Nuclear Energy Agency (NEA) of OECD. My task this morning is to tell you about the study and the resulting report. I will describe them and summarize conclusions and recommendations.

In 1995 the Committee on the Safety of Nuclear Installations of the NEA proposed a study of safety research needs for WER and RBMK reactors in order to expand and apply capability in nuclear power plant safety technology in the Central and Eastern European Countries (CEEC) and the Newly Independent States (NIS)

Mr. Gianni Frescura, Head of the NEA Nuclear Safety Division formed a Support Group to carry out the study. The Support Group included technical experts from Russia and NEA countries. The Support Group began work in July of 1995 in Paris, held working sessions in Moscow in May of 1996, and produced a working draft in July of 1996. The Support Group completed a final draft report in the fall of 1996. CSNI approved the report at its Annual Meeting in December, and the Steering Committee of NEA approved the report for publication last April. Copies of the report are available here.

The rationale for the study is that transfer of known technology in the last few years from NEA countries to the CEEC and NIS has helped to improve the safety of plant operations. It is reasonable to expect further improvement in the coming 5 to 10 years by doing research on these reactors. Thus safety research performed by the CEEC and NIS on their reactors is an important subject for the attention of NEA and other countries with nuclear power programs.

Objectives

The basic objectives of the study are to identify WER and RBMK safety issues, to point out safety research performed in NEA countries that applies to Russian-designed reactors, and to identify additional research needed to resolve remaining safety issues. A fourth objective is to make this information widely available to governments, appropriate funding agencies, and research institutions. Accomplishment of the objectives will encourage research proposals and programs, and help to justify funding by appropriate national and international authorities. The

60

expected research aims to improve safety of WER and RBMK reactors•

Accomplishment of objectives will also promote development of continuing program of safety research. A continuing research program will help to assure safe operation of CEEC and NIS reactors in the future.

Uses of Research

With these objectives in mind it is appropriate to point out the uses of nuclear safety research in OECD countries. The major uses are: (1) protection of public safety; (2) development of improved plant, systems, components, and fuel; and (3) improved reactor operations.

1. OECD countries have done research to help assure public safety for many years, both for prevention of accidents, and for mitigation of accidents that release radioactivity. Safety experts anticipated some problems, while others came to light during plant operation. Research findings often change assessment of safety, and make it possible to discover weaknesses and to evaluate effectiveness of proposed fixes, whether changes to plant systems, or changes to operating procedures. Regulatory authorities usually sponsor research for this purpose.

2. Research and development to improve plant design are leading to advances in performance and greater safety margins for next generation plants in Europe, Japan, and the U. S. Reactor manufacturers are the usual sponsors of this research.

3. Nuclear plant owners and operators are interested in improving plant performance, fuel cycles, and economics. Usually they are the sponsors of research for these purposes. There are many interests in safety research.

WER and RBMK Differences

NEA countries have a large base of operating experience and safety knowledge for Pressurized Water Reactors (PWR). Operating experience is important, because of events that shut plants down and challenge emergency safety systems, or because of events that cause accidents. Analysis of events and determination of root causes make it possible to uncover design weaknesses. Feedback of this experience leads to plant improvements that reduce the frequency of accidents in the future, or eliminate them. PWRs and WERs are similar in concept and systems design, and consequently much of

61

Light Water Reactor (LWR) operating experience in OECD countries is applicable to WERs, adding to the base of actual WER experience.

For the same reason, safety research carried out in OECD countries is also applicable, for the most part, to WERs, Application of OECD research findings to WERs is underway, but it lags behind application of operating experience. As a result, there are still unrealized benefits to achieve through application of research findings.

Because of major differences in concept and design between RBMK and WER, operating experience with light water reactors is not as directly applicable to RBMK as to WER. Also the base of RBMK operating experience is considerably less than the LWR/WER base, because of the lesser number of RBMK reactors. Feedback of RBMK operating experience to RBMK plants is therefore at an earlier stage than in the case of WERs. Pressure tube reactor technology exists in OECD countries, but is less extensive than Light Water Reactor safety technology. Thus the application of research findings pertinent to RBMKs is also at an earlier stage than in the case of WERs. For this reason, the study reflects significant differences in the scope and depth of research recommendations for the two reactor types, which are greater in the case of the WER.

With regard to RBMK safety research, the report focuses on work that can reduce risk in the near term. Specifically it recommends research to provide adequate technical bases for improved operating procedures in emergency conditions.

Organization of Study

At its first meeting, the Support Group experts organized the study in seven task teams:

Thermal-hydraulics/Plant Transients for WER Integrity of Plant Equipment and Structures for WER Severe Accidents for WER Thermal-hydraulics/Plant Transients for RBMK Integrity of Plant Equipment and Structures for RBMK Severe Accidents for RBMK Operating Safety Issues for WER and RBMK

This organization of disciplines provides a comprehensive scheme of safety research review. Thermal-hydraulics/Plant Transients includes the physical models and computer codes that describe plant behavior and indicate whether response

62

is normal, or whether conditions are unsafe and could lead to an accident, fuel damage and radioactive release. It also includes the analysis of containment system performance during accidents. Integrity of Plant Equipment relates to assessment of reactor pressure boundary integrity, and prevention of pressure vessel and piping ruptures that cause loss-of coolant accidents. The Severe Accidents task includes the knowledge of events and phenomena that could occur when core cooling fails and possibly leads to fuel damage, core meltdown, and release of radioactivity. Finally, Operating Safety Issues brings in the knowledge of operating events at many reactors that have caused • accidents.

In order to focus sharply and consistently on key safety research topics, the task groups each addressed the following questions:

What are the safety concerns? What are the open safety issues? What are the safety research needs?

Technical Conclusions of the Study

The study reaches technical conclusions on Operational Safety, the WER, and the RBMK.

1. Operational Safety for RBMK and WER: the study strongly urges research to improve human performance, because human error is a major contributor to accidents, and because it is possible to change procedures for operator action quickly and thereby achieve early safety improvement. Other important research includes powerful Probabilistic Risk Assessment (PRA) methods that help to identify plant safety weaknesses; better monitoring of readiness of safety-related equipment to operate; and development of technical bases for emergency operating procedures, i. e., the bases for important actions for reactor operators to take in order to terminate progression of reactor accidents.

2. WER: W E R conclusions in the study are summarized below. Thermal-hydraulics and reactor kinetics govern almost all accidents relating to reactor safety. Codes that describe these phenomena are therefore essential tools for safety evaluation, and they must be verified in order to use them with confidence. Both Russian and OECD codes require additional verification for this purpose. For verification of Russian codes, there is need for more experimental data. For OECD codes, there is need for adaptation to incorporate specific W E R features, and then validation of codes with

63

WER experimental data. With respect to Thermal-hydraulic performance of containment/confinement systems, additional verification and validation of Russian and OECD codes are needed. The WER Bubble Condenser requires additional testing for the purpose of validation.

Integrity of WER Equipment and Structures is the second important research area. Integrity of the reactor coolant boundary, and leak-tightness of containment/ confinement are both necessary to safety. Methods to evaluate satisfaction of these requirements must therefore be verified. For reactor pressure vessels, there is need for research to extend the materials property data bases, including welding material subject to irradiation. There are needs for improved models of other pressure boundary components, and for experimental data for code verification. Also, there is need for improved Non-Destructive-Testing (NDT) methods to monitor actual material properties to evaluate the remaining life of plant components.

Severe Accidents research is the third major area. Severe Accident research performed in OECD countries is in large part applicable to WERs. The most important task for the WER is to improve the codes that describe accidents in order to validate WER Accident Management procedures. There is also need for additional experiments and analyses related to unique WER design features.

3. RBMK: as in the case of WERs, safety improvement depends on the quality of analysis and validated Thermal-hydraulic and reactor kinetics codes. There is need for research to improve the neutronic data base, the coupling between neutronic and Thermal-hydraulic codes, and to validate the codes. There is need for research to improve the technical basis for safety criteria for initiation of fuel and fuel channel failure, and to determine hydrogen distribution after initiation of accidents. For confinement safety assessment, there is need for research to improve and validate codes that describe confinement performance.

Integrity of the primary coolant circuit, and especially the fuel channel, is a major safety issue for the RBMK, and methods of safety evaluation require verification. Research is urgently required to develop improved In-Service-Inspection systems, and monitoring of fuel channel integrity. Analyses are required to develop fuel channel pressure tube rupture models, and response to loads that are the result of pressure tube rupture and loads that are the result of earthquakes. There is need for experiments to

64

obtain materials property data in order to complete integrity assessment.

The RBMK does not have a containment that surrounds the reactor system, and it is therefore necessary to rely on Accident Management for risk reduction. In the case of slow accidents, such as loss-of-coolant, there is enough time for Accident Management to be effective in reducing severity. RBMK loss-of coolant accidents are slow, because of the large heat capacity of the graphite moderator in the core. The graphite prevents rapid increase of temperature in loss-of-coolant accidents. There is need for research to develop simple physical models and codes, based on existing Russian and NEA codes, that describe transients and are useful for developing effective Accident Management procedures at individual plants. There is need to perform integral experiments and separate effects tests in order to verify these codes. Accident Management is important, but cannot solve all problems. It. is not useful in the case of fast reactivity transients in the RBMK, because there is insufficient time to take effective action. For these events, development should focus on limiting fast reactivity increases or preventing them.

General Conclusions

The study reaches the following general conclusions:

1. Research is a major contributor to improving safety. Establishing the technical basis for Accident Management procedures for both WER and RBMK is the most important near term safety research task.

2. CEEC and NIS experts should have the opportunity to work at the leading edge of safety research so that there will be no significant gaps between East and West in this important technology.

3. There is every expectation that the research identified in this report will make a difference to safety of these plants, based on experience with application of safety technology research to OECD plants.

4. There is more experience with NEA PWR and similar WER technology than with RBMK technology. The NEA LWR experience is very beneficial to development of WER research needs. This explains the apparent lead in comparison with RBMK needs, which are in an earlier stage of development.

15

Recommendations

The report of the study recommends preparation of a Safety Research Strategic Plan that establishes goals, defines products required of safety research, and describes how and when work will be carried out. The Plan should also establish research priorities based on needs of the users of research. The users are the regulators, the reactor operators, and the reactor designers. Such a plan would become the organizing principle for the safety research program, and would help to answer the questions that funding authorities will ask when they are considering proposals.

The key players in the CEEC and NIS nuclear communities should get involved in planning and carrying out the research, and in applying the results to improving safety. These persons are the officials in government responsible for energy and safety regulation, the nuclear plant operators, the engineers in reactor design and construction organizations. Involvement of key players is essential to a successful and useful research program.

Nuclear safety is an international interest, and international cooperation is very important to it. The report recommends cooperation in reactor safety research, because of the advantages of sharing knowledge., technical contributions, and funding. Cooperation brings experts together and employs the best facilities world-wide, rather than nationally. This improves research quality. The report identifies a large scope that can benefit from cooperation. Cooperation with Eastern countries has increased substantially in the last several years, and should be increased more to prevent technical isolation in safety technology in the future. At the same time, total reliance on international research cannot substitute for healthy national programs, which are the prerequisite for strong international cooperation.

Finally, there is much research information available within OECD countries that is potentially applicable to Russian-designed reactors. The report recommends finding new approaches to information and technology transfer, such as establishing a forum for a specific technical topic for the purpose of transfer of technology and safety information.

Looking Back and Looking Ahead

The study proves that it is possible to bring together experts from NEA countries and Russia for the purpose of

studying and identifying safety research needs for Russian-designed reactors, to arrive at consensus, and to produce a /useful report of high quality in a rather short time.

Looking back, I believe readers will find that the strength of the report is in its technical breadth and depth. This stems from the competence of task group chairmen and contributors, all experts from NEA countries and from Russia. The availability of technical journal articles on safety technology to both sides, and meetings of technical specialists from both sides in the past few years axe also important factors. The experts understood each other on basic issues, and it was not necessary to spend much time in the study to arrive at common views of safety issues.

Several weaknesses in the study are apparent. One is that experts from CEEC countries did not participate in the original study. However, they are interested in it, and authorities in the Czech Republic and Hungary have reviewed the report, and are present and giving papers at this Seminar. Their interest and presence are important, and help to strengthen the report in this regard.

Another weakness is that Russian regulatory authorities and nuclear plant operators, have not, so far, undertaken a safety research program that responds to needs identified in the report. The study will not be successful until a commitment to undertaking safety research for Russian-designed reactors becomes a reality.

Looking ahead, I believe there is reason to hope for such a coraraitment as the result of work now underway to develop a strategic plan for safety research for Russian-designed reactors, as recommended in the report. Russian International Nuclear Safety Center for Russia, and DOB/Argonne National Laboratory for the U. S. are cooperatively working on the plan. NEA has offered assistance,by providing NEA country experience related to strategic planningrand by performing a peer review of the plan when it is ready for review.

Development of the strategic plan for Russian-designed reactor safety research is an important subject, and one that I hope will attract the attention and interest of participants at this Seminar. I propose that we discuss it further during the Panel Discussion that will take place tomorrow. I welcome your thoughts, and suggestions on this subject at that time.

ê?


Eric S. Beckjord International Symposium

Tokyo

July 8,1997

Introduction

CSNI: do study of SRN of RdR - 4/95

NEA Support Group formed 6/95

Support Group Meetings: 7/95, 5/96, 7/96 Report Draft approved: CSNI 12/96; NEA Steering Committee 4/97 Rationale: current NEA technology helps CEEC and NIS reactor safety today; safety research will help tomorrow

Objectives

• Identify VVER&RBMK Safety Issues

• Point out applicable NEA research

• Identify additional research needed • Write report of study ; make it available to

governments, funding agencies, and research institutions

NEA Uses of Research

• Assure protection of public

• Develop improved reactor systems and components

• Improve reactor operations, fuel cycles, and economics

69

VVER & RBMK Differences

• NEA LWR experience applies to VVER • NEA safety research also applies to VVER • RBMK experience less than VVER/LWR

experience

• Pressure Tube Reactor technology base less than VVER/LWR technology base

• Therefore VVER research needs are better defined than RBMK needs

Organization of Study Seven Task Teams:

VVER Thermohydraulics/Plant Transients

VVER Structural Integrity

VVER Severe Accidents

RBMK Thermohydraulics/Plant Transients

RBMK Structural Integrity

RBMK Severe Accidents

VVER & RBMK Operating Safety Issues

70

Organization of Study (more)

• In order to focus on main technical issues and justify needs, Each Task Team addressed three basic questions:

• What are the safety concerns?

• What are the open issues?

• What are the safety research needs?

Technical Conclusions

• VVER & RBMK Operational Safety: improve human performance; and use PRA

• VVER TH: verify & validate codes

• VVER Pressure Integrity: verify evaluation methods

• VVER Severe Accidents: improve codes and validate Accident Mgmnt procedures

71

Technical Conclusions (more)

RBMK TH: improve and validate codes

RBMK Confinement: improve and validate codes

Pressure Tube Integrity a major safety issue

RBMK Severe Accidents: develop better LOCA Accident Management codes RBMK Fast Reactivity Insertion: prevent or limit these events

General Conclusions

• Research contributes to reactor safety

• CEEC/NIS experts should work at leading edge of reactor safety to avoid technology gaps

• VVER & RBMK safety research will make a difference

• RBMK safety research at an earlier stage of development than VVER

72

Conclusions Today (more)

• Actual safety research by CEEC/NIS will make study successful

• Strategic Plan: development underway • Strategic Plan review by NEA: propose

more discussion in Panel Session tomorrow

73

STATUS AND PLANS FOR THE VVER-440/213 BUBBLER CONDENSER CONTAINMENT RESEARCH

Prof. Dr.lng.H. Karwat (Germany)

Consultant

Presentation on occasion of the

Seminar on

"The Safety Research Needs of Russian -Designed Reactors"

Tokyo, Japan

8-9 July 1997

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*—' OECD-SUPPORT GROUP ON

VVER-440/213 BUBBLE CONDENSER CONTAINMENT RESEARCH (HISTORY AND MANDATE)

• INITIATED AS MULTILATERAL WORKING GROUP UPON RECOMMENDATION OF VVER-SAFETY SYMPOSION IN COLOGNE (GERMANY) ON JULY 5TH - 7TH 1992

• ACTING AS OECD-SUPPORT GROUP ON VVER-440/213 BUBBLE CONDENSER CONTAINMENT RESEARCH AFTER CSNI ENDORSMENT OF A FRENCH-GERMAN PROPOSAL IN DECEMBER 1993

• ASSESSMENT OF THE EXISTING EXPERIMENTAL AND ANALYTICAL BACKGROUND OF THE BUBBLE CONDENSER SYSTEM

• EVALUATION OF RESEARCH PROPOSALS TO RESOLVE OPEN ISSUES

• TECHNICAL CONSULTATION SHARING EXPERIENCE OF WESTERN EXPERTS IN EARLIER BWR PRESSURE SUPPRESSION SYSTEM RESEARCH WITH EASTERN PARTNERS

• ASSISTANCE IN PREPARING APPLICATIONS FOR FUNDING AND IN ORGANISING PROJECT STRUCTURES

• REVIEW OF PROGRESS OBTAINED DURING PLANNING AND EXECUTION OF PROJECTS

• REPORTING TO OECD/NEA PRINCIPAL WORKING GROUPS PWG 2 AND 4

79

COUNTRIES PARTICIPATING IN OECD SUPPORT GROUP

MEETINGS

• CZECH REPUBLIC

• FRANCE

• GERMANY

• HUNGARY

• ITALY

• RUSSIAN FEDERATION

• SLOVAK REPUBLIC

• UKRAINE

• UNITED KINGDOM

• CLOSE CONSULTATION WITH THE EUROPEAN UNION AND THE INTERNATIONAL ATOMIC ENERGY AGENCY (IAEA)

80

? Working Meetings of the OECD-Support Group on

"VVER-440 Bubble Condenser Containment Research Work"

-Kiev /Ukraine (November 1992)

-Rez/Czech Republic (March 1993)

-Bratislava/Slovak Republic (September 1993)

-Budapest/Hungary (January 1994)

-Zvenigorod/Russia (April 1994)

-Budapest/Hungary (September 1994)

-Kiev/Ukraine (April 1995)

-Prague/Czech Republic (February 1996)

-Levice/Slovak Republic (May 1997)

S 1

Milestones of Development "VVER-440 Bubble Condenser Containment Research Project"

• Kiev (November 1992) Evaluation of phenomena not well known and on

existing test facilities (ZUETES)

• Rez (March 1993) Assessment of the utilization of existing test rigs and of

the status of analytical simulation models

• Bratislava (September 1993) Discussion of the need for a

supplementary test facility (NABUCCO)

• Budapest (January 1994) Proposal of a Unified Bubble Condenser

Research Project involving existing and supplementary test facilities (UBCRP)

• Zvenigorod (Apri l 1994) Adoption of the UBCRP- Testing Concept for

further consideration and possible funding

• Budapest (September 1994) Recommendations concerning the

analytical support for the planned experimental program.

• K i e v (April 1995) Recommendations concerning the need for developing

supplementary Instrumentation. First information on the EU Bubble Condenser

Feasibiity Study

• Prague (February 1996) Information and discussion of interim results of

the EU Feasibiity Study, introduction of a proposal for static structural bahavior

tests in response to findings of some IAEA studies

• Levice (May 1997) Discussion of the proposed PHARE/TACIS Bubble

Condenser Research Project and information on the Status of the EU-Project

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IMPORTANT PROCESSES AND PHENOMENA HITHERTO NOT FULLY COVERED:

DYNAMIC LOADING OF BUBBLE CONDENSER STRUCTURES AND GAP-CAP SYSTEMS IN PARTICULAR

INTERACTION OF POOL STRUCTURES WITH OSCILLATORY CONDENSATION PROCESSES

PROBABILITY OF DANGEROUS CHUGGING SITUATIONS CAUSED BY AIR-FREE STEAM FLOW UNCLEAR

* INTEGRAL CONDENSATION EFFICIENCY IN CASE OF PARTIAL FAILURE OF A TRAY

84

/ /

ANALYTIC SUPPORT and

INTERPRETATION (SVUSS/EREC)

TRANSIENT TH/SD-

INTERACTION (EREC)

STATIC STRUCTURAL!

TESTS (VUEZ)

SMALL BECHOVICE TEST RIG (SVUSS)

STRUCTURE OF THE PHARE-TACIS BUBBLE

CONDENSER RESEARCH PROJECT

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CONCLUDING REMARKS

• SUPPORT GROUP HAS REVIEWED THE EARLIER EXPERIMENTAL AND ANALYTICAL DEVELOPMENT WORK PERFORMED IN THE FORMER SOVIET UNION WHEN DESIGNING THE BUBBLE CONDENSER SYSTEM

• OPEN ISSUES IDENTIFIED AND NECESSARY COMPLEMENTARY RESEARCH WORK RECOMMENDED

• PROPOSED RESEARCH FOR THE BUBBLE CONDENSER CONTAINMENT CONSTITUTES A HIGH PRIORITY ISSUE TO ESTABLISH AN ADEQUATE LEVEL OF REACTOR SAFETY FOR EXISTING VVER-440/213 PWRs

• UNFORTUNATELY CONTRACTING AND EXECUTION OF RECOMMENDED EXPERIMENTAL WORK HEAVILY DELAYED

• OECD SUPPORT GROUP WILL CONTINUE TO PROVIDE TECHNICAL ADVICE ON THE PROJECT EVOLUTION AND MONITOR THE APPLICABILITY OF RESEARCH FINDINGS TO SAFETY ASSESSMENTS OF NUCLEAR POWER PLANTS

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WWER Thermal-Hydraulic Code Validation

Matrix for Reactors of the WWER Type

Klaus Liesch

GRS, Germany

Michel Reocreux

IPSN, France

International Seminar on the Safety Research Needs

for Russian - Designed Reactors

Kohku Kaikan, Tokio, 8 - 9 July 1997

93

2

Outline

1. Retrospect

• Initiative by BMBF

• Continuation by OECD/NEA

2. WWER Specific Code Validation Matrices

3. Data Base and Selection of tests

4. Future Activities

5. Summary

94

1. Retrospect

Initiative by BMBF, Germany, July 1992

The Federal Minister for Education, Research and

Technology of Germany (BMBF) initiated a Multilateral

Symposium on Safety Research for WWER Reactors,

Cologne7-9 July 1992

• Topic 1 (out of 12)

Verification Matrix for Thermalhydraulic System-Codes

applied for WWER Analyses

• Objective of cooperation

Formulation of an internationally agreed WWER

specific verification matrix as a supplement to the

CSNI matrix for PWRs with U-tube steam

generators

Establishment of an international Working Group

(Leader GRS in close cooperation with IPSN)

GSS 95

4

• Interested Partners

Czech Republic, Finland, France, Germany,

- Hungary, Russia, Slovak Republic, Poland, Ukraine

• 5 Meetings of the WG

Berlin 12-13 May 1993

Rez 23 - 24 September 1993

Borovoe 21 - 23 March 1994

Moscow 14-15 June 1994

Budapest 13-14 October 1994

• Final Report

K. Liesch (GRS) Mr. Reocreux (IPSN)

Verification Matrix for Thermalhydraulic System

Codes Applied for WWER Analysis

Common Report IPSN/GRS No 25

July 1995

• Major Findings and Results

• Cross Reference Matrices (Revision 5, July 1995) for

Large Breaks , Small and Intermediate Leaks and

Transients

• Descriptions of relevant test facilities

Separate Effect Test facilities (SET)

Integral (System) Test facilities (IT)

• Selection of individual experiments

From this basis the activités continued under the

auspices of OECD/NEA early 1995

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7

Continuation by OECD/NEA since June 1995

• Establishment of a OECD Support Group (SG) on

WWER TH Code Validation Matrix June 1995

• Participating Countries Bulgaria Czech Republic Finland France Germany Hungary Italy Poland Russia Slovak Republic Ukraine United Kingdom United States of Amerika

• OECD/NEA

8

• Review of Issues and Accomplishments

1. Meeting, 6 - 7 June 1995 at RRC-KI, Moscow

the SG identified need for

> Description of WWER specific phenomena

> Optimization of the matrices

>* Development of criteria for test selection

> Central data storage in Russia (accessible to all

participating countries)

2. Meeting, 19 -20 March 1996 at RRC-KI, Moscow

the SG addressed the identified issues organizing

work in three Task Teams (TT)

> Task Team 1 (TT 1 )

Description of WWER specific phenomena

Team Leader: S. Logvinov (EDO-GP, RF)

• characteristics of phenomenon • relevance to nuclear safety • data base -> reference NPP

107

9

> Task Team 2 (TT 2)

Optimization of matrices

Team Leder: L Szabodos (AEKI, Hu)

• representativty of test with regard to expected reactor conditions/parameter ranges

• boundary conditions • scaling considerations • data quality/adequacy of instrumentation,

uncertainties • completeness and quality of documentation • challenge to system codes

> Task Team 3 (TT 3)

Central data storage

Team Leader: H. Holmstroem (VTT, Fin)

• provision of a set of criteria of requirements for a central data storage: - guarantee of permanent storage - ensure easy access to data sets - facility and test documentation - data format/data storage/back up - release procedures/on-line access

GKS 108

10

Strategic Meeting 17 January 1997 at GRS, Garching

Review of progress and identification of next actions

• to date, no response from facility owners to TT 2's

requests for information

• decision to collect untranslated information for initial

evaluation of candidate facilities and tests

• urgent request for the reports of TT

(expected March '97)

• solving the data bank problem is a high priority issue

(TT3)

• NEA to send requests for support to Minatom and

GAN-RF

• agreements on schedules

G3S 109

11

Based on the results of the Strategic Meeting, the future

work was planned by the SG chairman and the NEA

Secretariat in three phases:

• Phase I

• revise the phenomena list forming the basis for the

validation matrices

• select tests for the matrices

• identify approach and establish a Russia TH data

bank which must be accessible to countries with

WWER reactors and all NEA-members

• Phase II

• revise the matrices

• collect test information and implement on data bank

• prepare draft final report

110

12

Phase III [in the future, possibly under Principal

Working Group 2 (PWG2)]

• organize International Standard Problems

• update the matrices as needed as new experimental

evidence and analyses are produced

111

13

3. Meeting, 3 - 4 June 1997 at Rez, Czech Republic

Status of Phase I

• TT 1 meeting was held 14-15 May 1997 at EDO GP,

Podolsk, Russia

• Cross References Matrices Revision 5, July 95

were reviewed and revised

• some phenomena were excluded (Zr-Steam

reaction, ECC injection into DC)

• ECC injection into UP was included

• description of phenomena have been extended

where necessary

Basis is the descriptions of phenomena as reported

OCDE/GD (94) 82

OCDE/GD(97)12

112

14

• TT 2 activities

• some facility descriptions have been supplied, some

' are still missing

• all test descriptions are outstanding

• NEA secretariat contacted Russian data owners

(Febr. 97) and requested support from Minatom and

GAN-RF

# TT 3 activities

• NEA identified the data bank established at the

Russian International Nuclear Safety Center

(RINSC) as the best suited data bank for the

WWER thermalhydraulic database

• agreement was reached with sponsoring

organisations (US-DOE, Minatom)

• data base will be located in two data banks

>• NEA data bank located in Paris

> RINSC/INSC data banks located in Russia and USA

113

2. WWER Specific Code Validation matrices

• Purpose of validation matrices

• Prediction of nuclear reactor TH behaviour under

off-normal conditions can be made by:

> analysis of TH behaviour in smaller scale facilities and

extrapolation to NPPs

>- assessment of computer codes and application of

validated versions to plant situations

• assessment of computer codes

(developmental/independent) requires qualified

experimental data base (information from NPPs

included)

• complex interrelation between phenomena and

experimental confirmation can be systematised by

matrix formats

• concept of validation matrices is based on two main

ideas:

> selection (minimized number of tests)

> completeness (in view of future code applications)

G2S 114

16

• Matrices were established in four steps:

>- Step 1 : classification in specific matrices

* LB LOCA, SB and Intermediate Leaks, Transients

' >- Step 2: elaboration of cross reference matrices

>- Step 3: selection of tests in the matrices (criteria refer

to CSNI report 17)

> Step 4: validation matrices: phenomena versus

specific tests serve as basis for code validation and

assessment

• Structure and format of the WWER specific matrices

are based on the well known CSNI Code Validation

Matrices for

• LWR LOCA and Transients (OECD/NEA 1987)

• PWR and BWR Thermalhydraulic System Codes

(Wolfert, Brittain, Nue E+D 1988)

• Reasons for decision

• practically approved by the users

• many main common features of PWRs (U-tube SG)

and WWER

(5SS 115

17

Based on the knowledge and experience of the SG

experts the relationships have been assessed for

relevance and suitability in six submatrices:

• phenomenon versus CSNI matrices (additional)

• phenomenon versus plant type (additional)

• phenomenon versus test type

• test facility versus phenomenon

• test type versus test facility

• plant type versus test facility (additional)

Interrelations between individual parameters (in the six

submatrices) are given in the same way as in the

CSNI matrices

116

Matrix I CROSS REFERENCE MATRIX

FOR LARGE BREAKS

CSNI + covered by o partially covered - not covered

Phenomenon vs plant type + fully specific to WWER o partially specific - not specific

Phenomenon vs test type + occuring o partially occuring - not in list

Test facility vs phenomenon + suitable for code assessment o limited suitability - not suitable x expected to be suitable

- Test type vs test facility + already performed o performed but of limited use - not performed

Plant type vs test facility + covered by o partially covered » not covered

117 Matrix I - Revision 5 - Status July 1995

18

3. Data Base and Selection of Tests

According to the critieria developed in TT 2 following

facilities and tests currently have been selected for

qualification for the WWER TH data base.

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4. Future Activities

• Completion of the plant and phenomena descriptions

• Completion of facility and test qualifications

• Preparation of the Final Report

• 4th meeting of the SG ,19 - 20 November 1997, AEKI,

Budapest

• April 1998 issue first version of the Final Report

120

5. Summary

• The establishment of WWER specific validation

matrices is still underway.However significant results

have already been obtained:

• WWER specific cross reference matrices

• lists of relevant test facilities and draft-descriptions

• lists of selected experiments and draft-descriptions

• drafts of phenomena descriptions

• Progress with descriptions of WWER specific

phenomena could be improved by financial support

remarkably

• To date situation of the optimization of the matrices

reveals:

• the relatively small number of relevant test data

suitable for code assessment

• the lack of larger scaled experimental facilities and

data

121

21

• Utilizing the existing structure of the INSC data bank

for establishment of the international WWER

thermalhydraulic data base under auspices of

OECD/NEA will provide to the international reactor

safety community easy and flexible access to

experimental data needed for safety analysis, code

development and validation

Support Group and NEA Secretariat optimally promote

the elaboration of the WWER specific supplements to

the CSNI validation matrices

DEVELOPMENT OF A WWER-SPECIFIC VALIDATION MATRIX FOR THE ASSESSMENT OF THERMALHYDRAULIC CODES

K. Liesch Thermalhydrauiîcs Division

Gesellschaît fur Anlagen- und Reaktorsicherheit (GRS) mbH Garching, Germany

M. Réocreux Safety Research Division

Nuclear Protection and Safety Institute Cadarache, France

A. Suslov

Institute for Nuclear Reactors Russian Research Centre - Kurchatov institute

Moscow, Russia

S. Zaytzev Codes and Algorithms Development Department

EDO Gidropress Moscow-Podolsk, Russia

ICONE-4, New Orleans, USA, March 10-14, 1996

C:\AMiPRO\DOCS\LIEIC0NE96.SAM

DEVELOPMENT OF A WWER-SPEC1FIC VALIDATION MATRIX FOR THE ASSESSMENT OF THERMALHYDRAULIC CODES

K. Liesch, GRS, Germany M. Réocreux, 1PSN, France A. Suslov, RRC-KI, Russia

S. Zaytzev, EDO-GP, Russia

(For full authors' affiliations, please refer to separate sheet.)

ABSTRACT A multi-national Working Group consisting of

experts from Czech Republic, Finland, France, Germany, Hungary, Russia, Slovac Republic, Poland and Ukraine has been formed on the initiative of the Federal Minister for Research and Technology (BMFT) of the Federal Republic of Germany and in close cooperation with the Nuclear Protection and Safety Institute (IPSN) of France in May 1993 to elaborate the topic "Verification Matrix for Thermaihydraulic System Codes applied for WWER Analysts".

The topic was combined with the objective of a cooperation to formulate an internationally agreed WWER-specific verification matrix as a supplement to the existing CSNI matrix for PWRs with U-tube steam generators.

Based on the CSNI cross reference matrices the lists of phenomena have been reviewed and adopted to the characteristics of WWER-440 and WWER-1000 systems respectively, and the lists of test facilities suitable for code assessment have been completed.

According to the marking and ranking within the matrices the situation of the experimental data base could be demonstrated.

The three cross reference matrices for large breaks, small and intermediate leaks and for transients presented in this paper describe as a

result of the Working Group the actual status based on available information. It became apparent that the validation matrices have to be reviewed and supplemented continuously, therefore a new Support Group has been installed under the auspices of the OECD/NEA, June 1995.

1. INTRODUCTION On the occasion of an invitation of the BMFT a

Multilateral Symposium on Safety Research for WWER Reactors was held at GRS, Cologne, July 7-9, 1992. An agreement was achieved on 12 topics for concerted actions on safety research for WWER-type reactors. Topic No. 1 "Verification Matrix for Thermaihydraulic System Codes applied for WWER Analysis" was to be developed as a supplement matrix to the CSNI matrix for PWRs with U-tube steam generators. The tasks to be elaborated by a multi-national Working Group were determined as: - specification of dominant WWER-specific

phenomena, - selection of appropriate test types addressing

these phenomena, - selection of appropriate test facilities, - selection of individual tests, - preparation of a first version of the matrix (the

matrix should be updated in new versions considering new WWER-specific test data).

C-.\AMIPR0\D0CS\L1E1C0NE96.SAM

124

The above tasks have been performed successfully by the Working Group under the leadership of GRS in close cooperation with IPSN during 1993 - 1995, and the results were published by Liesch and Réocreux (1995).

The main results of the joint development of WWER-specific validation matrices are presented in this paper.

2. STRUCTURE OF THE WWER-SPECIFIC MATRICES

The well-known CSNI Code Validation Matrix of Thermo-Hydrauiic Codes for LWR LOCA and Transients (OECD/NEA, 1987) and the CSNI Validation Matrix for PWR and BWR Thermal-Hydraulic System Codes (Wolfert and Brittain, 1988) were chosen as a basis and in particular the procedure for establishing and the format for writing the matrices were adopted. Reasons for these decisions have been mutually agreed upon: practically approved by the users and the presence of many main common features of PWRs with U-tube steam generators and WWER reactor systems.

In the Cross Reference Matrices the important phenomena which are believed to occur during a LOCA or a transient, the experimental facilities suitable for reproducing these effects, and the test types of interest are listed. In addition an indication is given which phenomenon is covered by the experimental data base available from the CSNI matrices and which is specific to WWER-440/213 and WWER-1000 systems respectively.

Based on the knowledge and experience of the Working Group experts, the relationships - phenomenon versus (covered by) CSNI matri

ces (which is additional), - phenomenon versus plant type (which is

additional), - phenomenon versus test type, - test facility versus phenomenon, * test type versus test facility, - plant type versus test facility (additional)

C:\AM1PRO\DOCS\LIEICONE96.SAM

have been assessed for relevance and suitability to the extent possible to date.

Assuming that the qualification of individual models and the verification o1 codes with respect to WWER reactor characteristics have not been completed satisfactorily, separate effects tests have been selected therefore not only to compensate cases where suitable integral system tests could not be found or are not yet operable to address a particular phenomenon.

The three individual matrices consisting of six sub-matrices each have been prepared, differentiating between - large breaks - small and intermediate leaks, and - transients.

3. WWER-SPECIFIC CROSS REFERENCE MATRICES

In figures 1 to 3 the Cross Reference Matrices are shown. The interrelations between the individual parameters which are fixed within the six sub-matrices are given in the same way as in the CSNI matrices.

The relationship phenomenon versus CSNl-matrices is rated at one of three levels: - covered by: which means that the particular

phenomenon is already evaluated in the CSNI data base satisfactorily (plus in the matrix),

- partially covered: only some aspects of the phenomenon have been treated (zero in the matrix),

- not covered (dash in the matrix). The relationship phenomenon versus plant type is rated at one of three levels: - fully specific to WWER: which means that this

phenomenon is specific to the individual WWER reactor system and to be treated with high priority

- partially specific: some aspects of the phenomenon can also be evaluated from PWR type experiments or analyses (zero in the matrix),

- not specific (dash in the matrix).

125

file://C:/AM1PRO/DOCS/LIEICONE96.SAM

The relationship phenomenon versus test type is rated at one of three levels: - occurring: which means that the particular

phenomenon is occurring in that kind of test (plus in the matrix),

- partially occurring: only some aspects of the phenomenon are occurring (zero in the matrix),

- not occurring (dash in the matrix). The relationship test facility versus phenomenon is rated at one of four levels: - suitable for code assessment: which means

that a facility is designed in a way to simulate the phenomenon assumed to occur in the plant and it is sufficiently instrumented to reveal the phenomenon (plus in the matrix),

- limited suitability: the same as above, with problems due to imperfect scaling or insufficient instrumentation (zero in the matrix),

- not suitable: obvious meaning, taking into account the two previous items (dash in the matrix),

- expected to be suitable: definition introduced in some cases to emphasize that new facility still under construction particularly addresses the simulation of this aspect; clearly a conclusive comment cannot be made at present (cross in the matrix).

The relationship test type versus test facility is rated at one of three levels: - already performed: or planned within 1995;

the test type is useful for code assessment purposes (plus in the matrix),

- performed but of limited use: this kind of test has been performed in the facility, but has limited usefulness for code assessment purposes, due to poor scaling or the lack of instrumentation (zero in the matrix),

- not performed (dash in the matrix). The relationship plant type versus test facility is rated at one of three levels: - covered by: which means that the plant char

acteristics and the overall system behaviour

C;\AM1PR0\D0CS\LIEIC0NE96.SAM

are adequately simulated by the facility (plus in the matrix),

- partially covered: only some of the characteristics have been treated (zero in the matrix),

- not covered (dash in the matrix). All spaces which have been left blank corre

spond to cases where experimental evidence is missing to date.

Additional explanations, boundary conditions and information as supposed to be necessary are given in the foot notes below the matrices.

3.1 CROSS REFERENCE MATRIX FOR LARGE BREAKS

The cross reference matrix for large breaks is given in Matrix I (Fig. 1). 17 phenomena, four test types, five system tests and 15 separate effects test facilities are included.

Four new phenomena in addition to the CSNI matrix have been supplemented to this matrix: - steam condensation in (horizontal) steam

generator primary side (may lead to plug formation and loop sealing)

- two-phase flow in steam generator primary and secondary side (two-phase flow may influence the heat transfer regime and thus the heat transfer from primary to secondary side of a SG)

- metal-steam reaction (according to the specific fuel rod cladding the metal-steam reaction could contribute to the heat generation even in LB LOCA)

- non-condenstble gas effects (the accumulation of non-condensible gases in the horizontal steam generator tube could result in a partial loss of heat sink).

As obvious from the sub-matrix test type versus test facility only a few experiments have been performed in the Hungarian PMK-2 and the Finnish PACTEL facilities respectively. The results are categorized as of limited use for the code assessment purposes. The Working Group, especially experts from Russia, expressed that large break scenarios are well understood for

126

WWER plants in operation and therefore limited experimental activities are planned at the moment.

3.2 CROSS REFERENCE MATRIX FOR SMALL AND INTERMEDIATE LEAKS

The cross reference matrix for small and intermediate leaks is given in Matrix II (Fig. 2). 28 phenomena, eight test types, eight system test-and 13 separate effects test-facilities are included.

Four new phenomena in addition to the CSNI matrix have been supplemented to this matrix: - natural circulation, core-gap-downcomer-

dummy element in core (the presence of dummy elements in the core periphery could result in an in-core natural circulation)

- recirculation in the (horizontal) steam generator primary side (the circulation between the SG vertical collectors could influence the heat transfer between primary and secondary side)

- water accumulation in steam generator tubes (the accumulation of water in the horizontal tubes of a SG could decrease the two-phase natural circulation in the primary circuit)

- dynamics of ECCS check valves (the opening/closing characteristics of a check valve could influence the ECC water delivery to the primary circuit).

The Russian test facility ISB-WWER, designed for modelling mainly a WWER-1000 system by a volumentric scale of 1:3000 is already in operation. This facility, however, will be improved by an adequate modelling of the horizontal steam generators. Therefore it is expected to be suitable for code assessment as well as the Russian PSB-WWER facility also modelling a WWER-1000 system by a volumetric scale of 1:300, the elevations 1:1, four loops and a design pressure of 25 MPa, and an electrical power of 15 MW. The PSB-WWER facility is planned to be operable end of 1996. The Russian KMS facility (volumetric scale 1:27, elevations 1:1, four loops, design pressure 18 MPa, containment

C:\AMIPRO\DOCS\LIEICONE96.SAM

scale 1:3) presently is in a feasibility study phase. Experiments of those facilities, together with PACTEL, PMK-2, REWET-lll, PM-5 and SB test facilities could contribute to the evaluation of scaling effects.

3.3 CROSS REFERENCE MATRIX FOR TRANSIENTS

The cross reference for transients is given in Matrix III (Fig. 3). 12 phenomena, nine test types, seven system test- and six separate effects test-facilities are included, besides, measurements from NPPs are incorporated by WWER 1:1 in the list of system test facilities.

The NPP data will be valuable for the investigation of phenomena such as "thermalhydrauiic nuclear feedback". The list of phenomena of this matrix remained unchanged compared to the CSNI matrix, which certainly reflects the actual status of July 1995.

However, there is an extensive discussion underway concerning optimization of operational procedures or investigation of accident management measures. So this matrix is expected to be modified and extended in the near future.

It should be noted that a detailed description of the phenomena or the test facilities cannot be given within the frame of this paper. This will be performed as a task under OECD/NEA auspices or is already documented (Liesch, Réocreux, 1995).

4. SELECTION OF TESTS As a basis for the selection of individual ex

periments a number of factors were considered: - typicality of facility and experiment to ex

pected reactor conditions, - quality and completeness of experimental

data, - availability of data.

In ail cases attempts were made to find plant results or integral experiments to address each of the phenomena of interest. Only in cases where suitable plant results or integral

127

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experiments could not be found, results from separate effects tests were used to some extent. Also questions relating to scaling and facility design compromises have been considered when identifying experimental data.

Proposals have been made by the experimentalists or owners of the data for: - Large breaks

26 individual tests, which have been performed in - the Russian test facilities

PM-5, SB, 4 tube model, SVD-1, SVD-2, TVC-440, ETVUS, KS

- the Czech Republic test facility GWP

- and the Finnish test facilities REWET-II, VEERA.

- Additionally, seven sets of experimental data are stored in data banks from the Russian facilities KS-1, SKN and SG-NPP and the Finnish IVOCCFL facility.

- Small and intermediate leaks 32 individual tests, which have been performed in - the Russian test facilities

PM-5, SB, ISB-WWER, Thermal Mixing, 4-tube model, KS-1, IF-NC

- the Finnish test facilities PACTEL, REWET-il

- and the Hungarian test facility PMK-2. - Additionally, five sets of experimental

data are stored in data banks from the Russian facilities KS-1, TF102, 24 MT and the SG-NPP and the Finnish IVO loop seal facility.

- Transients Four individual tests, which have been performed in the Russian test facilities BD and the Finnish test facilities PACTEL, VEERA, REWET-II.

- Additionally, two sets of experimental data are stored in data banks from Russian facilities TF 102 and 7 Assembly.

C:\AMIPRO\DOCS\UEICONE96.SAM

Numerous WWER-specific experiments have been performed and many WWER-specific data have been generated which could not be incorporated into the selection because of restrictions in availability, typicality, quality and completeness, however, for the present status, it could be considered as an encouraging result for the selection of tests that over 75 individual tests have been proposed which should be qualified according to the factors mentioned above.

5. FUTURE ACTIVITIES The available integral test facilities represent

different design rationals, nominal parameters, scaling ratios, etc. Therefore, results from counterpart tests at different test facilities could improve the investigation of the importance of key phenomena in a specific experiment and could contribute to the assessment of the capabilities of computer codes.

For this reason, the selection of tests from the large number of experiments proposed has to be continued, in order to get the ones which are the most suitable for code assessment with respect to a given phenomenon or test type. In order to support the selection, detailed explanations of the choices for the selected data have to be given as well as the definitions used for the examples.

As a consequence these activities will continue under the auspices of the OECD/NEA. Therefore, in June 1995 a new Support Group has been installed to continue with the further evaluation of the matrices, concentrating on three tasks: - description of WWER-specific phenomena

and safety relevance, - optimization of the WWER-specific code vali

dation matrices, - development of criteria for the data bank stor

age of experimental data valid for the matrices.

128

file://C:/AMIPRO/DOCS/UEICONE96.SAM

6. CONCLUSIONS The formulation of an internationally agreed

WWER-specific verification matrix as a supplement to the CSNI matrix for PWRs with U-tube steam generators resulted in establishing three cross reference matrices for large breaks, small and intermediate leaks and transients. Phenomena have been identified, which are assumed to occur in WWER-440/213 and WWER-1000 plants during accident conditions and facilities and test types suitable for code assessment have been discussed, individual tests have been selected and proposed which could be used for the validation of computer codes.

It should be noted that a detailed description of the phenomena or the test facilities or the lists of tests in the relevant matrices could not be given within the frame of this paper. This will be performed as a task under OECD/NEA auspices or is already documented (Uesch, Réocreux, 1995).

An updating of-the matrices will be necessary to include new experimental findings and better interpretation and understanding of existing data. The matrices also permit identification of areas where further research activities may be justified.

7. REFERENCES

K. Liesch, M. Réocreux, 1995

"Concerted Actions of Safety Research for WWER-Reactors". Report on Verification Matrix for Thermalhydraulic System Codes Applied for WWER Analysis. IPSN/GRS Report, Cadarache/ Garching

K. Wolfert, I. Britain, 1988 '"CSNI Validation Matrix for PWR and BWR Thermal-Hydraulic System Codes". Nuclear Engineering and Design, 108, pp. 107-119

OECD/NEA, 1987

"CSNI Code Validation Matrix of Thermo-Hydraulic Codes for LWR LOCA and Transients". Restricted CSNI Report 132

C:\AMJPno\DOCS\UEICONE96. SAM

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THE SAFETY RESEARCH NEEDS for RUSSIAN-DESIGNED REACTORS Kohku Kaikan, Tokyo, Japan

8 - 9 July 1997

Integrity of Equipment and Structures for RBMK; An Assessment of the Research Needs

Yoshitaka Hayamizu

Integrity of Equipment and Structurels for RBMK; An Assessment of the Research Needs

OUTLINE OF PRESENTATION

1 Reactor Coolant System of RBMK

2 Safety Concerns (1) Multiple Fuel Channel Rupture (2) Guillotine Break of Large Diameter Components (3) Seismic Qualification (4) Gap between Fuel Channel and Graphite

3 Safety Research Needs (1) Inspection Technology (2) LBB Technology (3) Seismic Analyses (4) Integrity of Reactor Cavity under Emergency Conditions (5) Evaluation Methods for Aging

133

Multiple Fuel Channel Rupture

1 Event Description 1) Causes

- Break of Pressure Header or, - Break of Distribution Group Header(DGH)

2) Event Sequence - Break of DGH - Temperature Rise of Fuel Cladding - Break of Fuel Channels - Pressure Rise of Reactor Cavity - Failure of Reactor Cavity

2 Safety Significance Control Rod Ejection

resulting in Reactivity Insertion

Guillotine Break of Large Diameter Components

SAFETY SIGNIFICANCE

First generation RBMK plants do not have a Pressure Confinement to endure a pipe break for a pipe over 300 mm in diameter

The design of the nozzles and joints in the Reactor Coolant System presents concerns as to their structural integrity and potential for cracking

134

Seismic Qualification

SAFETY CONCERNS

1 Fuel Channel Seismic qualification of the refueling machine is the subject of a discussion

2 Channel Inlet and Outlet Tubing Site specific seismic analysis for Leningrad NPP plant is completed while for the other plants the work is still in progress

Gap between Fuel Channel and Graphite

SAFETY CONCERNS

The gap between the graphite inside diameter and the fuel channel outside diameter is the criterion for replacement

Replacement of the fuel channel tubes is being performed on a routine basis

The study of creep should be studied

135

Safety Research Needs for Inspection Technology

1 SUBJECT FUEL CHANNEL 1) Diffusion Bonded Joint

Joint between Zirconium Pressure Tube and Ti-stainless Steel End-pieces

2) Deformation of Fuel Channel 3) Change of Material Properties

REACTOR COOLANT CIRCUIT COMPONENTS 1) Measurement of flaw sizing

2 RESEARCH 1) NDE Methods 2) Evaluation Methods

3 RELEVANT RESEARCH Russia, Canada, Japan

Safety Research Needs for LBB Technology

1 SUBJECT 1) Piping Elbows 2) T-joints of Branches 3) Zones of Perforation

2 RESEARCH 1) Development of Material Database 2) Development and validation of Codes for Dynamic Analysis 3) Improvement of Leak Detection System 4) Development of Monitoring System for Assessing the Condition

of the Metal 5) Development of Integrity Analysis of Pressure Tube in case of

Water Hammer due to the Shut-down of the Check Valves

3 RELEVANT RESEARCH Russia, UK, Canada, Japan

136

Safety Research Needs for Seismic Analysis

1 SUBJECT 1) Drum Separators with connected piping

2) Interaction with the Supporting Metal Structures

2 RESEARCH NEEDS 1) Development of Analytical Model

for the seismic assessment of the complex Drum Separator -Connected Piping - Support Structure System

Safety Research Needs for Integrity of Reactor Cavity

1 RESEARCH NEEDS 1) Experimental Study on the Performance of Suppression System

and Integrity of Reactor Cavity under Multiple Fuel Channel Rupture

2) Development and Verification of Thermal-mechanical Codes 3) Development and Verification of Analytical Models for Loading

of the Cavity 4) Development and Verification of Analytical Models for Fuel

Channel Loading


137

Safety Research Needs for Evaluation Methods on Aging

1 RESEARCH NEEDS 1) Development of Evaluation Methods for Degradation

of Fuel Channel Material

2) Development of Evaluation Methods for Deformation of Fuel Channel


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140

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142

International Seminar on Safety Research Needs for Russian-Designed Reactors

Kohku-Kaikan, Tokyo, Japan, 8-9 July, 1997

Operational Safety Issues Identified by the OECD Support Group of the Safety Research Needs for

Russian-Designed Reactors John R, Honekamp - Consultant and Robert L Moffitt -

Pacific Northwest National Laboratory

Operational Safety

• Fundamental concern is with impact of human errors on a wide range of activities from operations and maintenance to management and the attitudes toward safety

• Analysis of operational data and PSAs confirm the importance of human error in initiating accident sequences

• However, well-trained operators and improved man/machine interface technology (SBEOIs and SPDS) can be very effective in preventing or halting accident sequences

• For existing reactors, Operational Safety improvements are high priority because frequently fewer barriers to change exist in this area than those present in modifying the plant.

143

Operational Safety

• All Operational Safety research needs identified in SESAR report apply equally to Russian-designed reactors

• However, extent of existing research base applicable to Russian-designed reactors and relative priorities are different (for example, development of PSÂ data base versus extension of PSA methodologies)

• Important to include designers, operators, and regulators of Russian-designed reactors in these developing areas even though they may not be immediately applicable, to avoid institutionalizing the existing safety technology gap

Operational Safety

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• Assessment of Operational Safety * Collection, evaluation, reporting operating experience, both human

performance, and equipment reliability •:• Accident precursor analysis * Plant condition monitoring

• Human Factors * Use of western data for Russian-designed plants v Use of simulators to evaluate performance *> Analysis of operating experience

144

U.S. Bilateral Programs

The International Nuclear Safety Program (INSP) supports the operators and regulators of Russian-designed nuclear power plants in:

1. Management & Operational Safety 2. Engineering & Technology 3. Safety Evaluations 4. Fuel Cycle Safety 5. Safety Legislation & Regulation

Chornobyl Initiatives Conversion of Russian production reactors to power only International Nuclear Safety Center(s) - ( Dr. Hill & Prof. Bougaenko)


145

Operational Safety

non

Instrumentation and Control * Improved l&C equipment reliability data * Methods to improve immunity to EM noise •:• Development of improved set-point methodology * Include in International programs on l&C obsolescence

Probalistic Safety Assessment * Near-term priority is comprehensive, credible human performance and

equipment reliability data and; validated deterministic models needed to establish success criteria, core damage states, release fractions

* Longer term, joint research on improving PSA methodology as discussed in SESAR report

03-1147 970600» 5

Operational Safety

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• Accident Management * Near-term priority is validating and implementing symptom-based

emergency operating procedures and effective use of control room simulators to train operating crews in their use

* Longer term, joint research on improved simulator models capable of realistically depicting key control room signals during beyond design basis events

* Research on expert systems to improve staff response to off-normal and emergency events

06-11-97 97060093-6

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1 46


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Policy & l i ^ ^ l P W t ^ En**£? Guidance I K ^ ^ l ^ ^ ^Sfifenc* and Tédmolog?

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LMJrio D«Jd - Mamgw Dog HmAta - Deputy Manager

147

Management and Operational Safety

Objective • Increase the safety of day-to-day operations at

Soviet-designed nuclear power plants Projects • Increase the ability of plant personnel to operate

reactors safely Projects organized into following areas: • Conduct of operations • Operator exchanges • Simulators and training • Emergency operating instructions • Emergency management and planning • Maintenance technology transfer and training

06-11-97 97060093-11


Activities • Completed 16 draft guidelines based on "good

practices" standards developed by U.S. Institute of Nuclear Power Operations

• Balakovo (Russia) and Zaporizhzhya (Ukraine) designated as pilot plants

• Procedures are finalized, approved by regulator, and issued to other Russian-designed nuclear plants

• Three approved and issued to all Russian plants • Six approved for all Ukrainian plants

06-11-9? 970600» 12

148


Activities • Nuclear training centers established at Balakovo

(Russia) and Khmelnytskyy (Ukraine) plants. Joint effort with Russia and Ukraine.

• More than 800 nuclear power plant personnel have participated in training courses

• Paks Training Center (Hungary) is using evaluation tools transferred through the cooperative program to improve simulator training for control room operators

• Slovakia's Trnava Training Center identified training needs. U.S. experts have provided initial courses in instructor training.


Activities • Analytical Simulators

•:• Balakovo • Chornobyi • Novovoronezh

• Full Scope Simulators • Kola • Khmiinitskyy • South Ukraine Unit 1 • Rivne • Kalinin • South Ukraine Unit 3

U9


Activities: • The Novovoronezh (Russia) plant has

implemented the first set of 22 symptom-based EOls Zaporizhzhya (Ukraine) plant has drafted complete set of 48 EOls Rivne (Ukraine) plant has drafted 19 of 38 EOls Symptom-based EOls also being developed for Kozloduy (Bulgaria), Dukovany (Czech Republic), Paks (Hungary), Ignalina (Lithuania) and Bohunice (Slovakia) plants.


• B O B Activities: • Purchased vibration analyzers for

all RBMK reactors and converted their readouts to Russian (detect misalignment and imbalance in rotating machines)

• Laser shaft alignment equipment is scheduled for delivery to each plant by the end of 1997

• Reactors will receive valve seat resurfacing equipment

• Ignalina received an urgently needed pipe lathe/weld preparation machine for replacing corroded pipes

150

Engineering and Technology

Activities • Completed guidelines to assess

fire hazards and identify most important and cost-effective changes

• Provided f ire-retardant materials for the Smolensk (Russia) and Zaporizhzhya (Ukraine) plants (coat cables and seal room-to-room penetrations)

• Russian company is manufacturing 400 fire doors for Smolensk plant

• Ukrainian company is manufacturing 125 fire doors for the Zaporizhzhya plant

• Zaporizhzhya received 50 sets of fire brigade gear, 260 fire and smoke detectors and other fire equipment, including 1,200 sprinkler heads


Activities • Ukrainian companies are under contract to deliver fire

extinguishers and 300 units of a self-contained breathing apparatus

• Ukrainian personnel will receive fire hazards evaluation training

• Modern fire detection system installed at Smolensk and Leningrad (Russia) NPPs

• 2 fire trucks, firefighting equipment, protective suits, communications equipment to Kozloduy (Bulgaria)

• Fire-resistant floor coating, fire and smoke detectors, and fire-resistant doors to the Armenia Nuclear Power Plant.

151


Activities • Safety-grade direct current batteries mounted on seismically

qualified racks have been installed at Kola (Russia) and Kursk (Russia)

• Mobile pumping system for emergency water supplies was sent to Kursk (Russia)

• Emergency water supply system for Novovoronezh (Russia) under way

• Safety parameter display systems are being provided to 10 of the 15 operating RBMK reactors

• Safety parameter display system being developed for Novovoronezh Unit 3, a VVER-440/230 plant

• Additional SPDS projects are being developed for other VVERs

Fuel Cycle Safety

•BBS

Activities • Equipment has been delivered for

building three dry-cask storage units for Zaporizhzhya's spent fuel

• Zaporizhzhya staff poured a mockup concrete cask

• Transporter for moving filled casks to concrete storage pads

• Experts have trained Zaporizhzhya staff to operate the cask system and manufacture 12 casks per year

• Cooperative technology transfer agreements with the nuclear and radiation safety authority in Russia

152

VVER-1000 Large Scale Test Facility (PSB)

International Seminar on the Safely Research Needs for Russian-

Designed Reactors Tokyo, July 8-9, 199?

S.M. Modro OBCO - Nuctestr Energy Agency

PSB Facility - major issues and background

Safety concern: identified lack of experimental data on the thermal-hydraulic behavior of WER-IOÛ0 under accident conditions An integral test facility simulating VVER-1000 reactor system is being constructed at the Electrogprsk Research and Engineering Center (EREC) near Moscow, Russia An OECD Support Group was formed to provide advice and guidance to design, construction and experimental program

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153

PSB Programme Objective

* The primary purpose of the PSB test facility is to generate data for TH system code validation

• The PSB facility will be the first larger scale integral test facility addressing VVER-1000 behavior and suitable to produce valuable data for system code validation

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The OECD PSB Support Group

» The mandate of this OECD Support Group is to assess the design of the PSB facility and of the associated experimental program and, based on experience of western experimental programs, to provide advice and recommendation to EREC staff

* Over the period December 1993 through March 1996 the OECD experts met six times, conducted series of reviews of the facility design and experimental program and issued severqi recommendations which were implemented or are being planned.

154

The OECD PSB Support Group — • — — » M — — — T — •" i MII I — — M — i — —

• OECD expert recommendations addressed: - ûhangcB to the system hardware sueh as alignment of the upper plenum

with the lower part of the pressure vcôeej; installation of mora representative main coolant pumps', installation of secondary feud system and relieve system

- upgrade^ to the atandarf instrumentation - advanced two-phase flaw instrumentation - te$t$ data processing and qualification

- staff training - tcBt matrix - analytical support

- The Support Group monitored the facility construction progress and assessed schedule and costs

PSB Facility Design

* The P3B facility is designed to represent in a volumetric scale of 1:300 a WER-1000 system (modeled after the 30QO MWt Zaporozhskaya Unit 5 NPP). The facility is ftjl pressure (up to 25 MPa) and full height representation of the modeled system

• The PSB facility consists of : - pressure vessel housing electrically healed rod bundle (169 rods) - external do wTicomer - four circulation loops including pumps and horizontal steam

generators - safety systems including accumulators, low pressure injection

system - break system including break piping and a catch tank

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155

PSB Facility Design (cont.)

• PSB Instrumentation ~ 800 transducers - Température, pressure, differential pressure, flaw, local Void

fraction, he:atflux, electric power * FSB Datfc Acquisition System

- 1000 channel? - maximum sampling rate 20 Hz per channel

- PSB Process) Control System - originally selected: Siemens TELEPEBM system - transducers,, cross connectors* interfaces and actuators (to be

designed and made in Russia)

PSB Programme Test Matrix, Phase I

• Test selection based on the W E R TO Code Validation Matrix

• Characterization teats - heat losses - hydraulic resistances - natural circulation - single and two-phase - control of main coolant pumps

• Sinall Break tests - 0.9%, 4% 11% cold leg breaks - 6%pressurizerleqk - steam generator hot collector break (break of one SG tube)

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156

PSB Programme Analytical Support

• plans for analytical support are developed for design, construction commissioning and testing phases

• The analytical support includes development of thermal and fluid dynamic parameters based on characterization tests» preparation of code input models, evaluation of initial and boundary conditions, test planning, test analysis, post test calculations and code models assessment, data evaluation and quality assurance

• EREC has set up four analytical groups for TCH-M4, CATHARE, ATHLET and RELAP5 system codes

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Based on the Support Group review and recommendations the following external

assistance was provided - NEA sponsored training for EREC experts in the following areas:

- 1994/95, DAS and Process Control System at BETHSY France and FKL Germany - two persons

- 199S, management of experimental programs ai BETHSY France PKL Germany, PACTEL Finland - one person (1 week at each fnoilicy)

- 1996, operation, data qualification and documentation at BETHSY, France - one person three months

- 1997, two-phdPe flow instrumentation a t BETHSY, France - two pcrôûne, 3 months

• US DOE provides two-phase flow instrumentation for three primary loops and the break.

• Germany, France and ITS provided training to EREC staff in use of thermal-hydraulic system codes

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137

In May 1996 The OECD PSB Support Group issued a Status and Needs Report

and concluded: • the PSB facility is today the only integral facility suitable for

producing meaningful experimental results on the behavior of WER-1000,

• gome modifications are planned to ensure prototypical response, » there is a potential for significant delay of tile PSB ejiperimentat

programme, • the PSB programme as planned is Technically adequate for the first

phase, • the analytical support activities should accompany the experimental

programme, • in order to benefit tan previous Western experience in similar test

facilities:, PSB technical staff should continue receiving specific training.

DECC> SEA Ktioltu- mfcty PlWibn piH.ro! - 1

OECD Support Group Recommendation

The OECD Group strongly recommends that financial and technical support to be provided to the PSB project in order to complete the construction of the facility and to perform the experimental programme

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158

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Current Status of Facility Construction

• Facility construction continues at slow pace • System hardware almost complete • Basic instrumentation incomplete • DAS incomplete • Process control system needed • Two-phase flow instrumentation will be provided

in. August

OECD. îfEA Nutl.ir finely Pl"i>t°i piM.Pi* - 13

Support for the PSB Facility

• In September 1996 NEA Secretariat initiated solicitation for international support for the PSB Programme

* No commitments for further support were received so far * OECD Support Group on the Safety Research Needs for

Russian-designed Reactors hap identified generation of VVER-10Q0 experimental data as one of high priority needs

• Potential for PSB as OECD project is being evaluated

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Estimated cost of PSB Facility Completion

• Completion of the facility • $Q.4M for mechanical and electrical hardware

including basic instrumentation - $0.3M for data acquisition system • S0.5M for process control system • $O.SM for new primary pumps

* Operation - commissioning and characterization $ 1.0 M - on the basis of a 5 tests per year the annual estimate is

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Conclusions

• À large scale integral test facility for VVER-1000 is needed

• Small progress in facility contraction over last two years

• International funding support is needed to complete facility construction and to initiate test program

• OECD Support Group has provided valuable review and advice to EREC and promoted development of a programme plan

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Pres&nfeld by V. Asmoinv çrt International Seminar on the Safety Research Needs for Russiian-Disiqned Reactors TOKYO, Japan, 8-9 July, 1997

161 /

RASPLAV Phase I Major Activities * Large scale corlum tests

- RASPLAV-AW-20CM 9 of October 1996

- RASPLAV-AW-200-2 24 of May 1997

* Separate effect tests (Tulpan, Korpus, Tigel) * Material properties measurements * Salt tests * Code development and usage

Major G o a l s of O E C D R A S P L A V P r o j e c t

» Study behavior of prototypic molten core materials in the lower head of the RPV - Hoat t ransfer — Chemical interact ions of coro materials

- Crust formation — Core mater ial propert ies data at h igh temperatures

• Develop codes for consistent analysts of experimental data

To create a knowledge base to support analysis of core melt retention in the lower head of the RPV

162

Matrix of Corium Property Measurements

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Melting température

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up to 2840 K

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View of the RASP LAV»AW-200 Experimental Section

1 - Strap 2 - f tpd . 3 - Pi Pp a - Top lid 5,1 B - Inductor 0 - C-22 corium 7 - Thermal insulation

13 -Catcher 14 »- Support 1 5 - S t e e l lest wall 1$ -Water cooling units 17 - C-100 Insulation layer 19 - Heated side wall 20 - Tungsten rod 21 -Additional shield 22 - High temperature

transducer

RASPLAV-AW-200-1 Test: Location of Main Temperature Measurement Points

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Slice geometry Side wall inductive heating method to simulate Qv

Use of compatible materials to prevent Interactions of eorium with structural materials

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RASPLAV-AW-200 Facility

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3 - Gas-vacuum system 4 - Protective chamber

165

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First test : Post-Test Examinat ion and Thermal Analys is

Obtain physical picture as full a» possible!

What for? 1. Understand a phenomena corresponding with all te&t phases

(healing up, quasi-steady-state, cooling down)

2. Take into account alt obtained Knowledge and experience for RASF*LAV AW-200-2 test preparation:

• efficiency improvements of heath 19 system, side wall integrity, f r e e i n g seals design;

• pretest calculations; - power input scenario, » initial loading and boundary conditions; • instrumentation system improvement »

166

Post-Test Examinat ion

What wo would tike to know after the first test? 1. See how it looks like after disassembling:

- state of side walls (protector, subprotcctor, heaters); • state of corium ingot and preliminary estimation

of melt volume; • corium leakage; • state of thermocouples,

2> Examine specific features of the ingot itself; • typical macrozones; • mïcrostructure of different regions of the melt; • melt fraction; • element and phase composition and location; • density and porosity • melt properties after test

Post-Test Thermal Analys is

- Energy balance

- Heat flux distribution

- Behav ior of facil ity c o m p o n e n t s

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- Natural convect ion

- Post- test uncertainty analysis

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RASPLAV-AW-200-1 Test: Thermal Hydraulic Results

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RASPLAV-AW-20CM Test: Mmn Results

Convective molten pool was established Molten corium temperature was higher than 2600°C Molten fraction was about 70% Separation of initial melt - Upper layer is enriched with zirconium - Lower layer is depleted with zirconium

T liquidas after the test was higher than before test

168

RASPLAV-AW-200^2 Test

• Major goals — Increase heat flux through the test-wall

— Melting volume as much as possible

— Try to stay in quasi-steady-state as long as possible

• Test preparation — Dry tests to investigate components behaviour — Pre-test analysis

• Test was conducted in May 24-th, 1997

RASPLAV-AW-200-2 Test; Corium Temperature during Test

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RASPLAV-AW-200-2 Test: Left Side of Corium Ingot during Facility

Disassembling

RASPLAV-AW-200-2 Test: Edge Cross Section of Corium Ingot

170

11

RASPLAV-AW*200-2: Main Test Results

Melting a significant fraction of the eorium loading was achieved;

Peak heat f luxes greater than 200 kW/m p was obtianed;

Natural circulation flow regime was established for about 50 minutes before the power shut off, and approximately 3Q minutes after;

There was leakage of a significant amount of the molten corium but the test equipment did not suffer any damage;

The data obtained appears to be appropriate for validation of the analytical models;

Of greater interest are the data on chemical interactions and material stratification.

TULPAN Experiments Tests: Corium and Material Study

Aims of TULPAN tests: • Technical solutions and materials testing; • Study of corium components stratification; • Study of interaction:

- beetwen corium components; - corium with its interface»

It was completed Corium Stratification Test on 13 May» 1997,

Now Near-by Wall Crust Formation Test is prepared to carry out in June 1997

171

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* Technical issues to conduct large scale experiments were resolved — Methods of the heating up of core materials beyond

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3D temperature f ie ld Flux d is t r ibu t ion

177

The OECD SCORPIO-WER Project

A.Kohsaka, T.Suzudo Japan Atomic Energy Research Institute

Tokai-mura, Naka-gun,Ibataki-ken,Japan

International Seminar on

The Safety Reserach Needs for Russian-Designed Reactors

8-9 July 1997, Tokyo, Japan

179

1. Introduction

The core surveillance system SCORPIO developed at OECD Halden Reactor Project

(Norway) has been widely applied to Western type PWRs because of its user-friendly,

automated core monitoring and prediction capabilities for both normal and transient

conditions of a reactor, which provide reactor operators with a practical tool to improve

plant safety and efficiency of operations.

Under an initiative of Japan Atomic Energy Research Institute (JAERI), with support by

the Science and Technology Agency (STA) of Japanese government, a proposal for a new

project, namely SCORPIO-WER, to apply SCORPIO to WER type reactors was

studied and found to be valuable by experts from NEA, JAERI, Halden Reactor Project

(HRP) of Institutt for Energietheknikk (IFE, Norway) and Czech organizations such as

Ceske Energeticky Zavody Inc. (CEZ) and Nuclear Research Institute (NRI) who

expressed their strong interrests to this proposal.

It has been agreed that the project is funded by STA through NEA, the period is for 2

years starting in 1996, IFE/HRP performs principal works to develop a VYER version in

cooperation with Czech partners targeting the Dukovany plant (WER-440/213) as a

reference plant, a specially formed project committee reviews technical progress, etc..

Although the Dukovany plant was selected as the reference, the ultimate objective of the

project, is to develop a generic version of SCORPIO which is easily applicable to all types

of WERs including WER-1000s.

2. Outline of SCORPIO core surveillance system

The development of SCORPIO core surveillance system commenced in 1979 at OECD

Halden Reactor Project where a lot of experiences in development and operation of in-

core instrumentation had been accumulated. The main design features of SCORPIO,

ref[l] are:

• Combination of measurements and calculation to provide the best estimate of

the core state

• Efficient predictive capabilities to calculate the core behavior in planned

power transients.

• Modular design

• Fast access and display of information through high resolution graphics

• User friendly operation.

180

SCORPIO is operated in two modes, core follow mode and predictive mode (see Fig. 1).

In the core follow mode, the system produces a realistic estimate of the core state based

on instrument readings, and calculations with a 3-dimensional physics model including

Xenon dynamics. As soon as the core state data is available, the system checks and

displays the margins to the operating safety limits.

Core A Limit i ^ Checking ^ s ^

Core Follow Mode core state estimation x

Predictive Mode core prediction strategy gene rati or

Fig. 1 Basic structure of SCORPIO core surveillance system

In the predictive mode, the system calculates the core behavior during a planned power

transient. This is of great help for reactor operation in dynamic core state situations where

Xenon variations often have a complex influence on power distribution. Thus, the

operator can avoid control strategies that are unacceptable due to operational constraints,

by inspecting the predicted margins to these limits for different strategies.

The SCORPIO system has been in operation at the Swedish nuclear power plant

Ringhals-2 since 1987, ref[2]. In addition the system has been installed at the Sizewell B

of Nuclear Electric, UK sincel993, ref[3], and at Catawba 1,2 and McGuirel,2 of Duke

Power Co.,US sincel994, ref[4].

SCORPIO runs on Ùnix-based workstations. The user interface is generated using the

PICASSO-3 graphic system which has also been developed by the Halden Reactor

Project.

3. Current core-monitoring system in the target plant, VVER-Dukovany

The standard core surveillance system for the Russian-type reactors is called VK3, in

which core-exit temperature and in-core neutron-flux signals are monitored, and nodal

1 8 1

power peaking limit and fuel assembly temperature rise are calculated on the basis of

simple 3-dimensional coarse mesh core follow power distribution, as the detailed 3-

dimensional pin-wise power distribution is not available. There is no predictive function

included, thus it is not possible to check the safety-related parameters in advance when a

transient operation is planned. In addition to VK3, there is the pellet-clad interaction (PCI)

margin calculation system, called PES system. However, this function runs only at the

operators request and has no on-line monitoring. A detailed description of the differences

between VK3 and the new SCORPIO, is given below, ref[5].

4. Adaptation of SCORPIO to the target plant, VVER-Dukovany

Although SCORPIO was originally designed for Western-type PWRs, it is flexible

enough to be modified for Russian-type PWR (WER) thanks to its flexible design

approach, e.g. its modular design: The same framework can be used by replacing the 3-

dimensional physics model.

The final goal of the project is to replace VK3 with the SCORPIO system. For this

purpose, 3-dimensional power distribution calculations based on the physics model called

MOBY-DICK, which was used for the WER-Dukovany's core design and safety

analysis, will be installed in SCORPIO system so that the system can analyze the reactor

core state based on 3-dimensional pin-wise power distribution. Thus the new surveillance

system will include many new functions in addition to all functions in the VK3 system.

Figure 2 shows the main modules of SÇORPIO-WER system.

1) Core follow mode

In the core follow mode, as mentioned in section 2, the core state is monitored based on

a combination of instrument signals and a theoretical calculation of the core power

distribution.

The plant measurement data is collected by the existing data acquisition system,

Hindukus. First, SCORPIO reads the measured signals from Hindukus and pre-

processes them. This pre-process includes identification of the operational regime (e. g.

the number of loops in operation), signal validations, and determination of the coefficients

for transforming into physical units etc.. There are a number of core-exit thermocouples

in VVERs, and one pending problem of the current system is to validate the accuracy of

182

Core follow system Predictive system

Hindukus (data acquisition

.system) Pre-processing and signal validation

Strategy generator

3D power distribution

3D power distribution

Limit checking and thermal margin calc.

PCl-margin calc., PES

1 Limit checking and thermal margin calc.

PCl-margin calc., PES

primary coolant monitoring PEPA

primary coolant monitoring PEPA

Input by operator

Fig. 2 Main modules of SCORPIO-WER system

these measurements. In SCORPIO, however, the core-exit thermocouple measurements are corrected by the isothermal reactor state.

After the pre-processing, the 3-dimensional power distribution calculation module is triggered, the power distribution is calculated not only theoretically, but also reconstructed by the core-exit temperatures and the in-core neutron detectors (SPND). The best estimate power distribution is determined from the calculated and measured power distributions. The simulator (MOBY-DICK) is adapted taking the best estimate power distribution into consideration;

Next, based on the estimated power distribution; the 3-dimensional pin-wise power distribution is produced, and the power peaking factor and safety-related parameters such as departure from nuclear boiling ratio (DNBR) and saturation temperature are determined, and their margins are checked on the basis of subchannel analysis. The current

183

surveillance system, VK3, can check these safety margins only on the basis of coarse

mesh power distribution, and therefore improvement of this function is expected.

In case of large power changes, pellet clad interaction (PCI) may occur that might

damage the fuel. The largest limits of local and global power changes must be therefore

determined. A module to evaluate these parameters, called the PES module, is included in

SCORPIO. In the existing surveillance system this function runs independently of VK3.

This calculation is conducted only at the request of the operators and accordingly has no

on-line monitoring.

Measurement of radioactive nuclides' activity in the primary coolant enables the amount

of damaged feul to be estimated. The module to conduct this estimation, called PEP A, is

also included in SCORPIO. This module can give information on not only the amount of

damaged fuel, but also the type of damage.

2) Predictiye mode

This mode is a completely new feature in the core surveillance of WERs, as there is no

such system available in any WERs. The predictive capability has proven to be valuable

in core surveillance systems for Western PWRs and this should hold in case of WERs.

In addition to the actual operations, the predictive capability is a great benefit also for

operator training.

In the predictive mode, the operator can calculate reactor behavior for the next, at most,

48 hours taking the initial condition from the core follow mode. As shown in Fig. 2, the

predictive system has a similar software scheme to the core follow system. In the

predictive analysis, the operator first has to specify the desired power trend to be

produced. Next the strategy generator proposes a possible core control method during the

power transient on the basis of analyses by a simplified core model. The proposed control

strategy is verified by .the 3-dimensional power distribution calculation, i.e. the predictive

simulator. As with core follow mode, a detailed 3-dimensional is produced for

determining the power peaking factors and thermal margins on the basis of subchannel

analysis. The PCI margin is also calculated in the predictive mode based on the 3-

dimensidnal power distribution calculation and the acceptability of the planned transient is

checked. The PEP A module in the predictive mode calculate the radioactivity level in case

the planned transient is actually used.

184

3) Software frameworks The SCORPIO system is comprised of a number of plant and fuel specific data that need

to be maintained. The data must be updated as new signal calibrations are done or the

burnup changes. Therefore, an adequate set of tools to edit the database is essential.

Because a new type of fuel will be partly introduced in the target plant in 1998, this

function will be even more important.

Man-machine interface (MM) is of great importance for the operators, as SCORPIO is supposed to help and support the plant operation. Much effort need to be put into the simplifing MM; user input items must be reduced to a minimum and procedure must be as simple as possible, only key information shoud be displayed while the details is to be hidden unless requested.

The software structure is designed to be composed of many modules, as mentioned above. Each module has an object-oriented structure with, a class, an instance and an attribute, and is responsible for one distributed process. This modular structure enables flexibility in designing the software and is of great advantage for maintenance. The modules communicate with each other via a TCP/IP socket, connection and, as a result, a module can send information to other modules.

5. Summary The OECD SCORPIO-WER project has been launched since 1996 to develop a WER

version of the core surveillance system SCORPIO [originally developed at the OECD Halden Reactor Project (Norway)], aiming at improvements of WER plant safety and efficiency of operations. The major new features of the proposed surveillance system compared with existing WER monitoring systems will be; inclusion of on-line 3-D power distribution calculation and signal validation, improved limit checking and thermal margin calculation, on-line monitoring of pellet clad interaction, predictive and strategy planning capabilities, a user-friendly man-machine interface, etc.. The project is funded by STA of Japan through OECD/NEA, and IFE/HRP of Norway,

CEZ toghether with NRI of Czech and JAERI are particpating in the project. Although the Dukovany plant of Czech is currently targeted as the reference plant, a generic version of SCORPIO which is easily applicable to all types of WERs is expected to be established through this project.

185

References

1. O. Berg, T. Bodal, J. Porsmyr, K. A. Aadransvik (OECD Halden Reactor Project)

" SCORPIO-Core Monitoring System for PWRs; Operational Experience and New

Developments ", ANS Topical Meeting on Advances in Nuclear Fuel Management

H, Myrtle Beach, South Carolina, March 23-26, 1997

2. T. Anderson (Swedish State Power Board), O. Berg (OECD Halden Reactor Project),

K. Romslo (Scandpower, Halden, Norway), "Ringhals-2 Core Monitoring

Experience", INCORE s96, Oct. 14-18, Mito, Japan

3. O. Berg (OECD Halden Reactor Project, Norway), M. McEllin (Nuclear Electric Ltd.),

M. Javadi (Sizewell B Power Station), " Aplication of the core surveillance system

SCORPIO at Sizewell B", INCORE v96, Oct. 14-18, Mito, Japan

4. S. K. Gibby, S. C. Ballard (Duke Power Company), "Implementation of the Core

Surveillance System SCORPIO at Duke Power Company", ANS Topical Meeting on

Advances in Nuclear Fuel Management U, Myrtle Beach, South Carolina, March 23-

26, 1997

5. K.Zalesky (NRI), J. Svamy (Skoda), L. Novak (Chemcomex), J. Rosol (Dukovany),

A. Hornaes (JPE), "International Topical Meeting on W E R Instrumentation and

Control", Prague, Czech Republic, April 21-24, 1997

186

The OECD SCORPIO-WEE PROJECT

1. OBJECTVE of The PROJECT

2. GENERAL FRAMEWORK of The PROJECT

3. OUTLINE of CORE SURVEILLANCE SYSTEM SCORPIO

4. WHAT IS NEW WITH The WER VERSION of SCORPIO

5. MILESTONES of The PROJECT

OBJECTVE of The PROJECT

To Develop a W E R Version of The Core Surveillance

System SCORPIO ( developed at OECD / Halden

Reactor Project ) to be used for improvements of

W E R Plant Safety and Efficiency of Operations

187

GENERAL FRAMEWORK of THE PROJECT

1. Sponser : Science and Technology Agency of Japan

2. Period : 2 Years (1996 - 1997 )

3. Participants (Project Committee ) :

OECD / NEA IFE / Halden Reactor Project CEZ sharing with NEI JAERI Finland ( Chairman )

4. Target Plant : Dukovany ( WER-440/213 )

Project Committee

Signatories

JAERI IFE/HRP

CEZ(sharing member NRI)

Chairman (VTT, Finland)

OECD/NEA

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the /.S.T.C. early 1992 Agreement signed November rm

presidential order on a "temporary" basis ISTC is registered as an intergovernmental organization headquartered in Moscow.

* Ike ISTC ; Mew '^rj^.wg^.w^is^JMOTLyii''"»!;'*»!^

joining the European Union Georgia, Armenia, Belarus,

The Republic of Korea plans to join in 1997.

214

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The Center develops, approves, and monitors S <§ T projects for peaceful

m Projects carried out primarily locaU C\ I; £J\

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to peaceful activmes

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Support basic and applied research Promote integration of scientists in

215

M u K T * The ISTC ; Activities To<

As of 1 July 1997: m 1240 Project proposais m 450 projects approved and m $145 Million committed (US: $62 M;

BU: $56 M; Japan: $ 25 M; Other: $2

i > Mm institutions, > receive grants from tSTC.

A Brood Range of Projects *ifr* w w w w w ^ t ^ W > W ^ w • C T ^ ^ W W ' W W ^ ^

Environment assessment and monitoring Nuclear Power Plant Safety Nuclear Waste Disposal

S>» * §*?/•

Materials, Aircrafts, Computers and many

216

Non-nuclear Energy

Materials 2 11%

Manufacturing Technology

1%

Instrumentation 7%

Information and Communication

4%

Fusion 5%

Fission Reactor; 19%

[ * *

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Fission / Reactors: /

3ther Studies * $16,600(65%)

Space, Aircraft and Surface

Transportation

Biotechnology and Life Sciences

8%

Chemistry 2%

Environment 25%

• *r^'^ry^^<>>r^y^^y^^^^^Pi

Nuclear Reactor Safety

$9,100 (35°/

217

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project

1, ,1# Title

ptew Experiments

1 116

371

Verification of Reactor Data Bases

Pu-Utilization in LWR Experiments

New Mathematical Methods and Software

067

068

115

Simulation of Reactor Dynamics

Parallel Computer Programming

Reactor Kinetics

New Data for Reactor Physics

304

183

609

B-003

Accelerator-Based Measurements of Minor Actinides Cross-Sections Fission Spectra of Minor Actinides

Heavy Nuclear Neutron Cross-sections

Evaluation of Actintde Nuclear Data

Leading Institute

NIKIET (ENTEK), Moscow

FEI (IPPE)

<

Keldysh Institute of Applied Mathematics, Moscow VNIIEF. Sarov

Keidysh Institute of Applied Mathematics, Moscow

FEI (tPPE), Obninsk

Khlopin Radium Institute, St. Peterburg

Nuclear Physics Institute, Moscow

Institute of Radiation Physics and Chemistry Problems, Minsk

Cost ($ 1000)

350,00

90.00

960.00

922.01

510.00

600.00

360.00

274.00

169.00

Funding Parties

EU

EU, Japan

EU, Japan, US EU, US

Japan

Japan

EU, Japan

EU

Japan

^joyWffSB&PJ iff * *+V* P**wi<^*vfty^ftj»*^y>i^^ ^'t!nf*«^?w^^^^wr^w?3?^^ !W5^<inNC'T£^s?*^^tfV^^

Mathematics for Dynamic Processes (#

Verification of Codes and Data 0116,371)

and Fission Products) f# 116, 371) Analysis and Minimization of Error

1 More Accurate Cross-Section Libraries

1

218

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Title

É ^ Modeling Nuclear Axtdents

UOCA 408

Acdderi 065

355

683

Explosive Melt-Water interaction

t Development Reactor Axident Loacing

Fire and Explosive Safety

Benchmarks for Chernobyl SmJation Protection Materials

064

066

425

Concrete Protection of Reactors

Reactor Wall Loacing

Zrconiun Cerarrics for Nuclear Reactor

Leacfing Institute

MR, Ivbscow

Res-Eng. Centre of Nuclear Rants Safety, Bectrogorsk

IVTAN (HghTenperatures), Moscow ISIvV^H Chernogolovka

VMIAES, Moscow

IVT/W (Hç^ Temperatures), Moscow \VTAN (H^ i Terrperatures), Moscow NVm (HgJiTernperaîures), Moscow

Cost ($1000)

97.64

654.00

95.03

360.00

528.40

380.60

623.00

Funding Parties

EU, US

EU, US

SJ, Japan, US

EU, US

EU, Japan, US

EU, Japan, US BJ

Theory and Experiment as « result of LOCA (# 408)

mi

Behavior under Core ignMon, Detonation

* v w j

219

file:///VTAN

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220

273

290

650

Title

FastSoduriFteactors

RadaBon GTaracterisfcics of M3X Fuel

Rtforium Utilization in Mdear Fteactots

Safety of M/anoed Fast Fteactors

Leacfng Institute

FB QFPE), Cbrinsk

VMlMVlBochvar, Moscow

VNlMVlBxhvar, Nbsccw

FB (IFFE), Obninsk

Oost ($1000) 750.00

80.00

450.00

750.00

Funding Parties Japan

BJ

BJ, Japan

Japan

i , w w w r y ^ g y K ^ ^ igiwjgiBajrtyyw''*,i^

Experimental and Analytic Study of Safety Reactors Utilizing New Fuels:

Other Pu Containing Fuels

Metallic Fuels with Inert Matrix

220

to ISTG goals ' '"y À

(About 3-4 per year)

Sponsors travel of CIS foreign coiahomtors

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in-country project management

m Funds transfer, procurement direct payment m On-site monitoring and auditing m Tax and customs exemptions

Reasonable IPR terms

2

p rrn Conchtsi

i i The ISTC is expected to continue its activities for a while

m Governmental action should be relayed by other sources of funding

m Better programmatic coordination, especially for nuclear studies

m ISTC could organize a seminar focused on nuclear safety issues.

222

1

ME)WHAPO£HHfî[ HAytffîO-TEXHIFïECKHÏÏ IJEHTP JlyraHOcaa ymma,9 A/JI251115$l6MocKBa, PoccHÔCKaH >e^epaipiH Teœ (7-095)321-4665 4>aKe (7-095) 321-4744

The International Science and Technology Center (ISTC) and

ISTC Projects Related to Nuclear Reactor Safety Problems (Information Review)

A. Gerard, L V. Tocheniy, ISTC - International Science and Technology Center, Moscow, Russia

1. ISTC - history, state- of- art, perspectives.

The JSTC is operating under the auspices of an intergovernmental agreement between the Russian Federation, European Union, Japan and the United States. The Center started operation March 2nd, 1994. Since then Finland and Sweden have acceded as Funding Parties. CIS representation has expanded to join Georgia, Belarus, Armenia, Kazakstan and Kyrgyzia. All work of the ISTC is aimed at goals defined in the ISTC Agreement:

To give CIS weapons scientists, particularly those who possess knowledge and skills related to weapons of mass destruction and their delivery systems, the opportunities to redirect their talents to peaceful activities;

To contribute to solving national and international technical problems; To support the transition to market-based economies; To support basic and applied research; To help integrate CIS weapons scientists into the international scientific

community. The ISTC engages in a variety of activities aimed at meeting these goals:

Providing support to scientific research and development. Stimulating collaboration of CIS institutes with foreign institutes, industries

and universities. The Parties make great efforts to seek foreign partners that are interested in the proposed projects, to support meeting of the project recipients with foreign collaborators and their participation in scientific conferences.

Organizing of the ISTC Seminar Program on subjects of national and international interest. The projects may be funded both through governmental funds of the Funding Parties specified for the ISTC, and by organizations, nominated as Funding Partners of ISTC. According to ISTC Status, approved by appropriate national organizations, funds used within the ISTC projects are exempt from CIS taxes. Projects range from solving environmental problems related to nuclear industries and nuclear safety to the development of new vaccines for contagious diseases.

223

INTERNATIONAL SCIENCE AND TECHNOLOGY

CENTER Luganskaya uiltsa, 9

PO Box 25,115516 Moscow, Russian Federation

Têt (7-095) 321-4665 Fax (7-095) 321-4744

I S T C

M H T II

2

As of July 1997 above twelve hundred proposals had been submitted to the Center, of which 450 approved for funding, for a total value of approximately US$145 million. The number of scientists and engineers participating in the projects numbers about 20,000 / 1 / .

2. Projects Related to Nuclear Reactor Problems.

There are about 50 funded and as of yet unfunded projects related to various problems of nuclear reactors and nuclear power plants. Many of them address safety issues. The ISTC favors the coordination of the projects flow through participation at joint project workshops, seminars etc. Below is a short description of some groups of ISTC projects. 1. SAFETY OF NPP. a). New mathematical methods and computer codes: The joint teams of mathematicians, experts in the modelling of both nuclear explosive neutronics and nuclear reactors, develop in the framework of Projects #067, #068, #115, #250 and #259 new approaches for the calculation of neutron transport and nuclear reactor 3D dynamics. The methods previously used for solving fast processes in bombs will be used for improving the modelling of the reactor transfer processes. The projects develop more than a dozen new codes on the basis of parallel-type algorithms. b). Verification of codes and data: Two projects (#116 and #271) aim to develop methods, codes, data and experimental base for theoretical and experimental investigations of LWR(WER) Pu-content cores loaded with fresh fuel (#271 ) and fuel corresponded to m id-burn-up state (#116), It is planned to use both existing experimental subcortical and critical installations, previously used for weapon purposes, and to develop new ones. Neutron cross sections and data for actinides which are important for the U-Pu loaded reactors and fuel cycle with reprocessing and transmutation are being measured and updated by Projects #304, 305, 183,217,609. A group of projects (#453, #610) deals with the structure of nuclear reactor and power plants and plans to develop new codes for complex analysis of thermal, hydraulic and mechanical forces operating at the nuclear unit under normal and emergency regimes. Projects #100 and #109 compare the US and RF's approaches and norms of Strength Regulations for naval and power nuclear reactors. c). Severe accident analysis: Project #487 proposes the concept design of MIGR - a multichannel, pulse, experimental and test reactor for simulation of accident conditions within the reactor core including fuel melting. There is a set of projects developing a theory for ignition, fire and explosiveness of hydrogen content gaseous mixtures and studies of the impact of shock waves and corium on reactors and plant structures (#064, 065, 066, 355, 425). These projects coincide with the development of new protection materials - zirconium content concretes and ceramics for walls, melt core catcher and containment. Project # XXX includes the calculation 3D models for severe accident scenarios, including neutronic, thermal, chemical and mechanical stages. Project # 646 developes methods of risk analysis of reactor systems.

224

3

dlReactor Engineering Systems: Development, design and in-pile reactor tests of fast acting reactor control rods and control systems are planned within the Project #140. Fast acting emergency overlappings of the primary circuit and piping are suggested at the Proposal # 235, Project #502 will examine the hydrodynamic stability of hélicoïdal steam generators at two experimental installations. Projects #566, #595 aim to prolong the life-time of reactor vessels. Study of zirconium radiative swelling will be developed by project #681. The technology for production of materials with a defined content of 10B and 11B isotopes is suggested by the Georgian proposals #G-030 and #G-031.

NEW REACTOR AND REACTOR COMPONENTS CONCEPTS e).New Reactor Concepts. There is suggested to develop the conceptual design of power fast reactor for burning of Pu and minor actinides, computering and experimental analysis of reactor parameters and their self-regulating features for two reactor options: helium cooled (Project #269) and heavy-metal (lead) cooled (Proposal #071) ones. The Russian technical experience of using navy reactors is used within the latter project as well as within Project #141 ( reactors for remote areas). Experimental, engineering and analytical studies were carried out in the framework of the Project #220 for nuclear safety parameters of advanced sodium-cooled fast reactors with renovated oxide, carbide, nitride and metallic Pu-cbntent fuels. In-core tests and continuation of these investigations are planned within the next stage of activity (Project #721 ). The options of uranium-thorium fuel cycles with specific branches for weapon plutonium burning are suggested by Proposals #077, #313 and #723. Experimental modelling of thermo-hydraulics and study of other extremely high physical parameters of plutonium burning reactors with melted-fuel core are under way in Project #134. f). Fuels and Fuel Elements: Project #173 deals with the development of a technology and non-irradiation study of new fuel elements for LWR (WER)-type reactors based on metallic or silicide dispersion fuel for extremely high burn-up - up to 150 000 MW-d/1 (U). The radiation in in-core tests is planned for the next stage of the Project ( # 721). PLUTONIUM. MOXFUEL g). Plutonium ad actinides disposal: A study of various aspects and prospective technologies for utilization or for destruction of weapon and civilian plutonium is underway within a set of more than twenty projects. One group of projects has a particularly close relationship since the projects deal with the use of Pu as fuel for civil power reactors, and so it focuses on the specific reactor aspects. The other group attacks problems with an alternative approach based on subcritical systems (blankets) driven by accelerators of charged particles (#017, #442, # B-70). Project #369 would contribute to genera! strategies for the use of weapons-grade plutonium in Russia. It related to studies of the technological and economical feasibility of the use of plutonium in some definite scenaria of fuel cycles based on

225

4

both fast and thermal reactors and both existing and developing fuel manufacture technologies. h). MOX fuel: Project #534 plans to update the analytical and experimental parameters of MOX fuels (mechanical, thermal, radiative and so on) to verify models of fuel elements and to predict its behavior under working load. Project #290 would develop and test on a laboratory level a new compact technology to convert metallic plutonium to oxide form (MOX). The developed MOX fuel promises to increase burn-up indexes and to improve the efficiency of the fuel cycle. This technology fully corresponds to the industrial technology of the MAYAK plant (one of #369 items). In addition to the results listed above, the foliowingwill also be abtained: # physical, mechanical, corrosion and structure data of specific U-Pu metal alloys as a potential feed material for developed technology; alloys will be obtained and processed; # ceramic, from which MOX fuel granules will be made, with low dust properties. Proposal #565 plans to investigate properties of U-Pu-Zr metal fuels and fuel elements on this basis for fast reactors when steady and non-stationary loads take place. Project # 272 aims to develop a concept of nuclear reactor and fuel cycles for plutonium and actinides transmutation based on vibro-packed plutonium fuel. Methods for extraction of fission products from plutonium oxide fuels are developed by Project #279. RADIATION SAFETY AND DECOMMSSIONING D.Environment:: Project #273 will define data necessary to examine the out of core environmental impact of MOX during fuel fabrication and reprocessing, which coincides withl one of the objectives of Project #369, mentioned above. Project #281 aims to recommend approaches for improving reprocessing technologies in order to minimize technological radioactive wastes after the reprocessing of spent nuclear fuel. Project #330 develops a technology for the removal of plutonium and other transuranium elements from the silt of radioactive waste storage ponds. The joint Russian-Kazak Project #517 will measure the result of scattering of vertical neutron flux by atmospheric clouds and mist. This data will be used to accept the NPP and reprocessing plant sites when its are nearby the populated area. J).Storage: A feasibility study of methods and approaches to temporary storage of large masses of weapon-grade plutonium and uranium has been developed by the first stage of Project #332. A conceptial design of the storage is been fulfilled by the second Project stage. Proposal #538 proposes to develop a basis for the technology of compacting the non-reprocessed part of spent fuel of different types reactors for long-term environmentally-safe storage. k). Decomissionina: Principal aspects of the Decommissioning of transport and experimental reactors will be summarized in Project # 561.

226

5

Analysis of radioactive contamination of graphite of decommissioned weapon Pu productive reactors will be used for technology of incineration of RBMK graphite (Project #569).

5. Conclusion The fulfillment of ISTC projects shows their high level of scientifical and

technological potential. The majority of the proposed projects is of great interest for the international nuclear community. It it advisable to establish more active partnership between project recipients and foreign organizations and firms to define exactly project content and mutual interests, in order to find acceptable forms of funding and collaboration, using the advantages of the ISTC Agreement.

Reference I The ISTC Annual Reports - ISTC, Moscow, 1994, 1995, 1996.

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2. Russian Standard Safety Problem SSP-2

on Russian ISB-VVER Test Facility

Main Characteristics of ISB-VVER

Reference plant VVER-1000

Volume Scaling Factori : 3000

Elevation Scaling Factori : 1

Nominal Operation Pressure"! 6 Mpa

Number of Loops2 (1:3)

Boundary Conditions

Leak Opening at Upper Plenum (11% Leak)

ECCS not available

After 3 s >Start of Electrical Power Drop (954 kW to 100 kW)

>Start of Steam Generator Isolation

After 8 s >Main Circulation Pump Stoppage

End >Maximum Cladding Temperature archieves 723 K

Code Version used:

ATHLET Mod. 1.1 Cycle C

Conclusions

• The overall transient was calculated very well

• All essential phenomena are well matched

• In general parameter are in good agreement with the experimental ones

SSP-2 proved to be a successful and valuable exercise for thermohydraulic code user to get experience in the field of code handling and safety analyses

240

outlet 2 c

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Isometric view of ISB-VVER test facility

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PMK-Experiment Inadvertent Opening of Pressurizer Safety Valve

1. Initial conditions, measured and analysed for the zero-transient

Parameter

Core power

Pressure in upper plenum

Loop flow

Core inlet temperature

Coolant level in pressurizer

SIT-1 pressure (downcomer injection)

SlT-2 pressure (upper plenum injection)

SIT-1 level

SIT-2 level

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Feedwater flow

Feedwater temperature

Steamgenerator level

Unit

kW

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kg/s

K

m

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MPa

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m

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PR21

FL53

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PR92

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FL81

TE81

LE81

measured

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5,927

5,875

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10,054

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0,347

495

8,58

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535

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5,927

5,875

10,052

10,054

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0,347

495

8,25

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Action

Opening of pressurizer safety valve

Scram initiation

Closing of feedwater and steam line

HPIS starts

SIT's actuated at PR21

Pump coast down initiation

Steam relief valve opening

1. maximum of leak massflowrate

Pressurizer full filled

End of pressurizer top level

Parameter + Time

9,45 MPa

9,45 MPa + 6 s

9,45 MPa+ 17 s

6,036 MPa

-

5,065 MPa

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10,26 m

10,26 m

Time in sec

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102,8

113,8

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G3S

Coupling of ATHLET with BIPR8

The 3D neutronics code BIPR8 developed for WWER-

440 and WWER-1000 by Kurchatov-lnstitut was coupled

with ATHELT for transient analysis.

First validation by calculating a single pump coastdown

(1 out of 4) for a WWER-1000 from a startup experiment.

The event sequence:

• initial power 95 % Nnom

# pump in loop 1 is switched off • the power limitation device (ROM) start to reduce

reactor power to 71 % by inserting control rods • later, the second ROM action reduces power to 62 % # afterwards, power controller keeps reactor power

constant • according to the signal «ROM actuation" the turbine

control reduces turbine power, while keeping steam-line pressure nominal.

The comparison of calculated and measured parameters

is shown in the pictures.

CLMJap.Doc 04.07.97 (Zag)

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G2S

Analysis of RBMK Core Smolensk 3

• Analysis of reactivity balance for full core

• void effect

• fuel temperature effect

#* graphite temperature effect

• reactivity change between full power and hot zero

power condition

# Comparison of results from three 3D neutronic codes

On the basis of different nuclear cross-section libraries

• RRC-KI, two versions

• RDIPE

CLMJap.Doc 03.07.97 (Zag}

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/ H— V3Ï»

Void Reactivity Effect in RBMK Cores

The void reactivity as function of coolant density is

calculated for three RBMK plants. The core conditions

are characterised by a different number of additional

absorbers.

RBMK-1000

Smolensk-3 97 additional absorber

Leningrad-4 82 additional absorber

RBMK-1500

lgnalina-1 52 additional absorber

The dependency shows a strong nonlinear behaviour. It

depends on insertion of control rods, which determines

the axial positions of absorber element, graphite

displacer and water column, and the burnup distribution.

CLMJap-Doc 03.07.97 (Zag)

259

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260

Analysis of Unintended Withdrawal of Central

Control Rod in RBMK1500 at Full Power

The picture shows the power increase in the hot channel

during the first 12 s of the transient.

Variant 1 assumes that no other control rod is moving.

Variant 2 assumes that the power is controlled by

moving rods of the Local Automatic Control System

(LAC).

The LAC system compensates the local power change,

keeping the total power constant.

Without action of the LAC system the total power is rising

about 22 % in the given time intervall.

CLMJap.DûC 03,07.92 (Zag) 261

LB

Central Rod Wi thdrawal at 4 2 0 0 MW QUABOX/CUBBOX

1.4

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i—i—i—i—rh—i—J—i—i—rn—\—i—i—i—r 5.0 7.5 10.0 12.5 T i m e ( s e c )

262

BILATERAL RESEARCH PROGRAMMES OF WORK BETWEEN

FRANCE AND THE CEEC/NIS IN NUCLEAR SAFETY

M.REOCREUX (ISPM - FRANCE)

INTERNATIONAL SEMINAR ON THE SAFETY RESEARCH NEEDS

FOR RUSSIAN DESIGNED REACTORS

Kokhu Kalkan, TOKYO, 8-9 July 1997

263

CONTENT

BILATERAL PROGRAMMES WITH KURCHATOV INSTITUTE

O 1.ANALYSIS OF RIA TESTS PERFORMED ON IGR

• 2. HYDROGEN TESTS ON RUT

• 3. CODE MODELS DEVELOPMENT AND CODE VALIDATION

BILATERAL PROGRAMMES WITH IBRAE

• DEVELOPMENT OF MODELS FOR ICARE CODE

O CORIUM SPREADING STUDIES (CROCO CODE)

O PREPARATION OF ACCIDENT SCENARIOS FOR CRISIS EXERCISES

CONCLUSIONS

BILATERAL PROGRAMMES IPSN/KURCHATOV (1 )

O 1. ANALYSIS OF RIA TESTS PERFORMED ON IGR (Semipalaîïnsk)

• JOINT PROGRAMME WITH US-NRC

• TESTS PERFORMED IN FORMER TIME ON HIGH BURN UP FUEL

• PROGRAMME CONTENT

• STUDIES WITH COMPUTER CODES FRAPT6 and SCANAIR - Modification of codes (heat transfer, properties, specific models) - Code verification and analysis based on IGR tests results

• EXPERIMENTAL PROGRAMME - Mechanical properties tests - Post irradiation examination

• PLANT ANALYSIS OF W E R TRANSIENTS - Coupled thertnalhydraulics neutronics calculations

264 IJ1ISI m=. WEST

BILATERAL PROGRAMMES IPSN/KURCHATOV (2)

D 2. HYDROGEN TESTS ON RUT

• JOINT PROGRAMME WITH F2K and US-NRC

• OBJECTIVE: INVESTIGATION OF THE TRANSITION BETWEEN DETONATION DEFLAGRATION

• TESTS PERFORMED

• In different geometries, • For various steam air hydrogen mixtures, • Homogeneous or non homogeneous initial concentrations • With or without H2 injection during the test.

iWM.

WUUUIW. »>14! «Jtflfi-'JiCTdCSH^PPîBi

BILATERAL PROGRAMMES IPSN/KURCHATOV (3) MW»«nn«U^IJIIIIi»lttMMUMMll»|l|»WI»IM>MM«»B>«WWWMM»^^ • M l — IMIIIIBHI IHI I I IMI I l lUlim H T f f — « P — E H — I — W C B — P — — « — t

• 3. CODE MODELS DEVELOPMENT AND CODE VALIDATION

• 3.1. ICARE CODE

• Validation of ICARE on CORA WWER 2 (ISP36)

• Uncertainty and sensitivity analysis with SUNSET / iCARE

• Future activity: plant calculation with CATHARE-ICARE

• 3.2.ESCADRE CODE

• Calculation of Plant sequences and Code adaptation to WWER

• Adaptation for Portability on PCs

265

nmiif>i,wiiiin—wmi—niiiimnwrn» TIWW» ammmimmmmimimmmmmimmmmmiiimmiiimmmaKiimmmimm


BttJgPBMllMWWWnMJBBKWBWWWWWWMMjWWWI

• 3. CODE MODELS DEVELOPMENT AND CODE VALIDATION (Cont)

• 3.3 SCANAIR CODE

• DEVELOPMENT OF A THERMALHYDRAULIC MODULE

• 2D homogeneous modelling

• Heat transfer in fast transient and subcooled conditions

• Coupling with SCANAIR (tight coupling)

• DEVELOPMENT OF ON LINE GRAPHICAL VIZUALÏZATION OF THE RESULTS

m$

m


BmmimnmmamnzsKZ

D 3. CODE MODELS DEVELOPMENT AND CODE VALIDATION (Cont)

• APPLICATION TO:

• Plant Transients • NSRR Tests

• FUTURE ACTIVITIES:

• Introduction of the variation of cross section in the thermalhydraulic module in order to take into account the clad ballooning

• Continuation of code validation

BILATERAL PROGRAMMES IPSN/IBRAE (1)

• DEVELOPMENT OF MODELS FOR ICARE CODE

• DEVELOPMENT OF MECHANICAL MODEL FOR CLADDING

• Modelling of mechanical resistance of Zr02 and Zr (oc) layers

• MECHANISTIC MODEL OF INTERACTION BETWEEN CLADDING, FUEL PELLET and STEAM

• Modelling of oxidation, diffusion and dissolution phenomena

• SIMPLIFIED MODEL OF LOWER HEAD MECHANICAL BEHAVIOR

fljjjL

§13


D DEVELOPMENT OF MODELS FOR ICARE CODE (Cont)

• DEVELOPMENT OF PHADI

• Simplified data base for U-Zr-O-Fe mixtures developed from the GEMINI data base

• DEVELOPMENT OF MFPR (Mechanistic Fission Product Release Module)

D CORIUM SPREADING STUDIES (CROCO CODE)

• Littérature survey and synthesis on heat transfer correlation between corium/ water in stratified situation

• 2D-3D solver for asymptotic equations

• Littérature survey and synthesis on Solidification process

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O DETACHMENTS of SCIENTISTS in Cadarache

H PREPARATION OF ACCIDENT SCENARIOS FOR CRISIS EXERCISES • St Petersbourg • Kola • Becquerel

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CONCLUSIONS

n THESE BILATERAL PROGRAMMES ARE OF GREAT MUTUAL INTEREST

O THEY REPRESENT AN EFFICIENT WAY FOR TRANSFER OF TECHNOLOGY IN BOTH WAY

D IT SHOULD BE PURSUED IN THE FUTURE WITH EVENTUALLY INFLEXIONS IN ORDER TO MAXIMIZE THE BENEFITS FOR BOTH PARTNERS

BILATERAL RESEARCH PROGRAMMES OF WORK BETWEEN

IPSN/GRS AND THE CEEC/NIS IN NUCLEAR SAFETY WITHIN EU PROGRAMMES

K.LIESCH (GRS-Germany) M.REOCREUX (IPSN-France)

INTERNATIONAL SEMINAR ON THE SAFETY RESEARCH NEEDS

FOR RUSSIAN DESIGNED REACTORS

Kohku Kaikan, TOKYO, 8-9 July 1997

269

TACIS PROGRAMME WITH THE RUSSIAN SAFETY AUTHORITY

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• OBJECTIVE

Transfer of the Accident Analysis Codes to the Riussian Nuclear Safety Authority Gosatomnadzor and Technical Safety Organizations (TSOs) and Application of these Codes.

a CONTRACTOR RISKAUDIT

D BENEFICIARY GAN-RF

• SUBCONTRACTORS IPSN / GRS

® STARTED 03.09.93

# TIME PERIOD for execution Sept 93 - Feb 97

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GENERAL PROGRAMME CONTENT

0 1. TECHNICAL INFRASTRUCTURE

• 2. ASSISTANCE TO ACCIDENT AND TO ACCIDENT MANAGEMENT ANALYSES

• 3. ASSISTANCE TO SEVERE ACCIDENT ANALYSIS

© 4. ASSISTANCE FOR THE METHODOLOGY OF THE RUSSIAN CODE VALIDATION

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1. TECHNICAL INFRASTRUCTURE

• HARDWARE AND SOFTWARE SUPPLY

# Supply of additionally necessary hard-and software to interconnect the local area network of GAN-RF the already existing wide area network in Moscow

S Supply of 7 Work stations RISC 6000 and 10PCs with X terminal emulation software with uninterruptible power supplies (UPSs)

# Supply of related system software, graphics software products, peripheral devices

# Supply of communication and networking components for LAN/WAN

D TRAINING OF RUSSIAN SPECIALISTS

• MISSIONS 7 missions in Russia

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2. ASSISTANCE TO ACCIDENT AND ACCIDENT MANAGEMENT ANALYSES (1)

G SUPPORT TO CODE TRANSFER ATHLET CATHARE

RALOC DRASYS

• ASSISTANCE IN CODE MASTERING AND USE

• WORKSHOPS ON: • Code modelling and numerics • Code structure

• OVERVIEW ON CODE VALIDATION

• EXERCISES ON • PMK-NVH 4 (SB - WWER • PACTEL

• SB LOCAON ISB-WWER 1000 • Input data deck • Steady state calculations 4 Sensitivity studies (nodalization, physical models) • Complete sequence for validation (ATHLET / CATHARE)

271

2. ASSISTANCE TO ACCIDENT AND ACCIDENT MANAGEMENT ANALYSES (2)

D ACCIDENT AND AM ANALYSES

# Description of the data sets (ATHLET/CATHARE) to perform DBA calculations for WWER 440 and WWER 1000

# Sequences on WWER 1000 • AB initiated by the rupture of one accumulator discharge line with loss of off-site power + loss of ECCS • SGTR aggraved by loss of off-site power

# B.E. calculation of BALAKOVO 3 (results --> initialization of Sev. Ace. sequence) • Input data report • Steady state calculation report • Transient calculation report

• MISSIONS AND DETACHMENTS *M8 Missions in Russia © 23 Detachments in FRANCE /GERMANY

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a SUPPORT TO CODE TRANSFER • ESCADRE • ICARE2 • ATHLET CD

D KNOWLEDGE EXCHANGE AND ASSISTANCE TO CODE MASTERING AND USE

# WORKSHOPS ON: # Phenomenology # Description of the French/German codes * Status of validation

• EXERCISES ON SEQUENCES WITH THE 3 CODES

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O SEVERE ACCIDENT ANALYSIS

• PREPARATION OF ESCADRE/RALOC CALCULATION

# LB Loca with failure of electric power supply (AB seq)

# LB Loca with failure of ECCS and containment spray system (ACD seq)

# Analysis of the main processes - thermalhydraulic in primary circuit and containment - in vessel core degradation - FP behavior - Corium concrete interaction - H2 burning - Iodine behavior

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O SEVERE ACCIDENT ANALYSIS (cont)

* BALAKOVO 3 core degradation + FP release ICARE 2 and ATHLET CD

# BALAKOVO 3 whole accident sequence ESCADRE + RALOC

O MISSIONS AND DETACHMENTS G 12 Missions in Russia • 9 Detachments in France / Germany

273

4. ASSISTANCE TO RUSSIAN CODE VALIDATION

D PRESENTATION OF VALIDATION OF FRENCH/GERMAN CODES AGAINST CSNI Code Validation Matrices

Separate effect tests Integral tests

O PRESENTATION OF PKL AND BETHSY PROGRAMMES Preparation and interpretation of tests with ATHLETand CATHARE

O SEVERE ACCIDENT CODES VALIDATION

H UNCERTAINTY ANALYSIS

O SEC-NRS DATA BASE SYSTEM STRUCTURE

• MISSIONS 6 Missions in Russia

274

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CONCLUSIONS (1)

CONTINUATION OF TACfS PROGRAMMES

• FOLLOW UP PROJECT FOR CONTINUATION WITH GAN RF UNDER DISCUSSION

• START UP OF A TACIS PROGRAMME WITH NRA-UA

• PROGRAMME ON TECHNICAL INFRASTRUCTURE - Procurement and installation of necessary hardware and software - Support to data processing experts

• SUPPORT AND ASSISTANCE IN APPLICATION OF FRENCH GERMAN CODES -ATHLETSBLoca - CATHARE LB Loca - RALOC DRASYS WWER 440/213 confinements - ESCADRE ICARE Severe accident

• INFORMATION AND ASSISTANCE IN MASTERING AND APPLICATION OF CODES - Transfer of codes ATHLET,CATHARE, RALOC/DRASYS, ESCADRE/ICARE - Assistance in code mastering and use - Assistance to the analyses of accident sequences

CONCLUSIONS (2)

SOME STATISTICS ON TACIS GAN-RF

20 SEC-NRS Specialists trained by 52 IPSN/GRS Specialists with 43 workshops in Moscow and 32 medium term detachments in France and Germany

IMPORTANT HARDWARE AND SOFTWARE INFRASTRUCTURE INSTALLED

GOOD MASTERING OF THE MAIN EUROPEAN UNION CODES

GOOD RELATIONS BETWEEN THE SPECIALISTS OF THE 3 ORGANIZATIONS

NECESSITY TO CARRY ON WITH THESE RELATIONS

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International Nuclear Safety Center

Bilateral Program of Work Between the U.S. and CEC/NIS in Nuclear Safety

V, 0F « ^

D. J. Hill ANL/INSC

International Seminar on the Safety Research Needs for Russian-designed Reactors

July 8-9, 1997 Tokyo, Japan

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U.S. DOE Bilateral Programs

Programs directed by DOE, Office of Nuclear Energy, Science and Technology: Dr, T. Lash, Director

The major program is the, International Nuclear Safety Program (INSP) managed by Pacific Northwest National Laboratory (PNNL) includes the Chernobyl Initiative the focus is risk reduction for operating Nuclear Power Plants

The International Nuclear Safety Center (INSC) at Argonne National Laboratory has a research focus, no direct support for Nuclear Power Plants

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international Nuclear Safety Center

International Nuclear Safety Program

• International Nuclear Safety Program (INSP) Managed by Pacific Northwest National Laboratory (PNNL) Involves many contractors and other U.S. laboratories

• The goals of the INSP are: Risk reduction at Soviet-designed NPPs Safety infrastructure improvement in host countries

• Strategically this is carried out by work with: NPP operators, regulators and TSOs leading to self-sustaining nuclear safety improvement

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International Nuclear Safety Program * » Of

Plant Safety Evaluation is a key component of the INSP undertaking in-depth safety assessments for Kola, Novovoronezh, Leningrad, Kursk in Russia several plants in Ukraine teams are led by plant staff with management support from Western contractors technical support from institutes in-country

Infrastructure improvement is other component of Plant Safety Evaluation training, code acquisition, provision of computers, etc.

Assisting in the development of a regulatory infrastructure is another key goal for the INSP

U.S. insistence on regulator involvement in safety assessment a strengthened regulator will assist sustainable improvement in safety culture

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U.S. NRC Programs

U.S. NRC conducts programs complementary to USDQE projects

Emphasis on strengthening regulator development of regulations development of inspection program development of enforcement program

Technical areas PRA, fire protection, etc.

Example NRC Programs

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Exclusive collaboration with Government of Ukraine to assist in development of national nuclear legislation Collaboration with Gozatomnadzor in Russia to perform a PRA at Kalinin

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INSC/RINSC *»*Y o f

The International Nuclear Safety Center (INSC) was established at ANL announced by former Secretary of Energy OXeary in Vienna at the IAEA Congress in September 1994

Minister Mikhailov announced formation of Russian International Nuclear Safety Center (RINSC) in September 1995

finally established in July 1996

The role of the INSC is complementary to that of the program for improving the safety of Soviet-designed Nuclear Power Plants

safety research rather than plant safety upgrades

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INSC Vision * » Of C

Background

Safety requires a philosophy of continuous improvement Declining budgets everywhere

shrinkage in technical capability loss of staff, facilities and knowledge

Goal

• Build, sustain and enhance infrastructure to support safety technology for NPPs world-wide

• Prevent technical isolation as budgets shrink national programs will continue to be replaced by international programs

• Ideally multiple centers with common understanding of safety issues

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Role of Centers with Respect to Russia

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has an international reputation and presence prevent technical isolation

Impart Western standards, for example in code version control, documentation, etc.

RINSC must ultimately be self-sustaining

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Collaboration between INSC and RINSC

The INSC undertakes research for U.S. domestic LWRs as well as participating in international programs, »,

A strong domestic research program provides the basis for collaborations.

Domestic programs include: Reactor Operations Technology Risk Management (Safety)

Spent Fuel Minimization Reactor Materials Research

Collaboration takes the form of joint projects. Professor Bougaenko of RINSC will describe the projects.

Joint projects can have different motivations, but, all address identified research needs.

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Safety Research Needs for Russian-designed Reactors NEA/CSNI/R(96)12

Recommendations:

A Safety Research Strategic Plan should be developed. key players should be involved RINSC has begun to develop such a plan with INSC support

International cooperation in safety research should be encouraged... INSC/RINSC collaboration has begun (seven joint projects)

New approaches such as technical fora for specific technical topics... both INSP and INSC are sponsoring technical meetings need greater OECD involvement

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Safety Code Validation

Strong reasons for validation derived from U.S. bilateral programs in-depth safety assessments and regulatory infrastructure

Relationship to existing NEA activities U.S. staff delegated to NEA to assist in this area Research needs report also identifies this as a crucial area

INSP-funded program in safety code validation, verification and certification supported by INSC/RINSC task in validation task leader Blinkov (Electrogorsk)

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^^M^^ Safety Research Needs for Russian-designed Reactors NEA/CSNI/R(96)I2

S'TY OF

Thermal Hydraulics/Plant Transients for VVERs

Recommendations:

3.1.3.3. Improvement of VVER Code Validation Matrix and Development of Russian Data Bank (Test Facilities and NPPs)

3.1.3.4 Additional Validation of Thermal-Hydraulic Codes Joint project on code validation, closely coordinated with NEA activities (joint meeting was held in Russia, May 1997)

3.2.3.1 Improvement of Dynamic Reactor Physics Codes for VVER Accident Modelling

3.2.3.2 Validation of the Coupled 3-D Neutron Kinetic, Thermal-Hydraulic, Thermal-Mechanic Codes

Joint project on comparison of 3-D coupled codes

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Safety Research Needs for Russian-designed Reactors NEA/CSNI/R(9C>)12

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Integrity of Equipment and Structures for VVERs

Recommendation:

4.3.6 Model Development to Predict Containment Performance

Joint project on modelling for containment performance, especially pre-stressed concrete.

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Safety Research Needs for Russian-designed Reactors NEA/CSNI/R(96)12

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Severe Accidents for VVERs

Recommendation:

5.3.5 Code Development, Validation and Applications

Joint project on severe accident management

5.3.3 Material Properties and Interactions Database

Joint projects on materials properties in database Completed project on corium viscosity (support for RASPLAV)

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Operational Safety Issues "Mr of «"

Mainly the province of the INSP

presentation by J. Honekamp

Joint project to evaluate diagnostic capabilities

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Thermal Hydraulics/Plant Transients for RBMKs

Recommendation:

7.3,2 Code Improvement and Validation

Joint projects on coupled codes and code validation

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Integrity of Equipment and Structures for RBMKs

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Nothing within INSC

INSP has projects to evaluate inspection technologies for fuel channels and to address multiple fuel channel rupture effects.

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Severe Accidents for RMBKs

Recommendation:

Joint project on Severe Accident Management as applied to RBMKs.

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Mechanisms for Data Exchange

INSC Database Materials properties

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RINSC Database Thermal Hydraulic data

Meetings Severe Accident Information Exchange Meeting (INSC) Obninsk, September 15-19, 1997

Plant Safety Evaluation Information Exchange Meeting (1NSP) Obninsk, September 24-27, 1996 and September 8-12, 1997

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Summary

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The USINSC.and RINSC have begun a number of joint projects that address the technical recommendations in the NE A report.

• Other recommendations are addressed by INSP projects.

The RINSC with USINSC support, have begun to develop a strategic plan for safety research for MINATOM.

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Co-operative Programs of Work between the United States and Russia

in Nuclear Safety through the Russian International Nuclear Safety Center

Prepared by S. Bougaenko RINSC, ENTEK, Moscow, Russia

At the International Seminar on the Safety

Research Needs for Russian-Designed Reactors "Kohku Kaikan", Tokyo, Japan

8-9 July 1997

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289

INTERNATIONAL NUCLEAR SAFETY CENTER operated for Ministry of Russian Federation for Atomic Energy

International Nuclear Safety Center operated for Ministry of Russian Federation for Atomic Energy was established in accordance with the Joint Statement on establishment of International Nuclear Safety Centers signed in Washington in January 1996 in the course of Vlth session of the Russian-American Commission "Gore-Chernomyrdin" on economic and technological cooperation. The Russian INSC is located at ENTEK (RDIPE), the US INSC is located at Argonne National Laboratory. International Nuclear Safety Centers are supervised in their activities by Russian Minister of Atomic Energy and US Secretary of Energy, respectively. Cooperation between the Centers is maintained through the Joint Projects (JPs).

Russian International Nuclear Safety Center (RINSC) conducts R&D activities, calculations, experiments and design efforts to ensure safety of nuclear power plants, research reactors, civil nuclear facilities of space and marine applications. The scope of RINSC efforts covers practically all aspects relating to safety of civil nuclear facilities and technology.

Safety j

290

Russian International Nuclear Safety Center focuses its work in the following areas:

• maintaining database on safety of nuclear facilities; • safety upgrading for the nuclear facilities currently in

operation; • assistance to RF Ministry for Atomic Energy and US

Department of Energy in pursuance of the coordinated policy with respect to safety of nuclear facilities;

• coordination of efforts in safety ensuring of Russian nuclear facilities.

The most important and top-priority areas are the activities in NPP safety and, in particular, maintenance of the respective database. In the future, it is planned to extend these efforts to the research reactors and other nuclear facilities and also to the topics discussed at the Summit on Nuclear Safety held in Moscow in April 1996.

RINSC activities involve not only the leading organizations of RF Ministry for Atomic Energy but also the allied industries, research centers and institutes, universities and Russian Academy of Sciences. Noted scientists participate in JP activities. The results of the work are analyzed by peer review.

In compliance with the directive of RF Minister of Atomic Energy Russian Nuclear Safety Center reports on its subject activities to the Department of Development and Design of Nuclear Reactors and Laser Facilities, while the issues of its international cooperation are governed by the Department of International Relations. ENTEK provides managerial support to RINSC activities. The Coordinative Council approves the subjects of Joint Projects and reviews the results obtained.

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Cooperation with the US International Nuclear Safety Center is being carried out in the following topic areas:

• development of long-term strategic plans of actions for the Centers;

• discussion and agreement on Joint Projects;

• allocation of tasks and coordination of project activities;

• joint review and summarizing of the results;

• making of recommendations on implementation of developments;

• dissemination of experience and information, joint organization of the conferences and seminars, specialist training programs.

It is anticipated to embark on a broad cooperation with other international (IAEA, OECD, ISTC) and national (GAN, NRC, GRS, etc.) organizations.

Contents

• International Nuclear Safety Center of the Russian Minatom (RINSC).

• Co-operative Programs of Work of US INSC and RINSC.

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International Nuclear Safety Center of the Russian M in atom (RINSC)

RINSC Establishment

Scope of Activity

Main Purposes

Main Areas of Activity

Structure and Schemes of Interaction

Interaction with the US INSC

International Cooperation

' ** Nuclear * -Safety j

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Scope of Activity

General: R&D Activities, Calculations, Experiments and Design Works for Safety Ensuring of:

• Nuclear Power plants

• Research Reactors

• Civil Nuclear Facilities of Space and Marine Applications

The Current Most Important: NPP Safety and, in particular, Creation of the Respective Databases

Future Directions: • Research Reactors

• Other Nuclear Facilities

• Topics Discussed at the Moscow Summit on Nuclear Safety, April 1996

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Main Purposes

• Increase of Nuclear Facilities Operation Safety on the Base of Design, Operation and Safety Research Experience of the World Community

• Assistance of Russian Minatom and US DOE in Carrying out of Policy of Safety Ensuring of Nuclear Facilities

• Activity Coordination in Field of Nuclear Engineering Safety in Russia

• Development of Fundamental Knowledge Base in Field of Nuclear Technology and Establishing of Safety Information

• Development of Nuclear Safety Data Bases

• Development of Related Technologies

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Main Areas of Activity

Methods and Means, including Expert Systems, for Accident Diagnostics and Management

Nuclear Facilities Safety Systems and Safety Culture

Nuclear Safety Concepts and Criteria

Most Vital and Advanced Research on Safety

Systems and Methods of Information Exchange and Analysis

Scientific Knowledge and Technology, Related to Nuclear Safety

Safety and Reliability of Operation and Integrity of Nuclear Facilities

Organization and Technical Means for Information Collection for Data Bases

Radioactive Wastes and Fissile Materials Handling Technologies

Decommissioning Ecology Problems Related to Nuclear Energy Application

Carrying out of Independent Examinations of Russian and International Projects and Programs

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Structure and Schemes of Interaction

RF Minatom

RINSC H Coordinative Council

— Administrative Connection — Subject Link Connection

" * N u c l e a r ^ Safety

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Structure and Schemes of Interaction (com)

Coordinative Council

Minatom Gosatomnadzor ENTER OKB GP

OKB M VNII AES AEP

VNIPIET

NIIAR

FEI

VNIIEF

VNI1TF

VN1INM

Russian INSC

Organizations

Research Institutes

Universities ] Institutes of Russian Academy of Sciences

NIIA ENIC Rosenergoatom

LNPP

ZIO

PO "Izhora1 Works"

RNCKI

IBRAE RAN IMASH RAN MGTU

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Structure and Schemes of Interaction (cont)

Expertize Group

Leading Organization

Organization-Participants

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Interaction with US INSC

• Discussion and Coordination of Joint Projects

• Tasks Sharing and Coordination of Activities under Projects

• Joint Consideration and Generalization of Results

• Elaboration of Recommendations on Implementation of Developments

• Exchange of Experience and Information, Mutual Organization of Conferences and Seminars

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International Cooperation

• Participation in International Programs and Projects on Nuclear Safety

• Coordination of Activities with International and National Organizations Pertaining to Nuclear Safety

• Participation in International Scientific and Technical Conferences and Seminars

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RINSC Establishment

• RINSC was established in accordance with the Joint Statement signed in Washington in January 1996 in course of the VI th session of the Russian-American Commission "Gore-Chernomyrdin" on economic and technological cooperation

• International Nuclear Safety Centers are supervised by Russian Minister of Atomic Energy and US Secretary of Energy, respectively.

• Co-operation between the Centers is maintained through the Joint Projects

• The Centers are open to collaboration with other countries and international organizations.

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Co-operative Programs of Work of US INSC and RINSC

• Joint Projects Currently in Process

• Joint Projects Being Planned or Feasible in Future

• Works of the RINSC Pertained to Cooperative Programs

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Joint Projects Currently in Process

1.RINSC International Nuclear Safety Database

General Purpose: Establish an International Database on Nuclear Facilities and Nuclear Safety Technologies in the World

Current Goals: • Collect Site General Available Data for all Operational Russian NPPs

• Development of a Guide to Current Russian Regulations and Standards Related to Construction, Modification, and Operation of NPPs

• Development of a Description of Related Russian Organizations and their General Tasks, Functions and Goals

• Provide Readily Accessible Thermodynamic, Transport and Mechanical Properties of Reactor Materials under Normal, Transient, and Severe Accident Conditions

• Develop a Program Plan for Experimental Needs

' • * • N u c l o a r * ' . Safety

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Joint Projects Currently in Process (cont)

2.Coupled Neutronic Physical and Thermal Hydraulic and Thermal Mechanical Codes

General Purpose: Develop Coupled Codes Including Next Generation Codes

Current Goals: • Perform a Comprehensive Review of Relevant Methods and Computer Software Available in Russia, Including Possibilities of Parallelization Techniques

• Develop and Document Benchmark Problem and its Solution

' * * Nuclear^ . Safety i

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3. Monitoring and Diagnostics and Accident Management

General Purpose:

Current Goals:

Develop Methodology of Accident Analysis, Strategy of Accident Prevention and Consequence Mitigation

• Develop the Accident Management Program in the Capacity of the Technology Basis to Terminate, as Early as Possible, and Consequence Mitigate of Beyond-Design-Basis Accidents

• Document Status of Technology of Candidate Accident Management Features and Actions

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4.Monitoring and Diagnostics of Sensors, Systems, and Equipment, and Transient Management and Advanced Control

General Purpose: Develop Relevant Methods and Expert Systems Using Modern Techniques Like Artificial Intelligence, Fuzzy Logic, Pattern Recognition, and Artificial Neural Networks

Current Goals: Provide a Survey of Relevant Russian R&D Works Including the Validation and Verification Methods Investigated

Safety

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5. Identification of Research Thermal Hydraulic Facilities in the Russian Federation

General Purpose:

Current Goals:

Prepare Facilities Survey to Support the Accident Management and Code Validation and Verification Activities

Prepare a Systematic Compilation of Projects, Facilities Description, Experiments Data, and Computer Codes Pertinent to Safety Research and Development for Light Water Reactors

* * Nuetear^ , Safety •

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6.Computer Code Validation for Transient Analysis of Light Water Reactors

General Purpose:

Current Goals:

Increase the Confidence in the Result Obtained by the Use of Computer Codes in Operational and Safety Analyses of Reactors, and Support of Codes Certification by the Russian Regulatory Agency. These Computer Codes Include Neutron Kinetics and Thermohydraulic Analysis Codes, but Exclude Codes for the Analysis of Beyond Design Basis Accident Involving Core Melt Phenomena and Beyond

• Identify, Select, and Prioritize Phenomena, Facilities, Experiments, and Codes

• Select and Document the First Standard Problems for Application to Different Type Reactors, and Conduct the Analysis of the Standard Problems

' "*" Nuclear^ Safely

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7.Structural Analysis Software for the Evaluation of NPP Components

General Develop and Validate Modern Purpose: Three-Dimensional Software and

Models for the Analysis of NPP Components (such as, Containments, Steam Generators, Drum Separators, etc.) Subjected to Design and Beyond Design Basis Loadings including: Seismic, Thermal, Aircraft Crash, External and Internal Shock Waves, and Hydrodynamic. The Software will be Developed to Treat the Nonlinear Static and Transient Behavior of Metals, Reinforced Concrete, Prestressed Concrete and Liquids

Current Goals: • Provide a Review of Relevant Analytical Models, Methods, and Failure Criteria

• Compile the Major Structural Design Characteristics of Reactor Unit, and a Listing of the Type of Loads

• Provide a Survey of Pertinent to the Purpose Structural Analysis Computer Programs

• Develop a Plan for R&D of an Advanced Structural Analysis Computer Code Suite that Can Address All NPP Structural

311

Joint Project being Planned or Feasible in Future

• Experimental Needs and Database on Reactor Internals under Accident Conditions

• Improvement of NPP Simulation Methodology for Probabilistic Risk Assessment

• Methods and Means for Electrical Cables Diagnostics and Life Time Assessment

• Fracture Mechanics Approach to Pressure Vessel and Heat Exchanger Casing Applications

• Methods and Means of On-Line Diagnostics and Accident Prevention

ïrïïérà0œMmdezï Safe K Cerne.

312

Works of the R1NSC Pertained to Co-operative Programs

• Development of Safety Research Strategic Co-ordinative Program for Russian NPPs

• Development of RINSC Nuclear Safety Database

A Database on Index of Structure Engng Documents and Documents on Nuclear Safety being Made Available in the RINSC

À Access to IAEA Database on NPP Safety

À Development a Guide to Russian and Foreign Databases Pertinent to NPP Safety

• Development of Basic Software for Coupled Problems of Dynamics and NPP Safety Analysis

• Improvement of Models of W E R Thermal Hydraulics in Transient and Accidents

• Participation in International Programs, Conferences, Seminars, Practical Studies

• RINSC Co-ordinative activity and Providing of Respective Information

irgermiiofrnt Mo&ear $&fct? Center

313

JAPAN'S ASSISTANCE AND COOPERATION FOR THE COMMUNITY OF INDEPENDENT STATES AND

THE COUNTRIES OF EASTERN EUROPE

The Japanese nuclear safely assistance programme is carried out both by the

Japanese Ministry of International Trade and Industry, and by the Science and

Technology Agency.

A. MITI Technical assistance and cooperation.

CD(Bi-lateral cooperation)

Co-operation Programme for Nuclear Power Operation Training Centre

Since 1993, the Japanese Government has implemented the Co-operation Programme

for Nuclear power Operation Training Centre. This Programme included the

installation of a full scope simulator of WWER-1000 and the improvement of

training using this simulator, at the Novovoronezh NPP Operation Training Centre.

This Programme is conducted by MITI in the framework of the technical

cooperation agreed at the Munich Summit in 1992 to improve the operational

ability of the operators of WWER-1000 reactors. The installation of the full

scope simulator was completed in July 1996, and the arrangement of training

materials is now being performed.

CD(Bï-latcraî cooperation)

international Invitation Programme for Safety Management at NPPP's

The Japanese Government has carried out the International Invitation Programme

for Safety Management at NPPs since 1992 aimed at the enhancement of operators

safety culture and at the improvement of the safety of nuclear power generation

mainly in the Commonwealth of Independent States and the Central and Eastern

European countries. The approach of this MITI programme is to invite the

315

supervisors and managers, operators, or designers working for nuclear power

plants as trainees to Japan in order to cultivate a better understanding of

safety regulations and administration for NPPs, maintenance management of

facilities and equipment, safety design, etc. in Japan. Approximately 1,000

people are scheduled to be invited during the 10 years of this programme, and

the number of people invited in 1997 will be similar to that in 1996.

<D(Mult i-laieral cooperation)

Extrabudgetary Prograane on the IAEA Safety Pragraame of WWER and RBMK NPPs

The Japanese Government has contributed a large sum of money to the

Extrabudgetary Programme on the IAEA Safety Programme of ÏÏÏÏER and RBMK NPPs,

which was started in 1990 in order to pick up technical safety issues in former

Soviet type reactors and to recommend safety improvements. The results of this

Progrrame has also been used effectively in the bi-iateral assistance programmes.

The Japanese Government endorses the conclusion of this Programme by 1998 as

approved at the Advisory Group Meeting in December 1997.The Japanese

Government will continue to contribute to this Extrabudgetary Programme in the

two remaining years to complete this Programme, and will, in this way, make a

contribution to safety of NPPs in the Commonwealth of Independent States and the

countries of Eastern Europe.

B. Science and Technology Agency Assistance

@(Bi-lateral cooperation)

International Seuinar on Nuclear Safety

Seminars for nuclear engineers from FSU, CEEs and Asian countries are held in

Japan to improve the nuclear safety in those countries. Twenty persons were

316

invited from FSU and CEEs in FY(f i seal year) 1992, 23 persons in FY 1993, 31 in

FY 1994, and 36 in FY 1995 and 48 in FY 1996.

The Science and Technology Agency (STA) has planned to implement international

seminars annually from 1992, by assigning the Japan Atomic Energy Research

Institute (JAERI) as the executive organisation. In this context, JAERI

implements seminar on Nuclear Safety and Management of Radwaste and Spent fuel

for such countries as FSU, CEEs.

(D(Bi-lateral cooperation)

Nuclear Safety Experts Dispatching Programme (Technical Training on Nuclear

Safety)

This programme consists of sending nuclear safety experts to FSU and CEEs to

give technical training or advice on nuclear safety in these countries.

Nuclear safety experts are dispatched to FSU and CEEs to give technical training

or advice on nuclear safety, etc.

©(Bi-lateral cooperation)

Leak Detection System using microphones at Leningrad Nuclear Power Plant

Leak detection System developed for Fugen NPP has been modified and installed at

Leningrad Nuclear Power Plant Unit 1 (LNPP-1) in order to demonstrate the

applicability of the System to RBMK reactors. This system is to detect

the leakeage of cooling water from the reactor's pipe by detecting the

sound of leakage using microphones. The completion of this cooperation on

Leak Detection System will be by the end of March 1998.

317

®CBÏ-1ateral cooperation)

Co-operative Research on Deconnissioning Safety for NPP-A1

Co-operative Research between Japan and Slovak Republic has been conducted

on decommissioning safety for research reactor, NPP-A1, in Bohunice, Slovak

Republi c from 1995.

The purpose of this project is to assist the Slovak Republic to make a

decommissioning plan by using Japanese know-how.

Twelve experts from the Slovak Republic were invited to Japan in February

and March, 1996 to discuses this co-operative activity.

Seven Japanese experts visited the Slovak Republic to discuss a future co

operative research programme in January 1997.

®(Bi-lateral cooperation)

Operator Support System for Ignalina Nuclear Power Plant (in preparation)

Co-operation on the operator support system such as establishment of an

operating database system and providing equipment to measure the oxide layer

thickness of reactor pressure tubes are under consideration for I-NPP

in Li thuania for 1996.

(D(Mnlti-Iateral cooperation)

Special contribution to OECD/NEA Eitrabudgetary Programme

Special contributions are made through OECD/NEA for assistance of long term

nuclear safety of FSU and CEEs countries.

318

#(Multi-latcral cooperation)

Special Contribution to IAEA Extrabudgetary Progranme

The same as ® .

319

JAPAI'S ASSISTANCE AND COOPEIATI0I FO» THE COiiUIITT OF IIDEPEMDEIT STATES AID

THE COUITtlES OF EASTEII EUROPE

MITI

Tatsuya SHINKAWA Deputy Director, International Affairs on Nuclear Safety

Agency of Natural Resources and Energy Ministry of International Trade and Iadustiy

GOVERNMENT OP JAPAN

A. KIT I Technical assistance and cooperation.

(D(Bi-lateral cooperation)

Co-operation Prograiic for Nnclear Power Operation Training Centre

3 ( R e f . ) C o o p e r a t i o n Program. Expendi ture (10 USS, a t 105* /$ )

1993FY 1994PY 1995FY 1996FY 1997FY

17,580 5 ,340 6 ,370 2 ,420 (Budget) 2 ,37 0 (Budget)

(Bi-lateral coopératif*)

International Invitation Prograwe for Safety Management at HPPP*s

( R e f , ) Number of Accepted P e r s o n s and I t s Account Fiscal Year Expenditure Number

Russia Ukraine Hungary Czech Slovenia Bulgaria Lithuania Romania China Total

(10 3US$) 1992

1/810

23 5 8

8

8

13 65

19 93 3,970

25 21 8 5

. 7 8 5 2

14 95

1994 4,380

24 12 9 8

:-... B 11 5 7

14 98

1995 4,350

14 20 11 11 11 11 8 7

13 106

1996 4,330

20 15 11 11 10 11 7 7

13 105

Total 18 ,840

106 73 47 40 39 '49 25 23 67

469

1997 4,520

17 17 10 11 11 12 9 10 13

110 (*: planning)

©(Multi-lateral cooperation)

Eitrabndgetary Prograne on the IAEA Safety Pragrawe of H E R and RBMK NPPs

320

1. The International Seminar on Nuclear Safety.

2 . The Nuclear Safety Experts Dispatching Program.

3 . The Leak Detection System using microphones at Leningrad Nuclear Power Plant.

4 . The Cooperative Research on Decommissioning Safety for the Slovakian Bohunice NPP-A1.

»

5. The Operator Support System for Ignalina Nuclear Power Plant.

321

O Number of participants in 1996

1. The Seminar on Nuclear Safety for FSU and CEEC

Bulgaria 2 Lithuania 2 Slovak 2

Czech 2 Romania 2 Ukraine 3

Hungary 2 Russia 5

2, The Seminar on Management of Radwaste and Spent Fuel

Bulgaria 1 Lithuania 1 Ukraine 1

Czech 1 Russia 2

Hungary 1 Slovak 1

3• The Seminar on Nuclear Safety for Ukraine

Ukraine 1 0

4. The Regional Training Course on Safeguards.

Armenia 1 Russia 2 Estonia 1

Kazakhstan 1 Ukraine 2 Latvia 1

Lithuania 1 Belarus 1

322

The Nuclear Safety Experts Dispatching Program in FY 1996 .

(1) Russia (a (b (c

Duration: January 27 - 30, 1997 Japanese experts: 5 persons Contents: Decommissioning, Radwaste

(2) (a (b (c

Slovak Republic / Czech Duration: January 27 - 31, 1997 Japanese experts: 9 persons Contents: Nuclear safety

(3) (a (b (c

Hungary / Bulgaria Duration: February 3 - 5 , 7, 1997 Japanese experts: 6 persons Contents: Nuclear safety

(4) (a (b (c

Bulgaria Duration: December, 1996 - March, 1997 Japanese experts: 1 person Contents: Nuclear safety

(5) (a

(b (c

Ukraine Duration: October - November, 1996, and

February - March, 1997 Japanese experts: 3 persons Contents: Nuclear safety

(6) (a

(b (c

Lithuania Duration: August - November, 1996, and

February, 1997 Japanese experts: 5 persons Contents: Nuclear safety

323

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reactor pressure tubes.

327

SAFETY EVALUATION OF WWER TYPE REACTORS By Turkish Atomic Energy Authority (TAEK)

Introduction

The former Union of Soviet Socialist Republics developed two types of commercial nuclear power plants. One is RBMKfwater cooled graphite moderated nuclear power plant), the other is WWER(similar to western type pressurized water reactors). These reactors were also built in Eastern European Countries such as Hungary, Bulgaria, Poland, etc. Two of Turkey's neighbors, namely Bulgaria and Armenia, have WWER types reactors. In Bulgaria, there are 6 WWER type reactors (4 units WWER 440 model 230, 2 units WWER 1000) which are located in Kozloduy site. In Armenia, there are 2 WWER type reactors (2 units WWER 440 model 2SO) which are located 30 km. away from Turkish border. In the event of a large scale accident, Turkey will be affected as much as Armenia. This is the reason we are ititerested in this topic.

General Characteristics of WWER's

WWER-440S are always built in modules of two units. These two units are housed, m a single reactor building. All WWER-440s have six loop, isolation valves on each loop, horizontal steam generators, rack an pinion type control rod drives, use hexagonal fuel assemblies containing 126 fuel rod position, and with a single exception, all use two 220 MWe steam turbines. The key mechanical components are of a standardized design and are produced with standardized manufacturing procedures. Although the basic WWER design has undergone changes m engineering and institutional judgments, there has been little change in the basic component and system design. The WWER-440 model V230 relies on local area compartmentalization alone to prevent the release of fission products. The design basis accident is a pipe rupture with an effective 10 cm. diameter carrying a unidirectional flow with the flow reduced by special orifices to an amount corresponding to the flow through a 3.2 cm. opening. The V230 has makeup coolant pumps thai provide a limited capability for emergency core cooling but does not have an Emergency Core Cooling System (ECCS). The WWER-440 model V213 differs from the older model V230 in that the V2Î3 has both an ECCS and connects a "bubblier/condenser tower" to the accident localization compartments of. each unit to mitigate the effects of severe accidents. The model V230 sealed accident localization compartment contains a pressure release valve intended to relieve over-pressure and then close. The pressure vessel interior of the WWER-440 model V213 is clad with stainless steel, the model V230 's are not clad. Flywheels are incorporated into the primary coolant pumps of the WWER-440 model V213 to increase their coast down time during an emergency, the model V230 uses low inertia canned motor pumps.

329

Safety Problems of WWER-440 model V230

From safety point of view, these type of reactors have some deficiencies in that various' parts of the reactor don't have adequate redundancy, diversity and independence. The weak points of this type of reactors can be itemized as follows :

• Some systems of different units are situated in the same machine whole. • Service water system of different units are common. • Power cable and control of independent equipment are not planned separately. • Coast down time of primary coolant pump is relatively small • The desigti basis accident is a pipe rupture with an effective 10 cm. break which is

different from western approach. • Separate ECCS doesn 't exist. The cooling make-up system is designed to provide a

limited capability for ECCS. • The design does not include a containment in the western sense. Instead, a system of

sealed compartments is used in order to mitigate the escape of coolant and fission products following a small leak in piping of the primary system. Compared to western light water reactor, the size of the leak for which the sealed compartments are designed, are rather limited.

• Due to the small diameter of the vessel, there is only a relatively small gap between the reactor core and the vessel wall, resulting in a high flux of fast neutrons in that watl near the core. If impurities in the welding are high, this results in a comparatively strong tendency to embrittlement, especially in the weld material.

• Poor quality of instrumentation and control is a generic problem of the WWER-440 V230. Another significant deficiency is the limited capacity and redundancy of safely equipment. For instance rupture of large pipes of the primary system are not included in the design basis. Furthermore, interconnection between different systems-results in insufficient independence of safety functions.

• Structures which do not function as pressure boundary are designed like conventional industrial frame buildings.

• Precast concrete structural elements are extensively used. • The construction quality is poor. • Site-related external events were not properly considered in the original plant desigti.

Seismic Requirements for WWER's

Prior to the 1977 Vrancca earthquake, the practical application of seismic criteria in the Soviet Union simply prohibited the construction of nuclear power plants in regions of seismic activity greater than 5 points on the MSK- 64 scale. There were no special provisions in the design of the structures or in the design or installation of equipment for seismic loading ; standard building practices were used according to other criteria.

330

However the rule prohibiting construction in zones of intensity greater than 5 was not strictly adhered to. The Kozloduy (Bulgaria), Armenia, and Bohunice (Slovakia) plants were constructed at sites ofseismicity 8, 8 and 6 respectively. WWER plants, either not explicitly designed, to seismic loads or designed to lower earthquake than the one now defined for the site, are listed below.

Seismic Design Basis for (SDB) NPP 's in Eastern Europe

COUNTRY

BULGARIA SLOVAKIA SLOVAKIA BULGARIA HUNGARY CZECH REP. ARMENIA

PLANT

KOZLODUY BOHUNICB MOCHOVCE BELENE PAKS TEMELIN MEDZAHOR

TYPE

WWER-440/1000 WVER-440 WWER-440 WWER-1000 MWER-440 WWER-1000 Wn/ER-440

ORIGINAL SDB (PGA) NED NED 0.05g

OJg NED 0.06% No information

RE-ASSESSED SDB (PGA) 0-2g 0.25x

OJg Continuing 0.35g 0.1 g 0.4g

* NED : No Explicit Design Against Earthquakes. * PGA : Peak Ground Acceleration

It is clear from the above table that Eastern Europe WWER type reactors must be upgraded according to re-assessed design peak ground accelerations. As an example ; although the original SDB-PGA value for Medzamor plants was 0. Jg, the re-assessed SDB-PGA value for the same site is 0.35 g. It was declared that the seismic resistance of the plant itself has recently been upgraded with the objective to withstand earthquakes with PGA 's at least in the range of 0.2 g. It still needs some upgrading. • • -

Another example is Kozloduy Plants. On March 4, 1977, a severe earthquake occurred at Vrancca, Romania, about 300 km. from the Kozloduy site. It was found that one of the stream generators of the Unit 1 had been displaces by 12 cm during the earthquake as registered by a sliding damper. Such a displacement appears to have been extraordinarily high, even for flexible piping rims favored, at that time by the designers of the plants. Major equipment was replaced after the earthquake.

Conclusion

WWER type reactors were designed and constructed in a manner that is very different from western safety culture which includes international and national activities in order to upgrade reactors and to improve the education and qualification of the staff and operators. In western safety culture, rupture of large pipes is accepted as design basis

accidents and containment is designed to withstand against this type accidents and to mitigate the result of such accidents. However the confinement of the WWER's was not designed for such a design basis accident. In this respect, the result of this type of accident can be very severe. Another unresolved problem of these reactors is seismic events. Site-related design basis PGA were not properly considered in the original plant design. From setsmicity point of view, upgrading of some systems and components can be possible. On the other hand, upgrading of safety related buildings including confinements is limited.

Overall, we are proposing to our neighbor countries which have WWER type reactors to cooperate in carrying out the following activities :

* Seismic studies - exchange of data on historical and site specific earthquake data, - exchange of geological, geophysical and geotechnical data, - establishing common local network to monitor seismic activities, - exchange of instrumental earthquake data,

* Emergency planning - establish a direct telephone line betweeti the two countries for early notification, - exchange of information on respective environmental monitoring, - exchange of information on respective emergency plans, - carrying out joint exercises,

* Education and training - fellowshipsfor visiting professors of both countries and information exchange on

nuclear engineering education, -fellowships for exchange of students in nuclear engineering,

(especially in practical training on research and power reactors) - common workshops on nuclear and radiation safety.

332

V V

•UJV* NUCLEAR RESEARCH INSTITUTE REZ pic

Division of Integrity and Materials

INTERNATIONAL SEMINAR

ON

THE SAFETY RESEARCH NEEDS

FOR RUSSIAN-DESIGNED REACTORS

„Kohku Kaikan", Tokyo, Japan 8-9, July 1997

National Programme in the Czech Republic for VVER Reactors

by

JIRI ZDAREK NRI Rez, Czech Republic

333

Dukovany Nuclear power plant with four units of the WER 440/213C type has reached over 100,000 hours of operation. At present there are several key projects under consideration.

the modernisation of systems selected on the basis of systematic safety assessments increasing the power output the introduction of equipment qualification programmes

All of these projects require financial and human resources the extent of which has to be judged in relation to the returns on output and availability during the design lifetime and possible life extension.

For the WER type reactors the regulatory and utility policy for plant life management has not yet fully defined. For several years the management of Dukovany NPP together with the Nuclear Research Institute has been preparing individual methods of assessment and more recently have been working on the overall concept of reliability oriented plant life management. In the presentation, the principles of the concept will be summarised for the following systems and components:

• reactor pressure vessel • reactor internals • primary circuit piping components • cables • steam generator

334

Most of the individual programmes are either already underway or the detailed work plan is being prepared. By the end of 1997 the database of individual tasks will been prepared and matched with the other programmes specified earlier.

For each system or component, the structure is illustrated and described, a flow chart is presenter of the evaluation scheme together with a statement on the status of the programme. Finally the conclusions and recommendations for work through to the year 2000 are given.

Throughout each programme a significant feature is that of verification that the data, the analysis and any tests are of an acceptable standard. This is part of the quality assurance scheme applied to any programme of work in which NRI is involved. Some of the verification work, particularly that of NDT on piping, is being carried out within the framework of the PHARE programme.

The economic issues related to the plant life management have a significant influence on the decision making process. The prime objective is to meet the safety requirements. In doing this it is also desirable to maximise availability and output of the plant. The possibility of lifetime extension must always be considered. Plant life management requires a full understanding for planned maintenance, repair and replacement. When the time arrives for consideration of plant life extension all the necessary data are to hand and the extension case can be readily made.

Dukovany NPP is only ten years into its operating life, unlike a large proportion of plants which are now approaching twenty years old or more. The opportunity therefore exists to apply reliability orientated plant life management and provide the utility with maximum economic return on their capital investment.

335

RFY/POKOVANY NFF - Ageing Management

ORIGINAL AND SUPPLEMENTARY SURVEILLANCE

PROGRAMME

INDICATION AND SIZING OF DEFECTS

STRESS STATE ANALYSIS

FRACTURE MECHANICS

THERMO-HYDRAULIC ANALYSES

Downcomer

MATERIAL PROPERTIES

Base metal

-Cladding

336

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PTS* Pressurized Thermal Shock

337

RPV/DUKOVANY NPP « Conclusions and Recommendations

for time period of 97 to 2000

l .To establish the supplementary surveillance programme

2. To continue in thermo-hydraulic and stress state calculations for the selected PTS modes

3. To perform and evaluate the cladding irradiation program

4. To complete and maintain qualification of non-destructive inspections

5. To continue in experiments on cruciform specimens and in the measurement of instrumented hardness

6. To perform and evaluate temperature measurements of the HPI ECCS system including temperature measurements on the vessel

7. To define limiting factors in order to achieve extended RPV lifetime, including proposals for measures

338

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amage growth kinetics

339

RPV Internals -Conclusions and Recommendations

for time period 1997 - 2000

1.To perform more detailed static & dynamic calculations on the RPV internals,

2. To perform more detailed calculations of the neutron fluxes on the internals.

3. To evaluate the sensitivity of the RPV internals materials irradiated up to ftuences of design or extended lifetime, to corrosion cracking by measurement of the crack initiation threshold values and their growth kinetics,

4. To measure changes in mechanical properties of the RPV internals caused by radiation embrittlement

5. To perform calculation of the stress state (static loading, fatigue) for the RPV internals at the end of life.

6. To determine & verify in-situ measurement of the RPV internals mechanical properties change.

7. To determine possibility of the defect initiation and growth for the selected critical areas of the RPV internals with high mechanical stresses and high radiation damage.

340

Primary PÈpïng/Dukavany WPP Conclusions & Recommendations

1.To complete NDT verification in the framework of PHARE 1.02/94 and in proposed extension of PHARE 1.02 (98-99).

2.To start temperature measurements on the piping to the pressurizer. To modify the computing model for stratification with respect to the experimental results.

3.To start temperature measurements and the follow-up evaluation of the HPI system.

4.To perform an evaluation of the remaining piping systems.

• LP ECCS piping • Hydroacumulator piping to the RPV • Primary coolant continuous purification piping • Pressurizer by-pass piping to the pressurizer relief

tank

341

CABLES/DUKÔVANY NPP Conclusion and Recommendations

• Set up of the measuring device and performing tests on initial

and aged samples

• Analysis (modelling) of indentation diagrams and development

of the evaluation method of „primary results"

Calibration of the method

• Performance testr in the NPP environment

342

STEAM GENERATOR/ DUKOVANY HPP Conclusions & Recommendations

• critical locations identification with respect to cavity sludge concentration

residual life assessment in all critical locations

• predictions for specified water chemistry regimes

• sensitivity analysis to simulate efects

change in operation conditions (inerease of power, power regulation

non standard water chemistry

• development of „expert" system covering all above specified items.

343

OTHER SUPPORTING

RESEACH / APPLICATION AND

PHARE PROJECTS IN THE AREA

COMPONENT INTEGRITY

344

PLANT LIFE MANAGEMENT OF WER-440/V-213C COMPONENTS

Project is concentrated on main critical components of primary as well as secondary circuit and their ageing mechanisms. The most important parts are:

- preparation of standard procedures for assessment of residual lifetime of main components of primary circuit,

- elaboration of a software (connected with direct measurements) for automatic fatigue/corrosion-fatigue damage evaluation of components,

- proposal of lifetime management programmes for critical components,

- creation of databases of material properties of critical components, results of in-service inspection, operation and maintenance for lifetime evaluation,

- determination of necessary special mechanical properties for lifetime evaluation,

- corrosion-erosion problems and prediction procedures for secondary circuit piping.

345

REQUIREMENTS FOR LIFETIME ASSESSMENT OF WER REACTOR PRESSURE VESSELS AND

REACTOR INTERNALS DURING THEIR OPERATION

Project deals with preparation of requirements and procedures for lifetime assessment which will be approved and issued by the Czech State Office for Nuclear Safety as recommended materials.

This document will contain the following chapters, dealing with:

- principal requirements for lifetime evaluation of RPVs and internals,

- procedure for RPV non-ductile evaluation, - requirements for a determination of radiation field, - requirements for a fatigue assessment, - requirements for a corrosion-mechanical damage

assessment, - requirements for surveillance specimens testing, - requirements for evaluation of surveillance results, - requirements for defect allowability procedure, - requirements for instrumented hardness method

application, - QA requirements for lifetime evaluation, - requirements for documentation and reporting.

346

SUPPLEMENTARY SURVEILLANCE PROGRAMME FOR WER-440/V-213C

REACTOR PRESSURE VESSELS

This project substitutes a standard surveillance specimens programme which is practically finished after first five years of operation.

New programme is based on results from a standard programme as well as from re-evaluation of the standard programme. It contains only archive materials, particularly from the standard one. Main aim is on monitoring radiation field and neutron damage in RPV materials at least till the end of design lifetime with lead factor less than 2.

Programme is divided into three parts :

• irradiation of RPV archive materials in locations with low lead factors,

• irradiation of cladding materials (outer layer, inner layer, heat-affected zone) including effect of annealing and re-embrittlement rate,

• effect of annealing and re-embrittlement rate on RPV materials,

• dosimetry chains/containers for irradiation in time intervals between two withdrawals longer than two years.

347

RADIATION DAMAGE IN REACTOR PRESSURE

VESSEL AUSTENITIC CLADDING

Austenitic cladding is important not only as a corrosion

barrier, but also for a precise evaluation of RPV integrity

during PTS when cladding can serve as a bridge for under-

clad type postulated defects.

Special study is concentrated on determination of

radiation damage in all three important materials - outer

layer, inner layer and heat-affected zone. Charpy impact as

well as static fracture toughness tests specimens are

planned to be irradiated in the experimental reactor. Reason

for an accelerated irradiation is in necessity to know these

properties in advance for Periodic Safety Report

preparation.

348

PHARE 94 NUCLEAR SAFETY

project PH 1.02/94

In-service Inspections of Primary Circuit

Components

Contractor: consortium TECNATOM, S.A (Spain)

and VTT (Finland)

Western

subcontractors: JRC Petten (EU), IVO (Finalnd)

Local

subcontractor: NRI Rez, pic

(Division of Integrity and Materials)

Main Beneficiary: Dukovany NPP

(Co-B: Bohunice and Paks NPPs)

349

5 test assemblies being designed and manufactured:

• Pressurizer pipe to main circulation pipe weld (245 mm to 500 mm - BM: 08CH18N10T austenitic steel) implanted flaws: 5 PISC type A EDM notches

• Safety cooling pipe to main circulation pipe weld (108 mm to 500 mm - BM: 08CH18N10T austenitic steel) implanted flaws: 7 PISC type A EDM notches

• Steam Generator Transition weld No. 4.3.1. (upper part: steam generator shell -22K steel, lower part: primary collector - 08CH18N10T austenitic steel) implanted flaws: 4 PISC type A EDM notches

1 fatigue crack implanted

• Transition nozzle to main circulation pump elbow weld (transition nozzle and elbow: 08CH18N10T austenitic steel) implanted flaws: 8 PISC type A EDM notches

2 fatigue cracks implanted

• Transition nozzle to main circulation pipe weld (transition nozzle and piping: 08CH18N10T austenitic steel) implanted flaws: 8 PISC type A EDM notches

2 fatigue cracks implanted

350

5 calibration blocks with the original welds, notches and SDH being manufactured for dynamic calibration

Performance demonstration/practical open trials will be conducted as a part of NDT qualification of:

• improved UT procedure equipment for automated/mechanized inspections with 3 scanners and UT system (SIROCO+SUMIAD+WIASERA)

based on a written qualification procedure in compliance with the ENIQ methodology

Technical justification for the above austenitic and dissimilar welds include Czech RRT results, PISC and ENIQ pilot study experience

351

PHARE 93 project 4.1.2

W E R 440-213 In-service Inspection (RPV)

Contractor: consortium BELGATOM, S.A (Belgium),

EdF (France), Intercontrole (France) and

Siemens (Germany)

Western

subcontractor: JRC Petten (EU)

Local

subcontractor: consortium NRI Rez, pic

and SKODA (CZ)

Main Beneficiary: Dukovany NPP

(Co-B: Bohunice and Paks NPPs)

352

3 test assemblies designed and manufactured:

• VVER 440 nozzle with dissimilar weld (500 mm LD. - BM: 15Kh2MFA low alloy steel) • inspection area: nozzle inner radius

simulated defects: 4 underclad cracks (PISC type A ED M notches) 2 fatigue cracks in cladding (PISC type A EDM notches)

• inspection area: dissimilar weld simulated defects: 11 lack of fusion (narrow EDM notches) 8 fatigue cracks {PISC type A EDM notches)

• VVER 1000 nozzle (850 mm LD. - BM: 15Kh2NMFA low alloy steel) • inspection area: nozzle inner radius

simulated defects: 5 underclad cracks (PISC type A EDM notches) 4 fatigue cracks through cladding (PISC type A EDM notches)

• VVER 1000 simulated nozzle with homogenous weld (850 mm LD. - BM: 15Kh2NMFA low alloy steel) • inspection area: homogenous weld

simulated defects: 9 lack of fusion (narrow EDM notches) 8 fatigue cracks (PISC type A EDM notches)

353

other test assemblies with inspection areas designed for

manufacturing in the near future (project continuation):

• VVER 440 RPV circumferential weld (13 defects

designed)

• VVER 1000 RPV circumferential weld (13 defects

designed)

• repaired welds and cladding with implanted defects

qualification procedure for the above inspection areas

prepared for implementation in Beneficiary countries

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New Programme for the R&D Activity Sponsored by the Hungarian Nuclear Safety Authority

Lajos Vôrôss Director

Nuclear Safety Inspectorate Hungarian Atomic Energy Commission

International Seminar on

The Safety Research Needs for Russian-Designed Reactors

" Kohku Kaikan ", Tokyo, Japan 8-9 July 1997

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1. Short History and Background of R&D Programmes for Nuclear Energy in Hungary

In Hungary it was recognized in the late 1950's that nuclear energy research activities were a matter of importance for a country intending to enter into nuclear energetics in the not too distant future. In 1959 the first nuclear facility, the 5 MW research reactor came into operation and in the early 60's the zero reactor faculty ZR-6S hosted also by the Atomic Energy Research Institute (AEKI)3 Budapest has served as the technical and experimental basis of an extensive and comprehensive international research and development co-operation among the potential user countries of the - at the time developed - W E R - 1 0 0 0 reactors. Activities of code adaptations for reactorphysical and thermohydraulic analysis have been carried out and computerized monitoring and measurement systems for nuclear applications have been developed in different institutes. The early efforts were then concentrated and enlarged by organizing national coordinated R&D programmes to promote satisfactory financial resources.

The main goals of these programmes were threefold: - a firm scientific and technical basis should be established within the country with

experts being both professionally and morally capable of bearing the responsibility to keep the potential danger at an internationaUy acceptable level

- since the nuclear power plant covers a large part of the national energy demand, the high level of efficiency and economy of the plant should be supported scientifically and technically

- since the domestic industry shared a significant part in the erection and construction of the plant, and moreover produced several components for export purposes, appropriate scientific and technical support should be provided to ensure that the required high technological level and high quality standards be maintained.

The national coordinated R&D programmes can be divided in different time periods being characterized by different stages of the Hungarian nuclear energy programme.

The period 1981-85 was devoted to support the construction and commissioning of the first Hungarian nuclear units at Paks site. (The first unit was connected to the grid in Dec. 1982 and the last, fourth unit in Sept. 1987) The main targets concentrated on start-up tests and measurements, structure analysis,, analysis of operational and outage conditions, issues of corrosion and water chemistry, whereas safety analysis, environmental protection and radiobiology were not really in the foreground. It was recognized, however, that implementation of analytical and experimental tools for safety analysis is necessary, so computer codes and some large scale experimental facilities for their validation have been developed.

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In the period 1986-87 a concept aiming at building further nuclear units of W E R -1000 type reactors made it necessary to initiate a new R&D programme with participation of 85 research fellows of 11 institutions. However, because of the Chernobil accident occured in April 1986, the programme has changed and focused on the safety issues of W E R s in the limelight of the Chernobil accident. The nuclear programme has slowed down and finally terminated.

The period 1988-90 was characterized by the ever-growing international concern towards the safety of the Soviet type reactors. Therefore the national programme in this period was to extend the safety assessment of the WER-440/213 type reactors operated at Paks. Due to the stagnation and even decreasing of electric energy demand the construction of any base load power plant became unnecessary.

The years of 90Ts brought both positive and negative changes in the R&D work. The tendency that the domestic research activity can be more comprehensively integrated into the international processes is positively evaluated. However the reduction of the financial resources available for R&D generally in the country and the new financing methods from governmental resources made practically impossible to organize coordinated national programmes in any field including nuclear safety as well. The expertise of this area could be kept fortunatelly using it for reassessment of the safety level of Paks NPP in the framework of the AGNES-project and by participation in several international projects in the framework of PHARE-programme and of US NRC sponsored programmes like CAMP and CSARP.

It can be stated that Hungary is well provided with expertise in the nuclear field trained by nationally organized, continuously operated, governmentally sponsored R&D programmes. Some of the Hungarian research institutes achieved good international reputation and they are avilable today to successfully adress safety issues arises in Hungary or in any other countries operating Soviet-designed reactors.

2, Medium Term R&D Programme for Regulatory Assistance

In the national coordinated research programmes mentioned in the previous chapter the Hungarian regulatory body played relatively modest role. The main reason was that it did not have financial resources for use of its own purposes because it used to be operated until 1990 as a nuclear department of the state Office for Energy Supervision and Safety and its budget did not cover any R&D sponsorship. This situation did not change when this department was joint to the Hungarian Atomic Energy Commission as Nuclear Safety Inspectorate (NSI) in 1991.

Everybody knew that it would be necessary to make possible that the NSI be provided with financial resources to draw up its own priorities for R&D work. Hoping this opportunity the NSI initiated in 1995 a report to be prepared on this subject in wich R«BcX> pr io r i t i e s hor^c b«*sn def ined for r e g u l a t o r y app l ica t ions a s lOllOWS:

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- ageing and residual life assessment of equipment - PSA application for regulatory decision making process - containment behaviour during DBA and severe accidents - computer code validation and verification - regulatory aspects of emergency response - severe accident analysis - accident management - human factor including safety culture

Making a lot of effort by the HAEC management to receive money for launching a medium term R&D programme for regulatory assistance a success was achieved in 1996. A dedicated sum from the budget of the National Committee for Technical Development was made available for the nuclear safety authority to contract the most capable intititutes for conducting R&D work involving subcontractors in all areas mentioned above. Meanwhile, the new Atomic Energy Act No. CXVL passed by the Hungarian Parliament in Dec. 1996 and come into force in l.June 1997 ordered that appropriate financial sources have to be available for the regulatory authority to sponsor fundamental technical support necessary for its licensing and supervisory activities as requested also by the Nuclear Safety Convention. It means that a continuous regulatory assistance programme can be undertaken using the best expertise of the country as well as, if necessary, from abroad.

The regulatory priorities for R&D is being reviewed yearly and if necessary changed. An advisory committee serves for the authority in this regards to give advise in strategic questions and the Division of Technical Background of the NSI have been strengthened with three more staff members of several year research experiences to formulate concrete tasks and to help introduce the results into the regulatory work.

Another important step was taken to assist the everyday work of the nuclear safety regulatory staff in its decision making process: agreements were concluded with two institutes to serve as technical support organizations (TSO) of the NSI. These are the Atomic Enegy Research Institute (AEKI) of the Hungarian Academy of Sciences and the Institute for Electric Power Research Co. (VEIKI) which cover most of the fields connected to the nuclear safety. In cases they can not provide appropriate expertise, individual experts are being taken from universities or other iristitutions. The TSO-system seems to be very effective and flexible to use. Our TSOs are interested in giving immediate and good quality expert judgements helping in decision making process and, on the other hand,in receiving contract for long-medium term research and development work, so the mutual interest has been met

One part of the R&D work contracted in 1996-97 aims to equip the NSI's multipurpose centre CERTA (Centre of Emergency Response., Training and Analysis) with hardware and software tools. This centre has been envisaged to establish a suitable opportunity for regulatory siafFto be a b k to make a judgement independently

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frorn the utility about the severity of an accident situation (diagnosis) and about the possible source teim to the environment (prognosis) in an emergency case. Additionally, the centre has to be able to provide the NSI equipment as simulators, PSA-based tools as well as fast TH-estimation methods to predict possible escalation and consequences of a hypothetical or - hopefully never to occur - real accidents. The on-line data transfer to the centre from the Paks NPP has been accomplished to provide data for the Safety Parameter Display System of CERT A Comprehensive international assistance has helped realize the first phase as follows:

- UK governmental (Department of Trade and Industry) assistance by delivering computing and communication equipment and emergency diesel generator

- PHARE assistance in adaptation of SESAME (IPSN) software for diagnosis/prognosis estimation

- PHARE/RAMG assistance supplied HW tools - IAEA TC assistance providing expert missions, training and equipment for

integration CERT A into the regulatory workstations

To summarize this chapter it should be emphasized that the Hungarian Nuclear Safety Inspectorate is working in close cooperation with the national research and development organizations and has developed a medium term R&D programme relying on a solid financial basis and combined it with a reliable TSO-system based on mutual interest. Because of the limited recources, a multipurpose emergency response, Ixaining and analysis centre for the regulatory body has been established to enhance technical support activity for the regulatory staff.

It has to be mentioned that Paks NPP also has conducted R&D activity for support of its safety upgrading programme. There is a rather loose connection between the regulatory and the industrial R&D programme what is not cost-effective solution, therefore NSI intends to initiate movement towards a better co-operation between both programmes.

3- Some of the Newest Results and On-going Activities

It has to be emphasized that the newly initiated regulatory R&D programme has relied mostly on the previous results achieved in different institutes and only one of the subjects is a new project, namely the ageing and residual life assessment which has just very little precedents in Hungary.

Due to the limited extent of this paper just two recent products can be made aquainted with aiming to give an impression on the level and capabilities of the Hungarian nuclear safety research. The on-going activities are listed only by title.

A comprehensive on-line information system for the CERTA is under way to be hmstall@tL Tke Visual Information Trarusjfcr and Analysis (VITA) sysîem ttas been

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developed by the Atomic Energy Research Institute of the Hungarian Academy of Sciences (AEKI) in contract with HAEA The first phase of VITA is a Safety Parameter Display System (SPDS) which has been completed very recently.

The objective of the whole system is to receive and to process data of Paks NPP transferring on-line and to use them for different regulatory purposes.

There are three main parts of the system as follows: - on-line data transfer between the Plant Dispatcher Centre of Paks NPP and CERTA - data processing in CERTA - analysis and display of data

The VITA-system is - modular and flexible in its structure - expandable and open - using data-base controlled processing algorithms - sensitive for data changes/trends

For the emergency response tasks of the HAEA-NSI the SPDS will play the most important role together with the fast estimation computer codes giving possibilities for the diagnosis/prognosis activities of the CERTA

The display construction of the SPDS consists of four parts as follows; - the upper part shows the state-of - the-art of the critical safety functions (CSF) and of the protection systems

- also on the upper part the 12 most important safety parameters have been displayed - in the middle there is place for detailed analysis of the CSF displaying different

contents and forms - on the bottom part the functional control buttons have been installed

The SPDS distinguishes seven different CSF, as control of - reactivity - core cooling - secondary heat transfer - primary circuit integrity - primary coolant inventory - containment integrity - radioactive release

The first level gives an overview on all seven CSF containing the numerical values and trends of every parameter characterizing the CSF in question. The second level displays the detailed trend and time history of the parameters belonging to the CSF. Every CSF has 6 parameters, 2 of which are groupped together due to the convenient

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survey. The informations of the upper part mentioned earlier are always available on the display because those could be vital in any option of the SPDS.

The buttons located on the bottom part of the screen are available for selection of different options of the SPDS like print out the actual display or recording the actual status and displaying trends or calling the actual pressure-temperature (p-T) diagrammes. There are seven p-T diagrammes in the system, one for each of the six loops and one general picture with the mean values of the six. A "zoom-in" function serves for better recognition of the smaller changes occuring during normal operation. The diagrams also display the actual allowed range of operation.

As it was already mentioned, the VITA system aims to serve as a comprehensive information basis to the regulators not only in emergency situation but also in their everyday regulatory activity, e.g. in incident evaluation or safety assessment. Therefore a second phase of the project is now under discussion with the developer and the utility to extend the capability of the system using the PICASSO-3 software package which was developed by the OECD Halden Reactor Project and on the basis of which a project was offered to Paks NPP to support the development of the sympton oriented emergency operating procedures (EOPs).

The other important equipment installed in the CERTA as PSA-based softwer tool for regulatory decision making assistance is the risk supervisor.

The NSI has recognized that the probabilistic safety assessment (PSA) can be a valuable tool in the regulatory work as a complementary approach to be used together with the deterministic methods for evaluation of the safety. A PSA-based risk supervisor has been developed by the Institute for Electric Power Research Co. (VEIKI) for configuration control and evaluation of annual risk history of Paks units to be applied by the NSI staff in its decision making process.

The risk supervisor uses the technological model and data developed for the Paks units in the framework of the AGNES-project (Advanced General and New Evaluation of Safety) which was implemented in 1992-94 and of the periodic safely review of units Nr. 1.- 2,carried out in 1995-96. However, it was necessary to simplify the original model for the purpose of risk supervisor to accelerate the calculation time and to make the PSA-model more tranparent for the users.

A special SW has been developed for application of the general Risk Spectrum SW-package taking into account the two main objectives of the risk supervisor;

- to provide a tool for regulatory body to support the decision making process in case when the licensee submits a request for modification of operational conditions or for extension of the allowed outage time of a safety related system or component

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- to provide a PSA-based method for evaluation of the planned and actual testing and maintenance activity of the utility during a time period between two consecutive refueling.

The risk supervisor is able to calculate the mean numerical values of the core damage frequency and its contributors for the cases as follows:

- arbitrary operational mode of safety related normal operation systems - additional component failures in a given time interval - unavailability of any engineered safety system in a given time period - testing of trains of any engineered safety system - operational actions (testing, maintenance) during one fuel cycle campaign (some

11 months)

It is important to emphasize that for the risk supervisor a living PSA-model is necessary to be applied. It means that only the actual plant status gives ralistic results and consequently it has to be ensured in the applications. Important also, that the simplified model has to be validated by numerical tests which prove that the logical structure of the refence model has not been violated and the simplification did not cause significant changes either in the core damage frequence or in the ranking of the risk contributors.

The NSI intends to strengthen the application of PSA-based tools in its regulatory work taking always into consideration the limits and conditions of the probabilistic approach.

The on-going projects of the regulatory authority sponsored R&D programme are as follows; - coupled neutron kinetic - thermohydraulic system code development

(KIK03D+ATHLET) -evaluation of adapted TH-codes (RELAP5, CATHARE, SCDAP-RELAP,

MELCOR) by their uncertainty ranges, best-estimate characters;, validation, W E R -specific modelling capabilities etc.

- containment behaviour calculations used CONTAIN-code for pressure peak, H-distribution, code validation by large scale experiments to be carried out in PHARE-programme and assisted by NEA

- adaptation of MAAP4/WER version for fast estimation of TH-processes during an accident (PHARE regional project) and for accident management procedures evaluation

- adaptation of the SESAME code (PHARE regional project) - adaptation of TRANSURANUS code for fuel element behaviour (assistance by

IAEA in regional project) - analysis of incidents using PSA-based tool (precursor study) and development of

probabilistic performance indicators - aovore accident »ixr»*lat«r for training orregulaiory statTln CERT A

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-developments of regulatory guides for ageing management of mechanical components, (vessels, pipes, valves) cables and containment structures

- developments of methods and data bases for regulatory activity in technical radiological protection and waste management (corrosion, decontamination, liquid waste management, ALARA-requirements)

- development of regulatory guides on fracture mechanical acceptance criteria of indications monitored by in - sevice inspection methods

4. Capabilities of National Research Organizations for WER-speeific R&D Activities

In Chapter 2. it has been mentioned that the most important TSOs of the regulatory body are the AEKI and the VEIKI. The profile of both institutes covers R&D fields dealing with most of the issues arising in the nuclear safery authority's activity.

The AEKI has carried out R&D work hi the nuclear safety related areas as follows: - Fuel behaviour - Reactor physics - Thermohydramics - Severe accidents and radioactive releases - Structural integrity and ageing

The VEIKI has capabilities in the nuclear safety ralated areas as fallows: - Probabilistic safety assessment - Human factor - Containment behaviour - Severe accident analysis - Ageing of mechanical and electrical equipment

More deteiled information on the capabilities and activities of both institutes can be read in the Attachment of this paper.

5. Conclusions

- The adequate nuclear safety level can be ensured only by equistrong system: it is not allowed to neglect any important area. The best way for it is to conduct a coordinated R&D programme sponsored by the regulatory body.

- To achieve and to maintain the internationally acceptable level of the nuclear safety the maximal utilization and permanent development of the international cooperation is necessary. All possible opportunities have to be taken including both bi-lateral and multi-lateral contacts.

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-For countries intending to join the European Union there it is a precondition to use requirements being conform to those of EU. The new Hungarian legal system (Atomic Energy Act, regulatory norms and standards, governmental and ministerial decrees, etc.) has met this precondition. The Hungarian membership of the OECD-NEA has enhanced the opportunity to be involved in different coordinated R&D programmes, like Halden-project or RASPLAV and others.

- The results produced during the last decades in the Hungarian R&D programmes have contributed to the excellent operational and safety records and to the good reputation of the Paks NPP and these results could be benefitted also in any international cooperative action in the future.

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Attachment Nr. 1.

Programme of nuclear safety related R&D activities of KFKI Atomic Energy Research Institute

Budapest 114, P.CXBox 49, H-1525 Hungary

Fuel behaviour

*> Adaptation of fuel behaviour codes for W E R applications (TRANSURANUS for licensing purposes, FRAPCON, FRAP-T6) and development of PIN-MICRO

• Participation in the Halden Reactor Project and possibly in the CABRI measurements

•> Measuring W E R specific data (ballooning, creep rupture, corrosion)

•> Fuel behaviour expérimental tests for research reactor fuel

<• Fuel behaviour during wet and dry storage (WER and research reactor iuel)

Reactor physics

*> Completing the development of the coupled 3D kinetics code {last validation steps, coupling with a 3D vessel model, coupling with the model of the primary circuit)

*> Analysis of boron dilution accidents

*•• Development of the research reactor model (3D kinetics, kinetic effects of the beryllium reflector)

•> Development of burnup calculations (high burmip, spatial effects)

Thermohydraulics

•> Experiments at the PMK facility for WER-440 code validation

• Code validation (RELAP, ATHLET, CATHARE), quantification of uncertainties

• Introduction of CATHARE for licensing purposes

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*> Adaptation of a Computational Ffand Dynamics code, development of a 3D pressure vessel model

<• Coupling of the primary circuit model with the 3D vessel model and/or with the 3D kinetics code

•> Separate effect thermohydraulic tests for the investigation of specific problems (to be agreed with NEA programmes)

• Experimental and calculational investigation of W E R accident management problems

Severe accidents and radioactive releases

•> Fuel behaviour measurements for the beginning of core melt phase (separate effect tests and integral experiments)

•> Experimental investigation of air ingress (coordinated with PHEBUS-FP)

& Modelling of the nvvessel processes by using the ICARE code

• Participation in the RASPLAV project

•> Categorization of radioactive releases for later use in Level 3 PSA and in nuclear emergency preparedness

•> Development of a severe accident simulator based on the MELCOR code

Structural integrity and ageing

• Development of PTS methodology (effects of cladding, crack arrest, uncertainty analysis)

*> Experimental investigation of ageing of reactor pressure vessel materials (welding, cladding)

•> Application of the MARC-4 3D finite element code for non-linear problems

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3 -

Man-machJne connection, simulation

<• Critical safety function display systems

• Safety analyses for the development of symptom oriented emergency procedures

•> Development of 3D simulation techniques

•> Participation in HAMMLAB activities

*> Development of hardware in-the loop test methodology

*t* Verification and validation of softwares

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Attachment Nr. 2.

Company profile in the nuclear safety related field of VEIKI, Institute for Electric Power Research Co.

Budapest, P.O. Box 80. H 1251 Hungary

Probabilistic safety assessment and system reliability analysis

* calculation of core damage frequency due to different internal initiating events, failed human interactions, as well as internal and external hazards (fires, floods, earthquakes). The PSA activity covers risk quantification for all plant operational modes, i.e. operation on nominal power level, as well as under shutdowri/remeling/startup conditions, too.

* safety-related studies to support the engineering design activity to prepare safety upgrading plant modifications. These studies are aimed at the demonstration of the effects that result technological, electrical and I&C system modifications.

* development of new methods and tools for various PSA applications. On Hie one hand this work covers setup of new computerized procedures, algorithms, and software codes for PSA, on the other hand it is related to specific research objectives, e.g. to simulator aided human reliability analysis of the NPP control room crews.

* development of new Iraining materials to support the systematic training of NPP operators. These documents contain summary description of evolutionary results gained from completed safety analysis projects aimed at improving the basic engineering knowledge, as well as the safety culture level of the operating personnel. Some of the training documents are issued as lesson guides for simulator-based initial and continuing Iraining of the operators.

* flow-induced vibration analysis of primary circuit main equipment, as well as vibration and seismic qualification of selected I&C components (e.g.of relays and protection units) by analytical and experimental methods.

Safety analysis

* application of qualified western computer codes describing beyond design basis accident situations for WER-type NPPs. The application covers setup of model parameters and input data decks valid for WER-440/213 type NPPs, as well as their verification based on simulated and experimental data.(The important accident phenomena that might occur within the primary, the secondary systems and in the containment are treated from the initiating event to the determination of the fission product release to the environment.)

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* use of fast running accident analysis codes for a priori estimation of potential consequences of accidents to support the decision making tasks for mitigation. The consequences are given in the form of time-function of the potential radioactive release (source term) estimated by using the essential plant parameters received on-line in a crisis centre during a given accident.

* containment analysis including simulation and experimental investigation, too. Simulation covers calculation of its mechanical loads under design basis and severe accident situations. The experimental investigation as aimed at the verification of functional operation of its passive pressure suppression structure.

* study of possibilities for removal of hydrogen generated during design and beyond design basis accidents. The studies are directed to the calculation of hydrogen contcentration as a function of time and location, as well as to the selection of the most effective method and tool for hydrogen elimination, e.g.by passive catalytic recombiners.

Ageing of mechanical and electrical equipment

* assessment of life time limiting factors * development of monitoring methodologies to follow changes of conditions due to

ageing.

369

New Programme for the R&D Activity, Sponsored by the Hungarian Nuclear Safety Authority Lajos Vôrôss Director, Nuclear Safety Inspectorate, Hungarian Atomic Energy Commission

International Seminar on The Safety Research. Needs for Russian-Designed Reactors, " Kohku Kaikan ", Tokyo, Japan, 8-9 July 1997

OUITLINE

1. Short history and background of R&D programmes for nuclear energy in Hungary 2. Medium term R&D programme for regulatory assistance 3. Some of the new results and on-going activities 4. Capabilities of national research organizations for W E R - specific R&D activities 5. Conclusions

1. Short history and background

- Coordinated national R&D programmes to support * construction and commission of Paks NPP - 1981 - 85 * preparation of constructing further nuclear units of W E R - 1 0 0 0 (terminated) -

1986 - 87 * safety analysis of WER-440/213 - 1988 - 90

- Institutional and individual applications for reduced central budget resources - 1991 -95

* internationally integrated R&D * AGNES - project - safety reassessment of Paks unitNr. 3.

* *

2. Medium term R&D programme for regulatory assisstence

- no financial resources available for nuclear safety regulatory body until 1995.

- R&D priorities defined in 1995 * ageing and residual life of equipment * PSA applications for regulatory use * containment behaviour * computer code V&V * emergency response * severe accidents * accident management * human factor and safety culture

- 1996 - Nat'l Comm. for Tech. Development's sources - from 1997 - own budget ( new Atomic Energy Act)

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- 2 -2. Medium term R&D programme.—(cont'd)

- priorities yearly reviewed - TSOs for Nuclear Safety Inspectorate (NSI)

* agreements with AEKI and VEIKI - CERTA - concept (Centre for Emergency Response, Training and Analysis)

* multipurpose centre * international assistance

• UK Government EC/PHARE

• I A E A / T C * opened 1. July 1997

- loose connection with industrial R&D

* *

3. Some new results and on-going activities

- Visual Information Transfer and Analysis (VITA) in CERTA * developer: AEKI, Budapest * on-line data connection with Paks NPP * modular and flexible structure * SPDS - function ready to be used

• seven critical safety functions • 6 safety parameters to each CSF • annunciators of CSFs and time history of parameters • p - T diagrams

* second phase under discussion - use in normal operation (based on PICASSO -3 SW - package of OECD - Halden - project)

3 . Some new results...(cont'd)

- Risk supervisor * developer: VEIKI, Budapest * configuration control and evaluation of annual risk history * simplified technological model used * calculates CDF and contributors for cases with

• arbitrary operational mode of safety related normal operation systems • additional component failures • unavailability of any engineered safety system (ESF) • testing of trains of any ESF • operational action during fuel cycle

* living PSA necessary * s implif ied mnrlel validated

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3 . Some new resuIks...(cont'd)

- On-going projects * coupled neutron kinetic - TH code (KIK03D - ATHLET) - GRS * adapted TH-codes (RELAP, CATHARE, SCDAP/RELAP, MELCOR) - mainly

USNRC * containment investigations (H-problem, large scale experiment, CONTAIN) -

PHARE, NRC * MAAP 4 / W E R version, fast estimation - PHARE * SESAME, diagnosis/prognosis in emergency -PHARE * FE behaviour (TRANSURANUS) - IAEA, KfK * PSA - based event analysis * severe accident simulator * reg. guides for ageing (mech.& electr.equipm., structures) * reg. guides for techn. radiological protection

* *

4. Capabilities of Nat'1 Research Organizations

- AEKI - KFKI Atomic Energy Research Institute of Hungarian Academy of Sciences, Budapest

* 210 employee, fully nuclear profile * R&D areas

• Fuel behaviour • Reactor physics • Thennohydraulics • Severe accidents and radioactive release • Structural integrity and ageing

- VEIKI - Institute for Electric Power Research Co., Budapest

* 100 employee, only partly nuclear profile * Nuclear R&D areas

• PSA • Human factor • Containment behaviour • Severe accident analysis • Ageing of mech. & electr. equipm.

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5. Conclusions

- importance of coordinated, regulatory sponsored R&D programme for equistrong system in Hungary

- maximal utilization and permanent development of the int'l cooperation in every field

- EU -conform requirements in new Hungarian legal system - OECD-NEA membership enhances opportunities (Halden, RASPLAV., WGs3 etc.) - R&D results contribution to safety records and reputation of nuclear industry in

Hungary

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Operation Deepfreeze

CERTA IBM RISC

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IBM RISC

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PC

374

CERTA layout Video Display System

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Scenarios, Risk,Precursor

Training facilities

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ANALYSIS

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OfifN - NEA - International Nuclear Information System (INIS)

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