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CONGRESO INTERNACIONAL DE JÓVENES NUCLEARES 2000: JUVENTUD, FUTURO, NUCLEAR Международный Молодежный Ядерный Конгресс 2000: Молодежь, Будущее, Ядерные технологии sa ёш^§Ш: as, DI EH, Le Congrès Nucléaire International 2000 * de la Jeunesse: Les Jeunes, Avenir du Nucléaire Internationaler Kongress junger Kerntechniker 2000: "Young Generation" für die Zukunft der Kernenergie AMANSAN MMABUN WIEMU NHY1AMU 2000 (MFIE MPEMIENU), !' : *'4" .JMMABUN, WIEMU, DAAKYE

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2ÜÜU Y111 ülubkirorcisi Goiif:iik Nukloer Kongresi Welcome

Welcome to Bratislava, Slovakia for the first International Youth Nuclear Congress (IYNC 2000), organised by a new generation of professionals in the nuclear field from different countries to:

• develop new approaches to communicate benefits of nuclear power, as part of a balanced energy mix;

• promote further use of nuclear science and technology for the welfare of mankind;

• transfer knowledge from the current generation of leading scientists to the next generation;

• encourage the creation of a global network among young professionals.

The primary purpose of the Congress is to transfer knowledge from the current generation of leading scientists and engineers to the next generation. Scientific, political, public and corporate views regarding the development of different nuclear issues will be presented to provide comprehensive discussions on all sides of the subject. Since it is the youth who will go on to deal with all nuclear prospects and challenges in the future, we expect open and candid commentary from this younger generation.

For the latest Congress info and contact information please visit our Web site: http://www-nuen.tamu.edu/iync_2000 IYNC 2000 Sponsors and Supporting Organizations* Gold Sponsors

American Nuclear Society

COGEMA COGEMA

European Nuclear Society

Excel Services Corporation

International Atomic Energy Agency

Korean Nuclear Society

Ministry of Atomic Energy of Russia

Nuclear Society of Russia

Silver Sponsors. Al It It J"tlPIP ABB Atom

Bronze Sponsors

Xrcaiia CcmuCting, Inc. Arcadia Consulting BNFL BNFL Mayor of Bratislava

NBVAM Kwmnc Nevada Scientific Group

SIEMENS Siemens

SLOVENSKE Slovak Electric, Pic. # ELEKTRARNE Spanish Nuclear Society

* Alphabetical order in all categories Contributors Brewer & Associates The John W. Simpson Group Vedecko-technicka spolocnost pri VUJE U.S. Department of Energy

Supporting organisations ETCetera FORATOM French Nuclear Society International Nuclear Societies Council Nuclear Regulatory Authority of Slovakia Slovak Nuclear Society Uranium Institute Women in Nuclear

Executive Committee Supporting Decom Slovakia Ltd. Organisations Institute of Nuclear Power Engineering Inst. of Physics and Power Engineering Kurchatov Institute Mochovce NPP VUJE Trnava Inc. Utility.com

IYNC 2000 International Advisory Committee

Alexandre Brioussov, IAEA, Pierre-Louis Chometon, COGEMA, France Gerald E. CJa|p|jyr,aniunri Institute, Great Britain Andrej Gagary^i^^^ar Soeiejy

Yur ia Sergey Koltyshev, itre, Kazakhstan Yuri Serguei Ku ssia German Guido ntina Peter Lyons, assistant nici, USA Valery Mezguev, JSC Machi? Russia Victor Mourogov, IAEA, Austria Marc Nikolaev, IPPE, Russia Jacques Panossian, SFEN, France Goon-Cherl Park, Seoul National University, Korea N.N.Ponomarev-Stepnoi, Kurchatov Institute, Russia Agneta Rising, ENS and WIN, Sweden Anrej Stritar, Nuclear Society of Slovenia, Slovenia Jiri Suchomel, Slovak Nuclear Society, Slovakia Anatoly Zrodnikov, IPPE, Russia IYNC 2000 International Youth Executive Committee General Chair Alexandre Tsiboulia, IPPE, Russia Technical Program Chair Organizing Chair Florence Avezou Stanislav Rapavy COGEMA, France VUJE Trnava, Slovakia Public Relations Chair Finance Chair Emmy Roos Vladimir Mighal Rocky Mountain Remediation Services, USA VUJE TrnaVa, Slovakiai-: .,-[, Publications & Web Chair Reservations Chair Shannon Bragg-Sitton Marian Knstof University of Michigan, USA Nuclear Regulatory Authority §lovakia International Chair Technical Tours & Special Events Chair Serguei Klykov Robert Holy ' s INPE, Russia Mochovce NPR Slovakia Corporate Relations Chair Registration Chair August Pipkin-Fern Kristina Kristofova Utility.com, USA Decom, Slovakia Web-Master Sergei Tsyganov, Kurchatov Institute, Russia

Technical Program Sub-Committee Chair Florence Avezou, COGEMA, France Ignacio Sebastian Luppi Berlanga Gaston Meskens Yevgeniy Rozhikhin CNEA, Argentina SCK/CEN, IPPE, Russia Marko Cepin Stanislav Rapavy Donna Smith Jozef Stefan Institute, Slovenia VUJE Trnava, Slovakia MM., USA Anthony Hechanova Bradley Rearden Alexandre Tsiboulia Univ. of Nevada, Las Vegas, USA ORNL, USA IPPE, Russia

International Sub-Committee . Chair Serguei Klykov, INPE, Russia

Agrentlna Cuba Korea Slovakia Ignacio Sebastian Luppi Berlanga Saul Hernandez Valle Poong Hyun Seong Stanislav Rapavy

Armenia France Slovenia Stephan Gevorgyan Florence Avezou Audrius Jasiulevicius Marko Cepin

Australia Macedonia Tina Eddowes Osei Franklin Amoako Andrija Volkanovski Ezekiel Mohajane Obeng Samuel Mireku Belarus Mexico Sweden Vladimir Drozdovitch Rogelio Rea Soto Roger Carlsson Antoine Van de Velde Bulgaria Switzerland Dimitar Antonov Olusola Sokefun Urs Meyer Teodor Nochev Kristof Horvath Norway Spain Ijqp Moflfl Luis Garcia Monica Chandia R.K.Sinha Lluu IVIwOF 1 Ukraine China Japan Duan Liuyi Mihnea Anghelescu Olga Radovskaya Ryota Omori Jan-Ru Tang USA Russia Croatia Kazakhstan Shannon Bragg-Sitton Roman Tertytchnyi Ines Ana Jurkovic Zhanac Sabyrzhanova Emmy Roos About Bratislava Bratislava is a city with a rich history dating back to the Early Stone Age when first settlements were established in the area. Since this period the Bratislava region has been continuously settled. The Romans played an important role in the history of the city in the first half of the last millennium, when the Danube River became the frontier of the Roman Empire. By the end of the migration of nations, in the 5th century, the Slavs had arrived in the Bratislava region and in the beginning of the 9th century, definite boundaries between the Slav and German worlds were defined. The development of crafts, increasing social stratification, the establishment of tribal principalities and their mutual struggles led to the establishment of th first independent state, Great Moravia, in the first third of the 9th century. From 1536 to 1783 Bratislava was th capital of Hungary, the seat of the parliament and central bodies and the coronation town of Hungarian kings In 1993 Bratislava became the capital of the independent Slovak Republic. The nearest international airport (45 km from Bratislava) is Vienna International Airport but the local Bratislav airport offers regular flights from , and Munich. Bratislava can be reached from Vienn International Airport in approx. 90 minutes by shuttle-service.

Participants Participants in IYNC 2000 include both the new generation of nuclear scientists and engineers, and establishe professionals in the various fields of nuclear science and technology. It is anticipated that all participants w share a desire to think "outside the box" to consider innovative approaches to the issues facing the nudes industry now and in the future.

Active participation in IYNC 2000 is being solicited from students, young professionals and establishe professionals in the various fields of nuclear science and technology, and all those with an interesting view o the future of nuclear.

Conference Locatiol IYNC 2000 is hWcHn Bratislava in tri Facility of Ministry Of Foreign Affairs the Slovak Republic (SUZA), Drotarska cesta 46, 81104 Bratislava, phone: +421 7 5935 4114. We would like to extend to you an invitation to participate in a unique, first-of-its-kind event organized by an international group of young people. The International Youth Nuclear Congress 2000 (IYNC 2000) will bring together a new generation of nuclear scientists and engineers to discuss current and future issues in the nuclear field. In parallel to the estab- lishment of this international network, a common message on nuclear issues - distilled out of these days of unique interaction - will be sent to the outside world. Participants in IYNC 2000 will include both the new generation of nuclear scientists and engineers and established professionals in the various fields of nuclear science and technol- ogy. It is anticipated that all participants will share a desire to think «outside the box» to consider innovative approaches to the issues facing the nuclear industry now and in the future. Active participation in IYNC 2000 is being solicited from students, young professionals, and all those with an interesting view on the future of nuclear science and technology. Par- ticipation through paper or poster presentations, plenary sessions, and workshop discus- sions is encouraged from all those who are excited about the future of nuclear science and technology. ililli (I international Youth Nuclear Congress

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Dieser Tagungsband informiert Sie über ein besonderes Ereignis - das Erste seiner Art - organisiert von jungen Menschen aus aller Welt. Der IYNC führt eine neue Generation von Nuklearwissenschaftlern und Ingenieuren zusammen, um aktuelle Fragestellungen und zukünftige Herausforderungen im Nuklearbereich zu diskutieren. Neben der Knüpfung eines weltumspannenden Netzwerkes von Nachwuchskräften soll - als Ergebnis intensiver Zusammenarbeit während dieser Konferenz - eine gemeinsame Botschaft zu nuklearen Themen in die Welt gesandt werden. Der Teilnehmerkreis des IYNC setzt sich sowohl aus jungen Wissenschaftlern und Ingenieuren als auch aus erfahrenen Fachleuten unterschiedlicher Bereiche der Nuklearwissenschaft und -technologie zusammen. Von allen Teilnehmern wird erwartet, dass sie «über den Tellerrand hinausschauen» wollen, um innovative Lösungsansätze für die Aufgaben in der Nuklearindustrie zu entwickeln. Studenten, junge Fachleute und alle, die sich fbr die Fortentwicklung der Kerntechnik interessieren, werden zur regen Teilnahme an dem IYNC aufgefordert. Вейгдде in Form von Aufsfltzen oder Posterprflsentationen, sowie die Teilnahme an Plenarsitzungen und Diskussionsforen sind von allen erwbnscht, die zuversichtlich in die Zukunft der Kerntechnik schauen. Мы приглашаем Вас на уникальное мероприятие, впервые организуемое группой молодежи из разных стран. Международный Молодежный Ядерный Конгресс 2000 (IYNC 2000) соберет с 9 по 14 апреля 2000 года в Г.Братиславе, Словакия новое поколение специалистов и инженеров со всего мира для того чтобы обсудить перспективы и проблемы ядерной науки и техники. Среди участников Конгресса будет как новое поколение ученых и инженеров- ядерщиков, так и признанные специалисты в различных областях ядерной науки и техники. Ожидается открытое, не ограниченное рамками профессиональных стереотипов обсуждение новых подходов к вопросам, встающим перед ядерной отраслью сегодня и ожидаемым в будущем. Активное участие в Конгрессе ожидается от студентов, молодых учёных, равно как и заслуженных специалистов из различных областей ядерной науки и техники, а также всех тех, у кого имеются интересные взгляды на будущее ядерной отрасли. Мы приглашаем Вас принять участие в Конгрессе. Очень хотелось бы, чтобы все кто заинтересован в будущем ядерной науки и техники, приняли участие с пленарными и стендовыми докладами, выступлениями на дискуссионных заседаниях и семинарах. Dit document beschrijft een uniek event dat voor de eerste maal is ingericht door een internationale groep jongeren. De International Youth in Nuclear Congress 2000 (IYNC 2000) brengt een nieuwe generatie wetenschappers en ingénieurs in kernenergie samen om de huidige en toekomstige vraagstukken aangaande kernenergie te bespreken. Naast de oprichting van dit internationaal netwerk, zal tijdens deze dagen van unieke interactie ook een gemeenschappelijke boodschap over kernenergie naar de buitenwereld verstuurd worden.

Zowel de nieuwe als de meer gevestigde generatie ingénieurs en wetenschappers uit de verschiilende wetenschappelijke en technologische toepassingsgebieden van kernenergie nemen deel aan IYNC. Van alle deelnemers wordt verwacht dat ze een verlangen hebben om de problemen die de kernindustrie kent op een innoverende manier te benaderen, zowel nu als in de toekomst.

Studenten, jonge professionelen, en iedereen met een interessante kijk op en een enthousiasme voor de toekomst van de verschiilende wetenschappelijke en technologische toepassingsgebieden van kernenergie wordt verzocht actief deel te nemen in IYNC via een paper of een poster voorsteiling en via plenaire sessies en workshop discussies. 2000 (IYNC 2000)^]^^ KEYNOTE SPEAKER BIOGRAPHIES Agneta RISING

Name Agneta Rising Sex: Female Nationality: Swedish

Presidency President (1996-) and co-founder of Women In Nuclear (WIN) Global, a worldwide organisation with 1600 members in 48 countries.

President (year 2000-) The European Nuclear Society. Vice President (1998-1999) of European Nuclear Society.

Chairman Elect of The Uranium Institute, will become Chairman May 2000. Vice Chairman (1999-2000) of The Uranium Institute

President (1999-) of The Swedish Nuclear Society Vice President (1998) of Swedish Nuclear society

Board membership

Active member of Executive Committee of Uranium Institute, 1998 -

Board member of European Nuclear Society, 1998 -

Board member of Swedish Nuclear Society, 1988 - 1994, 1997 -

Board member of Women In Nuclear, WIN Global, 1993 - Member of Analysis Group at Nuclear Training and Safety Centre, 1987-, This group has responsibility for information in the field of nuclear safety, the effects of ionising radiation and comperative safety energy systems.

Former member of Board member (1990-96) of Nordic Society for Radiation Protection.

Member of Committee for Nuclear Energy and the Public, at the Uranium Institute.

Member of Information Committee of European Nuclear Society.

Current position Corporate senior adviser at Vattenfall AB. Since 1980 working at Vattenfall AB (formerly Swedish State Power Board) with radiation protection issues related to the nuclear fuel cycle. Appointed specialist, at the highest level within the Vattenfall Group, in the area of nuclear energy and environment. Working site: Ringhals Nuclear Power Plant.

Former Former Head of department for Radiological Safety at the head office of positions Swedish State Power Board.

Inspection of medical equipment at National Institute of Radiation Protection.

KEYNOTE SPEAKER BIOGRAPHIES 117 Other Project leader for Nordic projects in the safety program of Nordic liaison professional committe for atomic energy. records

Lecturing Lecturer for several different groups externally, such as teachers, politicians, students, universities, emergency personel, societies.

Teacher for many internal groups in environmental issues connected to the nuclear fuel cycle, radioactive releases from nuclear power reactors during normal operation and accidents: trainees, reactor operators, nurses, radiation protection personel, maintenance personel etc.

Advanced Participating in media interviews and debates, in newspapers, TV and radio communication programmes mostly in Sweden but also internationally in Polen, Australia, Argentina and Canada, Spain, Taiwan.

Participating in seminars and debates, both inside and outside the industry, also including anti-nuclear and green activist groups.

Auditor & Auditor for fulfilment of environmental criteria for companies in different parts of the nuclear fuel cycle that might be considered supplier to logivbA Vattenfall. These environmental audits have been fully designed and lead by the customer (Vattenfall together with other utilities in Sweden). This work has been unique in the world because the audits have been customer designed and performed. For example uranium mines in Uzbekistan, Kazakhstan, Canada and Russia have been assessed according to our criteria. Participating in the auditing team as expert in the radiation protection area for: employees, working environment, releases, environmental impact, emergency and education.

Advisor to the government's program analysing and reporting on the devel- opment of new Nuclear Power Reactors in the world. Appointed by the Minister of Energy.

Academic records Master of Science, Radiation Physics, Health Physicist.

Assistant researcher at the University.

Family Position Married, three children, born -81, -84 and -92. Living at the West coast of Sweden, near Gothenburg.

IB IKEYKOTE SPEAKER BIOGRAPHIES MARC DEFFRENNES

Mechanical Engineer in 1979, with complementary cycle in Nuclear Engineering in 1980 from the University of Louvain in Belgium.

Joined Westinghouse in 1982 and hold diverse positions :

- Safety Analyses engineer : Accidents Analyses, Setpoint Studies, Emergency Procedures,...

- Startup Engineer for the post critical startup test of the Tihange 3 plant in Belgium.

- Health Physics responsible for the W Maintenance Service Center in Nivelles.

- Responsible for the training services for customers (all W plants in ) : classroom and simulator training for plant operators and engineers, support ot plant operation.

Joined the European Commission in 1991 in the department Nuclear Energy of the General Directorate for Energy:

- follow and policy advice for all the activities of the European Commission in relation to nuclear energy issues in Eastern Europe, Russia and Ukraine : Phare and Tacis programmes, industrial cooperation initiatives, handling of the plutonium released from disarmament, Accession and Enlargment of the European Union.

Changed to the «Euratom Coordination)) department early 2000.

KEYNOTE SPfAKEB BIOGflAPHlES 110 Professor Andrei Yu. GAGARINSK! Director for Foreign Affairs Russian Research Centre «Kurchatov Institute»

Born in 1939. Graduate of Moscow Engineering & Physics Institute (1962). Since then till today works in Russian Research Centre «Kurchatov Institute.

Doctor of Physical & Mathematical Sciences, author of several monographs and over 100 scientific articles. Field of scientific activity: experimental nuclear reactor physics, nuclear criticality safety, pressurized water reactors, nuclear power engineering.

One of the founders of the USSR Nuclear Society; currently - Vice-President of the Nuclear Society of Russia. Member of Board of the European Nuclear Society, member of the American Nuclear Society.

Bearer of the Order for Courage of the Russian Federation for participation in the works on liquidation of the Chernobyl accident consequences.

In 1999 became Editor-in-chief of new magazine «Yadernoe Obshchestvo» (Nuclear Society) Russia. Member of editorial boards of the magazines: «Atomnaya Energiya» (Atomic Energy) - Russia; Nucleonika - Poland; Nuclear Europe Worldscan, Nuclear Engineering & Design - ENS.

201 KEYNOTE SPEAKER BIOGRAPHIES John GRAHAM

Partner

Tel: 303-770-7646 Fax: 303-804-9825 e-mail: [email protected]

Over 40 years of nuclear technology safety experience, in the development of reactor systems, in nuclear waste technology, in education and in regulation. Over 30 years of management experi- ence over technical and administrative staff with several-million dollar budgets for both private corporation and U.K., U.S. and Canadian government organizations.

EDUCATION 1951 - 1954 University College of North Wales, Bangor, Wales B.Sc. (Summa Cum Laude) in mathematics and physics, British State Scholarship, other honors 1954 - 1955 University of Illinois, Champaign-Urbana, Illinois, USA Graduate study in mathematics and quantum mechanics on a Fullbright and University of Illinois Fellowships 1955 - 1958 University College of London, London, England Doctoral research in theoretical high-energy physics on a Department Scientific & Industrial Research Scholarship

EXPERIENCE 1998 - Present: Partner, ETCetera Lip Partner with responsibility for training, for safety and quality assessments and technical program evaluation and development and Website evaluation and development. 1994 - 1996: American Nuclear Society, Vice President and President with responsibil- ity for all aspects of the business of the Society on behalf of the Board, as well as the Public and International Outreach. The Society had 16,000 members, and over 120 local units in domestic and foreign locations, plants and universities, with a budget of $7 million. 1991 - 1998: BNFL Inc. Vice President Managed Environmental Health and Safety and Quality Assurance functions for the company's engineering activities providing technical input to engineering as well as waste management and D&D proposals and contracts. In addition, I served as Manager of the Denver Office. 1988 -1991: Atomic Energy of Canada, Ltd. - Research Company. Directed the licens- ing activities for nuclear related facilities including small reactors, (high flux materials test or isotopic production facilities, and local energy systems), isotope use and production as well as new waste facilities. Specific reactors included the Slowpoke Demonstration Reactor at Whiteshell Nuclear Research Establishment, the MAPLE-X10 at the Chalk River Nuclear Research Labora- tories, and the NRU refurbishment to 2020. The responsibility included the development of small reactor safety and licensing criteria and interface with the Atomic Energy Control Board and U.S. State regulators. 1984 - 1988: Westinghouse Hanford Company (previously Rockwell Hanford Opera- tions). Managed the Regulatory and Institutional end-function, and the Licensing Group for the Basalt Waste Isolation Project (BWIP). This involved the production of major pre-licensing planning and evaluation documents such as the site Environmental Assessment and the Site Char- acterization Plan. The group also provided support for Department of Energy interfaces with the public, the State and the affected Indian Tribes. KEYNOTE 5PEAKE1 BIOGRAPHIES 121 1972 -1984: Westinghouse Advanced Reactors Division and its successor, the Advanced Energy Systems Division. Managed the safety and reliability department to ensure the licensability and public acceptability of the Clinch River Breeder Reactor (CRBRP). Developed the safety bases for the design, and coordinated the production of significant sections of the PSAR, leading to the award of the construction license in 1984. Managed the Advanced Reactor Division on-site (OSHA) safety program including the contin- ued clean-up of residual effects of the Westinghouse Test Reactor failure. 1968- 1972: Westinghouse Advanced Reactors Division. First as a senior scientist in safety, and then as a manager, developed the safety design bases for the Fast Flux Test Facility (FFTF). Developed and helped to implement the severe accident policy (fuel failure and hypo- thetical accident analysis) for the FSAR. Managed a multi-discipline engineering group provid- ing stress, thermal and hydraulic, design and safety analysis for the reactor design of FFTF. 1969 - 1974: Carnegie Mellon University, Pittsburgh, PA. Part time faculty for Fast Re- actor Safety, Nuclear Engineering Program. 1958 -1968: United Kingdom Atomic Energy Authority, Winfrith, Dorset, UK. Senior Scientist developed the safety principles for a wide variety of thermal and fast reactors, including the Advanced Gas-cooled Reactor and the Prototype Fast Reactor. Developed and lectured on methods in kinetic analysis. 1951 -1954: University of London. Lecturer in calculus and mathematical modeling at Northern Polytechnic, Chelsea Polytechnic, and the College of Rubber Technology.

MEMBERSHIPS 1971- Present: American Nuclear Society — Fellow (Vice President and President 1994-6) 1995 - 1997: American Association of Engineering Societies (Vice Chairman 1996-7) 1954 -1956: American Mathematical Society

PUBLICATIONS 45 technical papers latest 1998 (list on request), and over 1,000 general articles.

BOOKS «Fast Reactor Safety», Academic Press, New York and London, 1971 «Target 26-a definitive guide on marathon runnings, Macmillan Publishing Co., New York, NY, 1978, 2nd edition 1983; foreign edition, Collier Books, Toronto, Canada, 1980, «Target 42», (Russian translation of Target 26), Fitzcultura y Sport, Moscow, 1980 ((The Literature ofChess», McFarland and Co., Jefferson, NC, 1986 ((Women in Chess -players of the modern age», McFarland and Co., Jefferson, NC, 1987 4 other unpublished books.

HONORS Fellow of the American Nuclear Society Citizen of West Virginia — Governor's Award for work in energy Honorary Life Member of the US Blind Chess Association

22 I KEYNOTE SPEAKER BIOGRAPHIES ANDREW C. KADAK, Ph.D.

253 Rumstick Point Road, Barrington, RI02806 Tel: 401-245-0775 Fax:401-245-0385 Email: [email protected]

SUMMARY QUALIFICATIONS:

Dr. Kadak is a Professor of the Practice in the Nuclear Engineering Department at MIT. He is also President of Kadak Associates, a specialty consulting firm he recently established after 18 years of experience at Yankee Atomic Electric Company. At Yankee, Kadak was President and Chief Executive Officer. In this capacity, he was responsible for overseeing all Yankee operations, includ- ing the decommissioning of the Yankee plant in Rowe, Massachusetts and engineering, licensing, environmental and operational support to all eight nuclear plants in New England and many other national and international clients.

Dr. Kadak's expertise ranges from day to day operations of nuclear plants to senior executive management. In the past he has lead Yankee Atomic in license renewal of operating reactors, system- atic evaluation of older plants to allow them to demonstrate compliance to new regulations, financial rate proceedings to assure adequate capital for safe operation, innovative fuel purchase agreements, high level nuclear waste disposal and storage solutions.

Dr. Kadak is actively involved in industry efforts in seeking political solutions to the problem of high level waste disposal, improving the organizational structure of nuclear power companies to improve their competitiveness, decommissioning and regulatory reform.

Dr. Kadak is also President of the American Nuclear Society. He has served as a board and executive committee member of the Nuclear Energy Institute and the industry's Advisory Committee on High Level Waste. In 1995, he was a member of the Advisory Committee on External Regulation of Department of Energy Nuclear Safety. He has also conducted several audits of nuclear companies to assess management and operations. Dr. Kadak has made more than 50 lectures and speeches on topics related to the technical and business aspects of nuclear power. He is a continuing lecturer at the MIT summer safety course and the MIT and Institute of Nuclear Operations Utility Executive Training course.

Dr. Kadak's current consulting work is concentrated in the following areas: Safety Review Board of Siemens Nuclear Power Corporation; member of the Board of Directors of ATG, Inc. a low level radioactive waste services and processing company; a consultant to Southern California Edison on their decommissioning project. In addition, he consults with companies who are interested selling or acquiring nuclear plant assets.

EDUCATION: Massachusetts Institute of Technology - Ph.D., Nuclear Engineering (Reactor Physics) (1972) - M.S., Nuclear Engineering (1970) Northeastern University - M.B.A. (1983) Union College - B.S., Mechanical Engineering, Cum Laude (1967)

KEYNOTE SPEAKER BIOGRAPHIES 123 Chang Sun KANG

Date of Birth: April 13, 1943

Education: Seoul National University, B.S. in Nuclear Engineering, 1965 M.I.T., Sc.D. in Nuclear Engineering, 1972

Professional Engineering Resistration: State of Pennsylvania , State of California Republic of Korea

Summary of Experience 1980 to Date: Seoul National University; Professor, Department of Nuclear Engineering 1977 to 1980: Daewoo Engineering Company, Ltd.; Executive Managing Director 1971 to 1977: United Engineers and Constructors, Inc.; Manager of Reactor Systems 1981 to 1982: Associate Dean of College of Engineering, SNU 1993 to 1995: Director of Nuclear Safety Center in the IFEE 1993 to 1995: Vice-President of Korean Nuclear Society 1995 to 1997: Director of SNU Development Fund Foundation 1995 to 1997: Member of Presidential Council on Science and Technology 1997 to 1999: Editor-in-Chief for Journal of Korean Nuclear Society 1999 to Date: President of Korean Nuclear Society

Advisory Activities Ministry of Commerce, Industry & Energy: Steering Committee on KNGR Development Ministry of Labor: Review Committee on Technical Manpower System Korea Commercial Arbitration Board: Arbitrator Korea Institute of Nuclear Safety: Advisory Committee on Nuclear Safety Korea Electric Association: Committee on Nuclear Codes & Standards

Publications Introduction to Nuclear Engineering, KNS (1989) Modern Industrial Society & Energy, SNU Press (1992)

241 KEYNOTE SPEAKER BIOGRAPHIES Ann MACLACHLAN

European Bureau Chief of McGraw-Hill Nuclear Publications (Nucleonics Week, NuclearFuel, Inside NRC, Nuclear News Flashes). She has worked with the Nuclear Publications since 1982, first as European Editor, then European Bureau Chief for the past 13 years. Before that, she was European correspondent for King Publications (The Energy Daily, Defense Week, Metals Daily), based in . She began her journalistic career with Weekly Energy Report in Washington, D.C. in the middle of the 1973-74 oil crisis, and later served as managing editor of The Energy Daily, the U.S. daily newsletter created on the basis of Weekly Energy Report. She holds a master's degree in linguistics from the University of Michigan.

KEYNOTE SPEAKER BIOGRAPHIES 125 Professor Alexander MIKHALEVICH

Graduated in 1961 from Belarus Polytechnical Institute where he specialized in heat and power engineering. He completed a postgraduate course at the Physical and Technical Institute of the Acad- emy of Sciences of Belarus in 1966, after which he worked at the Institute of Nuclear Power Engineering of the Academy of Sciences of Belarus. In 1983 he became Vice Director of scientific work at this Institute, and in 1991 he became a Director of the Institute of Power Engineering Problems which succeeded it. Presently Prof. Mikhalevich is President of Belorussian Nuclear Society, full member of National Academy of Sciences of Belarus, International Engineering Academy and Academic Francophone d'Ingenieurs (le member foudateur). He is author of the 5 books, 20 patents and more than 130 papers. Area of expertise: energy strategy and planning nuclear power, radiological monitoring, energy saving, renewable energy sources.

261 KEYNOTE SPEAKS BIOGRAPHIES Victor V. ORLOV

Victor Victorovich Orlov,Dr.Sci. (Phys. & Math.), Professor, Academician of Russian Acad- emy of Natural Sciences. Chairman of the Scientific and Technical Council No. 1 of Minatom of Russia, Chief Scientist for Innovative Developments in RDIPE, Professor of Moscow Engineering and Physics Institute.

Graduated from Moscow State University (Physics Department, theoretical physics). Until 1976, worked in the Physics and Energy Institute in Obninsk.

Activities: Theory of reactors, input to the first NPP, development of the first fast power reactor BN-350, then BN-600 (Deputy to Chief Scientist A.I. Leipounsky).

1976-1988: Kurchatov Institute, development of the fundamentals for a demonstration fusion reactor, quest for a new fast reactor concept.

Since 1988 up to now: Research and Development Institute of Power Engineering (RDIPE), development of a concept of a «naturally safe» fast reactor and work on BREST design. Winner of Lenin and State awards of the USSR.

1990-1991: President of the USSR Nuclear Society.

KEYHDTE SPEAKER BIOGRAPHIES I ?7 Mr. James M. REED

Department of Technical Co-operation IAEA, Austria

Mr. James M. REED has worked in the Technical Co-operation Department since 1997 as a Senior Programme Management Officer responsible for nuclear safety and power in Europe. Before joining the agency Mr. Reed worked for the UK Nuclear Installations Inspectorate as a Senior Regu- lator. In this position he had close links with the EU's nuclear safety assistance programme for Easter Europe. He has been working in the nuclear industry for many years and is very much familiar with the role of nuclear energy for development.

28 j KEYHOTE SPEAKER BIOGRAPHIES Mr. Jean-Pierre ROUGEAU

Jean-Pierre ROUGEAU, born in 193 7, is Senior Advisor to the Chairman and CEO of COGEM A, President of FORATOM and Chairman of the Uranium Institute.

Jean-Pierre ROUGEAU graduated in Chemical Engineering from the Ecole Nationale Superieure des Mines de Paris, where he is currently a Professor.

He began his career in 1962 at the ISPRA Research Center of the European Union (EURATOM) to head the Core Components Section. In 1967 he joined the french Commissariat a l'Energie Atomique (CEA) where he served in various positions, from the Fuel Elements Control Department, to a super- visory position in economic and optimization studies in the fast reactor program.

In 1971-73, he managed economic and marketing studies of large international enrichment plants and when EURODIF was established in 1973 he was named Director responsible for the commercial development of the Company.

In 1983, he was appointed Vice President of COGEMA in charge of Marketing and Sales and, from 1995 to 1998, he took in charge the Corporate Strategy and International Development Division.

KEVMUTE SPEAKER BIDGflAPHJES 120 Ondrej STUDENEC

Mr.Ondrej Studenec is a Director General responsible for Energy, Mining and Metallurgy in the Ministry of Economy of the Slovak Republic. He won a tender for this position in 1999. He is 37 years old. In 1986 he graduated from the Slovak Technical University in Bratislava, specialisation Thermal and Nuclear Power Plants. He has been working for the Ministry of Economy since 1991, he was involved in Energy Charter Treaty ratification in 1994, preparing of Energy Law in 1998 and EU screening of the Slovak energy. He was leader of the Slovak delegation in the Working Groupe for the realistic date for decommissioning of the Nuclear Power Plant VI in Jaslovsk6 Bohunice in 1999. He was a team leader for preparation of a new Energy Policy for the Slovak Republic in 1999. Currently he is responsible for analysing the completion of the next Nuclear Power Plant in Mochovce. He is married and has three children.

301 KEYNOTE SPEAKER BIOGRAPHIES TABLE OF CONTENTS

Keynote Speaker Biographies

Agneta Rising, ENS, Sweden Marc Deffrennes, European Commission, Belgium Andrei Gagarinski, Kurchatov Institute, Russia John Graham, ETCetera, United States Andrew C. Kadak, American Nuclear Society, United States Chang Sun Kang, Seoul National University, Korea Ann MacLachlan, Nucleonics Week, Europe Alexander Mikhalevich, National Academy of Science, Belarus Victor V. Orlov, Russian Academy of Natural Sciences, Russia James M. Reed, IAEA, Austria Jean-Pierre Rougeau, Foratom, Belgium Ondrej Studenec, Ministry of Economy of the Slovak Republic

ORAL PRESENTATIONS

MONDAY, APRIL 10, 2000

YG Session A NEW MILLENNIUM, A NEW DESIRE: THE NUCLEAR DREAM OF THE YOUNG GENERATION

THE YOUNG GENERATION - GUARANTORS FOR THE FUTURE OF THE NUCLEAR INDUSTRY K Broy, Siemens AG/Power Generation Group (KWV), Erlangen, Germany 35 GENERATION "NEXT" AND NUCLEAR POWER A.A. Sergeev, Research and Development Institute of Power Engineering, Moscow, Russia 36 THE ENERGY MIX FOR THE NEXT GENERATION: WITH OR WITHOUT NUCLEAR? M.A. N'Diaye, COGEMA, Vdliy;, France 37 THE HUNGARIAN YOUTH'S KNOWLEDGE AND ATTITUDE IN THE NUCLEAR FIELD G. Petb'fi, G. L&grddl, Technical University of Budapest, Institute of Nuclear Techniques, Hungary .". 3s TftBLt BF mmu ! 293 AIMING AT THE REBIRTH OF THE NUCLEAR GENERATION M- M. Uematsu, The Japan Atomic Power Company, Corporate Planning Department, Tokyo, Japan ....39

Nuclear Education and Transfer of Know-How _____

CAN NUCLEAR SCIENCE OFFER A PROMISING CAREER? WILL TOMORROW'S SCIENTISTS HAVE THE KNOWLEDGE NEEDED FOR THE FUTURE?

Keynote Address: Charged to create the future (summary not included) A. Rising, ENS Keynote Address: Nuclear Education and International Nuclear University C. S. Kang, Seoul National University, Korea

THE DECLINE IN EDUCATIONAL OPPORTUNITIES FOR YOUNG PROFESSIONALS IN THE FIELD OF NUCLEAR SCIENCES U. Meyer, Nuclear Power Plant Leibstadt, Switzerland[ 44 THE ROMANIAN EDUCATIONAL SYSTEM IN NUCLEAR ENGINEERING FIELD - EXPERIENCE AND NEW APPROACHES O. Dragusin, A. Burghelea, "POLITEHNICA" University, Romania 45 THE SIEMENS GRADUATE PROGRAM /. Sch&ffler, Siemens AG/Power Generation Group (KWU), Erlangen, Germany •. 46

Keynote Address: Information is not Forever J. Graham, ETCetera

CERNAVODA NPP UNIT 1 - A PLANT OF SEVERAL GENERATIONS /. Rotaru, M. Metes, «Nuclearelectrica» National Company, Romania; M, Anghelescu, Center of Technology and Engineering for Nuclear Projects, \ 48 METHODS AND PROCEDURES OF SUCCESSION OF GENERATIONS A. Homann, R. Bendzko, Colenco Power Engineering AG; Baden; Switzerland 49 THE NECESSITY OF GENERATION-TO-GENERATION KNOWLEDGE TRANSFER V. Gupalo, VNIPIpromtechnologii, Moscow, Russia SO

Nuclear Technology I _____ WHAT DOES THE FUTURE HOLD?

Keynote Address: What Will it Take to Rejuvinate Nuclear Energy? A. Kadak, American Nuclear Society Keynote Address: Global Trends in Advanced Reactor Developments and the Role of the IAEA J. Kupitz and J. Cleveland, International Atomic Energy Agency Keynote Address: Nuclear Reactors and Technology in the Next Stage V. Orlov, Russian Academy of Natural Sciences, Russia

THE EPR - TECHNOLOGY FOR THE 3RD MILLENNIUM O. Bernstrauch, Siemens AG /Power Generation Group (KWU), Erlangen, Germany 56 MYRRHA : A MULTIPURPOSE ACCELERATOR DRIVEN SYSTEM FOR RESEARCH & DEVELOPMENT K. Van Tichelen, E. Malambu, Ph. Benoit, P. Kupschus, H. Ait Abderrahim, SCK/CEN, Belgium 57 2S41 TABLE OF CONTENTS PRELIMINARY DESIGN OF A GAS-COOLED ACCELERATOR DRIVEN SYSTEM DEMONSTRATOR B. Giraud, Framatome; Y. Poitevin, CEA/Saclay; G.Ritter, CEA/Cadaracke, France .....; 58 THE OECD HALDEN REACTOR PROJECT - INTERNATIONAL RESEARCH ON SAFETY AND RELIABILITY OF NUCLEAR POWER GENERATION 1. Moen, OECD Halden Reactor Project, Norway 59 BN-800 - HISTORY AND PERSPECTIVE I. Krivitski, Institute for Physics and Power Engineering, Russia , 60

TUESDAY, APRIL 11, 2000

Political Aspects IS NUCLEAR POLITICAL! Y CORRECT IN AN INTERNATIONAL ENVIRONMENT?

Keynote Address: Present and future role of nuclear power in the energy mix from the point of safety, competitiveness and climate protection O. Studenec, Ministry of Economy, Slovakia {summary not included) Keynote Address: EUNuclear Policy in the Wider European Context: Future Perspectives M. Defirenes, European Commission, Belgium (summary not included) Keynote Address: Non-Proliferation and Nuclear Disarmament M. Shea, IAEA, Austria

THE PLUTONIUM CHALLENGE FOR THE FUTURE L. Gray, Lawrence Livermore National Laboratory, USA , . 66 THE CONTRIBUTION OF CIVILIAN INDUSTRY TO MILITARY PU DISPOSITION J.A. de Montalembert, COGEMA, France .'. 67 RADIOECOLOGICAL DANGER OF FAILURES WITH NUCLEAR WEAPONS AND LIQUIDATION OF THEIR CONSEQUENCES ft Tetchko, VNIIEF, Russia 68 SIGMA: THE NOVEL APPROACH OF A NEW NON-PROLIFERATING URANIUM ENRICHMENT TECHNOLOGY M. Rivarola, P. Florido, D. Brasnarof, E. Bergallo, Comisidn Nacional de Energia Atdmica, Argentina 69

Nuclear Technology II WHAT DOES THE FUTURE HOLD?

Keynote Address: Nuclear Science & Technology: Applications for the Welfare of Mankind A. K. Padhy, IAEA, Austria

MEDICAL APPLICATIONS IN A NUCLEAR RESEARCH CENTRE E Vanhavere, G. Eggermont, Belgian Nuclear Research Centre, Belgium 74 TABLE Df CDR1ENIS j 295 OVERVIEW OF SOME PROJECTS OF SNPS FOR GLOBAL SPACE COMMUNICATION E. Ivanov, V. Ghitaykin, V Ionkin, A. Dubinin, A. Pyshko, Institute for Physics and Power Engineering, Russia 75 IODOBENZAMIDES: A POTENTIAL IMAGE AGENT FOR MALIGNANT MELANOMA DETECTION I.S. Luppi Berlanga, M. G. ArgUelles, E.A. Torres, Centro Atdmico Ezeiza - Comisidn Nacional de Energia Atdmica, Buenos Aires, Argentina. .. 76 THE ROLE OF COMPUTER SIMULATION IN NUCLEAR TECHNOLOGY DEVELOPMENT M. Yu. Tikhonchev, G.A. Shimansky, E.E. Lebedeva, V.V. Lichadeev, D.K. Ryazanov, A.I. Tellin, State Scientific Centre of Russia «Research Institute of Atomic Reactors», Russia 77 Environment & Safety WHA TMVST BE DONE TO ADDRESS THE SA FETY AND ENVIRONMENTAL CONCERNS FOR THE FUTURE?

Keynote Address: Bohunice V-l Safety Upgrading Process (summary not included) M. Lipar, Nuclear Regulatory Authority, Slovakia

OPTIMISATION OF TEST AND MAINTENANCE BASED ON PROBABILISTIC METHODS M. Cepin, «Josef Stefan» Institute, Ljubljana, Slovenia 81 HAEA NEPO TOOLS USED IN NUCLEAR EMERGENCY RESPONSE K. Horvath, Hungarian Atomic Energy Authority, Hungary 82 ESTIMATION OF RADIOLOGICAL CONSEQUENSES FROM ACCIDENTAL IODINE RELEASES AT NUCLEAR POWER PLANTS V. Drozdovitch, Institute of Power Engineering Problems, National Academy of Sciences ofBeiarus, Belarus 83 EMERGENCY RESPONSE TECHNICAL CENTRE OF THE IPSN R. Dallendre, IPSN, France 84 ACCELERATOR-DRIVEN TRANSMUTATION: A HIGH-TECH SOLUTION TO SOME NUCLEAR WASTE PROBLEMS A, Hechanova, Harry Reid Center for Environmental Studies, University of Nevada, USA 85 KARACHAY LAKE IS THE STORAGE OF THE RADIOACTIVE WASTES UNDER OPEN SKY A. Merkushkin, Ozyorsk Technological Institute of Moscow Physical Engineering Institute, Russia 86 "KOZLODUY" NPP GEOLOGICAL ENVIRONMENT AS A BARRIER AGAINST RADIONUCLIDE MIGRATION D. Antonov, Geological Institute of the Bulgarian Academy of Sciences, Bulgaria 87 MODERNIZATION AND SAFETY IMPROVEMENT PROJECT OF THE NPP V-2 JASLOVSKE BOHUNICE V. Mtchal, B. LoSonsky, J. Magdolen, Nuclear Power Plants Research Institute Trnava, Inc., Slovak Republic 88 Communication & Public Perception I HOW CAN WE IMPROVE PUBLIC COMMUNICATION EFFORTS ABOUT THE BENEFITS OF NUCLEAR POWER?

Keynote A ddress: Sustainable Nuclear Development and Public Confidence A. Gagarinski, Kurchatov Institute, Russia 2961TABIE OF CONTENTS Keynote Address: Public Communication about Nuclear Energy: Between Ideology and Pragmatism - It is Time for a New Dialog (summary not included) S. Feldhaus, Siemens, Germany

PUBLIC OPINION SURVEY "NUCLEAR ENERGY - THE PRESENT AND THE FUTURE" Renata Matani andJoslp Lebegner, Hrvatska elektroprivreda; Ines-Ana Jurkovi, Faculty of Electrical Engineering and Computing; MatJaPrah ENTECO; Croatia 92 NUCLEAR POWER AND PUBLIC OPINION /. A. Kazanikov, Moscow State Technical University, Moscow, Russia; and S.A.Klykov, Institute of Nuclear Power Engineering, Obninsk, Russia 93 NUCLEAR LITERACY IN LIGHT OF RADON /. Ldzdr, I. Cziegler, RAD Lauder Laboratory, Budapest, Hungary , 94 AREN HAS GOING INTO ACTION FOR NUCLEAR PROGRAM IN ROMANIA T. Chirica, Societatea Nationala " Nuclearelectrica" SA, Romania; T. Mauna, Romanian "NUCLEAR ENERGY" Association-AREN, 95 APPLICATION OF MEMETIC ENGINEERING TO THE STRUGGLE FOR PUBLIC ACCEPTANCE /. Whiilock, Canadian Nuclear Society, Canada 96

Communication & Public Perception II HOW CAN WE IMPROVE PUBLIC COMMUNICATION EFFORTS ABOUT THE BENEFITS OF NUCLEAR POWER?

Keynote Address: Media-nuclear industry communication: mission impossible? A. MacLachlan, Nucleonics Week, Europe (summary not included)

Keynote Address: Public Acceptance of Nuclear Power after Chernobyl A. Mikhalevitch, National Academy of Science, Belarus

COGEMA GIVES ITS COMMUNICATION A NEW IMPETUS: TRANSPARENCY TO CONDUCT A NEW DIALOG Katherine Graffln, COGEMA, France , 100 THE ROLE OF INFORMING SOCIETY AND INTERNATIONAL COOPERATION IN IMPROVING THE NUCLEAR "IMAGE" Y. Kazakevich, Pollna Biryukova, Ozyorsk Technological Institute of Moscow Physical Engineering Institute, Russia 101 SYMPATHY FOR THE DEVIL ©: COMMUNICATION INSIDE OUT G. Meskens, SCK/CEN-ENS-YGN, Belgium 102 PUBLIC ACCEPTANCE AND PUBLIC UNDERSTANDING: NEW APPROACHES D. Dobak, Bohunice NPPs, Slovak Republic , J03 THE ROLE OF THE YOUTH OF OZYORSK IN CREATION OF POSITIVE IMAGE OF NUCLEAR ENERGY IN THE CHELYABINSK REGION T. Kostareva, A. Teslov, Production Association "MAYAK", Ozyorsk, Russia „... 104

TABLE DF CONTENTS I 297 WEDNESDAY, APRIL 12, 2000

Nuclear Programs & Technical Cooperation

Keynote Address: IAEA's Technical Co-operation Programme anditsRole in Assisting Member States in the Safe Utilisation of Nuclear Power J. Reed, IAEA, Austria

INDIA'S POWER PROGRAMS AND ITS CONCERN OVER ENVIRONMENTAL SAFETY G.E. Prasad, J. Mittra,M. S. R. Sarma, Indian Nuclear Society, Mumbai, India HO GLOBALIZATION: PROSPECTS FOR FUTURE INTERNATIONAL COOPERATION /. P. Dinu, CNEPROD, Cernavoda, Romania nl EXPEDIENCY OF NUCLEAR POWER USE IN RUSSIA N. Babaev, Minatom of Russia, Russian Federation 112 NUCLEAR ENERGY IN ARMENIA 5". Gevorgyan, Vahe Kharazyan, Armenian State Engineering University, Armenia 113 NUCLEAR POWER: BENEFITS FOR THE FUTURE G. Vultur^C. Vultur, CNE-PROD CERNAVODA, Romania, u4 NUCLEAR POWER IN LONG TERM ENERGY STRATEGIES IN MACEDONIA A. Volkanovski, Research Center for Energy and Informatics, Macedonia I US

Economics WHAT ECONOMIC TARGETS MUST NUCLEAR MEET FOR THE FUTURE?

Keynote Address: Title to be determined (summary not included) M. Kosovan, Slovak Electric, Slovakia

DEREGULATION OF THE ELECTRIC UTILITY INDUSTRY - IMPLICATIONS FOR NUCLEAR POWER A. Fern, Utility.com, USA • 119 THE URANIUM INDUSTRY : LONG TERM PLANNING FOR SHORT TERM COMPETITION X. Voitero, COGEMA, France 120 FUEL COMPONENT OF ELECTRICITY GENERATION COST FOR THE BN-800 REACTOR WITH MOX FUEL AND URANIUM OXIDE FUEL, INCREASED FUEL BURNUP, AND REMOVAL OF RADIAL BREEDING BLANKET A. Raskach, Institute of Physics and Power Engineering (IPPE), Russia 121 EFFICIENCY IMPROVEMENT OF NUCLEAR POWER PLANT OPERATION: THE SIGNIFICANT ROLE OF ADVANCED NUCLEAR FUEL TECHNOLOGIES A. Van de Velde, F.Burtak, Siemens AG/Power Generation Group (KWU), Erlangen, Germany 122 TECHNICAL AND ECONOMICAL PROBLEMS OF DECOMMISSIONING NUCLEAR POWER PLANTS (NPP) M. Vaneev, MPEI, Russia 12S ZSB / TflBlE OF CONTElfrS Fuel Cycle Challenges HOW CAN THE PROCESS BE IMPROVED?

Keynote Address: Nuclear Fuel Cycle Systems: The Ever-Progressing Part of Nuclear Technology J. P. Rougeau, Foratom, France (summary not included)

THE EVOLUTION OF CANDUO FUEL CYCLES AND THEIR POTENTIAL CONTRIBUTION TO WORLD PEACE /. Whitlock, Canadian Nuclear Society, Canada 127 TECHNOLOGY OPTIONS FOR FUTURE RECYCLING T. Kikuchi, Mitsubishi Materials Corporation, Japan 128 SOME CONDITIONS AND PROSPECTS OF GOING TO A CLOSED FUEL CYCLE IN RUSSIA A. Lependin, V. Oussanov, E. Lependina, State Scientific Center Institute of Physics and Power Engineering, Russia 129 MANAGEMENT OF WASTE FROM FRENCH NUCLEAR FUEL CYCLE : WHAT ARE THE KEY ISSUES? V. Londres, SGN, International Nuclear Division;R. Do Quang, P. Fournier, COGEMA;France 130 THE FINAL DISPOSAL FACILITY OF SPENT NUCLEAR FUEL S. Prvdkova, V. Ne&as, Slovak University of Technology, Bratislava, Slovakia 131 THORIUM-FUEL FOR TWENTY FIRST CENTURY S. R. Sarnta, Indian Nuclear Society, Mumbai, India 133 "SCRAPYARD CHALLENGE" A. Hollick, British Nuclear Fuels pic, United Kingdom. 134 PERFORMANCES OF ACTINIDE TRANSMUTATION BASED ACCELERATOR-DRIVEN SYSTEMS (ADS) AT CIEMAT M. Embid, D. Cano, E. Gonzalez, D. Vdlamarln, CIEMAT, Madrid, Spain 136

POSTER SESSIONS

MONDAY, APRIL 10, 2000

YG Session A NEW MILLENNIUM, A NEW DESIRE: THE NUCLEAR DREAM OF THE YOUNG GENERATION

DYSNAI: FESTIVAL OF INTERNATIONAL YOUTH NUCLEAR ASSOCIATION A. Bolgarov, International Youth Nuclear Association, Lithuania 141

TABLE OF CQHTEHTS Nuclear Education and Transfer of Know-How CAN NUCLEAR SCIENCE OFFER A PROMISING CAREER? WILL TOMORROW'S SCIENTISTS HAVE THE KNOWLEDGE NEEDED FOR THE FUTURE?

RADIATION LITERACY FOR FUTURE ECOLOGISTS D. Belous, StPetersburg State Institute of Technology, Russia 145 TRAINING PROGRAM FOR OPERATORS AT THE BN600 POWER UNIT REACTOR DEPARTMENT D. Chichikin, O.A.Potapov^A.EAminev, Beloyarsk NPP, Russia \ 146 COMMUNICATIVE FOREIGN LANGUAGE EDUCATION FOR DEVELOPMENT OF INDIVIDUAL IN DIFFERENT CULTURES T. Solovieva, Obninsk Institute of Nuclear Power Engineering, Russia 147 IAEA ACTIVITIES IN NUCLEAR REACTOR SIMULATION FOR EDUCATIONAL PURPOSES A. Badulescu,R.B. Lyon, International Atomic Energy Agency, Austria 148

Nuclear Technology I WHA T DOES THE FUTURE HOLD?

ESTIMATION OF DOSES OF IONIZING RADIATION FROM PATIENTS TREATED WITH I3II Y. Chaban, A. Roziev, O. Mileshin, A. Klyopov, N. Shishkanov, E. Matusevich, Russia. 151 THERMOHYDRAULIC RELATIONSHIPS FOR ADVANCED WATER COOLED REACTORS AND THE ROLE OF THE IAEA A. Badulescu, IAEA, D. C. Groeneveld, AECL, Austria). 152 GRADIENT AND STEPWISE INFRARED (IR) FIBER LIGHT GUIDES BASED ON SILVER AND THALLIUM HALOGENIDES V. D. Fedorov, S. E. Orlov, A. A. Barmakova, A-RICT, Moscow, Russia 154 LOW-WASTE FABRICATION OF BARIUM, LITHIUM AND CALCIUM FLUORIDES FOR THERMOLUMINESCENSE DETECTORS AND HIGH DENSITY SCINTILLATORS A. A. Barmakova, S. E. Orlov, A-RICT, Moscow, Russia, 155 BN600 REACTIVITY DEFINITION V. Zheltyshev, A. Ivanov, Beloyarsk NPP, Russia 156 LIFETIME EVALUATION OF BOHUNICE NPP COMPONENTS L. Kupca, Bohunice NPP, Slovak Republic 157 CHILEAN EXPERIENCE IN PRODUCTION OF 18F-FDG FROM I8F IN A REACTOR M. Chandla, N.Godoy, X.Errazu, Hernandez, M.Figols, G.Firnau, F.Troncoso, Chilean Nuclear Energy Commission, Santiago, Chile \ 159 CURRENT PROBLEMS OF VVER-1000 REACTOR CORE OPERATION IN UKRAINE A. Bykov, National Atomic Powergenerating Company "Energoatom", Ukraine 160 PIN WISE CALCULATION BY MONTE CARLO CODES /. Popova, Institute for Nuclear Research and Nuclear Energy, Bulgaria; J.Bucholz, Oak Ridge National Laboratory, USA 162 NUMERICAL EFFECTS IN THE NEUTRON FLUX CALCULATIONS INTO WWER-TYPE REACTOR VESSELS USING THE MONTE CARLO METHOD E Garcia Tip, CM. Alvarez Cardona, M. Rodriguez Gual, S. Hernandez Valle, Nuclear Technology Center, Institute of Nuclear Sciences and Technology, Havana, Cuba.. 163 3001 TABLE OF CfiHTEKTS SPECTRAL NODAL METHOD FOR SOLVING THE NEUTRON DIFFUSION EQUATION D. Sanchez and C. R. Garcia, Institute Superior de Qenciasy Tecnologia Nuclear, Cuba; R. C. de Barros, Institute de Energia Nuclear, R(o de Janeiro, Brasil; andD. E. Milian, Centro de Tecnologia Nuclear 164 CASCADE ENERGY AMPLIFIER A. P. Barzilov, A. V. Gulevich, O.E Kukharchuk, State Scientific Center of the Russian Federation, IPPE, Obninsk, Russia 165 DIRECT UTILIZATION OF INFORMATION FROM NUCLEAR DATA FILES IN MONTE CARLO SIMULATION OF NEUTRON AND PHOTON TRANSPORT P. Androsenko, D. Zholudov, A. Kompaniyets, O. Smirnova, Institute of Nuclear Power Engineering, Obninsk, Russia • •• • 166 PHOTONUCLEAR REACTIONS DATABASE FOR FUNDAMENTAL RESEARCH AND APPLICATIONS S. Vedernlkov, Institute of Nuclear Power Engineering, Obninsk, Russia 167 EXACT AND NUMERICAL SOLUTIONS OF NONLINEAR KINETIC EQUATIONS M.A. Zaboudko, Institute of Nuclear Power Engineering, Obninsk, Russia 168 MEASUREMENT AND CALCULATION OF ACTIVITY IN VANADIUM AND V-TI-CR ALLOYS IRRADIATED IN BR-10 REACTOR Yr.R. Kondrashechkin, G.A.Birzhevoy, A.I.Blokhin, M.I.Zakharova, State Scientific Center of Russian Federation, Institute of Physics and Power Engineering, Obninsk, Russia 169 THE POLARIZATION OF FAST NEUTRONS V.V. Talov, State Scientific Center of Russian Federation, Institute of Physics and Power Engineering, Obninsk, Russia 170 DOSIMETRY METHOD FOR MEASUREMENTS OF ENERGY OUTCOME FROM GADOLINIUM AFTER NEUTRON CAPTURE S.A.,Klykov, Yu.A.Kuratchenko, E.S.Matusevitch, Institute of Nuclear Power Engineering, Obninsk, Russia; S.P.Kaptchigashev, V.LPotetnya, S.E.Oultanenko, Medical Radiological Research Center RAMS, Obninsk, Russia ; 171 CSI(TL) DETECTOR APPLICATION FOR LOW ENERGY MIXED RADIATION FIELD SPECTROSCOPY V.A. Khriachkov, M. V. Dunaev, N.N. Semenova, I. V. Dunaeva, Institute of Physics and Power Engineering, Obninsk, Russia, 172 APPLICATION OF DIGITAL SIGNAL PROCESSING FOR RADIATION SPECTROSCOPY V. A. Khriachkov, I. V. Dunaeva, M. V. Dunaev, N.N. Semenova, Institute of Physics and Power Engineering, Obninsk, Russia, 173 CALCULATIONS OF ACCELERATOR-BASED NEUTRON SOURCES' CHARACTERISTICS R.G. Tertytchnyi and V.S. Shorin, State Scientific Center of Russian Federation, Institute of Physics and Power Engineering, Obninsk, Russia 174 RBMK-1500 REACTOR PUMP TRIP EVENTS MODELLED USING THERMAL- HYDRAULIC CODE CATHARE2 VI.3L A. Jasiulevicius, ffelsinki University of Technology, Finland 17$ OPTIMIZATION OF IONIZATION CHAMBERS ARRANGEMENT IN VVER-1000 TYPE REACTOR FACILITIES S. Philippoff and A. Soidatov, MEPhI, YDRNS, Russia. .176 THE EPR LAYOUT DESIGN U. Mast, Siemens AG/Power Generation Group (KfVU), Erlangen, Germany 177 STUDY OF CORROSION PRODUCTS FROM SECONDARY CIRCUIT NUCLEAR POWER PLANT V-I BOHUNICE A. Zeman, V. Slugeii, andJ. Lipka, Slovak Technical University Bratislava, Slovak Republic 178 TABLE OF CDHTENfS I 301 CVD DIAMOND DETECTORS OF IONIZING RADIATION A. Perdochovd, V. NeSas, Slovak University of Technology, Slovak Republic, 179 NEUTRON DIFFRACTION MEASUREMENTS OF RESIDUAL STRESSES IN NPP CONSTRUCTION MATERIALS R. Hinca, Slovak University of Technology, Slovak Republic; and G. Bokuchava, Frank Laboratory of Neutron Physics, JINR Dubna, Russia* 180 INCREASING OF UNIT CAPACITY WHILE CONTINUING CAMPAIGN A. Slapakov, A. Mironova and V. Ivanov, Branch of StPetersburg State Technical University,. Sosnovy Bor, Russia • 181 THE ROLE OF CHANNEL TYPE REACTORS IN RUSSIAN NUCLEAR ENGINEERING A. Karev, A. Eperin, N. Eremtn, Branch of Saint Petersburg State Technical University, Sosnovy Bor, Russia 182 EXPLOITATION QUESTIONS REGARDING CHANNEL TYPE REACTORS: WATER GRAPHITE CHANNEL REACTOR -1000 (OPERATION, RECONSTRUCTION, ADVANTAGES, AND DISADVANTAGES) D.A. Chichindaev, Saint Petersburg State Technical University, Russia \ 183 EXAMINATION OF CONTROL ROD EJECTION IN WWER-440 TYPE REACTORS USING THE CODE DYN3D G. Petofi and A. Asi6di, Institute of Nuclear Techniques, Technical University of Budapest, Hungary... 184 MATHEMATICAL SIMULATION OF GAS DYNAMICS ON EULERIAN MESHES: A REALIZATION OF YOUNG'S METHOD FOR INTERFACE TRACKING IN A THREE- DIMENSIONAL CASE ON ORTHOGONAL MESHES R. Veselov, A.N. Bykov, B.L. Voronin, Russian federal nuclear center- VNIIEF, Russia 185 CONTAINMENT LEAK-TIGHTNESS ENHANCEMENT AT VVER 440 NPPS M Prandorfy, VVEZ, Levice, Slovakia 186 THE GRADUAL DEVELOPMENT STEPS OF THE EXTERNAL COUPLED RELAP5- DYN3D CODE C. Strmensky, VUJE Trnava, Slovakia 187 THE ANALYSIS OF PHYSICAL PERFORMANCES OF THE REACTOR RBMK-1500 V. Ledzinskas, Ignalina NPP, The Trainee of the Technological University, Lithuania 188

TUESDAY, APRIL 11, 2000

Nuclear Technology II W1IA T DOES THE FUTURE HOLD?

PERSPECTIVE PRODUCTION OF RADIOISOTOPES AND RADIOPHARMACEUTICALS IN DIVISIONS OF IPPE G. O. Terentyev, Institute of Physics and Power Engineering, Russia 193 UTILIZING HORIZONTAL REACTOR CHANNELS FOR NEUTRON THERAPY E. Yu. Stankovsky and Yu.A. Kurachenko, Institute of Nuclear Power Engineering, Russia. 194 188RE-MICROSPHERES OF ALBUMINE - THE POTENTIAL PREPERATION FOR RADIOTHERAPY D.N. Dyomin, V.M.Petriev, Medical Radiological Research Centre RAMS, Obninsk, Russia 195 RADIOIODTHERAPY: DOSIMETRY PLANNING A.Apyan, A.Roziev, O.Mileshin, A.Klyopov, NShishkanov, EMatusevlch, Russia 196 3021 TABLE OF COHMS Political Aspects IS,NUCLEAR POLITICALL Y CORRECT IN AN INTERNATIONAL ENVIRONMENT?

NON-PROLIFERATION ISSUES WITH WEAPONS-USABLE PLUTONIUM L W. Gray, Lawrence Livermore National Laboratory, USA 199 RUSSIAN FEDERAL NUCLEAR CENTER VNIIEF - POSSIBILITIES OF INTERNATIONAL COOPERATION V.M. Shaburov, JR. V.\Mozharov, Russian federal nuclear center- VNIIEF, Russia , 201 PREVENTION OF ACCIDENT CONSEQUENCES IN PLUTONIUM STORAGE O.L Tetchko, VNIIEF,Russia, 202

Environment & Safety WHAT MUST BE DONE TO ADDRESS THE SAFETY AND ENVIRONMENTAL CONCERNS FOR THE FUTURE?

THE CHARACTERISTICS OF INTAKING CS-137 AND SR-90 WITH FOOD BY OZYORSK'S INHABITANTS M. Dronova, Ozyorsk Technological Institute of Moscow Phistcal Engineering Institute, Russia 205 PROBABILISTIC SAFETY ASSESSMENT AS A STANDPOINT FOR DECISION-MAKING M. Cepln, "Jozef Stefan " Institute, Slovenia '. 206 SAFETY PREDICTION FOR BASIC COMPONENTS OF SAFETY-CRITICAL SOFTWARE BASED ON STATIC TESTING Han SeongSon, Poong Hyun Seong, Department of Nuclear Engineering, KAIST, 207 INFLUENCE OF HEAT LOSSES AND ACCUMULATED HEAT UPON THE ACCIDENT PROCESS EVOLUTION /. ti. Gashenko, Electrogorsk Research and Engineering Centre (EREC) on Nuclear Power Plants Safety, Moscow Region, Russia 208 POST-TEST CALCULATION OF THE LOOP SEAL 1% LEAK ON THE PSB-VVER TEST FACILITY WITH THE HELP OF THE CODE RELAP5/MOD3.2 A. V. Kapustln, Electrogorsk Research and Engineering Center, Electrogorsk, Moscow region, Russia .. 209 THERMAL-HYDRAULIC CODES VALIDATION FOR SAFETY ANALYSIS OF NPPS WITH RBMK N.A. Brus, O.E.\Ioussoupov, Electrogorsk Research and Engineering Center for NPP Safety, Russia. ...210 PROPAGATION OF A WAVE THERMAL DETONATION IN VIEW OF MICROINTERACTION MODEL O.I.Melikhov, V.LMelikhov, A. V.Sokolin, Electrogorsk Research & Engineering Center, Electrogorsk, Moscow region, Russia . 211 DETAILED SPECTRA DATA FOR THE INTERNATIONAL HANDBOOK OF EVALUATED CRITICALITY SAFETY BENCHMARK EXPERIMENTS Y. Rozhikhin, Institute of Physics and Power Engineering (IPPE), Obninsk, Russia 212 VERIFICATION OF THE CODES RELAP5MOD3.2 AND CATHARE 2 V1.4 BY MODELING THE LONG TERM COOLING IN THE VVER-640 POWER PLANT AFTER A LARGE BREAK LOCA ON THE PACTEL FACILITY A. Alexeev, Obninsk Institute of Nuclear Power Engineering, Russia; J. Bdnati, E. Virtanen, Lappeenranta University of Technology 213 CREATION OF CADASTRE OF GROUND CONCENTRATION OF CHEMICAL POLLUTING SUBSTANCES IN THE CITY OF OBNINSK TAKING INTO ACCOUNT AERODYNAMIC SHADOWS OF BUILDINGS O. Malik, Institute of Nuclear Power Engineering, Obninsk, Russia 214 TABLE OF CONTENTS I 303 DOSE LOAD AT ORAL ENTRANCE AND INJECTION OF RADIONUCLIDES IN HUMAN ORGANISM £ Korkoshko, Institute of Physics and Power Engineering, Obninsk, Russia 215 DEVELOPMENT OF NEW EXTRACTANT UTILIZED BY COORDINATION PROPERTIES OF DIGLYCOL AMIDE (DGA) TO TRIVALENT CURIUM AND LANTHANIDES T. Yaita, \ H.Narita, S.Tachimori, Japan Atomic Energy Research Institute, Japan; N.M. Edelstein, Lawrence Berkeley National Laboratory 216 A NEW SYSTEM FOR THE MEASUREMENT OF THE SPACE RADIATION T. Pdzmdndi, I. Apathy, S. Deme, KFKI Atomic Energy Research Institute; R. Beaujean, Kiel University, Hungary ••••• -277 ANALYSIS OF EFFECTS FILLING OF STEAM LINES WITH WATER AT IGNALINA NPP R. Urbonas, A. Kaliatka, Laboratory of Nuclear Installation Safety, Lithuania Energy Institute, Lithuania 218 THERMAL-HYDRAULIC ANALYSIS OF IGNALINA NPP COMPARTMENTS RESPONSE TO GROUP DISTRIBUTION HEADER RUPTURE USING RALOC4 CODE E. Urbonavicius, Lithuania Energy Institute, Lithuania 219 ACTIVATION OF BWR COOLANT IN TRANSIENTS S. Grachev, MPEI, Russia , 220 SAFETY CULTURE IN FUTURE E. Berishev, MPEI, Russia 221 MINERAL-LIKE URANYLSILICATES OF ONE-VALENCE METALS OF STRUCTURE

V.E. Kortikov and N.G. Chemorukov, Nlzhny Novgorod State University, Russia . , 222 SYNTHESIS AND INVESTIGATION OF URANYLPHOSPHATES, URANYLARSENANES AND URANYLVANADATES OF LANTHANIDES N.G. Chernorukov, A. V. Knyazev, S,V. Barch, O.V. Feoktlstova, E. V. Suleymanov, Nlzhny Novgorod State University, Russia 223 SOLID SOLUTIONS OF SOME URANYLPHOSPHATES-URANYLARSENATES. STRUCTURE AND THERMODYNAMICS N.G, Chernorukov, E.V. Suleymanov, S.A. Ermonin. Nlzhny Novgorod State University, Russia .224 MODELLING THE IMPACT OF CERNAVODA NPP ON THE ENVIRONMENT Angela -Ioana Constantln, Cernavoda NPP - CNE - PROD, Romania j ,. 225 ALARA AND RADIATION PROTECTION OPTIMIZATION STATUS OF CERNAVODA NPP D. I. Han, Cernavoda NPP, Romania 226 INCREASE NUCLEAR SAFETY OF WWER-440 T. Nochev andS. Sablnov, NPPKozlodul, Bulgarian 227 NEW INTERMEDIATE HEAT EXCHANGER FOR LOOP TYPE LMFBR K. Miyazaki, N. Uda, J.Toyooka andH. Horilke, Osaka University, Graduate School of Engineering, Osaka, Japan, , „. .,. , 228 THE TECHA RIVER: 50 YEARS OF RADIATION PROBLEMS D. Evlanov, Ozyorsk Technological Institute of Moscow Physical Engineering Institute, Russia 229 DYNAMIC OF AGE STRUCTURE AND THE NUMBER OF POPULATION IN OZYORSK AND AFFECTING FACTORS O. Panchenko andM. Rtischeva, Ozyorsk Technological Institute of Moscow Physical Engineering Institute, Russia 230 ANALYSIS OF MODES OF HEAT TRANSFER BY FUEL DAMAGED BY BREAKING A FUNCTIONING SIDEBAR OF A WATER-COOLED REACTOR A.V. Golubinsky, Russian federal nuclear center - VNIIEF, Russia 231 3D4j TABLE OF CONTENTS DYNAMIC ANALYSIS AND SAFETY MEASURES OF VVER-440 NPP PAKS 1 TO 4 DUE TO SEISMIC AND OPERATIONAL LOADING CONDITIONS T. Burjan, PAKS Nuclear Power Plant LTD, PAKS, Hungary; B. Schwemin, U. Erben,A. Halbritter, W. Trubnikov, Siemens AG, Power Generation (KWU), Offenbach, Germany. 232 CONFINEMENT STRENGTH AND INTEGRITY, NUCLEAR POWER PLANT V-l BOHUNICE SLOVAKIA M Berge, Siemens AG/Power Generation Group (KWU), Offenbach, Germany; J. Janik, EBO, Bohunice, Slovakia 233 PROVISIONS OF COMUNICATION BETWEEN NPP OPERATIONAL PERSONNEL: DRIFTS OF DEVELOPMENT S.A. Piskarev, V.R. Aksenov, Branch of St. Petersburg State Technical University, Sosnovy Bor, Russia 234 A SIMPLE EVALUATION OF CONTAINMENT INTEGRITY AGAINST EX-VESSEL STEAM EXPLOSION ON ADVANCED PWR K. Murayama, The Kansai Electric Power Co., Inc., Japan 235 AN ANALYSIS OF THE INTENT OF ENVIRONMENTAL STANDARDS IN THE UNITED STATES THAT APPLY TO WASTE DISPOSED AT THE NEVADA TEST SITE Anthony E. Hechanova, Brett T. Mattingly, Harry Reid Center for Environmental Studies, University of Nevada, USA 236 INVESTIGATION OF ELECTRO-KINETIC METHODS FOR SOIL DECONTAMINATION A.N. Shabanova, Urals State Technological University-UP[, Ekaterinburg, Russia 257 STATIONARY EQUIPMENT FOR DETECTING RADIOACTIVE MATERIAL ON PASSING PEDESTRIANS DEVELOPED AND MADE IN RFYC - VNIIEF D.S. Kapustin, CYRYRFYC- VNIIEF, Russian Federal Nuclear Center, Russia 238 THERMO-HYDRAULIC ANALYSES OF BOHUNICE NPP VVER-440 T.Kliment, VUJE Trnava Inc., Slovakia . : 239 RESEARCH OF CHEMICAL STRUCTURE OF ATMOSPHERIC PRECIPITATION D.A. Korenyac, T.K. Korenyac, Russia. 240 INTRODUCTION OF REGULATORY AND LICENSING PROCEDURES OF SOME OECD COUNTRIES P. Zagyvai, S. Czifrus, 0. Benedekfi, P. Ormai, G. Dankd, Technical University of Budapest, Institute of Nuclear Techniques, Hungary 241 DEVELOPMENT OF VENDOR INDEPENDENT SAFETY ANALYSIS CAPABILITY FOR NUCLEAR POWER PLANTS IN TAIWAN J. Tang, Institute of Nuclear Energy Research, Taiwan, Republic of China 242

Communication & Public Perception

HOW CAN WEIMPROVEPUBLIC COMMUNICATION EFFORTS ABOUT THE BENEFITS OF NUCLEAR POWER?

THE INFLUENCE OF NUCLEAR RISK SENSE TO ATOMIC INDUSTRY'S PUBLIC RELATION P. Biryukova, Y. Kazakevich, Ozyorsk Technological Institute of Moscow Physical Engineering Institute, Russia 245 DEMONSTRATION-INFORMATIVE CENTER BASED ON RESEARCH REACTOR IR-50 IN HEAT REGIME P. Krupenina, Research and Development Institute of Power Engineering (ENTEK), Moscow, Russia 246 PUBLIC OPINION SHIFTS TO THE FAVOUR OF NUCLEAR ENERGY DUE TO STEAM GENERATOR TRANSPORT /. Lengar, T. Nemec, Josef Stefan Institute, Slovenia .....247 TABIE OF CONTENTS 13115 ANALYSIS OF ACTIVITY OF INFORMATION INQUIRED GROUP ON RADIOECOLOGY AND PUBLIC COMMUNICATION IN OZYORSK(THE TOWN OF NUCLEAR INDUSTRY) £". Govyrina, Ozyorsk Technological Institute of Moscow Physical Engineering Institute, Russia 248 A RISK COMMUNICATION CASE STUDY: THE NEVADA RISK ASSESSMENT/ MANAGEMENT PROGRAM A.E. Hechanova, Harry Reid Center for Environmental Studies, University of Nevada, USA 249 PUBLIC ACCEPTANCE/PUBLIC UNDERSTANDING: NEW APPROACHES S. Ryabtsun, Russia 250 PUBLIC INFORMATION — NORTH WEST REGION OF RUSSIAN FEDERATION A. Saiapina, State Educational Center, St. Petersburg, Russia 251

WEDNESDAY, APRIL 12, 2000

Economics WHAT ECONOMIC TARGETS MUST NUCLEAR MEET FOR THE FUTURE?

THE STRATEGY OF IMPLEMENTATION OF AN INFORMATIONAL MANAGEMENT SYSTEM FOR A POWER STATION C. Isar, CNE-PROD CERNAVODA, Romania 257

Nuclear Programs & Technical Cooperation

DEVELOPMENT OF THE NUCLEAR ENERGY PRODUCING AS A MAIN FACTOR OF PRESERVING THE ECOLOGICAL BALANCE ON THE EARTH M. Yanusova, Russia • 261 THE KEY FOR COMPETITIVE NUCLEAR POWER, A VIEW FROM TAIWAN /. Lin, Nuclear Safety Department, Taiwan Power Company, Taiwan 262

Fuel Cycle Challenges

HOW CAN THE PROCESS BE IMPROVED?

ASPECTS REGARDING THE FUEL MANAGEMENT FOR PHWR NUCLEAR REACTOR O. Dragusin,A. Bobolea.A. Voicu, "POLITEHNICA" University,Romania 265 CALCULATIONS OF NON-STATIONARY PROCESSES AND SPATIAL DISTURBANCES IN THE BN-600 REACTOR CORE A.M. Tuchkov, A.I.Karpenko, Yu.A.Blinov, Beloyarsk NPP; E.F.Seleznev, VNIIAES, Russia '. 266 APPARENT STABILITY CONSTANTS OF THE COMPLEXES OF Am(III) AND Cm(III) WITH HUMIC ACID AS A FUNCTION OF pH BY THE SCHUBERT METHOD T. Sakuragi,, S. Sawa, S. Sato, and H. Ohashi, Graduate School of Engineering Hokkaido University, Japan; T. Mitsugasira, M. Haraand Y. Suzuki, Institute for Materials Tohoku University, Japan..: 267 AN UNEXPLORED FRONTIER: FLUORINATION OF URANIUM OXIDE BY CF4/O2 R.F. PLASMA K Kim, Y. Seo, S. Jeon, Hanyang University, Seoul, Korea; B. Cho, IAEA, Vienna, Austria 268 3D81 TABLE GF CONTENTS OPERATIONAL BENCHMARK FOR VVER-1000, KOZLODUY NPP UNIT 6 T. Apostolov, B. Petrov, Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences, Bulgaria 269 LEARNING FEATURES OF FISSION FRAGMENT INTERACTIONS WITH POLYMERIC COMPOUNDS S. Kosarev, Institute of Nuclear Power Engineering, Russia'. , 270 CALCULATION RESEARCH OF NEUTRONICS PARAMETERS OF THE VVER REACTOR WITH Th AND 233U DIOXIDE FUEL A.A. Belov, V.M. Decusar, A. G. Kalashnikov, Institute of Physics and Power Engineering, Russia 271 WIMS/ABBN LIBRARY BASED ON THE FOND-2 EVALUATED FILES G. Jerdev and S. Zabrodskaia, Institute of Physics and Power Engineering, Russia 272 RELATIVE ABUNDANCES AND PERIODS OF DELAYED NEUTRONS FROM FISSION OF H9PU BY FAST NEUTRONS. VM.Piksaikin, L.E.Kazakov, S.GJsaev, G.G.Korolev, V.A.Roshenko, M.Z.Tarasko, R.G.Tertytchnyi, Institute of Physics and Power Engineering, Russia 273 TRIGA FUEL ELEMENT BURNUP DETERMINATION BY MEASUREMENT AND CALCULATION T. Zagar, M. Ravnik, A. PerSie, R. Jeraj, Institute Jozef Stefan, Ljubljana, Slovenija..... 274 RADIOACTIVE WASTE TREATMENT IN SLOVAK REPUBLIK P. Dubovsky, NPP V-2, Jaslovske Bohunice, Slovak Republic 275 CURRENT RADIOACTIVE WASTE UTILIZATION AT PA "MAYAK" A. Merkushkin, Ozyorsk Technological Institute of Moscow Physical Engineering Institute, Russia 276 THE METHODS OF CONVERSION OF RADIOACTIVE WASTE: A LOOK AT THE PAST D. Rezchikov, Ozyorsk Technological Institute of Moscow Physical Engineering Institute, Russia 277 MANAGEMENT OF RADIOACTIVE WASTE IN NUCLEAR POWER: HANDLING OF IRRADIATED GRAPHITE FROM WATER-COOLED GRAPHITE REACTORS S.S. Anfimov, Research and Development Institute of Power Engineering, Moscow, Russia 278 CALCULATED INVESTIGATION OF ACTINIDE TRANSMUTATION IN THE BOR-60 REACTOR /. Yu. Zhemkov, O, V.Ishunina, I. V.Yakovleva, State Scientific Center of the Russian Federation Research Institute of Atomic Reactors, Dintitrovgrad, Russia 279 APPLICATION OF SELF-PROPAGATING HIGH-TEMPERATURE SYNTHESIS FOR IMMOBILIZATION OF HARD RADIOACTIVE WASTES IN CERMET MATERIALS I.Yu. Pashkeev, A.V. Senin, EM Ifyin, N.V. Gerasimova, South-Ural State University, Russia 280 SHADOW CORROSION EVALUATION IN THE STUDSVIK R2 REACTOR Charlotta Sanders, Gunnar Lysell, Studsvik Nuclear AB, Sweden 281 SYSTEM STUDY ON PARTITIONING AND TRANSMUTATION OF LONG-LIVED ISOTOPES M. Szieberth, Institute of Nuclear Techniques, Hungary 282 NOVEL TECHNIQUE FOR MANIPULATING MOX FUEL PARTICLES USING RADIATION PRESSURE OF A LASER LIGHT R. Omori, Department of Quantum Engineering and Systems Science, University of Tokyo, Japan 283 THE HIGH ORDER APPROXIMATION OF THREE-DIMENSIONAL NEUTRON DIFFUSION EQUATION BASED ON COMBINATION OF FINITE ELEMENTS AND FINITE DIFFERENCES SCHEMES IN KORAT 3D CODE XV. Shemyakina, O. Zvenigorodskaya, R.M. Shagaliev, Russian federal nuclear center - VNIIEF, Russia 284 STUDY OF MATERIAL STABILITY SURROUNDING WITH LOESS-CLAY-LOAM ROCKS ON AN EXAMPLE OF "OLVIYA" MONUMENT OF UKRAINIAN NORTHERN PRICHERNOMORYA B, Zlobenko, V. Kadoshnikov, L. Demchenko, T. Golovko, State Scientific Center of Environmental Radiogeochemistry, Kyiv, Ukraine; V. Krapivina, Institute of Archaeology of NAS of Ukraine, Kyiv 285 ME OF CONK i 1? NUCLIDES 2000: AN ELECTRONIC CHART OF THE NUCLIDES /. Gafy andJ. Magill, European Commission, Joint Research Centre, Karlsruhe, Germany 286 TAKING BURNUP CREDIT FOR INTERIM STORAGE AND TRANSPORTATION SYSTEM FOR BWR FUELS K. Yoshioka, Toshiba Nuclear Engineering Laboratory, Japan 288 RUSSIAN SYSTEM OF COMPUTERIZED ANALYSIS FOR LICENSING AT ATOMIC INDUSTRY (SCALA) AND ITS VALIDATION ON ICSBEP HANDBOOK DATA AND ON SOME BURNUP CALCULATIONS T. Ivanovo, M. Nikolaev, A. Polyakov, T. Saraeva, A. Tsiboulia, Institute for Physics and Power Engineering, Russian Federation 290 FUTURE PERSPECTIVE OF THORIUM BASED NUCLEAR FUELS AND THORIUM POTENTIAL OF TURKEY Turan UNAK, Yeliz YILDIRIM, Ege University, Turkey 291

3081TABIEBF CONTENTS AUTHOR INDEX

Chaban, Y. 151 Chandîa, M. 159 Abderrahim, H. Ait 57 Chernorukov, N.G 222, 223. 224 Aksenov, V.R 234 Chichikin, D. 146 Alexeev, A 213 Chichindaev, D.A 183 Alvarez Cardona, CM. 163 Chirica, T. 95 Aminev, A.F. 146 Cho, В 268 Androsenko, P. 166 Constantin, Angela-Ioana 225 Anfimov, S.S. 275 Anghelescu, M. 48 Cziegler, I. , 94 Antonov, D. 87 Czifrus, S. 241 Apostolov, T. 269 Apyan, A 196 D Apathy, I. 217 Arguelles, M. G. 76 Dallendre, R 84 Aszódi, A 184 Dankô, G 241 de Barros, R.C. 164 В Decusar, V.M. 271 Demchenko, L 285 Babaev, N. 112 Deme, S. 217 Badulescu, A 148, 152 de Montalembert, J.A 67 Barch, S. V 223 Dinu, LE Ul Barmakova, A. A. : 154, 155 DoQuang, R. 130 Barzilov, A.P. 165 Dobak, D. 103 Beaujean, R 217 Dragusin, O. 45, 265 Belous, D 145 Dronova, M. 205 Belov, A.A 271 Drozdovitch, V. 83 Bendzko, R. 49 Dubinin, A 75 Benedekfi, Ö ; 241 Dubovsky, P. 275 Benoit, Ph. ; 57 Dunaev, M.V. 172, 173 Bergallo, E. 69 DunaevaJ.V. 172, 173 Berge, M. 233 Dyomin, D.N. 195 Berishev, E. 221 Bernstrauch, О. 56 Biryukova, P. 101, 245 E Birzhevoy, G.A 169 Blinov, Yu.A 266 Edelstein, N.M. 216 Blokhin, A.I. 169 Eggermont, G 74 Bobolea, A 265 Embid, M. ¡36 Bokuchava, G. 180 Eperin, A 182 Bolgarov, A 141 Erben, U. 232 Brasnarof, D 69 Eremin, N. ¡82 Broy, Y. 35 Ermonin, S.A 224 Brus, N.A 210 Errazu, X. 159 Bucholz, J. 162 Evlanov, D. 229 Burghelea, A 45 Burjan, T. 232 F Burlak, F. 122 Bykov, A.N. 160, 185 Bánáti, J. 213 Fedorov, V.D ¡54 Feoktistova, O.V. 223 Fern, A //p С Figols, M. 159 Firnau, G. 159 Cano, D 136 Florido, P. 69 Cepin, M. 81, 206 Fournier, P. 130 AUTHOR INDEX I M G К Galy, J. 286 Kadoshnikov, K, 285 Garcia, C.R 164 Kalashnikov, A.G. 271 Garcia, YipF. 163 Kaliatka, A 218 Gashenko, i V. 208 Kaptchigashev, S.P. 171 Gerasimova, N.V. 280 Kapusttn, A. Y. 209 Gevorgyan, S. 113 Kapusttn, D.S. ; 238 Ghitaykin, V. 75 Karev, A 182 Giraud, В 58 Karpenko, A.I. 266 Godoy, N. 159 Kazakevich, Y. 101, 245 Golovko, T. 285 Kazakov, LE. 273 Golubimky, A. V. 231 Kazanikov, LA. '.. 93 González, E 136 Kharazyan Vahe • 113 Govyrina, E 248 Khriachkov, VA... 172, 173 Grachev, S., 220 Kikuchi, T. • 128 Graffin, Katherine 100 Klm, Y. 268 Gray, L. W. 66, 199 Klimmt, T. 239 Groeneveld, D. C. 152 Kfykov,S.A 93, 171 Gulevlch, A. V. 165 Klyopov, A 151, 196 Gupalo, V. 50 Knyazev,A.V 223 Kompaniyets, A 166 Kondrashechkin, Yr.R. 169 H Korenyac, D.A 240 Halbritter, A 232 Korenyac, Т.К. 240 Han, SeongSon 207 Korkoshko, S. 215 Hara, M. 267 Korolev, G.G 273 Hau, D. I, ¡ 226 Kortikov, V.E. 222 Hechanova, A.E. 85, 236, 249 Kosarev, S. ,... 270 Hernández, Valle S. 159, 163 Kostareva, T. 104 Hinka, R 750 Krapivma, V. 285 Hollick, A 134 Krivitski, L '..,. 60 Hamann, A , 49 Krupenina, P. 246 Horiike, H. 228 Horvath, K. 82 Kukharchuk, O.F. 165 Kupca, L 157 Kupschus, P. 57 l Kurachenko, Yu. A 171, 194 llyin, E.N. 280 îonkin, V. 75 ¡oussoupov, O.E. ; 210 L fsaev, S.G 273 Légrádi,G. 38 Isar, С 257 Lázár, I. 94 ishunina, O.V. 279 Lebedeva, E.E. 77 Ivanov, A 156 Lebegner, Josip Jurkovi 92 Ivanov, E. 75 Ledzinskas, V, 188 Ivanov, V. 181 Lengar, 1. 247 Ivanova, T. 290 Lependin, A 129 Lependina, E. 129 Lichadeev, V.V. 77 J Lin, J. 262 Janik, J. 233 lipka.J. 178 Jasiulevicius, A. : 175 Londres, V. 130 Jeon, S. 268 Losonsky, B. 88 Jeraj.R.. 274 Luppi Berlanga, l.S. 76 Jerdev, G 272 Lyon, R.B 148 Jurkovi Ïnes-Ana 92 Lysell, Gunnar 281 3101 AUTHOR IHDÏX M Philippoff, S. 176 Piksaikin, V.M. 273 Míchal, V. 88 Piskarev, S.A 234 Magdolen, J. -, 88 Poitevin, Y. 58 Maguí, J. • 286 Polyakov, A 290 Malambu, E. -, 57 Poong, Hyun Seong 207 Malik, О 214 Popova, I. 162 Mast, U. 177 Matani, Renata 92 Potapov, O.A 146 Mattingly, Brett T. 236 Potetnya, V.l. 171 Matusevitch, E.S. 151, 171, 196 Prah, Matja 92 Mauna, T. 95 Prandorfy, M. 186 Melikhov, O.I. 211 Prasad, G.E. 110 Melikhov, V.l. 211 Prvákova, S. ¡31 Merkushkin, A 86, 276 Pyshko, A 75 Meskens, G 102 Metes, M. 48 Meyer, U. 44 R Mileshin, О. 151, 196 Raskach, A 121 Milian, D. E. 164 Ravnik, M. 274 Mironova, A., 181 Rezchikov, D 277 Mitsugasira, T. 267 Mima, J. 110 Rivarola, M. 69 Miyazaki, К. 228 Rodríguez, GualM. 163 Moen, L 59 Roshenko, VA 273 Mozharov, R. V. 201 Rotaru, 1. 48 Murayama, K. 235 Rozhikhin, Y. 212 Roziev.A 151, 196 Rtischeva, M. 230 N Ryabtsun, S. 250 Ryazanov, D.K. 77 N'Diaye, M.A 37 Narita, H. 216 Neöas, V. 131, 179 Nemec, T. 247 S Nikolaev, M. 290 Sánchez, D 164 Nochev, T. 227 Sabinov, S. 227 Saiapina, A 251 Sakuragi, T. 267 О Sanders, Charlotta 281 Ohashi, H. 267 Saraeva, T. 290 Omori, R. 283 Sarma, M. S. R 110, 133 Orlov, S.E. 154, 155 Sato, S. 267 Ormai, P. 241 Sawa, S. 267 Oulianenko, S.E. 171 Schwemin, B..:. 232 Oussanov, V. 129 Schafften 1. 46 Seleznev, E.F. 266 Semenova, N.N. 172, 173 P Senin, A.V. 280 Pázmándi, T. 217 Seo, Y. 268 Panchenko, 0 230 Sergeev, A.A 36 Pashkeev, I.Yu 280 Shabanova, A.N. 237 Perdochovà, A 179 Shaburov, V.M. 201 Persiè, A 274 Shagaliev, R.M. 284 Petofi, G. 184 Shemyakina, T.V. 284 Petriev, V.M. 195 Shimansky, G.A 77 Petrov, B. 269 Shishkanov, N. 151, 196 Peton, G 38 Shorin, V.S. 174 AUTHOR INDEX 1311 Sîapakov, A 181 Slugeñ, V. . 178 Van de Velde.A. 122 Smirnova, О. 166 Vaneev, M. 123 Sokolin, A. V. 211 Vanhctvere, E 74 Soldatov, A 176 Vedernikov, S. 167 Solovieva, T. : 147 Veselov, R. 185 Stcmkovsky, E. Vu 194 Vdlamarln, D 136 Strmensky, С 187 Virtanen, E. 213 Suieymanov, E.V. 225, 224 Voicu, A 265 Suzuki, Y. 267 Volkanovski, A 115 Szieberth, M. 282 Voronin, B.L 185 Vottero, X. 120 Vultur, С ¡14 T Vuliur, G. 114

Tachimori, S. 216 Talov, V.V., 170 W Tang, J. 242 Tarasko, M.Z. 273 WhitlochJ. 96, 127 Teilin, A.I. 77 Terentyev, G.0 193 Y Tertytchnyi, R.G 174, 273 Teslov, A 104 Yaita, T. 216 Tetchko, 0.1 68, 202 Yakovleva, I.V. 279 Tichelen, K. Van 57 Yanusova, M. 261 Tikhonchev, M. Yu. 77 Yildirim, Yeliz 291 Yoshioka, K. 288 Torres, E.A 76 Toyooka.J. 228 Troncoso.F. 159 z Trubnikov, W. 232 Tsiboulia, A 290 Zaboudko, M.A 168 Tuchkov, A.M. 266 Zabrodskaia, S. 272 ¿agar, T. 274 Zagyvai, P. 241 и Zakharova, M.I. 169 Zeman, A 178 Udo, N. 228 Zheltyshev, V. 156 Uematsu, M. 39 Zhemkov, I.Yu 279 Unak, Turan 291 Zholudov.D. 166 Urbonas. R. 218 Zlobenko, B. 285 Urbonavicius, E. 219 Zvenigorodskaya, 0 284

312 ! ПУТНОЙ INDEX THE YOUNG GENERATION - SK01 K0002 GUARANTORS FOR THE FUTURE OF THE NUCLEAR INDUSTRY

Yvonne Broy

Siemens AG / Power Generation Group (KWU) Nuclear Fuel Cycle P.O. Box 3220 D-91050 Erlangen, Germany http ://www. Siemens .de/k wu

Yvonne .Brov(a),erll 1 .siemens.de

For several years the YOUNG GENERATION The German Young Generation also assists ac- has been attracting great interest all over Europe. tively all the ENS YGN activities. Based on the Young Generation Network (YGN) of The contribution of the management to support the European Nuclear Society (ENS) founded by Jan all our manifold efforts will be also discussed and the Runermark, in a lot of European countries a national Young Generation's demands from the management Young Generation has been established, as well in for the future will be emphasised. It is our future. We Germany. have to take it in our hands and convince the manage- Since October 1998 the Young Generation in Ger- ment it is worthwhile to speak up! many has been working in the frame of a difficult po- We - the young members of the staff in the nu- litical situation after the decision was made about phas- clear field - are going to work for more than 30 years! ing out of nuclear energy in Germany. And we - the present Young Generation - want Nowadays, our highly qualified and motivated to transfer once our knowledge and our experiences young people who have been working for a couple of to a next Young Generation^. years in the nuclear field and took already over a lot of knowledge and experiences, have to decide: Does Therefore, the main goal and our all request for our profession have a future in the nuclear industry? the future should be: The main purpose of this paper is to point out: Nuclear Power for Generations! There is a young generation who is ready to take over In the public we should discuss the important role the knowledge and the responsibility for the future. of nuclear energy for solving the environmental and The paper will briefly summarise the wide range energy problems! Nuclear energy must play a role in of activities of the German Young Generation. A se- the future world energy mix! lection of them will be chosen to highlight our fight In that case, there is no doubt about the future of for the future of nuclear energy in Germany, e.g. com- the nuclear industry, and obviously no doubt about munication with the public, know-how-transfer, im- provement of links between the fuel vendor and their further active assistance from the Young Generation customers. because we are an important factor in assuring nu- clear competence into the 21" century!

YE SESSISH 135 SK01K0003 GENERATION "NEXT" AND NUCLEAR POWER

A A, Sergeev

Research and Development Institute of Power Engineering, Moscow, Russia

My generation was labeled by Russian mass media NIKA-70 is a RDIPE-developed reactor facil- as generation "Next." My technical education is above ity of the next generation and is intended for use in average. My current position is as a mechanical engineer a floating cogeneration nuclear power plant. Due in the leading research and development institute for to specific design features of the facility, all Russian nuclear engineering for peaceful applications. It manufacturing activities and acceptance tests can is noteworthy to point out that many of our developments be performed on the manufacturer's premises. The were really first-of-a-kind in the history of engineering. facility can be delivered to the operating site as fully functional. However, it is difficult to grasp the importance of these accomplishments, especially since the progress Another advanced nuclear installation designed by of nuclear technologies is at a standstill. Can genera- RDIPE was given the name UNITHERM. This is a tion "Next" be independent in their attitude towards self-sustained small-size cogeneration nuclear plant in- nuclear power or shall we rely on the opinions of elder tended for electricity and heat supply to small towns colleagues in our industry? and industrial enterprises located in the regions that are poorly accessible for transportation of traditional fossil Humanists and philosophers claim that the main fuels, or laying pipelines for gac supply and construction purpose of a human is to reveal his capabilities as fully as of centralized power supply lines. Cogeneration NPP possible, to satisfy emotional and spiritual demands, and to UNITHERM was specially designed for operation in succeed in areas which best create a prosperous remote regions where there is no developed infrastructure environment. The world community has invented a measure and a lack of qualified personnel required for operation that indicates if living conditions meet these requirements. of the power systems traditionally based on large-size This measure of human development covers: thermal and nuclear power plants. • average life expectancy, My personal contribution in designing of these fa- • level of literacy or accessibility of higher education, cilities is moderate. But I am committed to the idea of developing safe, extremely reliable, and easy for op- • level of prosperity which is supposed to be in pro- eration and maintenance power systems that are capable portion to power consumption in a given country of generating cost-efficient electricity for many years. measured, for example, in kW-h per capita per year. 1 fully realize that implementation of the design is very Nuclear power is an area of human activity where dependent on the opinion of the nuclear community, the measure of human development is brought to a much so is the capability of nuclear power to affect economic higher level than in many other industries meant for sus- conditions in Russia and abroad. I am hoping that tenance of human population. The high-density source common sense will finally prevail and that technical of energy stimulate the capability for learning and progress will not be blocked. That is why I spend a lot replacement of traditional sources of energy such as oil, of effort to overcome the technical problems that are coal, and gas with nuclear fission and fusion energy. included in the scope of my responsibilities. I spare Combustion of fossil fuels involves a serious risk of glo- no effort to learn from my older colleagues who pos- bal warming due to air pollution with NOx and SO2 par- sess a huge engineering and research experience they ticles. Continued use of fossil fuels in the future will can share. My work is like climbing up a ladder to the increase a greenhouse effect that can be countered by future. I hope that my contribution will become more tightening of gas release monitoring and medical stan- important with time. dards, no matter how expensive these measures will be. I have to be fair about my future career. Probably, I work at the Research and Development Institute a correct choice of my way of life and my prosperity as of Power Engineering (RDIPE) which has recently un- a component of generation "Next" well-being can be dertaken an active effort to design advanced reactor fa- proved by the growing prestige and importance of cilities, of the "Next" generation with PWR reactors nuclear power among other power technologies of the having enhanced safety features. These developments new century. In that case reactor facilities NIKA-70 will have the benefit of RDIPE's long-time experience and UNITHERM will be needed to support the eco- in design of propulsion nuclear steam supply systems. nomic progress of Russia in the future. To make it Two of these projects will be discussed in detail below. true, it is the obligation of the young nuclear engineers, They are reactor facility NIKA-70 and nuclear power including myself, as the people of generation "Next" plant UNITHERM. to work intensively. 311 IMl PBESEHTATIIIS THE ENERGY MIX FOR THE NEXT GENERATION: WITH OR WITHOUT NUCLEAR?

Marie-Agnes N'Diaye

COGEMA V^lizy, France SK01K0004

This paper has been prepared as a contribution to tems mainly because it does not emit CO2. the ongoing debate on nuclear energy and sustainable Member countries of the International Energy development. Agency (IE A) recognize the potential contribution of nuclear power to a sustainable energy mix. The • Some of the advocators of sustainable energy sys- Nuclear Energy Agency of OECD recognizes the tems do not see nuclear power as part of the future: potential role of nuclear power in sustainable de- an important document from the UNDP (United velopment. In the framework of the United Nations Development Program) "Energy after Rio" Nations Convention on Climate Change, the suggests a role for nuclear power in a sustainable nuclear industry as a Non Governmental energy future in very doubtful terms; the Swedish Organization (NGO) involved in the climate ne- Parliament's February 1997 law launching the phase gotiations, calls for a role of nuclear power in out of nuclear power is entitled "Government Bill reducing the greenhouse gas effect. on a Sustainable Energy Supply;" many In this debate, radioactive waste is often first environmental organizations underlined the incom- pointed out as the main argument against the patibility of nuclear power and sustainable energy sustainability of nuclear power; the second argument systems; the European Parliament recently excluded that nuclear power does not produce emissions of nuclear power from the energy sources that can fit airborne pollutants or CO2 is completely correctly and into flexibility mechanisms because of its can therefore be a great contributor to sustainable unsustainability. energy systems.

• The supporters of nuclear power see the concern Our purpose is to go further in the debate: that about climate change as a lever to revitalize inter- sustainability is not only about climate change and the est in nuclear power. They call for a significant role of nuclear power in achieving a "sustainable role of nuclear power in sustainable energy sys- development" goes further than the reduction of greenhouse gas emissions.

B SESSION 137 SK01K0005 THE HUNGARIAN YOUTH'S KNOWLEDGE AND ATTITUDE IN THE NUCLEAR FIELD

Gdbor Petofi, Gdbor Legrddi

Technical University of Budapest, Institute of Nuclear Techniques, Hungary Abstract We constructed a nuclear TOTO, which fit the The Hungarian Youth for Nuclear (FINE) was atmosphere of the Student Island in style and in established in 1999 as the Hungarian branch of the intonation. The TOTO was made to insist on topics Young Generation Network. Our purpose is to remove that were interesting to the islanders. In connection with the misconceptions and fears that have arisen around the questions, we could assess the education of the the nuclear techniques, mainly nuclear energetics, and participants, Moreover, it was possibile to reward the to reply to the questions brought up by the Hungarian participants with high level issues, leaflets and toys youth on this topic. This year, our main activity was to which were able to attract attention to the Nuclear-tent. take part in the Student Island with a Nuclear-tent. The idea fulfilled our hopes; almost everyone was In this paper we delineate our experience that we willing to fill in the TOTO. have gained with the help of our programmes about During the five days we spent on the Student Island, the attitude and knowledge of the Hungarian youth. almost 350 TOTOs were filled out by the visitors of the Nuclear-tent. This means that more than 350 young Programmes of the Nuclear-tent persons spent at least twenty minutes with us, because usually more than one person filled out each test. We stayed on the island for five days. We communicated with the youth in the language of young The sample is small and not representative, people, our flag and our presentations were adjusted because the people who visited us are probably the to the atmosphere of the festival. The visitors of the more sensitive layer of the young generation for the Nuclear-tent were asked to fill a nuclear TOTO. problems of the country and the world. But the result is suitable to draw some conclusion and to get some Our visitors had the opportunity to control a reactor feedback about our work. simulator, and they could follow interesting demonstrations with a Geiger-Mtller counter. Every There were 13+1 questions in the TOTO in day invited experts arrived to the Nuclear-tent to give order to estimate the knowledge of the youth in the a lecture. nuclear field. It contained questions about the operation The culmination of the programmes was a public of reactors, radiation, radiation protection, accidents debate held on one of the biggest stages. Three and energetics. members from FINE and three activists from the Our experience was that usually the youth heard Hungarian Antinuclear Workgroup were debating the about only the largest accidents, but they were not clear opposite opinions. about the causes and consequences. On the other hand, on the other topics, the level of their knowledge was Lessons learned from the nuclear TOTO low or they have a lot of misconceptions, which can often be worse. Today in Hungary it is very hard to decide what portion of the population supports nuclear energy. Surveys of the public opinion - carried out mainly by Conclusions the Paks NPP on a small, but representative sample - In spite of the fore-mentioned experiences, we showed that the questionnaires must be compiled found that the youth was very open when they were carefully. addressed in their language. Since 15% of our visitors People usually don't trust atomic energy. This were from abroad, we think that in the future, it makes situation is the result of the misinterpretation of the sense to invite the activists of YGN organisations of media and the scientific explanations of nuclear experts other countries. that are difficult to understand, With the demonstrations we could make the young One way to solve this problem is to have us, the people pay attention to the problems related to atomic young experts, try to teach and inform the youth in a energy. Several interviews were taken with us and were clear and understandable manner in our own language. broadcast on TV and radio channels. The Nuclear-tent The critical point is to find a way to raise their interest. also appeared on Internet news pages. 381 ORAL PBESENTAT1DHS SK01K0006 AIMING AT THE REBIRTH OF THE NUCLEAR GENERATION

Man Marianne Uematsu

The Japan Atomic Power Company, Corporate Planning Department 1-6-1 Otemachi, Chiyodaku, Tokyo #100.0004 Japan Abstract In fact, the nuclear industries of today have varieties of branches from uranium mining, conversion, A half century has passed since Japan began an enrichment, reprocessing, plutonium management, industrialization of nuclear energy. The nuclear waste management, transportation, etc., each industry industries of today have a variety of branches and each functioning independently. Consequently there are less industry functions independently. Young professionals and less opportunities for communications among need opportunities for communications .among industries, utilities, and institutes. industries, utilities and institutes, and also nuclear experts. We, young professionals, have a serious necessity to create opportunities to communicate beyond the We, young professionals, are in the motion of barrier of organizations. We have built a free discussion organizing the "Young Generation Network (YGN) of meeting since April 1998. The young nuclear Japan," and also foresee to organize "YGN in Asia" in professionals of different fields are voluntarily the future. gathering frequently in order to get to know each other, to openly discuss, and each meeting we invite our Introduction predecessors. Our speakers are highly experienced nuclear experts. We believe that because of these A half century has passed since Japan first began meetings communication between young professionals the industrialization of nuclear energy, and nuclear and experts are established. safety is believed to be established.

Japan has experienced several accidents in recent 2. The Young Generation Network of Japan years. After a critical accident at a uranium conversion The young professionals are participating in this plant, a new organization for supplementing safety free discussion meeting voluntarily, but participants of review for the entire Japanese nuclear institutions, the this meeting are still limited in certain areas of Tokyo Nuclear Safety Network, was established in December and Tokai region. Therefore we are now in motion to 9, 1999. Almost all nuclear industries, utilities, and construct a more widely organized and officially research institutes are participating in this network. recognized group, namely "Young Generation Network A cause of prime importance of these accidents is (YGN) of Japan." an enfeeblement of the Japanese nuclear industry. As Through this YGN, we seek to emphasize a result of a worldwide nuclear bashing culture, interests communications between young professionals, to the nuclear technologies have been decreased. students, and experts of whole Japanese region, and Progressive students intend to lose interest in nuclear also with foreign young professionals. generation technologies, and professionals in nuclear industries tend to lose their pride and sense of responsibility. 3. Young Generation Network of Asia The necessity of nuclear technologies is growing Nuclear safety can be obtained not only by in the Asian region including South East Asia. Since technological advancement, but also by the the nuclear problem is worldwide, it is necessary to responsibility of human beings. The people's have a network in the Asian region. We hope to understandings of the nuclear generation can be collaborate with other Asian young professionals in obtained only by the results of nuclear safety. order to organize the YGN of Asia.

1. Free Discussion Meeting of Young Conclusion Professionals The reason we have chosen to be nuclear We recognize that the most efficient means to professionals in this era of nuclear ordeal is to overcome reactivate nuclear industries is communication, and the difficulties lying beyond the 21st century. These technological advancement can also be a result. are problems such as plutonium management, It is up to us to make this circumstance, and to environmental impacts, and radioactive waste. create a new nuclear generation.

YG SESSliM 139 SK01K0007

Keynote Address: Nuclear Education and International Nuclear University C. S. Kang, Seoul National University, Korea

There are more than 400 nuclear power units are two problems, an innovative system of human resources in operation world-widely at present. Even though development and disposition program should be future nuclear power programs do not look promising, initiated internationally. In this aspect, the concept of it is the truth that there continue to exist more than 400 establishing the International Nuclear University (INU) units in the world quite awhile. And it is the general would be one of the most viable options. consensus that the world nuclear power market will The INU would provide young professionals with revive in 10 years to comply with the UN Climate not only university-level education but also high-skill Convention. This means that we should keep and training in the fields of nuclear technology. The dispose a right size of well-qualified human resources program will emphasize on global and multi- to safely maintain existing nuclear power programs as disciplinary perspectives, which should offer our young well as to timely adopt advanced concepts of nuclear generation broader opportunities of advanced technology for the future. education and motivate professional staffs in the In this respect, the first problem that we are enhancement of their knowledge and skills. experiencing is the overall shortage of present nuclear The "World Council of Nuclear Education" could human resources over the world. This could be mainly be formed to steer the INU for close international caused by departure of many qualified persons from cooperation under the auspices of the IAEA. The INU nuclear profession in combination with reduction of would organize a world network of existing nuclear- new young people entering nuclear fields. We are related educational organizations and training centers actually experiencing the declination of number of which already exist in Member States. Existing facilities enrollment in the department of nuclear engineering at and teaching staffs should be utilized at maximum. Use the university. The second problem is the unbalanced of cyber-lecturing through Internet, cross-approval of disposition of nuclear manpower among nations. There credits among educational organizations in degree are big differences in man-power requirements and its work, certification of credits by the authorized body availability. Some nations could still maintain sufficient like IAEA, human resources placement services, etc. man-power while others could not. It is mainly due to are some of the activities that the INU could provide lack of globalization in the present educational system, in addition to its professional training and higher and lack of promoting the exchange program of human education. resources in the international level. To cope with these

NUCLEAR EDUCATION AND TRANSFER OF MOW HOW j 43 THE DECLINE IN EDUCATIONAL OPPORTUNITIES FOR YOUNG PROFESSIONALS IN THE FIELD OF NUCLEAR SCIENCES

Urs Meyer

Nuclear Power Plant Leibstadt, Switzerland

Over the last decade there have been signs of coincident decrease in nuclear facilities, with particular erosion in educational opportunities for young emphasis given to the use of neutron scattering, professionals in the field of nuclear sciences. The Efforts {o enc ^ generation are nuclear programmes and courses at umvers.t.es are djscussed jn ^ tQ meet ^ irements of the being merged and/or have simply been cancelled two jndu The ( ^ ides a yiew of of the mams reasons being the persistent public . _ .r ~ .. v, * , ,rvi[,v^vn , . ..,,,,.., . the European Young Generation Network (ENSYGN), scepticism and the grid-locked pohtica environment. ••*•:•* i • J ^ .. \ e c _, ... j . K , r , an initiative taken in order to ensure the transfer of These have led to a decrease in the number of students , , . .. „ , ,. ... , . . , know-how in the European nuclear world, and budget cuts; a comparison is made with the

SK01K0008

44 I ORAL PRESENTATIONS THE ROMANIAN EDUCATIONAL SYSTEM IN NUCLEAR ENGINEERING FIELD - EXPERIENCE AND NEW APPROACHES

Octavian Dragusin *) Andrei Burghelea *) SK01K0009

«POLITEHNICA» University, Romania

Abstract of "POLITEHNICA" University Bucharest was creditated Radioprotection and Nuclear Safety Master In this paper we would like to present the actual Program, offering to the students two directions of status of the education in the nuclear engineering study: Radioprotection and Nuclear Safety. As a result field at "POLITEHNICA" University Bucharest of co-operation and assistance offered by TEMPUS- (RO), Nuclear Power Plant Department, and also SENECA program, the new major is provided with a the efforts of integration of the educational system modern curriculum harmonized with the EU and IAEA of Romania into the international system and the requirements and with a modern state of the art development of new concepts concerning the Radioprotection Laboratory. education of the new specialists generation. At the present, the department is involved in two new programs: distance learning program and training Discussion and continuously education in the field of energy- The Nuclear Department courses in Power environment and is proposing new courses such as Engineering Faculty of "POLITEHNICA" University environmental engineering and waste management. Bucharest started in year 1970, concurrently with the Nuclear Program development start and the present Results departmental staff consists of six PhD professors and four assistant professors. At the present in Romania We also wish to establish a nuclear issues public there is operating one PH WR Nuclear Power Plant with information center and intend to develop a Young one unit operating and other four in various stages of Generation ... in our University. construction. Through this paper we would like to illustrate the We have graduate and post-graduate programs, the present potential of the Department and the opening base program is nuclear engineering with 30 students for future common programs. every year. The experience of the didactic corps in We consider that the academic environment training personnel for the Cernavoda PHWR NPP represents a strong nucleus of Young Generation and represents a plus. represents an essential juncture in the formation of the Starting fall 1995 at the Power Engineering faculty future generations of professionals in nuclear field.

•) "POLITEHNICA" University of Bucharest, Romania HOClfAB EDUCATIIK AND TRANSFER If INIWKOW j 45 THE SIEMENS GRADUATE PROGRAM

Isabelle Schaffler SK01K0010 Siemens AG / Power Generation Group (KWU) Nuclear Fuel Cycle P.O. Box 3220 D-91050 Erlangen, Germany http://www.siemens.de/kwu

Isabelle.Schaeffler(a),erl 19.siemens.de

Abstract The SGP normally extends over a period of two years and consists of three different assignments with SIEMENS is an international company acting in different tasks in various functions. The home different domains: power generation, communication company, which employs the SGP members after the and information, traffic, health, etc. To be more flexible program, determines where the assignments take place. and active in a world in constant evolution, the company proposes a graduate program where young people with One out of the three SGP assignments will be a special background have the possibility to start an abroad (i.e. not in the country of the home company). international career in all the domains of activity. This Trainees must be prepared to relocate several times. graduate program is especially important in the domain The aim of this abroad assignment is to have of nuclear energy, where the know-how transfer experienced professional and personal integration in a between the previous generation and the new one is a foreign country. constant point of interest. Supplementary training effectively supports the This article presents the conditions to be accepted goals of the Program: multi-disciplinary method in this graduate program, and the supplementary training (i.e., presentation techniques, teamwork skills, training supporting this program. The SIEMENS and personal working methods), selected management graduate program (SGP) proposes a global concept training, language courses, and intercultural with a main emphasis being international. instructions. Special training courses are offered in accordance Recruiting Methods with the demands of the different fields of activity. Participants report on their work and learn about The SGP applied to highly qualified university worldwide SIEMENS activities. graduates with a Master degree in fields such as electrical/electronic engineering (preferred), computer Results science, mechanical/ production engineering, engineering management, and economics/business After these two very intensive years, a young engineer administration. in the nuclear field tries different experiences, for example, The aim of this program is the qualification and in the fuel assemblies service, in the design area and in development of highly qualified junior engineers and management and sales. The assignment abroad allows junior business managers for future multifunctional and him to work more efficiently in an international international tasks. The program focuses preferably professional milieu. During the three assignments, the on the following fields: production/quality control, trainee has the opportunity to meet people, which can be project engineering/systems technology, sales/ a source of information for him in the future. marketing, logistics/organization, and personnel management. Conclusion The eligibility requirements for this program are a convincing personality with social competence, an The SGP is a good opportunity for young engineers excellent academic record, skills and potential for to start a career in the nuclear field. The program is planning, organization and communication, and being optimized so that the trainees acquire a broad overview fluent in English. A third language is actually learned of some activities in the nuclear field over a short period by trainees. of time. This graduate program is open for all the domains of activity from SIEMENS.

41 | ORAL PRESENTATIONS SK01K0011 Keynote A ddress: Information is not Forever J. Graham, ETCetera Assessments LLP

In 1968 AD, 25 BW (Before the Web), a fast re- 80's. To read the 'secure' electronic files one has to actor based on high-temperature gas-cooled silicon- also keep the software and the necessary hardware. clad pebble-bed fuel was designed. Fuel was tested in The latest software does not always read earlier Mol, Belgium, and the work was done in cooperation files. Machines using later operating systems don't rec- with the researchers at the Halden Test Reactor in Nor- ognize earlier software. For example, the latest MAC way. The design had great characteristics - its high tem- OS 8.5 operating system will not recognize the stand- perature gave it an efficiency that was 10% above that ard web software - Netscape 4.0.2 Thus, protecting of liquid-metal fast reactors and its silicon-clad fuel gave previous information is a multi-faceted problem involv- great safety margins. ing hard copies, electronic files as well as the appro- Today, US Nuclear Energy Research Initiatives is priate software and hardware for the those files. I have funding research and design work at MIT on a new a recording of my Father speaking in the year of my pebble-bed design for the next generation of NPPs. birth. Yet, the information remains sealed in the alu- The danger is that the 1999 workers, in the year 6 AW, minium disc. It requires a bamboo needle to play and a know nothing of the earlier research and the testing. 45-rpm gramophone machine to play it. In years past, the first research step was to look at Books from the sixteenth and seventeenth centu- the "literature" through extensive searches of Abstracts ries can be preserved and still only need a young eye to see what had been published and what had been dis- to read. Hard copy, although taking space, is still a cussed at Technical Societies. Such information saved good way to preserve vital information. It lasts at least heartbreak in reinventing things that had failed before as long as electronic images but it is difficult to search in and gave short cuts to new work. research mode. Museums have been involved in this prob- lem for some time. They know that photographs fade, Now - what is different? The new researcher still magnetic tape prints through to adjacent tape, paper be- sifts through the archives to see what has gone before. comes brittle, reproduction machines die and even mar- However, the archives are not large tomes of small type ble statues crumble. summarizing published papers, they have become the Internet — the web. What existed before the web al- So much expert information is being destroyed or most doesn't exist. Information now appears to have a exists only in the heads of an aging population. How time-zero. History in often is only six years long. The is that to be preserved for the younger generation? UK-Belgian-Norwegian work was news to an MIT manager. In summary, there are at least four problems: There is another information problem that we need • Understanding that information did not begin with to be aware of. the web — there is prior knowledge, In this decline of nuclear enterprise research • Preserving the expertise and knowledge of those projects have been cancelled. Major facilities, such as who will shortly not be available to the industry, EBR-2, have been closed. The paperwork is carefully • Preserving information presently in many forms in put into storage as the project team moves on or re- physical storage, and tires. Then another government hand issues a notice that old files must be cleared out in unused filing cabi- • Funding this work — it is not negligible. nets. Information is not forever. Researchers of the fu- The first part of the solution is to make the problem ture are excused in thinking that their work is brand known. Administrators must be persuaded that the issue new if there is no record of prior work. is important and that money needs to be allocated to It is a problem that was been recognized earlier so preserving your nuclear heritage. that some information has been transferred to micro- The second part of the solution is to make sure fiche, electronic files and now CD-ROMs to give them that the young generation has access to older a better chance of preservation. This introduces an- information and knows how to use it. other problem. A third part of the solution is to include this My first book is also protected on electronic files. subject - as a matter of international importance - in However, those files are in Super Scriptsit language, technical conferences, to raise the profile of the issue. which was used only on the TRS-80 computer in the

NUCLEAR EDUCATiU AND TRANSFER IF HflW HOW I 47 SK01K0012 CERNAVODA NPP UNIT 1 - A PLANT OF SEVERAL GENERATIONS

loan ROTARU*, Mircea METES*, Mihnea Serban ANGHELESCU**

* "Nuclearelectrica" National Company, 33 Magheru Blvd., Bucharest 1, Romania ** Center of Technology and Engineering for Nuclear Projects, POB 5204-MG-4, Bucharest-Magurele, Romania • accident rate for 200,000 man-hours worked Abstract • volume of low-level solid radioactive waste pro- Cernavoda NPP Unit 1, the first nuclear power duced unit in Romania, has a long and tormented history. It • rate of the fuel failures reported to the total number represents a rather unique case in Eastern Europe. The of irradiated bundles project started well before 1989 (the construction phase Some comments are included about the operation lasted 17 years and generations were involved in its costs of the plant, the behavior of the nuclear fuel manu- completion), but it is effectively based on western factured in Romania and the results of the unit's warranty technology (CANDU). Meanwhile, the national nuclear tests performed during the initial operation period. program underwent many changes, affecting the lives and careers of Romanian nuclear professionals. Finally, on As regards nuclear safety, some comments are December 2nd 1996, the unit began its "commercial included about the events having occurred in these 3 operation", being operated at its nominal power rating of years, the improvement of the unplanned event analysis 706 M We. It now provides a reliable source of electricity and the development of "safety culture" among the for Romanian economy, supplying to the national grid operating personnel. about 10% of the country's average annual demand. After 3 years of commercial operation, the The paper reflects some aspects related to the shift operational and nuclear safety performances achieved of generations during the project's development, including by Cernavoda NPP Unit 1 are at a good level, compared the present stage. The operational performances achieved with other nuclear power plants in the world. "in service" by Cernavoda NPP Unit I up to the end of Besides the power generation, nuclear safety is a 1999, are also presented. Reference to the electrical energy permanent concern of all the company personnel. The production, performance indicators, production costs, indicators related to the unit safety are closely nuclear safety, radiation protection, radioactive wastes, monitored and reviewed in accordance with the nuclear fuel consumption and nuclear fuel performances requirements of the license issued by regulatory body, are included. Comparisons are performed with similar as well as with the recommendation of international indicators reported by other worldwide nuclear power organizations (IAEA, WANO). plants, in order to assess our results. Conclusions Assessments methods and results Cernavoda NPP Unit 1 has proved the capacity of The assessment of the operational and nuclear the Romanian personnel, mainly young professionals, to safety performances achieved "in service" by operate a complex nuclear power plant, taking advantages Cernavoda NPP Unit 1 up to the end of 1999 is of the features of CANDU technology and of the high performed using the important World Association of quality of the equipment supplied by well-known Nuclear Operators (WANO) indicators and other manufacturers, with the due care to the nuclear safety significant indicators, such as: requirements. It is an example of a successful transfer of experience to the younger generation. The results are more • unit capability factor remarkable as they have been achieved in the first 3 years • unplanned capability loss factor of operation of the first nuclear power plant in our country, • number of unplanned automatic scrams per 7000 under unfavorable economical and social conditions, hours critical reactor specific for the actual transition process in Romania. • unavailability of the special safety systems At this time, despite the abnormal economic and social • collective radiation exposure for the profession- conditions in our country, the success of Cernavoda NPP ally exposed personnel Unit 1 is proof of our credibility. We believe the completion • volume of the radioactive emissions to the en-vironment and commissioning of Cernavoda NPP Unit 2 will challenge expressed through the equivalent do-se received by the the young people to pursue a career in the nuclear field. critical group of population 481 ORAL PRESENTATIONS SK01K0013

METHODS AND PROCEDURES OF SUCCESSION OF GENERATIONS

Alexandra Homann, RalfBendzko Colenco Power Engineering AG; Baden; Switzerland

The development of nuclear energy began in approach is required - the transfer by guided training Germany and Switzerland in the 1960's. The most on the job. experienced specialists started their career in those In nuclear technology, the first approach is widely years when we were born. They gained their knowledge in use. A comprehensive system of international in a time when nuclear energy developed in an (IAEA), national (for example KTA in Germany) and exploding manner. They founded the nuclear societies internal terms, standards and rules has been developed. and the related institutional framework. Today, the older But there is a real problem with the "Know-why"-part generation is moving toward retirement. of this codified knowledge. The codes and standards The young generation is prepared to take over the reach an inaccessible volume, trying to codify these responsibility. soft facts. The European nuclear landscape of the 1990's is We are proposing a programmed dual strategy of mature. The utilities decrease the staff to cut costs in knowledge transfer from the older generation to the order to survive in the liberalised energy market. This younger one: is only possible due to highly experienced personnel. - Plan the knowledge transfer like an investment. But even this personnel belongs to the older generation and has to be replaced by the younger generation. - Identify the repeatable knowledge of the know- how-type and store it in standards, rules and pro- It appears that we should be less experienced. Is cedures. that true? How can it be ensured that we are able to take over the responsibility? Is experience only a matter - Identify the knowledge of the know-why-type and of many years working on a job? What about the transfer it to the young generation. transfer of responsibility? Why do we have to solve a The first task is a challenge for the management, problem in the formerly considered way, rather than which has to ensure the required resources. The older follow our own inspiration and experience? And last generation used the "trial and error" approach to gain but not least: What is the future of nuclear energy in their knowledge. A lot of money was spent for this Europe? approach. Why not invest in knowledge transfer with- Knowledge is one of the most important human out errors? resources. The knowledge potential available in a A systematic approach should be used for the sec- company is often not very accessible. The structuring ond task. Experienced specialists have to lay down their of our own knowledge resources is therefore becoming knowledge, gained in long years of doing their jobs. a decisive key to succeed in knowledge transfer. The codification must be performed in an ergonomic From the viewpoint of knowledge transfer there manner. Training programmes have to be based on these are different types of knowledge. written standards and rules. For each type, a different strategy of knowledge The transfer of the "know-why" is the most daunt- transfer should be used: ing task. The young generation has to take over the responsibility - and the older generation has to give it 1. Codification: Hard facts and often-used "know- to us. But we need their knowledge and their experi- how-type"-knowledge can be codified. The main task ence to do our job. We need them as Mentors, as expe- of the transfer is the codification and use of the codes rienced people, who should help and guide us, when in training and performing the job. we ask for guidance. 2. Personification: Soft facts and "know-why- type"-knowledge is hardly to be codified. Whether the The success of the succession of generations de- codification will be very expensive from a cost-ben- pends on the transfer of responsibilities from the gen- efit point of view - whether it will be not possible at eration that developed nuclear power to the genera- all. For this type of knowledge a specialist-focussed tion that will operate it in the next millennium.

NUCLEAR EDUCATION AND TBANSFEB OF KNOW HOW I 49 THE NECESSITY OF GENERATION-TO-GENERATION KNOWLEDGE TRANSFER

VS. Gupalo

VNIPIpromtechnologii, Moscow, Russia

1 Abslrav; -. hard rock are moving slowly, the lifetime of a human being will not be enough to run the studies. Radioactive waste is generated at all stages of the nuclear fuel cycle, including ore mining, processing This paper will outline the examples of and enrichment, nuclear fuel production, NPP investigations on physical processes in the hard rock operation, fuel processing, production of weapon massif using a heat source that was acting during a 40- materials, and decommissioning of nuclear facilities. year period of facility operation. Today, the disposal of radioactive waste in geological formations is the only option for waste Discussion isolation accepted worldwide. Here, the hard rock A necessity of a Generation-to-Generation massif becomes the major safety barrier for the waste knowledge transfer to substantiate the radioactive waste containing long-lived radionuclides. safe isolation in geological formations will be For the purpose of objective safety evaluation for discussed. this isolation option, it is necessary to obtain the experience, instruments, and methods allowing a long- Conclusions term prediction of changes in the rock safety properties. Underground laboratories are created for this Currently, the generation that has created nuclear purpose. The more monitoring is conducted in these energy production is retiring, but its problems and tasks facilities, the greater the probability is in obtaining the remain for future generations. reliable data needed. To successfully resolve these tasks, it is necessary Taking into consideration that the processes in the to consider the experience, instruments, and methods obtained by previous generations.

SK01K0014

5D1ORAI PRESEMAIIONS SK01K0015

Keynote Address: What Will it Take to Rejuvinate Nuclear Energy? A. Kadak, American Nuclear Society

The issue of rejuvenation of nuclear energy on a The driving force for these changes is to make rldwide basis will require several fundamental new nuclear plants competitive without compromising inges. These changes include how we design them, safety. If nuclear plants are not competitive with snse them and operate them. While on the surface natural gas or local fuel sources that may be more ;se changes may seem overwhelming, they are abundant such as coal, nuclear energy will not be :requisites to nuclear energy's re-emergence as a widely used in either developed or developing nations. ible electric energy source. The requirements for Our challenge is to develop the new "Generation IV" M nuclear technology are that the plants must be technology that is on already on the drawing boards upetitive; they must be "demonstrably" safe; they at several institutions, and companies. An example ist be proliferation resistant; and finally they must of such a technology will be discussed. ist in the current political climate.

NUCifAR BimeiBEY 1153 SK01K0016 Keynote Address: Global Trends in Advanced Reactor Developments, and the Role of (he IAL J. Kupilz and./. Cleveland, Internationa. Atomic Energy Agency

Due to further increases in the world's population and enhancement of safety culture and international along with further industrialization and economic co-operation are highly important in preserving the development, global energy demand will surely potential of nuclear power to contribute to future continue to increase in the21" century. energy strategies. In the second half of the 20* century nuclear To assure that nuclear power remains a viable power has evolved from the research and development option in meeting energy demands in the near and environment to an industry that supplies medium terms, new reactor designs for all principle approximately 16% of the world's electricity. In these reactor lines and for different applications are being 50 years of nuclear development a great deal has been developed in a number of countries. Common goals achieved and many lessons have been learned. At the for these new designs are high availability, user- end of 1998, according to data reported in the Power friendly features, competitive economics and Reactor Information System, PRIS, of the IAEA, there compliance with internationally recognized safety were 434 nuclear power plants in operation and 36 objectives. under construction. About 9500 reactor-years of operating experience have been accumulated by today. World-wide, considerable efforts are being made to develop advanced nuclear power. Various The continued contribution of nuclear energy to organizations are involved, including governments, energy needs depends on several key issues. The industries, utilities, universities, national laboratories, degree of global commitment to sustainable energy and research institutes. Expenditures for development strategies and recognition of the role of nuclear energy of new designs, technology improvements, and the in sustainable strategies will impact its future use. related research for the major reactor types combined Technological maturity, economic competitiveness is estimated to exceed US$ 2 billion per year. and financing arrangements for new plants are key This paper gives an overview about nuclear factors in decision making. Public perception of power technology development programmes and energy options and related environmental issues as projects in Member States and the role of the IAEA well as public information and education will also play as a forum for informatic exchange and co-operative a key role in the introduction of advanced designs. research. Continued vigilance in nuclear power plant operation,

54 I ORAL PRESENTATIONS SK01K0017 Keynote Address: Nuclear Reactors and Technology in the Next Stage V. Orlov, Russian Academy of Natural Sciences, Russia

In the next stage of its development, nuclear • to hit a balance between the radiotoxicity of power can aspire to reach the goals which the eminent waste and that of feed uranium, by providing physicists had in mind as they initiated in the mid- neutron transmutation; century a civil nuclear line when developing nuclear • to create power reactors and fuel cycle tech- weapons. Being cheap, inexhaustible in its resources and producing no greenhouse gases, nuclear fuel was nology that would not afford extraction of meant to substitute conventional chemical fuels on a weapon-grade materials. large-scale. However, the reactors created in the first To fulfil all these requirements, it is necessary to stage of the nuclear era and resulting from the provide substantial neutron excess in a chain reaction conversion of military developments, failed to fulfil for Pu breeding, to use fuel with an equilibrium these goals, which is the root cause of the current composition, to bum actinides and LLFPs. All this stagnation in nuclear power and of its projected falling can be done only in fast reactors. Fast reactors can contribution to world energy production. also provide fuel for thermal reactors that might still A At the same time, with the solid expertise gained be used for some applications, operating in a Th U in the area, it is possible to create in a fairly short time cycle, which is the best option for such facilities. Novel reactors and fuel technology that would meet the main engineering solutions will be necessary: high-density requirements for large-scale power production, i.e.: heat-conductive fuel (UPuN), chemically inert high- boiling coolant (Pb), dry reprocessing. These issues • to afford a 100-fold reduction in the specific have been studied well enough to allow embarking consumption of uranium, by utilizing thou- on the development of advanced fast reactors. sands of tonnes of Pu accumulated in the spent Minatom institutions are finalizing a detailed fuel from the reactors of the first stage; design of a demonstration BREST-300 plant, • to rule out nuclear disasters, by taking advan- complete with an on-site fuel cycle that will meet the tage of the intrinsic properties and behavior requirements of large-scale nuclear power. Hopefully, of reactor, coolant, fuel, etc., with the plants construction of this plant at Beloyarsk site with its made simpler and cheaper; subsequent trial operation would open a door to the next stage in nuclear power development.

NIICLEAIl TECHNBlflCY 1155 SK01K0018

THE EPR - TECHNOLOGY FOR THE 3RD MILLENNIUM

Olaf Bernstrauch

Siemens AG / Power Generation Group (KWU) P.O. Box 3220 D-91050 Erlangen, Germany http://www.siemens.de/kwu

Olaf.Bernstrauch(5),erl 11 .siemens.de

The Basic Design of the European Pressurized More comprehensively, the safety of the EPR was Water Reactor (EPR) was completed 1997, the Basic maximized both to prevent hypothetical accidents - Design Optimization Phase 1998 and the Detailed even severe ones - and to reduce their consequences Design Phase will start in the near future. With these (spreading area, pre-stressed double containment with milestones, a new generation of PWRs is moving liner, four trains safety systems,..). This n+2 strategy forward. Most of all, this is another story of a (in the event of a problem with one train and even if a successful Franco-German cooperation (see atso the second train is undergoing maintenance, the remaining Airbus and the Ariane program). It is a rundown of two are always available) allows to perform the history of the EPR, before a decision is made to maintenance during operation which leads launch the lead-unit construction. consequently to short outage periods, highest availabilities and economical operation. The EPR project was launched in 1992 by Nuclear Power International (NPI), a joint company Moreover, most of the EPR components are the of FRAMATOME and SIEMENS KWU, supported result of evolution. The reactor vessel, as a key element by EDF and nine German electric utilities. Each step for the reactor service life, is designed to be in service of the development of the EPR was harmonized with for 60 years, the core is large (241 fuel elements the Nuclear Safety Authorities both in France and instead of 205 in the N4 or 193 in the Konvoi) and Germany to reach an early approval. the steam generators have higher efficiency. Along the same line, the core and core barrel design has been What was the objective? The EPR is expected to modified to allow a reduced uranium enrichment and replace progressively the existing nuclear power plants the increase to 65 Gigawatt-day per metric-ton of the which reach the end of their service life: this process fuel discharge burn-up. starts in France between 2015 and 2020. Moreover, many nuclear reactors elsewhere in Europe and in the Finally, the instrumentation and control systems U.S. will be reaching the end of their service lives minimize human error by giving the operators a grace around the year 2010. Competitiveness with regard period of at least 30 minutes to make decisions. to other power plants, environmental aspects (green Presently, Siemens KWU and FRAMATOME are house effect), saving of natural resources (gas, oil and preparing the detailed design phase and the following coal) and security of national energy supply are other construction and commissioning phase. The decision main objectives to keep the nuclear option open. to build an EPR is not yet made either by the German The EPR integrates the latest technological electric utilities or by EdF, but it will be expected advances, especially in safety and operational aspects within the next months as a strong statement to follow and comprises more than 30 years operating experience, the nuclear way and to ensure the know-how transfer. Thus, the EPR combines the qualities of its predecessors, the French N4 and the German Konvoi.

581 ORAL PRESENTATIONS SK01K0019 MYRRHA : A MULTIPURPOSE ACCELERATOR DRIVEN SYSTEM FOR RESEARCH & DEVELOPMENT

K. Van Tichelen, E. Malambu, Ph. Benoit, P. Kupschus, H. Ait Abderrahim

SCK'CEN, Boeretang 200, B-2400 Mol, Belgium E-mail: [email protected]

Abstract sign of the Myrrha system is going on and the basic engi- The development of a new nuclear installation that neering phase has started. It deals primarily with ADS re- is able to fulfil the economical, social, environmental lated research, i.e. materials and fuel research, liquid metals and technological demands, is of first importance for and associated aspects, reactor physics and subsequently the future of sustainable energy provision. Accelerator with applications such as transmutation and safety research. Driven Systems can pave the way for a more environ- 2. Technical description mentally safe and acceptable nuclear energy production. Fundamental and applied R&D are crucial in the devel- The Myrrha concept, as it is today, is based on opment of ADS technologies and demand the availabil- the coupling of a commercial proton accelerator with ity of appropriate prototype installations. In answer to a spallation target surrounded by a subcritical neu- this need and in order to update its current irradiation tron multiplying medium. Its design is determined by potential, the Belgian Nuclear Research Centre the versatility in applications that should be made pos- (SCK«CEN) has launched the Myrrha project. It is sible. Further technical and/or strategic developments fbcussed on the design, development and realisation of of the project might change the concept. amodular and flexible irradiation facility based on ADS. A cyclotron, based on positive ion acceleration tech- This paper describes the concept, the applications fore- nology brings the protons up to an energy level of350 MeV. seen in the Myrrha installation and the accompanying The nominal current is 2 mA. The protons hit a liquid Lead- design activities currently being performed at SCK'CEN. Bismuth spallation target and produce the neutrons needed to sustain the chain reaction. The spallation target is made 1. Introduction of a double concentric cylindrical circuit with a dump tank All over the world nuclear energy has to cope with at the lower end of the circuit. At the upper part of the target the economical question of the increasing demand for system, a free surface is in contact with the incoming pro- energy. In resolving this question, nuclear energy has ton beam. No conventional window is foreseen between to satisfy the conditions of public acceptability: increas- the free surface and the beam in order to keep the energy ing the absolute safety of the installations and manag- losses at their minimum. Besides, the window is a very vul- ing more efficiently the nuclear waste. The develop- nerable part, decreasing the reliability. The windowless con- ment of a new nuclear installation that is able to fulfil cept distinguishes Myrrha from other ADS design activi- the economical, social, environmental and technologi- ties and makes the design very challenging. cal demands, is of first importance for the future of The spallation target is surrounded by a subcritical sustainable energy provision. Accelerator Driven Sys- assembly. The design of this assembly is application de- tems can pave the way for a more environmentally safe pendent. To meet our goals, the subcritical assembly must and acceptable nuclear energy production. consist of two spectral zones: a fast neutron spectrum zone In contrast with a classic nuclear reactor, an ADS is and a thermal one. The central fast spectrum zone con- a subcritical system in which the deficit on the neutron sists of fast breeder-type fuel rods with a Plutonium con- balance is counterbalanced with externally produced tent between 20 and 30% in a Lead environment. This neutrons arising during the process of spallation when a zone is very suitable for material testing and transmuta- proton beam, descending from an accelerator, is fired tion studies due to the high fast fluxes attainable. The ther- upon a spallation target. The fission chain reaction thus mal zone surrounding this fast zone, also has two irradia- cannot sustain itself and the safety margin is enlarged. tion channels, enabling irradiation experiments in light Moreover, the conditions reached in an ADS allow the water-like conditions as well as production of radioiso- transmutation of long-lived nuclear waste in short -lived topes and extraction of neutron beams for applied research. waste, reducing the time scale of final disposal. It consists of LWR-type fuel rods and is water-cooled. Fundamental and applied R&D are crucial in the 3. Conclusion development of ADS technologies and demand the avail- ability of appropriate prototype installations. In answer Accelerator Driven Systems can become the perfect to this need and in order to update its current irradiation solution to the major remaining problems of nuclear en- potential, the Belgian Nuclear Research Centre ergy production. The development of these systems re- (SCK'CEN) has launched the Myrrha project. It is quire a thorough study and experimental verification in focussed on the design, development and realisation of which SCK»CEN can play a major role, in collaboration a modular and flexible irradiation facility based on ADS, with the Belgian universities, research institutes, engineer- well-matched to SCK'CEN R&D needs and international ing bureaus and industry. Moreover, the Myrrha system fundamental research programs in the ADS domain. provides the indispensible installation for the continuation Currently the study and preliminary conceptual de- and extension of current R&D programmes at SCK»CEN. NUCLEAR TECHNOLOGY I I 57 PRELIMINARY DESIGN OF A GAS-COOLED ACCELERATOR DRIVEN SYSTEM DEMONSTRATOR

B. Giraud Framatome, Y. Poitevin SK01K0020 CEA/Saclay, G.Ritter CEA/Cadarache, France

At the present time, nuclear power appears to be The practicality on an industrial scale c the best solution for producing a large amount of partitioning and transmutation through ADS fc electricity from both economical and ecological reducing the amount of long life radio-nuclides has t viewpoints, provided that acceptable answers to the be evaluated. nuclear waste concern are found. In France, this is It was recognised that the most efficient way, i the subject of the law 91-1381 (December 30 th , terms of cost and planning, to conclusively assess th 1991). potential and the feasibility of a full scale industrii The transmutation of most of the long-lived programme on ADS was to design and operate an AD radioactive wastes is a promising solution which could Demonstration Facility (DF). play a substantial role for the safety of the fuel cycle. After a description of the spallation mechanism Sub-critical Accelerator Driven System (ADS), the justification of the interest of ADS is provided. coupling an ion accelerator and a sub-critical reactor, The paper then focuses on ADS DF features (tf seems to have a high capacity for the fission of minor sub-critical reactor and interfaces with the accelerate actinides and transmutation of long life fission including the process for the selection of desig products. options. This potential has been shown in particular by The main ADS DF characteristics, defined with: the studies performed by the French Institutes CEA a joint working group (CEA/CNRS/FRAMATOME and CNRS in the frame of the research group are described and justified. GEDEON (CEA/CNRS/FRAMATOME/EDF). Then, general conceptual principles for the desij The ADS is the combination of three main of an ADS DF are provided, including the reasons f< systems: an ion accelerator, generally protons are used; the selection of the gas for the cooling medium. a spallation target which generates an external neutron source; and a sub-critical fissile assembly producing A description of ADS DF basic options the fission power. The sub-critical assembly is very provided and the gas cooled DF currently under stuc similar to the core of the classical critical fission by FRAMATOME and CEA is presented, includir nuclear reactors. the spallation target, the sub-critical core and ti primary cooling system. The ion beam is focused on the spallation target, which is located in the centre of the sub-critical core. For each of these sub-systems, interfaces with tl The interactions between the accelerated particles and accelerator part and adjacent sub-systems ai the heavy metal nuclei generate a neutron source, highlighted and main issues which call for resean which maintains the fission reactions in the sub-critical and development support are identified. core. In this manner, the sub-critical core amplifies the neutron source.

581 ORAL PRESENTATIONS THE OECD HALDEN REACTOR PROJECT - INTERNATIONAL RESEARCH ON SAFETY AND RELIABILITY OF NUCLEAR POWER GENERATION

LiseA. Moen

OECD Halden Reactor Project - Institutt for energiteknikk Norway

The Halden Reactor was one of the first The reactor core contains about 110 fuel experimental reactors in the world, built into a assemblies and 30 control rods located in a so called mountain cave in the middle of the town of Halden open hexagonal lattice configuration (i.e. each fuel located 120 km south of Oslo. The reactor went critical channel is surrounded by 3 control rod positions and for the first time in June 1959 with a core containing 3 fuel positions). There are about 75 driver fuel rigs natural metallic Uranium only. The OECD Halden and about 35 instrumented fuel assemblies (IFA's) in Reactor Project was established in 1958 as a joint the core. undertaking of the OECD Nuclear Energy Agency. An IFA or a so called test rig can be extensively The Project is an international collaboration and today instrumented with detectors for temperature, pressure, 21 participating countries are sponsoring the flow and elongation measurements as well as gamma experimental program. The reactor facility is owned and neutron flux detectors. by the Institute for Energy Technology (IFE) in Norway. These advanced in-core sensors are used for monitoring fuel and material parameters during reactor The main program covers fuel performance with operation such as fission gas release, fuel rod thermal emphasis on extended fuel utilization, studies of plant behavior, fuel swelling and pellet cladding interaction, material behavior under water chemistry and nuclear clad creep properties, Zircaloy corrosion and conditions similar to commercial power plants and hydriding and Irradiation Assisted Stress Corrosion also development of advanced computerized Cracking (IASCC). surveillance systems in support of upgrading control rooms. The paper gives an overview of: The Halden Reactor is a boiling heavy water • the Halden Reactor and its associated experimen- reactor based on natural circulation of the 14 tons of tal facilities, heavy water used both as coolant and moderator. The • the current research program on fuels and materi- maximum power is 25 MWth and the water als testing, including some selected results, temperature is 240° C, corresponding to an operating pressure of about 34 bar. • an outline of the planned research activities in the period 2000 - 2002.

SK01K0021

NUCLEAR IECHHOLOGS j j ® SK01K0022 BN-800 - HISTORY AND PERSPECTIVE

/. Yu. Krivitski

Institute for Physics and Power Engineering, Russia, e-mail:[email protected]

Abstract SVRE value in the reactor core by introduction < sodium plenum above the core. The sodium cooled fast reactors are one of the most developed and advanced directions of future The analysis of the numerous ways to reduce tr nuclear engineering. Russia is the first among other SVRE value allowed to choose the most optimal coi countries in field of fast reactor development. design. The idea of fast reactor designing was proposed The design project of BN-800 core WE in the former by Dr. A.I. Leipunski at developed in 1992. During next 5 years the compte the end of 40th. The successful operation of Russian justification of reactor physics was carried out base fast reactors (BOR-60, BN-350 and BN-600) and the on calculational analysis and experiment; world experience proved the feasibility, reliability and investigations at the MFS facility critical assemblk safety of this direction of nuclear engineering and and based on results of international benchmar allowed to begin the development of the BN-800 analysis of BN-800 reactor core with sodium plenun reactor project as the commercial fast reactor. At the end of 1998 we got the license for reactc In 1992 Russian Government confirmed the construction. construction of BN-800 reactors on South Ural NPP in Chelyabinsk region and on Beloyarskaya NPP. BN-800 Reactor designs for nuclear wastes utilization History of BN-800 design During last five years the numerous investigation The first design of BN-800 reactor was developed were carried out on possibility to use the different cor and was undergone an examination in 1985. It fulfilled modifications based on BN-800 reactor project fo the demands made of the reactors in that time. effective utilization of plutonium (including weapoi grade plutonium) and minor actinides. But last time (after the serious accidents on Chernobyl NPP and Three Mile Island NPP) the The main directions of the core modifications ari aspects of safety increase of NPP play a leading role as follows: when designing the new reactors serving the economic - abandonment of fertile blankets; competitiveness. - use of MOX-fuel of increased enrichment; All these aspects were introduced into the new - use of uranium-free fuel; Nuclear Safety Rules, adopted in our country in 1989. These Rules include the requirement of guaranteeing - use of special devices for long-lived nucleai of negative reactivity coefficients on reactor power wastes transmutation; and coolant temperature. - closed fuel cycle organization with thermal reac- tors. After having adopted the new Safety Rules the commission of Russian Academy of Science headed Conclusion by Dr. V. Subbotin made new examination of BN- 800 reactor project. The commission noted the large Thus on the basis of BN-800 reactor project it is positive value of sodium void reactivity effect (SVRE) possible to design the universal fast reactor permitting as a main disadvantage of this project. The to solve rather effectively the different problems oi recommendation was to develop a new reactor core nuclear fuel cycle: from high breeding of secondary design with negative value of SVRE. nuclear fuel to the effective transmutation of long- The first investigations in the end of 80th showed lived nuclear wastes depending on state of nuclear the principal possibility to achieve zero (or negative) market.

BO I ORAL PfitSIHfYflTIOMS SK01K0023 Keynote A ddress: Non-Proliferation and Nuclear Disarmament M. Shea, IAEA, Austria

International security rests upon a balance of equipment or diversion of materials, and intervene relations, some complementary, some in competition. where necessary to prevent an errant State from The possession of nuclear weapons remains a key acquiring nuclear weapons. A State deciding to element in the security balance. Any increase in the launch a nuclear weapons program today would need number of States possessing nuclear weapons de- to find motivation sufficient to offset the penalties of stabilizes security and brings alive new threats of discovery, the possibilities that the enterprise might conflict and shifting interests and influence. Progress not succeed, and costs that might be prohibitive; those towards disarmament can and should serve to diminish deterrents are more significant than ever before: the the prospects for the use of nuclear weapons in question is, are they now and will they remain nuclear-armed States and create an environment to sufficient? encourage further reductions. Sudden changes or changes which are not planned in conjunction with Conversely, progress towards nuclear related steps may encourage adventurism and thereby disarmament depends upon a host of factors, a security contradict the intended purpose. Disarmament in an balance and the lack of threats, a clear cut and environment in which proliferation is not perceived supported rationale and an infrastructure which would to be under control is unlikely; accepting intensified nurture and encourage some actions, and a motivation non-proliferation in an environment in which to proceed, which might take on different progressive steps towards the elimination of existing characteristics with time. Progress in disarmament arsenals is not evident is equally unlikely. depends in part on the prospects for proliferation at any moment, and seeking progress towards non- To supply an increasing share of ftiture world proliferation is one motivation for States to seek ways electricity requirements, nuclear power programs must in which measured progress can be made. not contribute in any appreciable way to the Fissionable materials are common to all nuclear proliferation of nuclear weapons. There are potential weapons and controls on the production, storage, technical links between nuclear power and nuclear processing and use of fissionable materials provides weapons, and to enable States to realize the benefits one means to address non-proliferation and of nuclear power, a multi-faceted effort is required to disarmament. In this article, the relevance of such minimize the motivations for proliferation, control controls is examined and the current situation and commerce in sensitive facilities, equipment and future prospects are assessed. materials, detect any misuse of such facilities or

POLITICAL ASPECTS IG5 SK01K0024 THE PLUTONIUM CHALLENGE FOR THE FUTURE

Leonard W. Gray

Lawrence Livermore 'National Laboratory Livermore CA 94551 United States of America

Designated as No Longer Required for Defense Purposes." Abstract In this joint statement the Presidents "affirm the intention Glenn T. Seaborg states that "plutonium is unique of each country to remove by stages approximately 50 among the chemical elements: metric tons of plutonium from the nuclear weapons pro- • It is a synthetic element, the first realization of the grams, and to convert this material so that it can never be alchemist's dream of large-scale transmutation; used in nuclear weapons... and., .to ensure that these mate- • It was the first synthetic element produced in vis- rials do not become a proliferation risk." ible amounts; These 100 tonnes of plutonium must be managed in • It has unusual and interesting chemical and metal- a proper way such that it becomes neither a proliferation \urg,\caV properties; nor an environmental risk. The United States has pro- • It is one of the most dangerous poisons that man posed that it manage it's 50 tonnes by a dual approach- must learn to handle; once through MOX burning of a portion of the plutonium • It was discovered and methods for its production and immobilization in a ceramic matrix followed by en- were developed in secrecy during World war II;" casement in high level waste glass. Russia has proposed • Its discovery was inextricably intertwined with that it manage its full 50 tonnes by burning in a reactor. the discovery and study of nuclear fission; The MOX program in the United States would burn • Its naming was inextricably intertwined with the the cleaner plutonium metal and residues. Weapons naming of the outer planets; and components would be converted to plutonium oxide, • Production of weapons-grade plutonium has been followed by dissolution utilizing an Ag(II)-nitric acid inextricably intertwined with the politics of the method. The plutonium would be purified by one cycle "Cold War." of solvent extraction, precipitated as plutonium oxalate Since its discovery in 1941 and its dramatic and then converted to oxide. This would not only remove emergence at Nagasaki, plutonium has altered the course the gallium but would provide the correct morphology of history, changed the concepts and consequences of war, for preparation of the MOX fuel. and paradoxically has become a powerful instrument for The Immobilization program in the United States peace. Since 1944, approximately 1600 to 1700 tonnes would mineralize the plutonium contained in a variety of weapons-usable plutonium have been generated in the of residues that were left in place when the weapons world's reactors. .Between 1944 and 1988, the United production complex was shut-down at the end of the States built and operated 14 plutonium-production cold war. These residues have a wide range of impurity reactors, nine at Hanford and five at Savannah River, pro- contents, typically from a few parts per million to > 90 ducing approximately 100 metric tons of plutonium. With wt. %. Plutonium in these residues would be blended to the fall of the Berlin Wall in 1989, which signaled the end levelize the impurities and thereby avoid reprocessing of the "Cold War", production of weapons-grade pluto- of this plutonium. This blended plutonium would then nium has stopped in the United States. be mineralized at high temperature in a titanate ceramic Russia has not disclosed the amount of plutonium followed by canning of the ceramic pucks. These cans produced, but various estimates indicate that the pro- would by loaded into a magazine and then locked into duction was about 130 tonnes. Production has been place within a stainless steel canister. The plutonium curtailed in Russia; three dual-purpose reactors still pro- ceramic would then be encased in high level waste glass. duce weapons-grade plutonium - two at Tomsk-7 A high radiation field would protect the plutonium (renamed Seversk) and one at Krasnoyarsk-26 (renamed in the spent MOX fuel and the immobilized ceramic Zheleznogorsk Mining and Chemical Combine). In a form for some period of time. Within the time period 1994 United States -Russian agreement that has yet to of the high radiation field, these forms would be en- enter into force, Russia agreed to close the remaining tombed in an underground repository. operating reactors by the year 2000. This, however, disposes of only 50 tonnes of the Treaties between the United States and Russia have approximately 1700 tonnes of the worldwide weapons- already cut the number of nuclear warheads from more usable plutonium. Russia will dispose of another 50 than 10,000 to about 6,000 under START 1, which has tonnes via the MOX burning route. Worldwide approxi- been ratified, and to about 3,500 under START 2, which mately 1600 tonnes of weapons-usable plutonium is still still awaits approval. If Russia and the United States available either as spent fuel or as separate plutonium conclude START 3, that number could drop to between oxide. Is society prepared to deal with these 1600 tonnes 2,000 and 2,500. of plutonium? At the present growth rate, before the On September 2,1998, the Presidents of the United United States and Russia completes the disposition of States and Russia signed the "Joint statement of prin- the 100 tonnes of weapons-grade plutonium, the ciples for Management and Disposition of Plutonium worldwide stockpile will exceed 2000 tonnes. 6SI OR/11 PRESENTATIONS SK01K0025 THE CONTRIBUTION OF CIVILIAN INDUSTRY TO MILITARY PU DISPOSITION JArthur de MONTALEMBERT

COGEMA, France

Russian and United States governments jointly The more advanced - and readily available - decided in 1993, following the START 1 & 2 nuclear solution is the fabrication of plutonium into MOX fuel disarmament Treaties, to dispose of the weapon-grade and its irradiation in existing Light Water Reactors. plutonium (Wg-Pu) declared in excess of defense This solution also has the obvious advantage of burning needs. A bilateral framework agreement now under some of the plutonium (30 to 40 %) and degrading the negotiation will define the general conditions under remaining portion of it, while generating electricity. In which each country will dispose of 34.5 tons of Wg- the longer term, the burning of plutonium in new types Pu, i.e. 69 tons in total corresponding roughly to 7000 of dedicated reactors, like High Temperature Reactor warheads. Additionally, the United States government or Fast Neutron Reactor, can be envisaged, but the time has decided to also dispose of some non-weapon-grade and costs associated with deploying such reactors and plutonium derived from its defense programs. their fuel cycle must be taken into account. Weapon-grade plutonium must be disposed of in On the other hand, ((immobilization)) in ceramics a safe and secure manner. Disposition programs should and in radioactive glass - although yet never proven start as soon as possible and be properly monitored to with plutonium and thus needing additional achieve disarmament and non-proliferation objectives. demonstration - will also rely heavily on technology Finally, for obvious economic and environmental developed and mastered by the civilian nuclear power reasons, such programs should make use, as much as industry, either for the fabrication of MOX fuel or for possible, of proven processes and technology as well the conditioning of high-level radioactive waste. as of available equipment and existing sites. The disposition of military plutonium declared in The United States has already decided to adopt an excess of defense needs therefore presents the civilian hybrid solution, with the bulk of its pure Wg-Pu to be nuclear power industry with another opportunity, after fabricated into nuclear fuel and burned in existing the dilution and recycling of HEU (Highly Enriched civilian reactors while less pure or non-weapon-grade Uranium), to significantly contribute to nuclear material is to be «immobilized» in a ceramic matrix disarmament programs, by bringing its technology, encased in a radioactive glass according to the «can- expertise and experience to safely and securely dispose in-canister» approach. of a direct weapon-usable material. This experience includes non-proliferation dispositions, such as national Russia has made clear its determination to burn and international safeguards and physical protection its Wg-Pu in reactors in order to take advantage of the measures currently enforced in civilian facilities, which high energy value of plutonium. It cannot be totally will be put to use for the disposition of weapons excluded however that a small quantity of Russian plutonium. All it takes now is the political will by plutonium may be «immobilized» as well, for the sake concerned States to reach agreement on basic principles of parallelism, depending on the outcome of current negotiations with the United States. and provide the necessary public financing for those disarmament programs. Whatever the final decisions will be, both options - plutonium burning and ((immobilization)) - will rely heavily on technology developed and currently used by the civilian nuclear power industry.

POLITICAL ASPECTS 117 SK01K0026 RADIOECOLOGICAL DANGER OF FAILURES WITH NUCLEAR WEAPONS AND LIQUIDATION OF THEIR CONSEQUENCES Oksana L. Tetchko

Mira Avenue 37, VNIIEF, 607190 (Arzamas - 16)

Nizhny Novgorod Region, Russian Federation

Nuclear warheads possess a great deal of potential For each type of safety concern, there is a strict energy. They contain conventional high-explosives, hierarchical system of documentation, governing and divided materials, radioactive substances, and also determining all requirements and norms at every level toxic materials and active chemical junctions in the of safety. Accidents at MINATOM facilities reveal construction of the warhead. During storage, the that many aspects were undervalued in the safety warhead can be a potential source of danger should an assurance system that were in place at the time, emergency arise involving fire, caustic and explosive especially those aspects involving the aftermath oi chemicals, or damaging mechanical shock. major accidents, including determination of their scenarios (their initial events and how they evolved) A serious emergency situation with warheads can and the subsequent protection of the population and potentially result in the dispersion of plutonium and other adjoining territories. The degree of reasonable danger radionuclides, resulting in district-wide contamination. and the acceptable level of risk have been the subjects Environmentally, plutonium is the chief concern. Where of serious scientific investigations and discussions, multiple warheads are collectively stored, for example, recently leading to recommendations formulated by a in the Trident Missile nose cone (which can hold eight number of authoritative international organizations. warheads), even damage to one warhead containing plutonium can potentially contaminate an area of To help expedite the cleanup of emergency approximately 38 square miles (about 100 sq. km) with situations that involve radiation at faci Ikies under the serious environmental consequences. The "cleaning" nuclear complex of the Russian Atomic Ministry, five of this territory would require more than 500 million emergency technical centers were created in dollars, which is comparable to the damage caused by MINATOM in 1993 by decision of the Council of the failure of the Chernoble Nuclear Power Plant. This Ministers of the Russian Federation. report also presents information on elements of the The emergency technical centers at MINATOM assurance system for nuclear and radiation safety at are in a constant state of readiness, whose mission is to MINATOM facilities and statistics on traumatic carry out work within their established specialization, mortality, various accidents, and their consequences. such as transportation accidents and related incidents, MINATOM is a scientific-industrial complex of including performing hazardous radiation work on the technology-related enterprises for the mining of territory of their home base, and whenever the Ministry materials, production of fissile materials, and decides they can be used to perform other chores fabrication of nuclear fuels for atomic power involving the presence of a radiation issues. engineering and other items for military technology. Work is also being done on preparation and creation Additional work is done on industrial and experimental of sector systems involving seismic and geodynamic reactors, processing of spent fuel for radioactive waste networks, including a system of training and certification burial, and sector response systems for possible for rescue workers. Thus, MINATOM is striving toward emergency situations. improving and developing the sector system of accident In a complex system there are various approaches prevention and operations during emergency situations, to safety assurance, but they also have some things in in which two specific directives should be pointed out: common. The level of safety at an installation is largely 1. Prevention and cleanup of the aftermath of dictated by the scientific ideas and technical solutions accidents and emergency situations at stationary behind its design, and to what degree the components facilities. have been corroborated and tested, including the quality of equipment manufacture, construction, assembly, and 2. Prevention and cleanup of the aftermath of level of operation. transportation accidents involving radioactive materials. An important feature of the safety assurance system of MINATOM is centralization and The upgrading of systems at MINATOM also coordination of activity for all types of facilities, which places much importance on the sector's readiness to includes the Ministry level and the Department of handle accidents at the very early stages. Thus, the Safety and Emergency Situations, and incorporates all system of safety assurance at the Ministry is oriented types of safety concerns at the individual sites which toward prevention of possible violations and toward are handled by the site senior engineer. effective operations in the event of radiation accidents. 681 ORAL PRESENTATIONS SK01K0027 SIGMA: THE NOVEL APPROACH OF A NEW NON-PROLIFERATING URANIUM ENRICHMENT TECHNOLOGY

Af. Rivarola, P. Florido, D. Brasnarof, E. Bergallo

Grupo de Diseflos Avanzados y Evaluaci6n Econ6mica Complejo Tecnologico Pilcaniyeu - Comisi6n Nacional de Energia At6mica, Republica Argentina Abstract poses. These plants were used primarily for this pur- pose through 1964. From 1959 through 1968, uranium The SIGMA concept, under development by enrichment production shifted primarily to supply the Argentina, represents the evolution of the Uranium nuclear power industry. Enrichment Gaseous Diffusion technology, updated to Increase in capital charge, electric power supply face the challenge of the new economic-based and and interest rates, shifted competitiveness to centrifuge competitive world frame. enrichment plants. The Enrichment technology has been historically This economic behavior is mainly produced by considered as a highly proliferating activity in the the component scale economy and the decrease of nuclear field, and central countries limited the access compressor efficiency at low flow stages. A GDP of the developing countries to this technology. cascade has thousand of stages in series, in which higher The SIGMA concept incorporates innovative prolif- enrichments have lower mass flow. The smallest flow eration resistant criteria at the beginning of the design pro- stages are the most expensive per SWU (higher cess, and inherits all the non-proliferation features of (he component costs with less compressor efficiency). gaseous diffusion plants (GDPs). The radical new prolif- Thus a GDP will be competitive when the smallest eration resistance approach of the SIGMA technology sug- stages are competitive. This usually happens in GDP gests a new kind of global control of the uranium enrichment market, where some developing countries might access an cascade configuration for a 3 MSWU/year capacity. Enrichment plant without access to the technology itself. This behavior could be seen as a cascade para- In this paper, we investigate the economy of the digm, because it is generated for the cascade concept SIGMA plants, and the implications of this technol- itself, with small dependence on technology. ogy on the Uranium Global Market. S1GMA Concept Introduction At present, the excellent experience in gaseous It is a well known issue that in the future, a strong diffusion cannot be used for new plants due to the strong demand for nuclear energy will be focused in develop- capital restrictions of large economical plants and the ing countries, and the enrichment demand of these relative flexibility of ultracentrifuge technologies. countries will produce a political interference between Then, in order to obtain a competitive gaseous diffu- natural market forces and political constrains. sion technology, efforts must be focused on capital cost Market forces in developing countries usually try to reduction and economy at low scales production, short move commercial activities with high capital requirements construction time, and lower energy consumption. to demand centers (developing countries too for nuclear The SIGMA concept solves all these targets intro- energy). Then proliferation issues will be under strong ducing several technological solutions, but using the well- pressure to take into account that present projections pre- known gaseous diffusion experience. It overcomes the dicts for the year 2040 nuclear energy in developing problem of large volumes needed to obtain good eco- counties will be higher than in developed countries. nomic performance at the smallest stage sizes (the cascade This issue could evolve in the worst case scenario paradigm). A new modular design is able to manufacture, due to technology evolution. assemble and commission into a proper installation, to be Present commercial Uranium enrichment technolo- finally send to the final site, The SIGMA concept allows gies were developed for military Highly Enriched Uranium optimum size for major component standardization. (HEU) production. Ultracentrifuges came from Russian For the SIGMA approach, reduction of the prolif- HEU technology, GDP came from US HEU production, eration risk is a central issue in the future nuclear energy and laser AVLIS technology came from the PuAVLIS US market and then has been included in the design. The project. All the designers, trying to improve the econom- Gaseous Diffusion Technology has low potential attrac- ics for the new generation of enrichment technologies, tiveness for HEU production, but relatively high diver- look to increase the separative factor and reduce the sion capabilities ofhigh LEU mass. The SIGMA concept Uranium hold up, increasing the potential proliferation has been developed in order to easily reduce the prolif- risk of all these future technologies. eration risk to a near-zero value. The design includes a Gaseous diffusions plants low cost detector for total neutron count and an optimized geometry for gamma-neutron detection. A commercial Construction of GDPs was started during World scale module is designed to obtain a maximum enrichment War II to produce enriched uranium for defense pur- grade of 5%, so it could not be used for HEU production. POLITICAL ASPECTS IEB SK01K0028 Keynote Address: Nuclear Science & Technology: Applications for the Welfare of Mankind A. K. Padhy, IAEA, Austria Human knowledge and progress go hand in hand. As destined for long-term storage. It also inhibits sprouting knowledge expands, the potential for development grows. and extends the shelf-life of fresh produce that can spoil Over the last century, nuclear science and technology have in transit. In poultry and red meat, irradiation destroys played an increasingly important role in improving human bacteria associated with food borne diseases. well-being and the world around us. Today, nuclear appli- Nuclear technologies are making sizable contribu- cations can be found in almost every social and economic tions in human health through their applications in the sector and in virtually every comer of the globe. fields ofNuclear Medicine, Radiotherapy and Radiology. Yet as the world becomes more reliant on scientific Principally nuclear medicine investigations are highly advances, society appears less and less aware of, or per- cost-effective diagnostic procedures used for assessment haps indifferent to, its achievements. If wise use of science of organ function and early detection of diseases. To this in future is to be assured, key applications for everyday use effect, radionuclides or compounds that are labeled with need to be better understood, not only by decision makers, radionuclides (radiopharmaceuticals) are administered to but by the people whose whole lives they touch. the patients and a specific organ function can be evalu- People around the world today are deepry concerned ated by tracing the dynamic bio-distribution of such com- about the dangers posed by nuclear weapons, and some re- pounds in specific organs. Tracing of the compound is main skeptical about the safety of nuclear power. But most achieved by external monitoring of the photon emitted are unaware of the many positive contributions that nuclear from the radionuclides by using the gamma cameras. On science and technology are making to everyday life. the other hand Therapeutic nuclear medicine is rapidly Food, health, energy, industry and environment - developing as an additional treatment modality in oncol- the fields could not be more diverse, yet these are just ogy and few benign clinical conditions. The treatment of a few of many examples where nuclear based technolo- thyroid cancer or hyperthyroidism with I-131 has been in gies are helping to understand and provide solutions use for more then 50 years. to to-day's economic and social problems. There has been a dramatic increase in cancer cases Many factors affecting plant growth and crop quality, world-wide, especially in industrialized nations. The number such as drought, insects and diseases, are often beyond farm- of new cases is expected to climb to 15 million by the year ers' control. While it is difficult to create ideal growing 2015, and roughly two thirds of these will occur in the de- conditions, for the past three decades it has become possi- veloping countries, where the average life span is quickly ble to produce plants that are better tailored to thrive under increasing. About half of all cancer patients today receive imperfect conditions by drawing on the skills of the plant radiotherapy as part of their treatment. geneticists and maximizing opportunities offered by nu- Nuclear techniques during the last decade have clear technology for inducing subtle changes or mutations been extensively used in human nutrition research. in the genetic make up (DNA) in plant seeds, buds or tis- Using such techniques, it has been possible to analyze sues. Thanks to the controlled use of ionizing radiation for micro-nutrient status of humans in health and disease producing mutation, to-day it has been possible to produce and also provide advisory on diet therapy. more than 1800 varieties of 164 plant species including To-day millions of tonnes of one time use medical rice, wheat, maize, barley and banana considered as staple products, ranging from scalpels to syringes, are being steri- food in most of the developing countries of the world. lized in more than 200 facilities in 50 countries all over Nuclear science also plays an important role in understand- the globe. Radiation kills disease producing bacteria with- ing soil-plant interactions and in helping to increase crop out leaving a residue. Penetrating gamma radiation al- production through better soil management. Insects reduce lows products to be sterilized on-line, in bulk and in their , food production through both direct damage and the dis- final packaging. Irradiated products are not radioactive, ease they spread to crops and livestock. Pesticide which and they can be used straight from the treatment unit. are used routinely for pest control pollute the environment, Nuclear techniques play an important role in envi- contaminate food and water, destroy beneficial insects and ronmental protection by providing assistance in promot- increase resistance in target species. Nuclear based tech- ing alternate sources of energy, reducing air pollution, niques like sterile insect technique (SIT) can reduce and in managing fresh water resources, controlling water pol- some cases eradicate insect pests that are harmful. lution and guarding the oceans and seas. They are also used to analyze minerals, soils, gases, water and other Nuclear technology provides unique information substances used in industry, and the results often influ- about the nutrition of farm animals- data that can be ence economic, ecological, medical and legal decisions. used to find better ways of converting animal feeds The International Atomic Energy Agency works into milk and meat. In countries where isotope-based to foster the role of nuclear science and technology in concepts have been widely adopted, research has paid support of sustainable human development. This in- handsome dividends in terms of higher animal produc- volves both advancing knowledge and exploiting this tivity and improved nutrition and health of consumers. knowledge to tackle pressing world-wide challenges - Food irradiation allows the complete dis-infesta- hunger, disease, natural resources management, envi- iion of grain, spices, vegetable seasonings and dried fruit ronmental pollution and industrial quality control. NUCLEAR lHUHUEYII173 SK01K0029 MEDICAL APPLICATIONS IN A NUCLEAR RESEARCH CENTRE

E Vanhavere and G. Eggermont

SCK-CEN, Belgian Nuclear Research Centre, Boeretang 200, 2400 Mol, Belgium

Abstract cessful treatment of the thyroid gland with 1-131, creat- ing however a number of radiation protection challenges In these days of public aversion to nuclear power, it is for the hospital, the family, and even the environment. essential to stress the importance of medical applications of ionising radiation. Not only to the general public, but the authorities and research centres with potential public Discussion health risk have to be aware of these medical applications The Belgian Nuclear Research Centre began medi- as well. Now that funding for nuclear research is declin- cal applications research some years ago. For a long time, ing, an opening of the medical world can give new oppor- isotopes have been produced in the BR2 high flux re- tunities to a nuclear research centre. A lot of research could search reactor for use in nuclear medicine (Tc-99m). A be done where the tools developed for the nuclear power new project is currently investigating the possibility of world can be very useful. Even new applications for the producing Re-188, a daughter product of W-18 8, for such research reactors like BNCT (boron neutron capture purposes. The W-188/Re-188 would make up a genera- therapy) can be envisioned in the near future. tor that can provide a sufficient amount of activity for different applications. The most important would be the Methods intra-arterial irradiation for the inhibition of restenosis after percutaneous transluminal coronary angioplasty. Other The average dose from ionising radiation to the gen- possible applications concern the treatment of painful bone eral public in Belgium is about 3.7 mSv per year. From metastasis, and radiosynoviorthesis. An international con- this, 2.6 mSv comes from natural sources, 1.37 mSv from sortium has been set up for the introduction of Re-188 in medical applications, and less than only 0.05 mSv from the medical world. This includes hospitals for the clini- other man-made sources such as nuclear energy produc- cal trials, radiopharmaceutical research centres to pro- tion. So when determining risk to the average popula- duce the required carriers, and our research centre for the tion, the medical exposures are much more important production of the generator. Since Re-188 is fairly new than the nuclear power plants. Still, radiation protec- to the hospital environment, our centre is also involved in tion is far less developed in the medical world than it is setting up the radiation protection side of the project. A in nuclear power plants and research centres. There is a great amount of experience is present to solve problems general lack of knowledge and awareness in hospitals, of handling and transportation of radioactiye materials, both by patients and by doctors. The implementation of of measuring activities and doses, and particularly in help- ALARA in the medical field has been almost non-exist- ing to develop a reasonable policy of risk governance ent up until now, while the highest dose at work occurs with ALARA tools and culture development. in medical facilities and deterministic effects are reported on patients in intervention radiology. A second project concerns the optimisation and inventarisation of doses to workers and patients in radio- The most common effects are from the diagnosis. X-ray investigations are very common and examinations of patients with X-rays. In Belgium are conducted at several different places and with many everybody has on average 1.1 radiodiagnostic different machines. The exact regulation and geometry investigations each year. Although QA is slowly of the machines can differ substantially from hospital to finding its way in the largest hospitals, patient doses hospital, just as the maintenance and the QA prescriptions. continue to rise due to the rising number of high dose This causes the doses to the patients to differ with a factor investigations like CT-scans. In the investigation of of 50 from one place to another, without a gain in image dynamical phenomena, fluoroscopy is used during quality. The dose to patients moreover is only surgical interventions. These kinds of interventions approximately known. No large dose scale evaluations give very high doses to the patient, which must be have been done in Belgium. Using their dose measuring weighed against the advantages derived from them. experience the Nuclear Research Centre could be helpful The surgeons themselves can also receive doses above in the numerous investigations needed to manage medical the limits on their hands for example. exposure as low as reasonably achievable. Radiopharmaceuticals are primarily used for di- agnostic purposes. At present, about 60 different phar- maceuticals are used, the most important being Tc-99m. Conclusions Another technique that is gaining applications is the By helping with training, dosimetry facilities, and PET-scan (Positron Emitted Tomography). Such a scan ALARA tools and research, the SCK could finally can give high doses to the patient. The therapeutic use contribute to the medical sector to avoid similar risk of radiopharmaceuticals generally concerns the suc- perception problems that the nuclear energy sector meets.

741 DUAL PRESENTATION SK01K0030

OVERVIEW OF SOME PROJECTS OF SNPS FOR GLOBAL SPACE COMMUNICATION C . E. Ivanov, V.j&hitaykin, V.Ionkin,A. Dubinin, A. Pyshko

State Scientific Center of Russian Federation "The Institute for Physics and Power Engineering", named after academician A.I.Leypunsky Bondarenko Sq. I, Obninsk, Kaluga reg., Russia, 249020 tel: 7 (08439) 98504, e-mail: [email protected]

Modern concepts of the application of power - space power unit must be able to maintain the technology in space believe in using an onboard source dual - mode regime of vehicle operation (self - of energy for maintenance of self-transportation of the transportation and long life in GEO); vehicle into working geosynchronous orbit (GEO). This - nuclear rector of unit must be safety and it must can result in reduction of the cost of all installation be designed in such a way that it will ensure mini- and the number of rocket starts. mum size of the complete system; Here we present several conceptual projects of - the elements of the considered technology can be nuclear power installations. Power units developed with used for the creation of NPPI and with other IPPE's participation are intended for long term supply sources of heat (for example, radioisotope); of electricity to the vehicle in GEO and to maintain - the degree of technical and technological readi- self-transportation of capabilities. ness of units of the thermal and power circuit of Considered units include a tiny, fast neutron installation is estimated to be high and is defined nuclear reactor, radiation shield, control system, by a number of technological developments in air, cooling system (radiator) and power conversion system. space and nuclear branches; Additionally, they must also include a rocket engine to - nuclear reactor and heat transfer equipment should maintain the dual-mode regime. This may be an work in a normal mode, which can be very reli- electric - stationary ion jet (with Xe as working body) ably confirmed for other high-temperature nuclear and a thermal engine such as hybrid or NRE. systems; There are three more prospective systems as - if used, the gaseous turbine must be designed not follows: only for nuclear application; - radiation shielding criteria are to comply with re- - gas cooled nuclear reactor with hybrid thermal quirements for allowable levels of neutron and engine and machine power converter; photon irradiation on an instrument compartment - nuclear reactor cooled by liquid metal and with a of a space vehicle (SV) fluence of neutrons with thermoelectric power generating system; E> 0.1 MeV -10 neutr/cm ; the absorbed dose - nuclear reactor with Li cooling and a thermionic of photons - 10 R. and thermoelectric power generator on board. All considered schemes have two thermal circuits The systematic approach to configuration of the and use thermal pipes in the radiator. Two final concepts reactor - power generating block and application of are based on a Xe electric jet; these differ from first some other approaches have allowed us to supply concept by having a lower working body mass but a acceptable weights of the entire system, including longer transportation period from basic orbit into GEO. weight of the reactor, shield and systems of heat The choice of a concept must fit strong conversion. The application of technologies taking requirements such as: place in a high degree of readiness allows us to plan the creation of such an installation in the near fiiture at - space nuclear power unit is aimed to be used in a a rather low cost. powerful mission;

MUCIEAR TECMMOGVII175 SK01K0031 IODOBENZAMIDES: A POTENTIAL IMAGE AGENT FOR MALIGNANT MELANOMA DETECTION

Luppi Berlanga I.S., Argiielles M.G., Torres E.A.

Centro Atdmico Ezeiza - Comisi6n Nacional de Energia At6mica Av. del Libertador 8250 - Buenos Aires - C.P. 1420 Repiiblica Argentina Introduction: purity was evaluated and also was considered as a poten- tial separation method because of the good retention times The aggressive nature of the malignant melanoma of the products. Labelling efficiency was compared. Non makes early detection a fact of great importance. In the isotopic methods showed poor results. The iodination melt early 90!s the first results using aminoaromatic compounds a method was chosen. In order to improve it, some varia- labelled with' I were published. Our Institution had been tions were studied: 150 °C*and 5 mg (NH<) SO / mg gathering experience on production and quality control 2 4 125 I3I benzamide were the best conditions for the process. Al- of I and I labelled compounds. Therefore, we have most 100% yield was achieved. Biodistribution studies the necessary infrastructure and well-trained staff in this (Table II and III) showed brain uptake to be more than subject. As the cyclotron is started, very soon we would 1% of the injected dose 15 minutes post injection. be ready to produce 123I regularly. Thus, it is possible to produce worthy SPECT compounds of I23I. Because of Table I. Melting Points that, we synthesize I25I and I3II labelled benzamides. Substance Melting Point Materials and methods: IDAB 72-73 °C The synthesis of benzamides was based on Brandau's BrDAB 75-77 °C method. A modification of it was made in order to synthe- IPAB decompose 81 °C size N-(2-diethylaminoethyl)-4-iodo-benzamide (IDAB) BrPAB decompose 85 °C and N-(2-piperidinylaminoethyl)-4-iodobenzamide (IPAB). Correspondient bromides derivatives were synthesized for Table 11. Biodistribution of *IDAB non-isotopic labelling methods (BrPAB, BrDAB). The 4- TIME (minutes) halobenzoyl chloride dissolved in dioxane/triethylamine reacts with 2-diethylaminoethylamine or 1 -(2-aminoethyl)- Tissue 5 10 15 30 piperidine. An alkali solution was added to the reaction Brain 2,5+0,2 2,2+0,2 1,4+0,4 1,0+0,1 product and then extracted with dichloromethane. The prod- L+S 13,1+2,5 14,4+1,9 14,2+1,5 12,2+0,6 uct was characterized by: melting point; FTIR, UV-VIS Urine 6,2+0,8 12,3+0,7 19,6+2,2 29,1+0,2 and H'RMN spectra, 200 MHz; TLC (aluminium sheets Gastr 15,2+1,2 17,1+1,0 18,0+1,1 19,1+0,6 silicagel 60) and HPLC (isocratic conditions, reverse phase Lung 3,5+0,4 3,9+0,4 3,0+0,8 2,4+0,1 column, photo diode array detector). A new benzamide synthesis method was studied. This one consist in using a L+S: Live+Spleen; Gastr.: Gastrointestinal coupling agent, such as DCC, in the reaction of the acid and the amine. The labelling was carried out with 125I and Table II. Biodistribution of *IPAB 131 1. The first one was supplied by New England Nuc (17 TIME (minutes) Ci/mg specific activity) no carrier added.l311 was produced as Nal in RA-3 reactor (Centro Atomico Ezeiza) by irra- Tissue 5 10 15 30 diation of TeO via ""Te^n.y) 131I nuclear reaction ( 10 Ci/ 2 Brain 2,5±0,2 1,8+0,3 l,4±0,2 mg specific activity). Several methods of iodination via u±o,i L+S 18,4+1,2 17,-9+0,9 18,3+1,2 isotopic and non isotopic exchange were studied: melt 16,2+0,6 Urine 9,5+0,7 11,5+0,9 16,2+0,8 method, metallic catalyzed method and phase transfer cata- 25,1+0,4 Gastr. 19,7+1,0 20,8+1,2 21,3+0,4 lyst method. Radiochemical purity was determined by TLC 22,9±1,6 Lung 2,5+0,1 2,4+0,2 1,8+0,3 using ethanol as the eluent. Two kinds of in vivo studies 1,7+0,3 were carried out: DL 50 and biodistribution. In both cases L+S: Live+Spleen; Gastr.: Gastrointestinal NIH mice were utilized. Conclusions:

Results: Benzamides have been obtained and characterised by spectroscopic and chromatograflc methods. Several Melting points are shown in Table I. IR and NMR 1311 labelling methods have been tested. The results of spectra have a good relationship with bibliographic data. studies in mice demonstrate the brain affinity of the com- After studying several chromatographic systems for puri- pounds. Biological behaviour makes evident that the fication, preparative scale were chosen instead of expected biological affinity has been reached. dichloromethane extraction. By means of isocratic HPLC 7fi I ORAL PRESENTATIONS SK01K0032 THE ROLE OF COMPUTER SIMULATION IN NUCLEAR TECHNOLOGY DEVELOPMENT

M. Yu. Tikhonchev, G.A. Shimansky, E.E. Lebedeva, V.V. Lichadeev, D.K. Ryazanov andA.1. Tellin

State Scientific Centre of Russia "Research Institute of Atomic Reactors" Dimitrovgrad, Ulyanovsk region, Russia

The second half of the twentieth century will come • Optimization of technical and economic parameters into history as a time of fast development of scientific of acting nuclear plant technologies. These technologies are widely applied • Planning and support of reactor experiments in practically all phases of human activity. • Research and design new devices and technologies • Design and development of "simulators" for oper- One of the important achievements of modern ating personnel training science is the appearance and wide application of computer simulation. Computer simulation is a powerful tool for scientific research. It has opened a Among marked applications, the following aspects new approach to research, given a new qualitative push of computer simulation are discussed in the report: for science and scientific technology development. • Neutron-physical, thermal and hydrodynamics In the report, the role and purpose of computer models simulation in nuclear technology development is • Simulation of isotope structure change and dam- discussed. The authors consider such applications of age dose accumulation for materials under computer simulation as: irradiation •Simulation of reactor control structures •Nuclear safety research

NUCLEAR TEGHNOLBGYII177 SK01K0033 OPTIMISATION OF TEST AND MAINTENANCE BASED ON PROBABILISTIC METHODS

M. Cepin

«Josef Stefan» Institute, Reactor Engineering Division, Jamova 39, Ljubljana, Slovenia [email protected], http://www2.ijs.si/~~cepin Introduction Algorithms for calculation of optimal equipment outages arrangement are faced with an extremely high Probabilistic safety assessment (PSA) is a number of possible configurations. Therefore special standardised method, which integrates fault tree and attention has to be placed to differ the local minimum event treo analysis. Both, fault tree analysis and event from a global one. tree analysis are widely used methods, which standalone or integrated together into probabilistic safety assessment serve well for a number of Conclusions applications connected with improvement of nuclear It has been recognised that the dynamic fault tree power plants safety and systems reliability [1,2,3]. is a useful method for evaluation of arrangement of Because the classical fault tree is a static tool, it equipment outages. has to be extended with the time, to become capable to The results of considered examples have shown evaluate arrangements of safety equipment outages. that there may exist more or even many equipment This paper presents the dynamic fault tree, which arrangements, of which differences in system extends the classical fault tree with time. A unavailabi lity among them may be neglected. The most mathematical model of the dynamic fault tree is important result of the analysis is not to select the most described, example calculations are shown and their suitable arrangement of equipment outages among results are discussed. those with similarly low unavailability, but it is to prevent such equipment arrangements with high Method unavailabilities. At the moment the optimisation algorithm is based The dynamic fault tree is represented by a set of only on evaluation of minimal unavailability over equations, which include time requirements defined in selected time interval. The future work may be focused the house event matrix. House events matrix is a to upgrade the algorithm with the constraints penalising representation of house events switched on and off short and high unavailability peaks. through the discrete points of time. It includes house events, which timely switch on and off parts of the fault tree in accordance with the status of the plant References configuration. Basic event time dependent failure 1. M. Cepin, A. Gomez Cobo, S. Martorell, P. probabilities are calculated as a function of several Samanta, Methods For Testing And Maintenance of parameters. Timing is introduced by outage placement Safety Related Equipment: Examples from an IAEA time for equipment modelled in basic event r (Tpr). Research Project, Proceedings of ESREL99: Safety and Time dependent top event probability is calculated by Reliability, 1999, Vol. 1, pp. 247-251 the fault tree evaluation. Optimal arrangement of components outages is determined on base of 2. M. Cepin, B. Mavko, Probabilistic Safety As- minimisation of mean system unavailability. sessment Improves Surveillance Requirements in Tech- nical Specifications, Reliability Engineering and Sys- Results tems Safety, 1997, Vol. 56, pp. 69-77 3. M. Cepin, B. Mavko, Probabilistic Safety As- Examples of two and ten components are exam- sessment Improves Technical Specifications, Interna- ined. System unavailabilities are calculated as a func- tional Topical Meeting on PSA, PSA96, Proceedings, tion of outage placement times Tpr. The values of Park City, Utah, September 29 - October 3,1996, Vol. parameters Tpr (r C\ {1 ..R]; R... number of compo- I, pages 385-392 nents considered) at minimal system unavailability give the optimal arrangement of outages of equipment.

ENVIRONMENT & SAFETY 12) SK01K0034 HAEA NEPO TOOLS USED IN NUCLEAR EMERGENCY RESPONSE

Kristof Horvath

Hungarian Atomic Energy Authority, Hungary

Abstract There are two software packages for source term evaluation: InterRAS developed by the IAEA and At first .1 brief overview of the Hungarian Nuclear SESAME developed by the French IPSN Institute. The Emergency Response System and the tasks of the InterRAS is fast estimator software, which requires little participating organisations are given. Later the information. Therefore this software is used in the first emergency centre of the Hungarian Atomic Energy phase of an emergency, wh. i the source term is needed Authority and the tools used in accidental situations is to send to the EIC as soon as possible. introduced. Finally, (he scenario and the results of the INEX2-HUN nuclear exercise are describee!. The SESAME is more sophisticated software, which requires more than hundred data every minute; Discussion therefore it is on-line connected to the VITA system. The SESAME software has five modules: acquisition, The principal organisation of the off-site diagnosis and prognosis, break size evaluator, time of Hungarian Nuclear Emergency Response System is (he core uncovery estimator, source term evaluator. The Governmental Committee of Nuclear Emergency source term evaluation is divided into two parts, based Preparedness. The decisions of the committee are on the type of the accident (LOCA or SGTR case). supported by the Information Centre (EIC) and the The first international challenge for the CERTA Emergency Centre (CERTA) of the Hungarian Atomic was (he INEX2-HUN exercise. In this exercise, besides Energy Authority (HAEA). The EIC is responsible the Hungarian emergency organisations, more than 30 for dose calculations based on the source term and the international organisations took part. All hardware and meteorological data. The CERTA is responsible for software tools were used during the exercise like in a plant state assessment, source term evaluation, real situation. The initiator of the exercise was a answering the questions of international experts, and collector cover that lifted off. The failures of the main providing the IAEA and the bilateral countries with gate valve and one of the safety valves belonging to expert information. the injured steam generator made the event more severe. CERTA is home of the HAEA Nuclear Emergency Direct release occurred into the environment from the Preparedness Organisation. Three groups work in the primary system through the open safety valve. The CERTA: analysing group, information group and plant state could be assessed by evaluating the trends logistic group. The emergency leader coordinates the of some significant parameters, for example primary work of the groups. The information group has to and secondary pressure and temperature, the water provide expert information and answer international levels of the injured and the intact steam generators. questions. The logistic group is responsible for Because of the continuous loss of the primary coolant, communication, catering, etc. The task of plant state the core first uncovered and later core-melt occurred. assessment and source term evaluation belongs to the Based on the support of the CERTA and the EIC, the analysing group. governmental committee decided on some countermeasures (iodine blocking, sheltering, To fulfill these tasks there are tools installed and evacuating) in the surrounding villages. information comes from the plant. The information can come through MLLN line, phone or fax. There is also a site observer delegated to the on-site emergency Conclusions team to provide the analysing group further information The appropriate tools made it possible to evaluate if required. the state of the plant and tn'e source term accurately. The most significant program is an on-line data Due to sufficient working of the groups, all the transfer software, which provides the expertr. with more questions were answered, reports sent, and decision lhan 500 data from each of the units of the NPP every made on time during the exercise. Based on the success ten seconds and displays them at the centre. In of the INEX2-HUN and the national exercises, the exercises, the software can connect to the full scope suitability of the Hungarian emergency preparedness simulator of the NPP. The software is called CERTA was proven. VITA and was developed by the HAERI.

821 BRAl PRESENTATIONS SK01K0035 ESTIMATION OF RADIOLOGICAL CONSEQUENSES FROM ACCIDENTAL IODINE RELEASES AT NUCLEAR POWER PLANTS

Vladimir V. Drozdovitch

Institute of Power Engineering Problems, National Academy of Sciences of Belarus Sosny, Minsk, 220109, Belarus

Radiation protection of the population is one of prognosis of population exposure and risk of exposure the key problems in nuclear power engineering. The with consideration of countermeasures. Republic of Belarus has no nuclear power engineering, To evaluate the radiological consequences of a however NPPs are located in close vicinity to the beyond design-basis accident at an NPP for a WWER- Republic's borders in Lithuania, the Russian Federation 1000 reactor, a preliminary investigation was and the Ukraine. The territory of Belarus enters a 30- conducted. The scenario of the hypothetical accident km or 100-km zone of the Ignalinskaya, Rovenskaya, was a fast loss of coolant as a result of a rupture in the Smolenskaya and Chernobylskaya NPPs. Accidents at primary circuit pipeline with diameter of 850mm and one of these NPPs can lead to wide-scale radioactive minimum radiation consequences when containment contamination of the territory and potential exposure integrity is preserved. of the population of Belarus living tens to hundreds of kilometers from the site. Using the software code COSYMA and accounting for radioiodine activity release to the Thyroid exposure from accidental iodine releases atmosphere at the early stage of the accident, the from an NPP is one of the main concerns of radiation radionuclide concentration in air and ground deposition danger for the population. The Chernobyl accident densities were estimated. Using modern knowledge has demonstrated how harmful is a "beyond design- about the risk of radiation-induced cancer, the basis" accident when all protective barriers against radiological danger of radionuclides released is radioactive releases are destroyed. The early stage of evaluated fordifferent exposure pathways: (a) external the accident is the most dangerous. It is characterized exposure from radionuclides in air; (b) external by the incidence, absence of complete information on exposure from radionuclides deposited on the ground the accident type, characteristics of the radioactive surface; (c) internal exposures from inhalation of release, and time deficit for decision-making. radionuclides in air; and (d) internal exposure from For preliminary development of the radiation ingestion of radionuclides with contaminated protection plans for a population and of the decision- foodstuffs. Risk was estimated for populations living making required in case of an accident at an NPP, it is at different distances from the source. necessary to simulate the different accidental scenarios and to obtain information about radionuclide releases, Preliminary results show that the methodology and contamination of territories, population exposures, and models can be applied to the evaluation of radiological radiation risks. The given problem has a large scientific danger of an accident at an NPP. Experience from the and practical significance, as the simulation of the Chernobyl accident has shown that the damage from a different accidental scenarios allows the prediction of radiation accident can be vast. Expenditures on radiological consequences of the accidental release of advanced simulations of the different accident scenarios radionuclides, the evaluation of radiation danger from are worth the cost because they allow radiation the accident, and the justification of necessary protection of populations for these low probability countermeasures. Such a simulation also allows the cases, but nevertheless possible accidents at an NPP.

[NYIRQNMEM & SAFETY I S3 EMERGENCY RESPONSE TECHNICAL CENTRE OF THE IPSN

Robert DALLENDRE Nuclear safety engineer INSTITUT DE PROTECTION ET DE SURETE NUCLEAIRE DEPARTEMENT DE PREVENTION ET D'ETUDE DES ACCIDENTS

Service d"Etudes des situations de CRise et d'Informatique BP 6, 92265 FONTENAY-AUX-ROSES CEDEX, France - Te1e"copie : 01.46.54.39.89 - Telephone : 01.46.54.84.40

Due to the potential consequences of an accident - the concerned site via audio-conference system and in a nuclear facility, a national emergency organization telecopies, was constituted in France which has the capacity to - the security panels of the nuclear plant (in the.case implement countermeasures necessary to protect the of a PWR) via networks (direct link to the compu- surrounding population from the consequences of ter of the stricken plant). radioactive releases. The Institute for Nuclear Safety To perform its missions, the CTC, which has to be and Protection (IPSN), the technical support of the both safe (a failure must not prevent the management French nuclear safety authority, provides the technical of crisis situation) and secure, uses multiple support needed in this decision-making process. telecommunication resources to dialogue with partners So, in the event of an accident arising at a nuclear and also mapping computer systems, data bases and facility, the IPSN would set up an Emergency Response software tools; Technical Centre (CTC) at Fontenay-aux-Roses. The - the SESAME system, which gives, during an acci- IPSN's objectives are: dent of a PWR, a calculation method for the diag- - to diagnose the state of the nuclear facility and nosis-prognosis aforesaid, monitor its development, - the CONRAD system, which calculates the atmos- - to prepare prognosis for the evolution of the acci- pheric dispersal of radioactive substances and con- dent and to give an estimation of the associated sequences in the environment in the early phase consequences according to the situation evolution, of an accident, - to estimate the risk of radioactive releases and the - the ASTRAL code, which allows to cope with long- consequences on man and on the environment, lasting situations. mainly on the basis of weather forecasts and on In order to be operational, the IPSN experts the prognosis. regularly undergo training in emergency situation This diagnosis-prognosis approach is build-up with management and participate in exercices organised b> the information on the state of the installation given by : the government authorities.

SK01K0036

84 I ORAL PRESEMAIIDNS SK01K0037 ACCELERATOR-DRIVEN TRANSMUTATION: A HIGH-TECH SOLUTION TO SOME NUCLEAR WASTE PROBLEMS

by Anthony E. Hechanova

Harry Reid Center for Environmental Studies, University of Nevada, Las Vegas

Summary in the U.S. (presuming reprocessing of spent nuclear fuel and waste form optimization would occur first). This paper discusses current technical and pol itical issues regarding the innovative concept of using In June 1991, a 19-member interdisciplinary accelerator-driven transmutation processes for nuclear committee sponsored by the U.S. government felt that waste management. Two complex and related issues any transmutation process would not eliminate the need are addressed. First, the evolution and improvements of for a high-level radioactive waste repository and would the design technologies are identified to indicate that there be less economically attractive than the current once- has been sufficient technological advancement with regard through fuel cycle. They recommended that any to a 1991 scientific peer review to warrant the advent of a research undertaken by the U.S. should be modest. large-scale national research and development program. In defense of the committee, they acknowledged that Second, the economics and politics of the transmutation accelerator-driven transmutation systems were in a far less system are examined to identify non-technical barriers to developed state than those of light-water and liquid-metal the implementation of the program. fission reactors they evaluated. In fact, the system they Transmutation of waste has been historically viewed reviewed was changed so significantly that the accelerator- by nuclear engineers as one of those technologies that is driven transmutation concept was reviewed again in 1998 too good to be true and probably too expensive to be by the Nuclear Engineering Department of the feasible. The concept discussed in the present paper Massachusetts Institute of Technology which included uses neutrons (which result from protons accelerated into some of the original committee reviewers. spallation targets) to transmute the maj or very long-lived The previous design reviewed by the committee hazardous materials such as the radioactive isotopes of used a critical nuclear assembly, a thermal neutron technetium, iodine, neptunium, plutonium, americium, spectrum, liquid fuels, molten salt, and centrifuge and curium. Although not a new concept, accelerator- separations. The current design has evolved significantly driven transmutation technology (ADTT) lead by a team and now uses a subcritical assembly, a fast neutron at Los Alamos National Laboratory (LANL) has made spectrum, liquid lead-bismuth coolant and spallation some significant advances which are discussed in the target, solid fuel, and pyrochemical processing. present paper. As discussed in this paper, the current design has The major attributes of the ADTT concept are: matured to the point that it is now based on engineered (1) it addresses issues of waste management storage concepts as opposed to its 1991 predecessor. There is capacity by efficiently destroying transuranics, much optimism among eight national laboratories and transmuting long-lived fission products to more benign many at research universities that the interfacing of the or stable isotopes, and partitioning of all fission ADTT's complex engineered components is products for optimal disposition, (2) it is reactor-like achievable. Unfortunately, the U.S. Congress did not in scale and function and is able to be economically provide the necessary funding to support the required viable by producing usable energy by destroying research in 1998 and 1999. hazardous components of spent nuclear fuel, (3) its components are based on proven technology, and (4) In 1998, the U.S. Congress did mandate that a the radiotoxicity of the residual material from a roadmap for the development of accelerator transmutation facility after 300 years is tower than direct- transmutation of waste be prepared. The method and disposal of spent nuclear fuel after 100,000 years. results of these investigations will be reviewed. Transmutation is not a new technology, in fact, The success of accelerator-driven transmutation scientific knowledge of transmutation has been around technology will have a great impact on the State of since 1919 when Ernest Rutherford bombarded Nevada more so than any other state because it will nitrogen with alpha particles and converted it to greatly mitigate the hazards of the proposed Yucca hydrogen and oxygen, fn 1980, the U.S. Department Mountain repository. Living in the host state and the of Energy began to look at strategies for high-level communities most likely to be affected by a high-level radioactive waste management, including transmutation radioactive waste repository, Nevadans have shown a of waste to a more benign form, and selected mined remarkable interest in this technology and may be on geologic disposal as the approach for waste isolation the front lines fighting for its funding and success. ENVIRONMENT & SAFETY 115 SK01K0038 KARACHAY LAKE IS THE STORAGE OF THE RADIOACTIVE WASTES UNDER OPEN SKY

Alexey Merkushkin

Ozyorsk Technological Institute of Moscow Physical Engineering Institute, Russia

I. Look into the past extreme phenomena. There is also a real possibility of contamination of groundwater due to migration of Lake Karachay was one of the gloomy symbols of radionuclides. While using Lake Karachay as storage for dangerous ecological situations on Earth. This lake is radioactive wastes, about 5 million m' of industrial solu- situated directly in the area of PA "Mayak," in the tions passed from the lake into the groundwater. A volume central part of the water-parting area of Lakes Ulagach, of polluted underground water was formed under the lake Tatysh, Malaya Nanoga, Kyzyltash, and the Mishelyak with a thickness of about 100 meters and area of about 10 River. Originally, Karachay was a natural square bog km2, swelling hourly. Contaminated water migrated with a of 26.5 hectare, a length of 750 meters and a width of speed of about 100 meters per year. In 1993, the edge of 450 meters. The maximum depth of the Karachay was the polluted water came close to the Mishelyak River. If only 1.25 meters, and the total volume of water in the J further movement of contamination continues, lake was 400,000 m . There is also considerable radionuclides will begin to penetrate into the Mishelyak variation of the water level in the lake, in rainless and Techa rivers. summer seasons the lake often dries up. Since October 1951, PA "Mayak" began to use III. Solving of the problems Lake Karachay for the storage of radioactive wastes, in order to stop the waste discharge into the Techa River. Around 1970, the closing of Lake Karachay was As a result, the water level in the lake rose up and flooded accepted. From 1978 to 1986, the lake was covered 51 hectares in autumn 1962. Between 1963 and 1964 with ground and empty concrete blocks, allowing the measures were taken to decrease the volume of water in mobile bottom sediments to mobilize. Since 1986 this order to reduce migration of radionuclides into the method has been used and includes three stages: groundwater. From 1926 to 1966, the area experienced little rain. The level of water in the Karachay was 1. The covering of the northeastern part of the lake lowered; 2.3 hectares of coast and 2-3 hectares of the and creating partitioned dams. Partition of the lake with lakebed were left exposed. As a result, in spring 1967 dams allows the reduction of the spread radionuclides from about 600 Ci of radioactive bottom sediments were the surface and decreases the effects of extreme cases such spread by the wind from exposed lakebed. as a tornado. This was implemented from 1988 to 1999. As a result of this work, about 60% of mobile bottom sedi- Because of this incident, new measures were taken ments (approximately 70% total quantity of radionuclides) to prevent future accidents. From 1967 to 1971 some were localized. work was carried out to cover the shoals. This caused 2. Complete closure of the lake and technical shores to rise throughout the perimeter of the lake and repair of the territory to "green lawn." This will remove the area was reduced to 36 hectares. Stringent control any possibility of contamination of the air. This is in of water levels was implemented. The lake was fed progress today. clean water to avoid depression of the level below the minima! permissible mark. 3. Localization and cleaning of polluted ground and groundwater at Lake Karachay. II. Today's problem As a consequence of completed work, in 1996 the area of Lake Karachay contained only 13 hectares, and its radiation Presently Lake Karachay contains about 120 was reduced significantly. About 80% of radionuclides are million Ci of p-radionuclides. Among them, about 40% localized in the covered part of the lake. is Sr-90 and 60% is Cs-137. Initially, radionuclides were allocated as follows: approximately 7% in water, Total closure of Lake Karachay is complicated 41% in loams of the bed and 52% in mobile bottom because radioactive wastes are still being dropped into sediments. the lake, although in much smaller quantities. One of the main problems related to Lake Therefore, we see Lake Karachay's problems are to Karachay is the contamination of air and adjoining ter- be solved using a complex approach that includes cessa- ritories with radionuclides. This was caused by the tion radioactive solution dumping, isolation of the lake from wind spreading small drops from the water surface and the atmosphere and aquiferous layers, capture of migrating aerosols containing radioactive substances. Lake radioactive wastes and its isolating in aquiferous zones. Karachay could potentially become the source of But even if Lake Karachay disappears forever from the widespread pollution as of the result of tornadoes or other Earth, problems related to it remain. 86 I ORAL PRESENTATIONS SK01K0039 "KOZLODUY" NPP GEOLOGICAL ENVIRONMENT AS A BARRIER AGAINST RADIONUCLIDE MIGRATION

Dimitar Antonov

PhD student, Geological Institute of the Bulgarian Academy of Sciences "Acad. G. Bonchev" Str., Bl. 24, 1113 Sofia

Abstract VII degree of seismic intensity according to the MSK scale and no active faults have been determined in it. The aim of this report is to present an analysis of Thick Neogene clays (more than 600 meters), some the geological settings along "Kozlodyu" NPP area with sand layers, covered by Quaternary clays and loess from the viewpoint of a natural, protective barrier have been deposited on the top of Jurassic and Creta- against unacceptable radionuclides migration in the ceous sediments. Beyond the range of the river Dan- environment. Possible sources of such migration could ube tributaries, the relief has low segmentation. The be an eventual accident in an active nuclear plant; ra- low Danube terraces covered by loess and the relief dioactive releases from decommissioned Power Units lowerings in the loess cover represent an interest in or from temporary or permanent radwaste repositories. searching for prospective sites. The report is directed mainly to the last case, and espe- cially to the site selection for near surface short lived The positive and negative qualities of the geologi- low and intermediate level (LILW) radwaste repository. cal environment as a host media for LILW disposal is made. Special attention is paid to the properties of loess In Bulgaria a preliminary choice of prospective soils in which "Kozloduy" NPP is built up. sites in different geological formations has been made. The terrains in the region of Kozloduy are among these The main conclusion of the geological settings sites since they offer advantages from the viewpoint of assessment and of the many years monitoring is that the local population reaction, of the hazards related to the "Kozloduy" NPP area offers good possibilities for RAW transport and of the natural conditions. This re- site selection of LILW repository. gion is calm in tectonic aspect. It is characterised by

ENVIROHMEIIT & SAFETY 18/ SK01K0040 MODERNIZATION AND SAFETY IMPROVEMENT PROJECT OF THE NPP V-2 JASLOVSKE BOHUNICE

Vladimir Michal, Blazej Losonsky, Jozef Magdoien

Nuclear Power Plants Research Institute Tmava, Inc., Okruzna 5, 918 64 Trnava, Slovak Republic

Abstract of NPP V-2 Modernization Project will be initiated after a period of gradual safety upgrading of NPP V-1. This contribution deals with the form, present state, Multiple improvements were performed in co-opera- and results of the Nuclear Power Plants Research In- tion with VUJE within the guidelines of NPP V-2 nu- stitute (the Slovak acronym is VUJE - Vyskumny Ustav clear safety, including seismic-hardening in the past Jadrovych Elektrarnf) participation in the NPP V-2 years. Safety Report recommendations on the 10-year Jaslovske Bohunice Modernization and Safety Im- operation of the NPP V-2, including enforcement of provement Project. A short description of VUJE his- the Nuclear Regulatory Authority of the Slovak Re- tory, activity, and results is also presented as well as public, along with significant advancements in safety NPPs Jaslovske Bohunice characterization. improvements, as well as recommendations by IAEA VUJE was established in 1977 and deals with sci- and NPP V-2 personnel, will be implemented in the entific and research needs of nuclear power plants, such Modernization Project. At the present time a as design, construction, commissioning and operation. summarization report, containing NPP V-2 safety prob- The next fields of VUJE activity are, NPP reconstruc- lems, has already been completed. The report reflects tion, NPP personnel training, radioactive waste man- the current state of the plant and includes description agement technology, and NPP decommissioning. The of operational experiences, and forwards recommen- nuclear power plant, Jaslovske" Bohunice, is situated dations by the operator. To the point, this initial report approximately 15 km from the district town of Trnava documents the preparation necessity of the Moderni- in the southwestern region of the Slovak Republic. The zation Project. construction of the first Czechoslovak NPP A-1 began on this site in 1957. The construction of the double- Results unit NPP V-I with VVER-440 (type V-230) reactor began in 1972. The first unit of NPP V-l began op- Organization of safety improvements into various eration in 1978 and the second in 1980. NPPs con- categories and priorities: Accident analysis; Compo- struction on the Bohunice site continued with NPP V- nent integrity; External hazards; Electrical power sup- 2, which has two units with VVER-440 (type V-213) ply; General Internal hazards; Containment Operational reactors. Unit 1 and Unit 2 of NPP V-2 were commis- Safety Issues; Reactor Core System; Instrumentation sioned in 1984 and 1985, respectively. Slovak electric and Control (and additionally more than 80 items) are utility Slovenske elektrarne (SE) is the owner/commis- sited in the paper. sioner of NPP V-2. This NPP is responsible for more than 20% of the total electrical energy production of Discussion and Conclusion SE, making it an essential supporter of the Slovak economy. Preparatory phase of the Modernization and Safety Improvement Project should be completed by the end Methods of this year. The engineering preparation of the project should be finished during the year 2003. Realization The main task of the Modernization and Safety of the specific safety improvements in the NPP V-2 Improvement Project is to maintain the safe, reliable, should be achieved in that time. The main part of the and economical operation of NPP V-2 during its de- modernization task is projected to achieve its goals signed lifetime, including extended life. The main part during the period between years 2002 to 2010.

88 I ORAL PRESENTATIONS SK01K0041

Keynote Address: Sustainable Nuclear Development and Public Confidence A.Gagarinski, Kurchatov Institute, Russia

According to the definition introduced into the limited and all possible energy sources, including world practice by the Brundtland Commission, nuclear ones, should be used. sustainable development is «the development that meets • Development of the «third-world» countries, and aspirations of the present generation without creating additional energy demand, which cannot compromising the ability of future generations to meet be met without nuclear power. their own needs». • Global (and influencing the plans of each country) The level of requirements to nuclear energy, which, need of availability and acceptable costs together from the very beginning, was considered a hazardous with reliability and safety of energy supply, and, technology (and still is a hazardous technology in the consequently, the interest to energy sources' di- society's eyes), considerably exceeds the same level versification in order to eliminate the dependence for other technologies. If other energy technologies still of fossil fuels import. have to develop in order to satisfy sustainable In order that the society could make a reasonable, development requirements, nuclear energy is already objective choice between energy technology options, practically ready to meet them. it should be provided with proper complete, objective Nuclear energy is intended for long-term and large- and accessible information, which could help it to make scale applications, and its accumulated potential, as this choice correctly. Here the possibility of objective well as presently proposed new promising nuclear comparison of all the energy technologies' advantages developments, will form the base for selection of and deficiencies should also be provided. development options and concrete technological The paper considers the ways to solve this strategic solutions for future generations. This would, to some task. Its solution could take a long time (several extent, compensate for these generations the non- decades) and should be properly perceived by the renewable resources, presently consumed without their generation of specialists now starting their career in permission. nuclear science and industry. Now it is a good time for However, to preserve and improve the role of the new generation of nuclear specialists to solve this nuclear power in energy supply, the society should problem - the large-scale NPP development is not yet deliberately accept nuclear technology. needed, there is a large accumulated experience and perspective ideas, and there is enough time to analyze This report discusses the objective preconditions, the problems in detail, propose and prepare the which would lead the world community to acceptance solutions and convince the general public, that these of nuclear energy. The following conditions deserve solutions are correct. And then the next phase of nuclear special emphasis: energy development would be based not only on correct * Demographic growth, resulting in the increase of technical solutions, but also on a favourable social energy demand and promoting the understanding environment. of the fact, that the world energy resources are

COMMUNICATIBH & PUBLIC PERCEPTION 1191 SK01K0042 PUBLIC OPINION SURVEY NUCLEAR ENERGY - THE PRESENT AND THE FUTURE"

Renata Matani and Josip Lebegner Hrvatska elektroprivreda, 10000 , Ulica grada Vukovara 37, Croatia

Ines~Ana Jurkovi Faculty of Electrical Engineering and Computing, 10000 Zagreb, Unska 3, Croatia

Matja Prah ENTECO, 10000 Zagreb, Trg Marka Maruli a, Croatia

Abstract Discussion

As a part of Croatian Nuclear Society Young The survey questions cover several different fields, Generation Network efforts in improving relations with all with a connection to nuclear issues. Among them the general public, a survey on nuclear energy issues are: among a student population (18-22 years old) has been • present and future energy resources planned. The survey, although somewhat modified, is based on a similar one that has been conducted at high • acceptability of different fuel type power plants schools as a part of organized educational program by • environmental protection and global warming the Museum of Technical Sciences and Hrvatska • radioactivity Elektroprivreda. • waste issues Methods • reliable information sources

One of the ways that we can measure public Such a wide range of subjects has been chosen in opinion concerning different energy sources is by polls. order to illustrate which one(s) cause(s) the most public Two faculties of University of Zagreb are included in concern and require(s) further efforts in presenting the this project: Faculty of Electrical Engineering and problem to the general public. Since the survey is Computing and Faculty of Medicine. The total number conceived as a general guidance to future actions of of students undertaking this public opinion poll is Young Generation Network, the results will offer a clear estimated to be around 1000. After the gathering of picture of current public opinion and mark a good all completed forms, statistical analysis will be starting point. It will also be interesting to see the performed. difference between the way of thinking among the students from two different faculties. Results Conclusions Brief result analyses pointed out that most of the questioned individuals are in favor of nuclear power Preliminary results show a positive altitude among plants, but against the radioactive waste repositories. the student population when it comes to nuclear power They are all aware of the fact that Croatia does not plants. As it can be seen from the obtained data have enough of its own energy sources, but most of (preliminary analysis), this opinion is mostly built on them are against coal, oil, or gas powered plants. The student awareness of nuclear power plants as a clean survey is still underway, so we can include as many energy source. students as possible and at the same time raise the credibility of obtained data.

921 ORAL PRESENTATIONS SK01K0043 NUCLEAR POWER AND PUBLIC OPINION

LA. Kazanikov

Moscow State Technical University, Moscow, Russia

S.A. Klykov

Institute of Nuclear Power Engineering, Obninsk, Russia The public opinion on Nuclear Power is not 1) comparison of contamination characteristics from favorable. A purposeful work with public perception chemical and nuclear industries; is necessary. 2) radiation environmental control; One way to create a positive image of the nuclear 3) radiation as permanent natural factor of environ- industry is to improve public radiological education. This ment; challenge can be resolved in the close cooperation with 4) radiation technology contribution to the welfare state school and preschool education. The information of society; about nuclear power should be simple and symbolical. 5) difference between military and peaceful nuclear Another method is «public relations)) itself. energy. People of different occupations consider nuclear A feedback system with the public is needed to power in different ways; therefore, methods of working assess action effectiveness. The system cannot be with the people should be different. realized without the following main elements: Our society can be divided into 4 parts which can • clear, printed address and telephone number in all be called as «target groups». distributed materials; • local branches making public opinion poll in their First group - People from the nuclear industry regions; with special education working at nuclear facilities or related to the industry. Some actions to increase • regular questionnaire design of «target groups» on corporate solidarity should be performed in the group. the nuclear power issues, the questionnaire design Member of the first group can be sources of information should be performed during public actions with- on nuclear power for the people from other groups. out fail. Second group - People working in the fields • press, TV and broadcasting analysis, quantitative connected with nuclear power. and qualitative changes assessment concerned nuclear industry. Third group - People not related to nuclear power Fourth group - The number of this group's or even with negative impression to the industry. This members is the least, but it has strong influence on group is the largest and the work required is the most public opinion. «Greens» and a broad spectrum of difficult. ecological organizations can be included in this group. The main principals of working with this group The information issued by them make negative public can be emphasized: relation to nuclear industry. • Creation of an organization that can lead this work To counteract such actions, the following steps can and can represent the whole nuclear industry. be carried out: • Creation of corporate and emotional symbolics, • measures aimed to harmful and false information well thought out slogans. obviation; • Permanent organization of seminars inside nuclear • taking the lead over opponents in the interpreta- societies on public relation issues. tion of nuclear power occurrences; • Influence on public opinion through any possible • actions having an emotional-positive influence on way: personal contacts, visual forms and mass the public; media. • official disproof of unaudited information or mis- • Actions aimed to external impressions - meetings, information. runs, demonstration, during which leaflets and The desired result of the actions described above other printed material should be distributed. is to make public think that nuclear power does not The following information can be in the leaflets have malicious intent, but it was created by man and and placards: for mankind. GBMMUKiCATIBH & PUBLIC PERCEPTION 1133 SK01K0044 NUCLEAR LITERACY IN LIGHT OF RADON

Istvdn Ldzdr and Istvdn Cziegler

RAD Lauder Laboratory, Budapest, Hungary Abstract Discussion

Since 1992 the RAD Lauder Laboratory has With our nation-wide radon measuring, we want carried out a survey of indoor radon levels all over to find houses with high radon level because these levels Hungary. The co-workers of RAD Lauder were pupils may cause health damage or a greater chance of cancer. and teachers in local schools. More than 50,000 people On average, we can find 20 new houses with high ra- have taken the survey and received detailed information don levels each year, and mitigate them in cooperation on the radon levels in their homes. with the dwellers. However, of greatest importance is that through Methods this radon-survey many people can get direct experience with radioactivity. Getting familiar with The tool of the survey was the CR-39 track- their own radon level they are able to realize that detector. We measured the indoor radon levels at pillow radioactivity is a natural element of their environment, level in three seasons, in three month-long periods. The and that it can be handled well. detectors were assembled in our laboratory in Budapest by our students. The distribution and collection of the Besides, the great size and the precision of our detectors was handled by the local students and teachers data-set makes it possible to use it in different in each settlement. The method of the measurement is interdisciplinary researches, such as drawing parallels simple enough for any pupil, and elementary school between radon levels and house-structures, geology, teacher to understand. Therefore the teachers and the meteorological changes, or analyzing cancer-risk in the pupils of the RAD laboratory went to each settlement low dose radiation. For the latter topic, we found that newly joining the campaign for a short presentation. medium high radon (between 110 and 170 Bq/m3) In these presentations we played together with the local causes lower cancer risk among women younger than pupils, showing them the GM-counter, the ionization 61 years, independent of the type of cancer. chamber and various common radioactive materials. Through these demonstration lessons they could easily Conclusions grasp the use of the detectors and some elements of nuclear literacy. The detectors were evaluated after An aim of RAD Lauder Laboratory is to find the each period. The final "radon-level" ordered to one houses with high radon level and help people reduce flat or house was the yearly average estimated from the radon activity concentrations of their homes to a the seasonal results. healthy level. We can achieve these goals only by teaching Results nuclear literacy and helping people to be able to make their own right decisions in this topic as well. In the last six years RAD Lauder Laboratory has measured more than 16,000 homes with more than Even the most interesting issues of our researches 60,000 track-detectors. From this data set some 11,500 are less important than the real aim: only if all children single-story houses were in villages or suburbs, where can learn about radioactivity like about air-pressure or higher radon levels are expected. The number of temperature, only if all children can understand and settlements involved was 121, from which 85 not fear radioactivity can we claim success. settlements' population was less than 5,000 people. The total population of these villages is about 3 million people (almost one third of the total population of our "People fear only the unknown. " country). Edward Teller During the whole survey, 109 houses were found with radon activity-concentrations higher than 600 Bq/ m>. As the sorting of the data revealed, these homes are mainly in the mountains of Hungary and are single- story village houses without exception.

94 I BRA! PRESENTATIONS SK01K0045 AREN HAS GOING INTO ACTION FOR NUCLEAR PROGRAM IN ROMANIA

Teodor Chirica" and Traian Mauna"

' Societatea Nationala "Nuclearelectrica" SA, 33 Magheru Blvd., Bucharest 1, Romania " Romanian "NUCLEAR ENERGY" Association - AREN, P.O. Box 53, 76900 Bucharest-Magurele, Romania

Abstract 3. Round Tables dedicated to different aspects of the nuclear energy, mainly addressed to the specialists. Romania has been a member of world nuclear power community since December 1996 when the first From the human resources point of view, attention CANDU type nuclear unit became fully in service in was paid to the transfer of nuclear knowledge from the Cernavoda NPP. The nominal power rating of the Unit older generation to the younger generation. is 706 MWe, covering about 10% of country's annual consumption. Now, the major issues of the Romanian Methods and results nuclear sector is to promote and develop the activities regarding completion of the second unit at the At the beginning of every year, AREN releases a Cernavoda site, connected to the natural uranium chain refreshment of its own strategy and issues a yearly Plan. and waste management. For the end of the century and millennium, AREN The Romanian "Nuclear Energy" Association included into its Plan-2000, the following programs: (AREN) operates as a non-governmental and non-profit • EduNucRo 2000; organization (NGO) member of the European Nuclear + Society and has been involved since 1990 in the nuclear • EduNucRo 2000 -initiation of-; field as a professional society and members of Council • Nuclear bridge over the years = NBOYRO-2000. of AREN work as volunteers. These programs make a link with the previous By cooperating with other Romanian sister NGOs, AREN's programs based on accumulated experience AREN developed and monitored specific programs and connected to the Romanian transition. regarding the proper understanding of specific issues in the nuclear field, in order to better inform the decision Conclusions makers and the local authorities of the country. AREN issued many specific documents in order to create a Compared with the prospective of years before public knowledge regarding the nuclear field, sustain 1990 when the Romanian nuclear program included at the continuity of nuclear power in Romania based on least 5 units on Cernavoda site and another 10 units CANDU-type reactors, as well as expose real and sited on the internal rivers, in 1991 the program was consistent information to the main players of Romanian adapted to the reality of financial means to complete decisional stage: the President, Prime Minister, it. The program first concentrated on the completion ministers, political parties leaders, and parliamentarians. of Cernavoda NPP - Unit 1 and now Romania is looking to find financial support to complete the second The main programs developed by AREN are: unit. The remaining units from Cernavoda site are now 1. Nuclear Energy Days, organized in the last in a conservation stage. quarter of the year, having three components: The re-evaluation of the Romanian nuclear • Children's drawing contest, organized into three program induces a lot of disturbance along of nuclear groups by age, including prizes for most signifi- chain of specific education, starting from the dedicated cant drawings. high-school system, through the universities and finally including the post-university training. This situation • AREN's annual award offered to a personality hav- carries away a lot of negative phenomena, the most ing a major contribution to the nuclear field in important being the gap created between generations Romania. by the break in continuity in the transfer of experience • Exhibition of Supporting Member companies of and knowledge to the next generation. AREN, who's purpose was to create proper ad- vertising for AREN, but also to provide public The intention of this paper is to discuss the impact information about the nuclear energy aspects. of the reduction of the Romanian nuclear program on the different categories of people and companies 2. International Nuclear Energy Symposium - involved in nuclear field, including the difficult aspect SIEN, organized every two years. of nuclear brain drain. COMMUNICATION & PUBLIC PERCEPTION 113S SK01K0046 APPLICATION OF MEMETIC ENGINEERING TO THE STRUGGLE FOR PUBLIC ACCEPTANCE

Jeremy Whit lock

Education and Communication Committee, Canadian Nuclear Society1 It is well-recognized that public opposition to public perception, memes involving popular nuclear power is largely based upon an irrational exaggerations of the truth may be considered in this response to real or imagined risks. Efforts to alleviate category. For example, that Chernobyl was a this through education and communication have met horrendous accident caused by foolishness and leading with some success, but inevitably encounter a barrier to death, is a truthful meme spread globally. That that some have called the "Dread Syndrome": a deep Chernobyl caused thousands of deaths, is a mutant form and almost visceral fear initiating an immediate of this meme, as is the misconception that the accident negative response, independent of external stimuli. The could happen again anywhere in the world. response is subconscious, and therefore unlike public As with genetic therapy, the trick is in the concern for many other technologies. A successful delivery - usually achieved by attaching the corrected approach to dealing with this response will likewise gene to a virus capable of entering cells and integrating be unlike that used with many other technologies. its own genetic information with that of the cell's Others have written extensively on the origin and chromosomes. One looks for similar transport evolution of nuclear power's perception problem, as mechanisms in the case of "memetic therapy". well as the various factors affecting popular risk Celebrity endorsements and television shows are perception. This paper proposes that the current perfect examples, but probably unrealistic. Well- popular perception of nuclear power can be publicized statements by respected scientific authorities characterized as a "meme". A meme, as coined by (professors, Nobel laureates, etc.) might suffice. Oxford zoologist Richard Dawkins in his 1976 book The metaphorical application of genetic therapy The Selfish Gene, is a fundamental unit of thought to public acceptance of nuclear power is useful for that drives cultural evolution in the same manner that gaining insight into the dilemma, as well as for solving genes drive biological evolution. Nuclear power's it. We may speak of "causal" versus "susceptible" perception displays classic memetic traits in that it is memes: technophobia and feelings of scientific contagious, replicating, mutable, and has significantly inferiority are precursor memes that make one affected cultural evolution since its first "infection" of susceptible to a host of social perceptions, including the collective public mind. That it is a particularly nuclearphobia. We may think of "weaker" or "down- strong and resilient meme, and therefore one that regulated" memes requiring higher "expression": most experiences increasing rates of growth, is self-evident. college-level students in the US are known to support This paper summarizes the events that created this nuclear power, but incorrectly perceive that they are in meme, and the reasons for its robust and contagious the minority. Anti-nuclear memes are inherently more nature. A method is proposed for increasing nuclear robust than pro-nuclear, since they are based upon fear. power's public acceptance, not by attacking the Unlike with genetics, the potential for individual "nuclearphobia" meme directly, but by employing its contribution to macroscopic change is much greater own ability to survive and replicate in an indirect with memetic replication. Memes spread much faster campaign against itself. An analogy is made to genetic and can affect sociocultural change within days of their engineering ("memetic engineering"), from which first introduction. This is cause for sober thought. strategies may be borrowed . It is suggested that "memetic engineering" has a One example is "meme-splicing" - the insertion higher potential for success than more direct methods of a foreign, but compatible, pro-nuclear meme currently practiced (albeit requiring more patience, amongst existing memes known to possess favourable subtlety, and time for implementation). Examples are replication characteristics. In this context Global given of pro-nuclear memes, currently replicating in Climate Change is discussed, and in particular the various environments, which have the nature of a direct international effort to raise awareness of nuclear "counter-attack", but which have lower potentials for power's potential contribution to this cause. success. The memes that have developed around Much of genetic therapy involves the delivery of certain positive and negative perceptions of the health "corrected" genes into cells known to contain mutated effects of exposure to low doses of radiation will be versions of the same gene. From the viewpoint of discussed in some detail.

1 The views presented here are those of the author. 961 ORAL PRESENTATIONS SK01K0047 Keynote Address: Public Acceptance of Nuclear Power after Chernobyl A.Mikhalevich, National Academy of Science, Belarus

During the 90s the growth in nuclear capacity Three polls on public acceptance on nuclear power throughout the world is essentially slowing down in development were provided in Belarus during 1995- comparison with previous decades. In some countries, 1998, including the regions contaminated after the such as Germany, Austria, Russia, there are nuclear Chernobyl accident. Consideration of the results shows units which are completely or almost completely the general number of supporters of nuclear power constructed but are not put into operation. Non- decreased from 49.5% to 17.2% during this period. synonymous public acceptance of nuclear power in part But at the same time, the number of the supporters of this situation, especially after the Chernobyl NPP increased among the highly educated people, young accident. people and other categories of the population. The Republic of Belarus has suffered from the The program of forming a favorable public Chernobyl accident most of all, as have other countries opinion of nuclear power has been developed in the including Russia and Ukraine. Therefore, public Republic of Belarus. opinion on nuclear power development in Belarus is very important in forming public acceptance in this field all over the world.

COMHUHICAnON & PUBLIC PERCEPTION II199 SK01K0048 COGEMA GIVES ITS COMMUNICATION A NEW IMPETUS: TRANSPARENCY TO CONDUCT A NEW DIALOG

Katherine Grqffin

COGEMA, France

COGEMA launched in November 1999 a mass employees who represent the professional and human public communication campaign and created an diversity of the plant. The second, in 45-second format, Internet site equipped with cameras (webcams) to make presents the questions to which public opinion wants everyone familiar with the COGEMA plant at La answers. These questions are also repeated in the press Hague. This system is designed to serve as a ads. communication policy that is resolutely open and • To ensure that everyone obtains all the answers to their attentive to French public concerns. questions, the TV spots and press ads refer to the • The COGEMA plant at La Hague is often per- website: www.cogemalahague.fr. ceived as a mystery, an occult, and dehumanized Cybernauts can witness live, by means of a dozen world. This communication campaign, entitled webcams, what actually happens in different places at "We have nothing to hide," illustrates COGEMA's COGEMA La Hague: general view of the site, spent determination to inform the citizens in with the fuel unloading installations, storage ponds, Valognes greatest possible transparency and its desire to bring rail terminal, etc. the Group's industrial operations and the workers closer to the public. The gist of this first step in the new dialog that COGEMA wants to establish with public opinion is to The campaign includes TV commercials and press get beyond irrational fears through transparency, and ads. The underlying principle is to work on issues that to show that COGEMA's men and women are fully have made the news. The televised system includes responsible and determined to contribute actively to two films shot at La Hague. The first, lasting 90 the information of the public at large. seconds, consists of interviews and testimonies of

100! ORAL PRESENTATIONS THE ROLE OF INFORMING SOCIETY AND INIERNATIONAL COOPERATION IN IMPROVING THE NUCLEAR "IMAGE"

Yuliya Kazakevich andPolina Bityukova

Ozyorsk Technological Institute of Moscow Physical Engineering Institute, Russia It is well known that there is a negative relation- also acquaints them with the experience of working with ship between Russian society and modern nuclear en- leading Russian scientific schools that are comparable ergy. Upon analysing this situation we may come to to schools in France, Great Britain, USA, Japan, etc. the conclusion that most people have a one-sided point A striking example of the use of international co- of view. Due to the pressure of the "Greens," the mass operation is the building of the divided materials' re- media depicts only the negative sides of the nuclear pository in the South Ural. The project created a unique industry such as nuclear weapons testing and nuclear highly protected repository and was built by Russian industry accidents. Without a doubt, accidents at the specialists in association with American specialists. chemical production plant "Mayak" in 1957, the trag- edy of Chernobyl's NPP, and fear of potential nuclear International contacts have very important goals, war, still prevent people from seeing the positive side such as the carrying out of joint investigations, elabo- of the atomic industry. It is very hard to ease society's ration of projects dealing with both military and peace- fear of atomic energy. Specialists, administrators of ful problems conducted in mutual confidence of the plants, and local administrations are all combining their Western countries and Russia, and consolidation of all efforts to reach this goal. the world's atomic specialists. One of the general goals of the atomic specialists' concord is to dismantle the The Chelyabinsk region is one of the most eco- wall of public distrust. People knowing only the nega- logically damaged regions in Russia. Appreciation of tive side of atomic energy cannot refer to it as friendly. the inherent value of the ecology combined with effec- That is why it is very important to tell people about the tive engineered environmental cleanup is required for positive side of atomic energy, which includes peaceful ecological sanitation. The decision concerning this investigations, the elaboration of medical techniques, problem cannot be made without calling upon some of etc. This information must be found to be simple and the atomic industry's specialists. They understand the understandable. The only way of being successful in danger of accidents produced by atomic energy. This improving the nuclear "image" is by public awareness. is why the specialists of the atomic industry are sig- nificant contributors in the effort to elaborate safety During the last decade in Russian cities dealing with measures, technologies for radioactive waste manage- atomic industry, some success has already been achieved. ment, and the handling of spent nuclear fuel. For example, since 1989 the service of Production Asso- ciation "Mayak" (Ozyorsk) dealing with public aware- While the application of nuclear technologies are ness has been active. Their main tasks include: public managed using accepted radiation standards, such as awareness about the activity of atomic industry plants, maintaining harmless radiation levels in natural and the ecological situation in the regions of specific plants, potable water, the users of radioactive material, such and the process of the work for rehabilitation of industry as for electrical energy production and many others, and nature objectives, etc. The activity of this service did not received the attention of the mass media. Since must be developed in a number of ways. They include scientists in various countries conduct such studies, the the development of relationships with the press, organi- establishment and expansion of international contact sation excursions to plants, public outreach in the form of is very important. Thus, in the Russian Federal Nu- lectures, and an organisation made up of Russian and for- clear Centre (), more than ten international eign radio-ecology scientists that promotes cooperation conferences and symposiums with the world's best sci- in dealing with issues related to atomic energy. entists invited were held. Beginning in 1990, about 100 foreign delegations have visited Snezhinsk. Dur- The modern computer's world-wide web internet, ing such visits, the directions for cooperation were in which most of the plants have their web-sites, the determined and contract projects were conformed. publication of different informative booklets and maga- zines that explain the life of people living in cities with International scientific and technical cooperation atomic industry, and working at atomic plants also in- has materialised by the activity of the International creases public awareness. Science-Technical Centre, intergovernmental and in- ter-departmental conventions, treaties and contracts. We hope that with the success of further interna- The studies of international projects not only allows tional cooperation and the development of communi- the extension of the spectrum of scientific investiga- cation services, the fear facing the atomic industry will tions, including application-oriented investigations, but be overcome. tf m* s, PUBLIC PERCEPTION III 111

SK01K0049 SK01K0050 SYMPATHY FOR THE DEVIL ©: COMMUNICATION INSIDE OUT

Gaston Meskens

SCK/CEN - ENS YGN Belgium

Modern time prides itself on it's strict rational the opportunity to become informed on all aspects of reasoning: helped by modern science and philosophy, the problems. Both parties must also have the ability this has led to the creation of our intelligent society. to open their minds to other's cultural, scientific and The more mankind performs cutting-edge research and societal backgrounds. generates innovative scientific applications, and the The above mentioned "criteria" for good problem inevitable way these impacts on nature influence our solving and communication are far from simple. The daily life, the more we seem to have to cope with new full paper and the presentation will deal with some kinds of problems and questions. On top of that, it is problems and questions which still remain (and obvious that not everyone feels as comfortable with probably will always exist ?) in todays communication all the possible solutions that are evolving. Based on on "hot" nuclear issues : rationalism, one can present simple solutions for issues such as global warming, the apparent uncontrollable population growth, the disposal of radioactive waste Inside : and the concerns regarding the possible applications - Synergy between young and older generations is of genetic manipulation. However, it is clear that the one and only pure rational solution to those kind of essential, when dealing with nuclear issues ; problems doesn't exist, because people are simply not - Even when we are "fighting for the same case", able to keep values related to their personal sphere there will always be different views and percep- out of reasonings which should lead to a solution in tions on certain issues within the nuclear world. function of the common interest, and this is not always These views and perceptions should be fine tuned a bad thing ... before starting to communicate "inside out".

Whether we are trying to solve a local conflict or Out: a world-scale problem, the final solution will always be one or another kind of concensus (even if there is a - Even being "nuclear experts", we don't know it winner "who takes it all" at the end of a conflict, the all. Problems which cannot be solved by pure tech- time of his/her life he or she spent on the conflict is nical reasoning need a transdisciplinary approach. definitively gone ...). Typical to a good and intelligent This makes the problem even more interesting. ..; concensus solution is that, although the solution is not - Effective communication is impossible without the ideal, all parties understand (deep inside themselves) ability to see beyond facades and defences. We that this was the best way - and only the brave ones have to be able to feel empathy for the people we dare to admit this "in public" then. are talking to. A good and intelligent concensus solution requires fair play of all the parties involved in the first place, Questions : and good communication secondly. Communication on a problem can only lead to a positive way forward if - What is the "public opinion" ? Does it reallyexists? those who are well informed and who want to "defend Can it be influenced ? a case" are able to position themselves more "down to - How "neutral" can we be, while discussing or com- earth" - on the same level as those with the concerns municating nuclear issues ? and the prejudices - and are able to think free from all - Do we really want to "solve problems" while com- kinds of self-interest. This last point also counts for municating on nuclear, or do we sometimes just those with the concerns, who, in addition, should have want to see our own idea's confirmed ?

1021 ORAL PRESENTATIONS PUBLIC ACCEPTANCE AND PUBLIC UNDERSTANDING: NEW APPROACHES

Dobroslav Dobak

Foreign Affairs and Public Relations Manager, Bohunice NPPs

1. Bohunice NPPs Experience • International This paper will describe PR activities at Bohunice With description of basic activities and special NPPs. aspects of each of these activities Until 1990, there was no need to provide information to the public. Since 1990, the need for PR 3. Present and future features has been increased. Globalisation of communication, new possibilities The initiation of this kind of acitivity, our first and tools used, the power of mass media. experience, establishment of a department, transfer of Someone or some idea can easily be tainted by knowledge from the west countries, and definition in a media information, even if it is wrong. We, therefore, QA system are described. need to persuade the politicians and the general public of our purpose; since the public votes for politicians, 2. Orientation of activities these are truly the same entity. • Local/Regional • National

SK01K0051

CfiMMIiMCATIOK & PUBUC PERCEPTION II1103 11 SK01K0052 THE ROLE OF THE YOUTH OF OZYORSK IN CREATION OF POSITIVE IMAGE OF NUCLEAR ENERGY IN THE CHELYABINSK REGION

Tatiana Kostareva, Alexei Teslov

Production Association "MAYAK", Ozyorsk, Russia

The town of Ozyorsk is situated in the picturesque Thus, a negative attitude toward factories of the place of South Ural. This is the land of mountains, pure atomic industry and nuclear energy was formed. "Green" lakes and beautiful forests. movements exercise great influence on public opinion. Members of these movements spread negative information The main factory of Ozyorsk is the Production through mass media, different conferences and so on. Association "MAYAK" (PA "MAYAK"). It was founded for aims of Russian defenses in the beginning Evidently one of the main problems of the atomic of the nuclear industry's development. In first years of industry in the region is to orient public opinion to have existence the activity of PA "MAYAK" led to large- a good attitude about objects of the atomic industry scale incidents, causing great harm to the environment and to their development. It is necessary to strive for and to the health of the population of Chelyabinsk benevolence and trust of the population. region. There are two large factories of the nuclear The Youth has to taken an active participation in industry apart from PA "MAYAK" in Chelyabinsk such work. It is possible to organize the work with region. The activity of all these factories causes the pupils, students and young specialists. For example, specification of the ecological situation in the region. we can organize the work in the following ways: A sudden transition from the regime of strong 1) With pupils: Conducting educational games, secrecy to an unmanaged lack of restrictions in the discussions, excursions, competitions, and drawing to circumstances of ecological ignorance of the public, joint actions. supplemented with mass media activity, resulted in degradation of public opinion. Today it is maintained 2) With students and young professionals: that the cause of all ecological problems of the region Conducting excursions, lectures, youth conferences, is PA "MAYAK" and not the great number of factories drawing participation in youth movements, etc. with obsolete technology in metallurgical, chemical, Thus, the young generation is able to lead active and energy industries that are situated in the work with the population for the social-economical Chelyabinsk region. progress of our country.

1041 ORAL PRESENTATIONS SK01K0053

Keynote Address: IAEA's Technical Co-operation Programme and Us Role in Assisting Member States in the Safe Utilbation of Nuclear Power J. Reed, IAEA, Austria

According to its Statute, the IAEA's objectives The Agency is aware of this and therefore acts in are «to seek to accelerate and enlarge the contribution response to the needs of its Members States. The of atomic energy to peace, health and prosperity activities and projects that it assists are expected to throughout the world» and to «ensure, so far as it is be national endeavours, that are supported by the able, that assistance provided by it or at its request or government and counterpart institutions in the country. under its supervision or control is not used in such a National support is considered to be essential to the way as to further any military purpose.» Thus, the success of the projects assisted by the Agency. Agency does more than just promote nuclear power Project planning and preparation are important and nuclear safety; it is more than a just watch-dog aspects to a project's success. The Agency's TC for nuclear activities around the world and it is programme attempts to provide assistance to projects concerned with more than just safeguards. The Agency with well defined objectives, expected results, and is also concerned with many broad areas of nuclear previously identified target beneficiaries with a view applications and provides assistance in these to maximizing impact and success. In recognition of applications to countries from Jamaica to Paraguay, the need to have well designed projects, and where from to Madagascar and from Cyprus to Vietnam, with the purpose of contributing to the there is evidence reflecting insufficient planning, the well-being of the population in its recipient Member Agency provides assistance at the project preparation States. The Agency can and does assist with the stage through various established mechanisms, such introduction and application of technologies that could as pre-project missions, country programming review have a significant impact in sectors as varied as missions and ad hoc advice that is provided to agriculture, human health and industry. In addition to counterparts whenever possible. These mechanisms all this, the technologies that are within the realm of provide opportunities for IAEA staff and local competence of the Agency are not only development- authorities to discuss and prepare plans that are oriented, but also environment-friendly, which makes effective and can be successfully executed. The them even more attractive in a world that has become Agency is thus a participant, a partner, and a aware of the damages that are being caused to the contributor to projects in which it and Member States environment. Nuclear power is obviously one of the share a vital interest. best known environment-friendly technologies for The Agency's TC programme is a most welcome today's energy needs and a solution for the future. mechanism for the transfer of nuclear technology, and The Agency has wide ranging programme related to to developing countries it is certainly the most the introduction and safe and reliable use of nuclear attractive side of the Agency, since it is through this power. programme that thelAEA can contribute to the solution of their problems through the provision of A country can, however, benefit from the transfer know-how, technology and training. of a new technology such as nuclear power if it Corresponds to a real need and if it solves a problem.

WICLEAR PJHGRAMS & TECHNICAL «RATIfl» | W SK01K0054 INDIA'S POWER PROGRAMS AND ITS CONCERN OVER ENVIRONMENTAL SAFETY

G. E. Prasad, JoyMittra andM. S. R. Sarma

Indian Nuclear Society, Shed A, Project Square, Near Niyamak Bhavan (AERB), Anushaktinagar, Mumbai, India - 400 094 233 Abstract effectively converted to obtain fissile U , which then can be used as fuel in a suitable reactor in the third stage. India's need for electrical power is enormous and per capita consumption of power is to be increased at In mis approach, reprocessing of the spent fuel, which least by 10 times to reach the level of the world average. is a vital resource material, obtains increasing significance. 239 233 Thermal power generation faces two-fold problems. Both the fissile materials, Pu and U , thus obtained First, there is scarcity of good quality fuel and second, after reprocessing when used in FBR and Advanced Heavy increasing environmental pollution. India's self reliant, Water Reactor (AHWR) will result in reduction of three stage, "closed-fuel-cycle" nuclear power program Uranium input in the materials management. As an 241 is promising a better solution to the above problems. inherent advantage the Am , which is obtained in the To ensure Radiation Protection and Safety of Radiation spent fuel, is not of proliferation concern and eventually Sources, the Indian Nuclear Power program emphasizes minimizes complication in the waste management. upon design and engineering safety by incorporating necessary safety features in the design, operational safety through a structured training program and Safety and Environmental Protection typically through software packages to handle rare India has maintained an excellent record of safety unsafe events and regulation by complying safety during all through its 140 reactor-years operation. Strict directives. A health survey among the radiation workers adherence to safety norms is the reason for this high indicates that there is no extra threat to the public from safety standard. An independent regulatory board called the nuclear power program. Based on the latest Atomic Energy Regulatory Board (AERB), constantly technology, as available in case of the nuclear power monitors design and engineering safety, operational option, it is quite possible to meet high energy safety and methodology to ensure compliance with requirements with least impact on the environment. established rules and regulations. Detailed health studies indicate that there is no extra risk of cancer or other radiation damage on the biological cell among Indian Power Scenario the radiation workers at the operating power reactors India is a large country, with nearly 1 billion people of this country. Investigations by the team of doctors and a per capita electricity consumption of in the Uranium mining region indicated that none of approximately 350 kWh, even lacking uniform the typical cases are due to radioactivity. distribution in all parts of the country. To reach the world average, this is to be raised at least 10 times. For posing the least threat to the ecological balance and towards fulfilling the environmental issues, like the Coal in this land has a high ash content and its- Kyoto protocol, nuclear energy has a significant role to reserves are quickly diminishing. Apart from the play. For further commitment and safeguarding the emission of CO2! and fly ash in the environment, solid environment reforestation is necessary. To mark this wastes from thermal power stations give four to six measure for environment protection, Nuclear Power times higher radiation compared to the environment of Corporation of India Limited (NPCIL) has planted a nuclear power station. Oil and natural gas power 150,000 trees around the power plants at Kaiga alone plants are costly options, as they are largely import- and also at Narora and Kakrapara nuclear power plants. dependent at present. Considering all the disadvantages of power plants based on fossil fuel and looking at the rich thorium Conclusion deposits in this land, nuclear power appears suitable to In view of what is stated above, nuclear power the Indian context. At present nuclear power in India is becomes the most eligible candidate in terms plant safety, established to be economically viable, reliable and safe. personnel, public and the environment. Controlling pollution is to be inherent with the program. This consciousness was there among our ancestors and well Three Stage Power Program reflected in the oldest document of the world, which India's self-reliant nuclear power program, which says that, "O pure earth, may we utilize your soil well is based on a "closed fuel cycle" has three stages based without causing injury or harm and disturbing any vital on the type of fissilemateria l used in the reactor. In the first elements in you". Based on advanced technology, it will stage natural uranium is used in PHWR. The plutonium, be possible to implement the safety and environmental which is obtained after reprocessing the spent foel of PHWR norms in the future more easily in the case of nuclear is used in the Fast Breeder Reactor, where Thorium can be power plants than the fossil fuel power plants. 1101 ORAL PRESENTATIONS SK01K0055 GLOBALIZATION: PROSPECTS FOR FUTURE INTERNATIONAL COOPERATION

Irina Paula Dinu

CNE PROD, Cernavoda, Romania Abstract environmental impact, like waste management, and to create a new generation of reactor designs. When I say "globalization," I think to that golden This concern also requires international cooperation. beginning when President Eisenhower gave his Convention on Climate Change from Kyoto established historical speech, "Atomic Power for Peace," to the to reduce greenhouse gas emissions by 2008-2012. General Assembly of U.N.O. in 1953. He proposed, International cooperation is a vital part of the for the first time, an international cooperation for world's science and technology program. The sustaining the peaceful application of nuclear energy. responsible transfer of energy technologies will also play Years later, the global nuclear dream was shaken by an important role in international cooperative activities. Chernobyl. Humankind had seen the reverse of There is much on-going collaboration in science and globalization: any lack in project, execution, or energy related fields. These include the Russian- operation of anNPP has global consequences. American Fuel Cell Consortium, the International Still, why globalization? Globalization because Thermonuclear Experimental Reactor Project, and the global urbanization trends are an important factor for Large Hadrort Collider program collaboration. energy planners and this debate is vital for fueling the bigger cities of tomorrow. Discussion Methods Creation of international organizations and Internet technology are well known methods in generating a global There were some efforts to aid international view of nuclear energy. I would like to discuss the other cooperation, such as the creation of the International methods. Companies doing business in the global Atomic Energy Agency (IAEA) in 1957, the highest nuc- marketplace are looking for new solutions to market lear forum. Many other international organizations were specific challenges. In this competition, the nuclear sector created after it. Methods to help globalization include: took its share: in the nuclear market, CO (one of the • Internet technology and its application to internal greenhouse gases) emissions are nonexistent, while networks (to become Intranet technology). electricity production costs are competitive with coal. »Creating an open, competitive international energy All concerns about competition and CO2 emissions market in electricity where energy can be sold and are overshadowed by public concern over safety. This bought like any other product. fact does not mean that nuclear is less safe than other • Promoting the development of clean and safe reli- power generation systems. This only shows that we need, able nuclear energy systems. for correct judgement, a system for comparing the global • Encouraging international collaboration in science environmental impact to various power alternatives. There and technology to avoid duplication and maximize is a desire to create a global approach, also in nuclear global benefits. power: an idea by the International Waste Authority, an IAEA initiative on proliferation resistant reactors and fuel Results cycles, and international standards for new generation re- Since the beginning, IAEA was mediating and actor designs. Now, we turn to science and technology financing the free exchange of information, scientific re-search, which has many challenges to face: the environ- and technical reports, conferences, and visits of expert ment, safety, nuclearwaste disposal, nonproliferation, and delegations. It encouraged research in nuclear reactor fusion energy. Regarding this field, our hope is that fusi- technology, plasma physics, radiation chemistry, biology on energy has the potential to provide an economical and and environment, waste disposal, and many others. Its environmental option that gives a long-term attraction. example was followed by other regional organizations, all having the same purpose: a global overview and an Conclusions opportunity for each country to attain maximum progress in the benefits from nuclear technology. As the world is moving into a new millenium, its energy needs are increasing. Next to nuclear, there is no Development of Intranets and the spread of Internet power generation alternative that creates more concern services will allow an increased connection between because of the many global and public issues that were nuclear companies. There is another project called Next apparent at the end of this millenium. The globalization Generation Internet (NGI), which will connect thousands concept is not an invention of our days, but today, it of teams of researchers spread across the world. needs more enthusiasm and support for the future. In February 1999, the European Union Electricity Directive was implemented. It discussed electricity We sustain a globalization of diversities, meaning market liberalization, an idea that will bring both that there are options that involve many parameters we challenges and opportunities for the nuclear market as have already mentioned. The most important fact is well as for all players in the market. Cost is not the that energy choices of today must not impede the only major factor though. Another one is how to reduce choices of future generations. NUCLEAR PROGRAMS & TECHNICAL COOPERATION I 111 SK01K0056 EXPEDIENCY OF NUCLEAR POWER USE IN RUSSIA

Nikolay S. Babaev

Minatom of Russia, Russian Federation

Nuclear plants incorporated in the Russia's United The competitiveness of NPPs in the wholesale Power System play a significant role, since the powerful energy market compared to fossil-fuel thermal power and economical NPPs like Kola, Leningrad, Kalinin, stations (TPS) is ascertained by relatively low Smolensk, Kursk, Novovoronezh and Balakovo electricity tariffs (on average, 35% lower). determine the structure of high-voltage power Many years of observations of the radiological transmission lines of the Russia's European part. They status of environmental objects in NPPs' vicinity are located in the key points of the power system and including the monitoring of potential changes in this ensure reliable functioning of the country's power status - of gas-aerosol releases and liquid discharges industry as a whole. Being located in crucial points of from NPPs' - allow us to conclude that nowhere has the grid, in the vicinity of state borders, they ensure the radiation situation changed from that which existed electricity export from the wholesale energy market before NPPs' commissioning. Thus, the overall via high-voltage lines to Finland, CIS countries and experience with Russian NPPs' operation demonstrates Baltic states. While producing 13% of electric power their ecological acceptability and safety. in Russia, NPPs supply 37% of electricity to the federal- Call-Russian) wholesale energy and power market (FOREM), and the nuclear share in the electricity export from the wholesale market is equal.

112 [ ORAL PRESEHTATIDHS SK01K0057

NUCLEAR ENERGY IN ARMENIA

Stephan Gevorgyan and Vahe Kharazyan,

Students of Armenian State Engineering University This summary represents an overview of the energy economy, but also its ecology. To provide the population situation in Armenia and, in particular, the nuclear energy of Armenia with electric energy, even for two orthree hours development during the last period of time. a day, the hydropower plants ofthe Sevan-Hrazdan cascade The energy sector of Armenia is one of the most had to operate at their higher level of capacity. Another developed economy branches of the country. The main crucial factor which contributed to the crisis was the almost sources of energy are oil products, natural gas, nuclear complete dried-up ofthe lake Sevan. The drying-up ofthe energy, hydropower, and coal. In the period of 1985- lake could lead to forest mining, extermination of flora and 1988 the consumption of these energy resources varied fauna, as well as to climate change. During the severe between 12-13 million tons per year of oil equivalent. winters of 1992-1995, Armenia was suffering from the Imported energy sources accounted for 96% of the absence of forest, which was an important element used by consumption. During the period 1993-1995 the the people to heat their houses. Throughout the energy consumption dropped to 3 million tons per year. crisis, the economy of Armenia was in deep decline; almost - Electricity in Armenia is produced by three all the industrial enterprises were closed. The only way out thermal, one nuclear, and two major hydroelectric from the crisis for Armenia was to find primary sources of cascades together with a number of small hydro units. energy. In April 1993, the Government ofthe Republic of The total installed capacity is 3558 MW. Armenia decided to reopen the ANPPUnit2. After5years Nuclear energy in Armenia began its development and 6 months of outage, the technical and financial during the late 1960's. Since the republic was not rich in assistance ofthe Russian Federation, over two years and natural reserves of primary energy sources and the only five months of period, make it possible to prepare and restart domestic source of energy was hydro resource, it was Unit 2 ofthe ANPP in November 1995. The considerable decided to build a nuclear power plant in Armenia. The methodical assistance was rendered by the IAEA and the Armenian Nuclear Power Plant (ANPP) Unit 1 was French firm FRAMATOM. Armenian specialists have commissioned in 1996 and Unit 2 in 1980. The design of developed the program of Unit 2 safety upgraded. That the ANPP was developed in 1968-1969 and was based program considered the experience of countries with the on the project of Units 3 and 4 of the Novovoronezh NPP. same reactors, and included the suggestions of various Both units of the plant are equipped with reactors WWER- international organizations as IAEA and WANO. For that 440 (V-270) type, which are also in use in some power program realization, the financial assistance was rendered stations in Russian Federation, Bulgaria, and Slovakia. by the Russian Federation, USA, FRANCE, and EC in the Currently in Armenia, 36% of the total electricity frame of TACIS program. The items of this program have production is nuclear power electricity. For the new been constantly observed and elaborated year by year. The constructed ANPP was necessary to have well educated international assistance ofthe ANPP was not limited by scientists and technical experts. So at the State Engineering the financial support. However, Ihe only time when this University of Armenia the new specially, Nuclear Energy, happened was when the Unit 2 safety upgrading measures was established. At that time the Scientific-Research were being implemented. Many Armenian specialists Institute ofNuclear Power Plants Operation Study was also attended the training courses on die special simulators in opened, as well as several montage-installing and adjusting foreign countries and participated in many international institutions were created. After the ANPP was seminars organized by the IAEA. commissioned and connected to the National Grid, a surplus Annually, the ANPP is producing about 30 % of quantity of electric energy appeared in Armenia, which was the total energy generated m the country. Since its transmitted to the neighbor countries. In 1988, the annual recommissioning on January 1, 2000, the plant has electricity production was about 15 billion kWh. The already produced about 7.4 billion kWh of electric consumption within the republic was about 12bHlionkWh, energy. Well-organized operation ofthe nuclear station with 3 billion being exported. has become possible mostly due to many good The destructive Spitak earthquake of December specialists graduating from the SEU of Armenia and 1988 was the crucial factor for the Government to make most ofthe them had expertise in "Nuclear Energy." the decision to shut-down ANPP, since the industrial Currently, Armenia has the whole spectrum of plant site was located within the seismic active area, and infrastructure for nuclear energy developed. Several the rate of risk of its further operation was high. So, the laws of great importance, concerning the use of nuclear Council of Ministers of the former Soviet Union decreed energy, have been adopted during the past years. In to stop the operation of the ANPP. Unit 1 was shut down . 1993 the Armenian Nuclear Regulatory Authority was on 25 of February .1989, Unit 2 on 18 of March 1989. created. Its functions included the IAEA requirements The ANPP outage had a very negative effect on the related with the ANPP safe operation. energy sector and on the country's economy as a whole. We are sure that the peaceful atom will yet serve Bad times came when the republic regained its much for the boon of mankind. We think that the future independence in 1991 and found itself in an energy of energy all over the world is tightly bound with the blockade, which was damaging not only the republic's safe use of nuclear energy for peaceful purposes. MCIEM NIEUMS &1ECH1CM. CHPE8ATIBN 1113 SK01K0058 NUCLEAR POWER: BENEFITS FOR THE FUTURE Geta Vultur and Cezar Vultur

CNE-PROD CERNAVODA, Maintenance Support Section

Abstract ule for NPP personnel and for the improvement and licensing of operation personnel. This paper explains how nuclear power was implemented in Romania, why Romania chose nuclear • The operation of a nuclear station requires an open energy, and what the impact of building a power plant culture with open discussions and the awareness is on the industry and environment of Romania. of the necessity to perform all work and tasks, meeting all requirements. In the 1960's, Romania started discussions with different partners to cooperate in the development and application of atomic energy for a peaceful purpose. Results and Discussion In 1977, the Romanian Government decided that the ZIRCATEC and AECL, the Canadian fuel CANDU-600 would be the basic unit for its nuclear suppliers, have qualified the Romanian fuel factory in program. The contract between Romania and Canada Pitesti to produce Candu fuel. Two hundred qualified was for 5 units. In 1979, the construction of the first fuel bundles have been produced and will form part of Candu unit started in Cernavoda, on the Danube 160 the initial fuel load for Cernavoda 1. km east of Bucharest. One-third of the initial charge of heavy water for Unit 1 is from Romanian production. Methods Romanian efforts were stepped up by The nuclear program development led to the manufacturing many components within the country, development of the industrial infrastructure that These included nuclear components such as the reactor provided reliable and safe equipment and participation end fittings, nuclear tanks and vessels, and nuclear and coordination among many different areas, fittings. especially in research. The methods used for The Romanian-Canadian specialists did a detailed implementing this program were: analysis of Candu security after the Chernobyl accident. • The Candu reactor used natural uranium as fuel This detailed analysis permitted a realistic evaluation and heavy water as the coolant and moderator. of the Cernavoda unit operation, as well as measures Both the fuel and heavy water can be obtained in to be taken in order to improve it. Such as to continue Romania, at FCN Pitesti and Drobeta Turnu nuclear power research for a better use of fuel existing Severin. in our country like natural uranium and thorium. • The great attention paid to safety matters (e.g. con- tainment, seismic design). Conclusions • Research development in the nuclear area of Ro- Using nuclear power in Romania involved the manian institutes that involved the manufacturing transfer of different ways to obtain power with of many components within the country. minimum costs, clean and non-toxic emissions, the use • The constraint of a new project management sys- of high technology, new jobs, and changing the outlook tem (document control, material control, critical of people. This alternative to produce electrical power path scheduling, lack of detailed planning sched- led to the energetic independence of Romania. ules for work performance) despite many years of community regime and central planning charac- The coming on-line of Unit 1 will be only the first terized by making decisions in higher management step of a project which is certainly the most challenging using dictatorial methods. and largest undertaking existing in Romania, not only in the power sector. The second step will be the • In the first step, training of operation personnel in completion of Unit 2, which is presently kept in Romania followed by training skills in Canadian preservation. Cernavoda is a key step for the NPP's. socioeconomic transition to a new future in Romania • Construction of a new training center and the simu- and Eastern Europe. lator necessary for implementing a training sched-

1141 ORAL PRESENTATIONS SK01K0059

NUCLEAR POWER IN LONG TERM ENERGY STRATEGIES IN MACEDONIA

Andrija Volkanovski

Research Center for Energy and Informatics, Macedonia

Abstract result from them is that nuclear power plants remain as a only viable option for replacing coal-fired power jThe Macedonian Power System was an integral plants and covering increased electricity demand. part of the European power transmission system. At the present time, it works isolated from the main portion Conclusions of the UCPTE network, connected only with the neighboring power systems After independence, the Considering the problems associated with fuel electricity generation in Macedonia was on the level transport, fossil fuels cost, and market instability, pf its needs. The dominant contribution was by the Macedonia is not in a position to use liquid fuels as thermal power plants with about 80%, and the hydro basic energy source for electricity production. power plants with 20% in total electricity supply Construction of oil pipe line does not resolve the energy covering the peaking part of load curves. Nowadays, problem in Macedonia.. In this way, the reduced the electric power system in Macedonia has only 3 production of crude oil in Macedonia will create an fossil fuel thermal power plants: Negotino, Oslomej opportunity for using gas as energy source for heating. and Bitola. Both the electricity production and the very high capacity factors for the lignite fired power plants, Consequently, rejecting the option for using new show that the maximum production possibilities of oil-fired power plants in Macedonia, the only two mines and power plants have already been achieved. options for electricity production available in In addition to that, at the beginning of the 21 -st century, Macedonia are: using coal from domestic production, as a consequence of the depletion of lignite reserves, and using nuclear power plants. The remaining coal Macedonia must start with activities for substitution reserves amount approximately to 7,5 milion tones of of the existing thermal power plants. lignite per year, which satisfies the current Macedonian needs. However, any increase in the rate of utilizing The options that are at our disposal are the the coal reserves in Macedonia by constructing introduction of gas for electricity generation by utilizing additional coal-fired power plants, will reduce the the already built pipeline through Bulgaria, and the available coal reserves. With existing trend of introduction of nuclear power. For the later, in the area exploitation, active reserve's life time is estimated to of Mariovo at the confluence of the river Crna, there is about 20 years. The predicted life of the new coal mines an ideal location for construction of a hydro-nuclear with significantly worse exploitation conditions than complex. This complex, according to some sources, may the existingjs about 20 years, include up to three nuclear plants, each one with 600 MW and two hydro-plants with pumped storage facility. Assuming that the nuclear option will be selected in Macedonia, following the period 2010-2015, the Methods shear in eletricity production will be: 40% from lignite power plants, 40% from nuclear power plants, and 20% IAEA methodologies within Decades, WASP and from hydro, gas and oil power plants together. ENPEP packages are used for this analysis of the iUsing nuclear power plants in Macedonia at the possibilities of involving the nuclear energy production beginning of 21 st\ century, will results into several in next 20 years. advantages: it will prolong the coal lignite exploitation, and will provide a smooth introduction of nuclear Results technology in Macedonia. This is compromise that has an additional environmental advantage, thus bringing I have analyzed several possible scenarios for Macedonia in line with the current environmental trends expansion of the Macedonian electricity system. Final in the world.

HUGIEAR PROGRAMS & TECHNICAL COOPERATION SK01K0060 DEREGULATION OF THE ELECTRIC UTILITY INDUSTRY - IMPLICATIONS FOR NUCLEAR POWER

August Rose Fern

Utility.com, USA Abstract As one of the first states to enact complete deregulation of the electric utility generation market, The deregulation movement sweeping the California is often looked at as a sample case of international electric utility community represents a deregulation implementation. Although it is one of dramatic shift from the traditional business model of the largest states in the U.S., only 3 large investor- utilities. This paper will focus on deregulation in the owned utility (IOU) companies serve most pnited States and the new challenges for nuclear power Californians. The California market was officially plant operators. An overview of the new operating opened to competition on April 1, 1998 and consists models being implemented in the US will lead into a of several key new features. discussion on new economic and operating concerns for nuclear power plant operators. A new corporation, the Independent System Operator (ISO), now operates the entire electrical grid Discussion in the state of California. The ISO is the responsible organization for insuring grid reliability and All electric utility systems consist of three main availability. To this end, the ISO runs competitive components: Generation, the production of kilowatts; markets for Ancillary Services and Real Time Energy Transmission, high-voltage transportation of the power; while maintaining contracts with key generators and Distribution, delivery of the power to customer throughout the state. All generation from the 3 large facilities. The traditional United States utility company IOUs must pass through a new non-profit market house, was vertically integrated, or composed of all three the Power Exchange (PX). The PX runs Day-ahead, components. Since the customers living in the utility's Day-of and Forward markets to determine a market service area had no choice but to receive service from clearing price for each hour based purely on supply the local utility, electric rates were determined through 9nd demand. To encourage competition, the 3 IOUs a formal "Rate Case" proceeding and approved by a in California were instructed to divest a percentage of Public Utility Commission or similar organization generation capacity. charged with balancing the interests of the utility with The changes implemented in the California market the interests of the general public. Utility companies have created several new players in the state. were almost always guaranteed an acceptable rate of Deregulated affiliates of regulated utilities have return on any capital investments made for purpose of purchased generation capacity in the state and operate serving customers, even when this led to the through the PX, bilateral contracts, or alternative construction of excess capacity. marketplaces. Companies formerly discouraged by the difficulties in building new generation in California are The past 25 years have seen the passage of several now considering building new generation capacity to key regulations that have set the stage for the current take advantage of what can be high market clearing changes in the electric utility market. These regulations prices. Another type of company finding a niche in include the Public Utility Regulatory Policy Act of this market are power marketers, companies that buy 1978 (PURPA), the Energy Policy Act of 1992 (EPAct), and sell power to end-use customers with the intention and FERC Orders 888 & 889 in 1996. Additionally, of providing a value-added service to the customer 24 out of 50 states have either passed or considered while making a profit on the margin between purchase legislation to create a competitive retail market. and selling price.

One issue of particular concern to utility companies These market changes force the Nuclear Power is stranded cost recovery. Since many utility companies Plant owner to review all current operating assumptions made large capital invests (such as nuclear power to assess the potential for success in. the new market. plants) under the traditional market model of We have already seen consolidation among nuclear guaranteed capital recovery, the switch to a competitive power plant owners looking to achieve economies of market leaves utility companies with large "stranded" scale. Existing operators face new challenges in costs that may not be recovered in a purely competitive meeting regulatory requirements created for a vertically marketplace. Each state is determining the appropriate integrated environment. Finally, any proposed new method to compensate traditional utility companies construction in the U.S. market must meet an even ihrough the use of a Competitive Transition Charge. stricter set of financial targets. SK01K0061

THE URANIUM INDUSTRY : LONG TERM PLANNING FOR SHOR TERM COMPETITION

Xavier VOTTERO

Uranium Branch COGEMA Velizy France

Abstract Secondly, the supply of other sources of uranium (uranium derived from nuclear weapons, uranium Today, uranium producers face new challenges in produced in CIS countries, ...) involve other risks, terms of both production (new regulatory, mainly related to politics and commercial restrictions. environmental and social constraints) and market conditions (new sources of uranium supply, very low Consequently, competitive uranium supply prices and tough competition). requires not only technical competence but also financial strength and good marketing capabilities in In such a context, long-term planning is not just a order to anticipate long-term market trends, in terms prerequisite to survive in the nuclear fuel cycle industry. of both demand and supply. It also requires taking into In fact, it also contributes to sustaining nuclear account new parameters such as politics, environment, electricity generation facing fierce competition from regulations, etc. other energy sources in increasingly deregulated markets. Conclusions

Discussion Today, a supplier dedicated to the sustainable production of nuclear electricity must manage a broad Firstly, the risk of investing in new mining projects range of long-term risks inherent to the procurement in western countries is growing because, on the one of uranium. hand, of very erratic market conditions and, on the other Taking into account all these parameters in a hand, of increasingly lengthy, complex and context of short-term, fast-changing market is a great unpredictable regulatory conditions. challenge for the future generation.

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1?0! ORAL PRESENTATIONS SK01K0062

FUEL COMPONENT OF ELECTRICITY GENERATION COST FOR THE BN-800 REACTOR WITH MOX FUEL AND URANIUM OXIDE FUEL, INCREASED FUEL BURNUP, AND REMOVAL OF RADIAL BREEDING BLANKET

Anastassia Raskach

Institute of Physics and Power Engineering (IPPE) 1 Bondarenko Square, 249020 Obninsk, Russian Federation

Abstract It was shown that increasing fuel burnup by -2% and removing the radial breeding blanket of the BN- There are two completed design concepts of NPP 800 reactor reduce the fuel component value by about with BN-800 type reactors developed with due regard 24%. Besides, it should be noted that the MOX fuel for enhanced safety requirements. They have been production scale is significantly affected by the MOX- created for the 3rd unit of BeloyarskNPP and for three fueled subassembly cost for the BN-800 reactors. Sac- units of South Ural NPP. Both concepts are proposed rificing that, the fuel component of electricity genera- to use mixed oxide fuel (MOX) based on civil tion for two BN-800 reactors is reduced by about 20%, plutonium. At this moment economical estimations for three reactors - by about 35% and for four reactors - carried out for these projects need to be revised in by more than 40% in comparison with fuel component connection with the changes of economical situation for one reactor. in Russia and the world nuclear market structure. It is also essential to take into account the existing problem The option of using uranium oxide fuel in the BN- of the excess ex-weapons plutonium utilization and the 800 reactor yielded similar allowances to the possibility of using this plutonium to fabricate MOX considered options for the MOX-fueled BN-800 fuel for the BN-800 reactors. reactors (fuel burnup increasing up to 12% and removing of radial blanket). The fuel component for Methods, Results and Discussion uranium option is reduced by more than 50% compared with the fuel component for the design option of the Construction of a MOX fuel production plant BN-800 reactor with MOX fuel. (CompIex-300) and the BN-800 reactors at the site of Beloyarsk and South Ural NPPs began in the early Conclusions . 1980s. Today it has halted from economical difficult lies and the funding problem is still indefinite. Mutual The obtained result proves the assumption of dependence of the BN-800 reactors and Comlex-300 preference to use the uranium core for BN-800 in the complicates the attraction of investments to complete early stage of its operation. In the course of time the construction of just one BN-800 reactor. relation among the fuel components will change in Along with the use of MOX fuel in the BN-800 favor of MOX fuel using the options as a result of the reactor, it has been considered an option to use uranium inevitable uranium price increase. foe! until Complex-300 is put into operation. This in- In prospect, it makes sense to consider a possibility creases its capacity at an adequate level to provide the to use reprocessed uranium that will obviously provide MOX subassembly demand of the BN-800 reactor. a significant reduction of fuel component for the option In the economical studies, the fuel component of of the BN-800 reactor with uranium fuel. electricity generation cost is an essential parameter and Another way to reduce BN-800 fuel component its analyses are very important in making a decision in might be to accumulate isotopes in assemblies that are foe opening mode of the BN-800 reactor operation. posed instead of radial breeding blanket assemblies. The cost of the fuel components of electricity gen- Now only radial breeding blanket replacement with eration for the BN-800 reactor was assessed for op- steel assemblies is under consideration. tions of BN-800 operation with MOX fuel based on Moreover, in evaluating the fuel component it civil and ex-weapons plutonium and with uranium ox-* might be reasonable to regard joint generation, both ide fuel. The initial data included MOX fuel produc- electric and thermal power, as thermal delivery is a fion costs at the Complex-300 with regard to current very big problem for Russia, especially for the regions status of construction works on the plant, modern esti- where BN-800 reactors are planed to be constructed. nates of purchasing prices of nuclear fuel, and changes of technical parameters of the BN-800 reactor. AH tost data are given for the basis date of January 1,1999. ECIMMICS! l?l SK01K0063 EFFICIENCY IMPROVEMENT OF NUCLEAR POWER PLANT OPERATION: THE SIGNIFICANT ROLE OF ADVANCED NUCLEAR FUEL TECHNOLOGIES

\intoine Van de Velde & Friedrich Burtak

Siemens AG / Power Generation Group (KWU) Nuclear Fuel Cycle P.O. Box 3220 D-91050 Erlangen, Germany http://www.siemens.de/kwu

Antoine. VandeVelde@,erl 19.siemens.de

For a number of years the power producers' market expertise; of both the power plant operator and its has been experiencing far-reaching changes world- suppliers is being optimally utilised through jointly wide. Increasing liberalisation of the power markets is initiated developments, resulting in innovative products resulting in fiercer competition in sectors which were with a high level of acceptance. once monopolies for power producers. At Siemens, the developments focusing on the Nuclear power generation in particular is being reduction of fuel cycle costs are currently directed on exposed to high cost reduction pressure. This is being caused by the improved efficiency of fossil-fuelled • further batch average burnup increase, power generating plants and by decreasing nuclear fuel • improvement of fuel reliability, prices. The cost situation is being further burdened by • enlargement of fuel operation margins, the high expenditures for licensing, monitoring and the • improvement of methods for fuel design and core establishment of reserves, and by plant depreciation. analysis. Therefore, cost reduction in nuclear power generation is the common goal of the utilities and their suppliers. These items will be presented in detail in the full paper and illustrated by the global operating experience Nuclear fuel cycle costs account for 25-40% of of Siemens fuel for both PWRs and B WRs. Highlights the total power generation costs. Although the cost of are (i) the replacement of Inconel with Zirkaloy as nuclear fuel fabrication only amounts to around 10% spacer and guide tube material, (ii) the transition to of the fuel costs, advancements in nuclear fuel low-leakage core loads, based on the development of technology are the key factor to cost savings and Gadolinum absorbers, (iii) the integrated code system efficiency enhancement in the entire nuclear fuel cycle. for in-core fuel management and accident analysis, and Further developments in nuclear fuel technology are (iv) the advanced product lines FOCUS and HTP (High therefore focused not only on the reduction of fuel Thermal Performance) for PWRs, and ATRIUM 10 fabrication costs but primarily on the by far greater for BWRs, with a bumup potential of 60 and 50 MWd/ savings potential which can be tapped in the uranium kg HM, respectively. supply sector, and in the management and disposal of spent fuel. Here, the trend in development is clearly As a result, the nuclear fuel cycle costs for a typical being dictated by the more efficient utilisation of fuel, LWR have been reduced during the past decades by so that significant savings can be achieved via the almost US$3 8 million per year. contingent volume effect. The estimated impact of further burnup increases The boundary conditions for these further on the fuel cycle costs is expected to be an additional developments represent a rather complex network of saving of US$ 10-15 million per year. Due to the fact requirements which always combine economic benefit that the fuel will operate closer to design limits, a careful with safety-related aspects. This demands for both fuel approach is required when introducing advanced fuel manufacturers and nuclear power plant operators a features in reload quantities. Trust and co-operation comprehensive knowledge of all technical, economic between the fuel vendors and the utilities is a and licensing details of the entire plant system. The prerequisite for the common success.

1221BBAL PBESENTATIBMS SK01K0064 TECHNICAL AND ECONOMICAL PROBLEMS OF DECOMMISSIONING NUCLEAR POWER PLANTS (NPP)

Mikhail Vaneev

MPEI, Russia

program was elaborated for calculation of the primary ~. . ... , ^ ., ^ , . . , technical and economical indexes of the process This paper will evaluate the technical and , ... , . , , ,™ , .„ . ,...... decommissionin g ocf such type s oef mtNPPs as WR-440, economical indexe s ofc process decommissioning of a , „„,;„ .,...... XIDD A * ui • J*- innn1000 and RBMK-1000 innn , taking into consideration NPP under unstable economic conditions. , ,, . , ..?. „ unstable economical conditions. Now we are Summarv ielaborating on a new module of this program. This module can evaluate the subscription disposal of the m. ,, , ... ^ . _, . waste. To elaborate on this module, we are using The proble m ocf decommissioning a NPmT>P in Russia . , , ' . . . ,,& , „ ., ., . . . . . scientifi tvr c researche s ocf our XTnNPnP department in the and all over the world is becoming very important. ., _. T- ••• .•*.••«. u Almost all of the NPPs in Russia win have completed MoS

ECONOMICS 1123 SK01K0065 THE EVOLUTION OF CANDUO FUEL CYCLES AND THEIR POTENTIAL CONTRIBUTION TO WORLD PEACE

Jeremy Whitlock

Atomic Energy of Canada Ltd. (AECL) Chalk River, Ontario, Canada, KOJ 1J0

The CANDU* reactor is the most versatile The use of slightly-enriched uranium (SEU) can commercial power reactor in the world. It has the improve uranium utilization by 30%, furthering the potential to extend resource utilization significantly, advantage in resource extraction that CANDU reactors to allow countries with developing industrial already have over LWR designs. The use of SEU can infrastructures access to clean and abundant energy, also reduce fuel cycle costs by 20-30%, and allow and to destroy long-lived nuclear waste or surplus greater flexibility in reactor operation that would lead weapons plutonium. These benefits are available by to further cost savings. The quantity of spent fuel choosing from an array of possible fuel cycles. produced would also be reduced by a factor of 2 to 3. The fuel cycle versatility associated with today's The ability to run on low levels of uranium mature CANDU technology has its roots in more enrichment, suggests that CANDU reactors could be prosaic goals faced by Canada's nuclear reactor employed to generate energy from the by-products of designers forty years ago. Technical feasibility, other reactor technologies. It becomes, in effect, a tool operational economy, and fuel-cycle sustainability were for "recycling" the waste products of electricity more important considerations when the CANDU production from other reactor types. reactor was being conceived. The "Recovered Uranium" (RU) fuel cycle This paper describes how several factors, including involves the use of uranium at about 0.9% enrichment, Canada's early focus on heavy-water reactor produced by the reprocessing of conventional LWR technology, limited heavy-industry infrastructure, and fuel. The enrichment over natural U-235 abundance desire for both technological autonomy and energy self- (0.72%) means that all of the benefits of SEU fuel sufficiency, contributed to the creation of the first cycles described above are realized. There is the CANDU reactor in 1962. potential for dramatic savings in fuel cycle costs. As a natural consequence, other countries with The "Direct Use of PWR fuel in CANDU" similar levels of industrial development expressed (DUPIC) concept skips the uranium-separation step of interest in the CANDU design. Their ability to sidestep reprocessing, and imports used LWR fuel directly into the same onerous manufacturing and operational a CANDU reactor core. Only dry re-configuration is requirements as Canada (the need for uranium applied, and volatile fission products are removed. enrichment, a large pressure vessel, and a complex fuel Again the benefits of SEU fueling are achieved, and design) is both a tactical and strategic economic overall efficiency is improved. advantage. The level of technology transfer thus Several fuel cycles involving thorium fuel have permitted was unprecedented in the growing nuclear been developed. Thorium's high abundance in the reactor market, and remains today a key attraction of earth's crust has lead many reactor vendors to study its the technology. application as a "breeder" of reactor fuel. The CANDU With the maturation of the CANDU industry, the reactor, with its on-line refuelling and remarkable unique design features of the now-familiar product - neutron economy, is uniquely suited to the thorium on-line refuelling, high neutron economy, simple fuel option. design, high level of passive and engineered safety - have lead to serious plans for realizing the full potential In addition, advanced fuel-cycle concepts enable of its fuel-cycle versatility. A diverse line of products CANDU reactors to operate in a "waste-burning" can be offered to suit any international market, while at mode, destroying long-lived actinides without creating the same time the natural-uranium "once-through" fuel more at the same time. A simpler form of this concept cycle remains the reference CANDU configuration, and is a CANDU-MOX fuel cycle that burns weapons- remains an important feature for many clients. grade plutonium. AECL is currently cooperating with the US and Russia to test this concept, with potential Several fuel-cycle options currently under near-term application to the disposition of some of the development by AECL are discussed in this paper. world's surplus military plutonium.

® CANDU is a registered trademark of Atomic Energy of Canada Limited (AECL).r»r. «y»r PBIIIC|P» j « SK01K0066

TECHNOLOGY OPTIONS FOR FUTURE RECYCLING

Toshiaki Kikuchi

Mitsubishi Materials Corporation', Japan Summary decontamination purification' for both uranium and plutonium. In the case of utilization of uranium and It goes without saying that recycling of nuclear plutonium for MOX fuel, high decontamination at the material is indispensable, not only for the effective use reprocessing plant is not necessary. of valuable resources but also to reduce the debt which •Further R&D work will be necessary in order to de- we may leave to the next generations. velop practical applications of this process. In particular in Many developments in advanced reprocessing the fields of equipment development to enable continuous technologies have been carried out in several countries process operation including development of process reliabil- to deal with the diversification of nuclear fuels. Also ity, safety examinations (e.g. criticality safety design for many technologies derived from reprocessing or other fuel types of fuel treatments) and evaluation of the safeguards cycle areas have continued to be developed in terms of recycling. Cost effectiveness and waste-free processing approach (some research has already been started in Japan). are increasingly important factors in the applicable of Relevant considerations of fuel recycling an alternate recycling policy. fib realize a proper fuel cycle strategy, the technolo- This paper introduces an example of the studies in gies applicable to many fuel cycle options are important. this field, which has been conducted in Japan and considers One of the problems for recycling relates to the the establishment of effective recycling methodologies uranium recovered from the reprocessing plant. Even taking into account the uncertainty of future policy. in the case of complete plutonium utilization, some of Reprocessing using Crystallization the recovered uranium still remains. As to the conversion technique for UO3 to UF^ for the The application of Crystallization techniques to utilization as nuclear fuel via re-enrichment, there are some reprocessing was first introduced in 1980s. Since then established methods, which are being used or ready for use research has been conducted in several countries, in- (in Japan, a pilot plant test has been carried out). cluding Japan. This technique, using solubility differ- The utilization of recovered uranium depends upon ences between uranium, plutonium, and other elements the contents of a2U and "'U. The232 U daughter nuclides, at reduced temperature, can recover moderately pure 208TVI2Bi that emit high-energy gamma rays, can be puri- uranium nitrate (UNH) from contaminated solutions. fied at the fluorination stage in a conversion process. How- The bench-scale test results show relatively high de- ever because M2U cannot be eliminated, these daughters contamination factors and ideal grain size suitable for will continue to build up. This limits re-use of high burn- sedimentation and filtration which indicates that this tech- up fuel because of low MSU and high M2U content. nique is applicable to any stage of separation processes. In this sense, if the irradiated fuel is low burn-up, re- In addition, this application, combined with solvent ex- cycling via an enrichment plant and a fuel fabrication plant, traction, may also be used in the following applications: which are designed for natural-origin uranium, can be done Example A: with high purification by a combination of crystallization • Rough recovery of uranium by crystallization from a and solvent extraction. In the case of high burn-up (low solution of dissolved nuclear fuel. More than 80% of MSU, high m\J), storage of recovered uranium may be con- uranium will be separated as nitrate (UNH) crystals. sidered a better strategy. The crystallization technique is • Consequent recovery of the remaining uranium and also applicable for low-decontamination reprocessing. plutonium using solvent extraction procedure. Another issue to be considered is die use of de- pleted uranium. In some cases, interim storage or final Example B: storage of spent fuel might be necessary. The depleted • Refining of uranium and plutonium by crystalli- uranium could be effectively used for the spent fuel con- zation after solvent extraction procedure. tainer; using its radiation shielding characteristics to Since the crystallization technique requires no ex- advantages. Research has been conducted in Japan in- tracting solvents and it can treat a high concentration cluding the development of the conversion processes. of uranium and plutonium solution, such as MOX fuel Conclusions or other advanced type fuels, the volumeof liquid waste is expected to be significantly reduced when compared The Crystallization process has many advantageous to the Purex process. Although the solvent extraction characteristics when compared to presently used reprocess- process cannot be eliminated in the above example, it ing technologies and is considered to be a promising candi- has the benefit that the greater part of separation work date for the establishment of innovative nuclear fuel cycles. can be replaced by this simplified process. Many technologies for future recycling options have been Another feature of this technique is flexibility. discussed and proposed and, by the proper combination of Several applications can be considered according to the technologies, the future strategy for the effective use of the situations. One example is an application for 'low- resources can be established in timely manner. ' Author is now working for the International Atomic Energy Agency as a cost free expert 1ZS1OBAL PRESEHTATURS SK01K0067

i ; • • SOME CONDITIONS AND PROSPECTS OF GOING TO A CLOSED FUEL CYCLE IN RUSSIA

Alexey V.Lependin, Vladimir I. Oussanov and Elena V.Lependina

State Scientific Center Institute of Physics and Power Engineering 1 Bondarenko Sq., Obninsk, Kaluga region, 249020, Russia

Nuclear policy in Russia is based on the necessity The authors have considered a model including of closure of the nuclear fuel cycle. At the same time, not only external costs but also total resource the schedule of such a move is not yet defined. expenditures with long-term power development. In the framework of such a method, it can be demonstrated In this study, some conditions and possible time that a closed fuel cycle has some important advantages frames of taking the nuclear fuel cycle of Russia to taking into account not only tasks of the immediate closure are discussed. Naturally, the main condition is future, but power development strategy for the next the revival of the Russian economy wherein nuclear 30-50 years. Under conditions that nuclear capacities power will turn out to be necessary in a number of increase (to 30-50 GW) limitation on cheap uranium Russian regions. The question is whether the closure resources available in Russia will assume a new of nuclear cycle strategy will be implemented in the significance. As the prices at the backend stages of near future or nuclear power will develop based on the nuclear fuel cycle approach the Western Europe level, open fuel cycle over a long period of time? this will also favor going to a closed fuel cycle. More At present, economic circumstances in Russia have severe ecological requirements answering to a formed in such a way that the economics of current sustainable development concept also will make a projects is not favorable to closure of the cycle due to contribution. Closure of the fuel cycle can be high capital investment costs and low fuel component significantly accelerated in the case of implementation of costs, due to a low cost of natural uranium. of a weapons plutonium utilization program. The Ecological analysis performed within the framework factors mentioned above facilitate fairly going to a of external cost models also does not suggest that the closed nuclear fuel cycle in Russia. closed cycle has essential advantages at present, but they are in sight.

flUEl CYCtE CUUEISES1120 SK01K0068

MANAGEMENT OF WASTE FROM FRENCH NUCLEAR FUEL CYCLE : WHAT ARE THE KEY ISSUES?

Vironique LONDRES (1), Richard DO QUANG (2), Philippe FOURNIER (2)

(1) SGN, International Nuclear Division -1 rue des HeYons, 78 182, Saint QuentinYvelines, France (2) COGEMA, 2 rue Paul Dautier, 78 140 Velizy Villacoublay, France Abstract ,.,,.. Depending on their origin and characteristics, Like any other industry, the nuclear industry waste are conditioned as follows: HLW vitrification generates waste. This waste arises in the different (fission product solutions and minor actinides), ILW successive stages of the fuel cycle, including nuclear bituminisation (precipitates from liquid effluent power plants, and its physical and chemical properties treatment), ILW supercompacting (spent fuel structure vary greatly. What is special about it is the radioactivity waste such as hulls & end-pieces, technological waste) it contains. and ILW cementation (technological waste). Waste has to comply with specifications laid down Management of waste generated by spent fuel by the producer and approved by the safety authority. conditioning in nuclear reprocessing facilities, and Waste package production is covered by a Quality which cannot be stored in surface repositories, Assurance system which can be audited by a third according to current French regulations (ILW and party (the safety authority, the repository operator or HLW), is specifically discussed in this paper. the client), with the producer remaining responsible Discussion for the waste throughout the process. The French National Agency for Radioactive Waste Management (the repository operator) and the producers have joined As the nuclear industry has developed, the need to set up a {sequential waste package approval for optimised management of this waste has become process: whereby the producer is charged with a major concern for all those involved. Optimisation is supplying reports detailing the number of packages based on properly designing the installations generating produced and their physical, chemical and radiological waste, minimising arisings of waste & sorting waste at characteristics. the source, and appropriate treatment and conditioning of waste prior to final disposal. The management Waste package behaviour, considered by the process is always continuous from the production site repository operator for the physical design of the to the repository so as to avoid any dispersal of repository and the way it is managed, is assessed in radioactive material. order to determine when, in what form, in what quantity and with what kinetics radionuclides and The best treatment for waste is determined on the other products (gases and degradation products in basis of its properties so as to guarantee the long-term general) are likely to be released and return to the stability of waste conditioned in the appropriate biosphere. matrix, after volume reduction, while complying with the conventional safety criteria for nuclear installations. Knowledge of waste package behaviour is The cost of disposal has led to the development of ways transposed to operational models which aim to of reducing the volume of waste to a minimum. represent the main phenomena influencing the release of radionuclides and degradation products. In France, nuclear waste is divided into 3 categories : Conclusions (i) Low Level Waste (LLW), which is short lived waste, Due to its fuel cycle activities, particularly in the (ii) Intermediate Level Waste (ILW), containing fields of reprocessing and recycling, and the work it long lived alpha emitters, but no significant ther- carries out to determine how volumes of waste can be mal release, often called TRU waste. Much has reduced and identify the long-term behavioural, already been performed, management improve- properties of packages, COGEMA is committed to ments are still to be performed among possible maintaining its involvement in the use and development options, of a nuclear industry which preserves the environment now and for future generations. (iii) High Level Waste (HLW), having long lived radionuclides and significant thermal release.

1311IRAL PHESENTATHHS . SK01K0069 THE FINAL DISPOSAL FACILITY OF SPENT NUCLEARFUEL

Sldvka Prvdkovd andVladimir Nedas

Department of Nuclear Physics and Technology, Faculty of Electrical Engineering and Information Technology, Slovak University of Technology, Ilkbvicova 3,8 i2 19 Bratislava, Slovakia

Nuclear power plants are not operated on the basis Situation in different countries: of fossil fuel and, therefore, they do not produce Slovakia is in an investigation phase of the selec- greenhouse gases. On the other hand, nuclear plants tion of the final disposal site. There is only wet interim are producers of the other form of environmental load: storage in Jaslovske Bohunice. nuclear waste. Due to its radioactivity the spent nuclear fuel has to be isolated from both people and nature. In the Czech Republic, spent fuel is stored in a This problem could be solved by final disposal into dry interim storage of the Dukovany area. The test geological formations. localiry.is situated in Melechovsky bedrock. The final repository is scheduled to be constructed in 2030- The spent fuel has to be disposed of in compliance 2040 [1]. with the obligation issued by EIA. First, it is stored under water to decrease the radioactivity and the heat The best experience with the storing of spent waste production. After this, the fuel rods are transferred into is available in Finland and France. an interim storage. High-level radioactive waste is Finland started research of suitability of the melted with a glass material, sealed in a metal canister Finnish bedrock at the end of the 1970's. The final and stored in a solid form. For the final disposal, this repository is planned for construction in 2010 and for vitrified waste is put into special tight canisters. The commissioning in 2020 [2]. canisters are made of two parts; the outer one is copper and the inner one is modular cast iron. France has more than 25 years of experience in this field. The first disposal site for low-level waste The project of a final disposal facility starts was opened at La Hague in 1969. In December 1993, with an investigation phase after the selection three possible sites were proposed in clay formations of the disposal site and the construction of the and one in granite rock. The underground laboratory investigation shafts is planned to start operation in 2006 [3]. The geological criteria of the final disposal facility Japan is a country with frequent earthquakes and are: volcanic activity and, therefore, it is necessary to - stability of the ground for 10 -10 years optimise the design of final repository. A general - no occurrence of natural resources which might concept of geological disposal started in the 1950's. have significance for future use of the area Disposal of high-level waste is planned to take place around 2040 [4]. - a few hundred meters away from the hydrological cycle of groundwater For the place of final disposal in the USA, Yucca The nuclear waste in the final repository is isolated mountain in Nevada has been chosen. It is planned to by multiple safety barriers: construct a final repository 350 m under ground, and it is designed for 70000 t of radioactive waste. - the solid state of the spent fuel Investigative works began 16 years ago, and the - twofold canisters research bores were finished in 1997 [5]. If the Yuccea - engineering barrier (bentonit) mountain is found unsuitable, studies will be stopped and a new direction will be found by Congress. - several hundred meter deep geological barrier . The project ends with the de,commisioning and Germany, compared to the USA, has stricter laws sealing of the final disposal facility. for spent nuclear fuel treatment. The emphasis is put on the environment. The present considerations of government can lead to the closing of the final FUEL CYCLE CHALLENGES 1131 References: repository in Oorftlten, Changes ia th«-project of liquidationof higMewet wastewere evofeeiByu6« [l uloffSte radioaktivn^ch odpadu, on the stability of the salty mine to Oorleben. The sprava, tTstav jaderndho v^skumu, ft.e2, 1996 (in underground research is not going to continue; regions Gzech) with different geological formations will be [2]Environmental imact assessment programei investigated. The final disposal...,.PosivaQY, Finland, 1997 Today, there is no final disposal facility throughout [3]YKaluzny, Chfitenay-Malabry, Internationale the world. There are only investigation shafts in Zeitschrift f(Sr Kernenergie, No/iarp631,1996 Finland, Belgium, and France. [4] Research deep underground, Tono Frontier Science Research City, Japan, 1998 The preparation and building of final disposal [5JM,K.Petr©#> ftrteraationaie Zeitschrift fflr facilities requires cooperatiottofall countries involved, Keraenergie, No. 1 lv p. 641,1999

1321 URAL PRESENTATIONS THORIUM-FUEL FOR TWENTY FIRST CENTURY

M.S.R. Sarma,

President, SK01K0070 Indian Nuclear Society, Anushakti Nagar, Mumbai-400 094.

Abstract so called non-conventional sources like sun, wind, waves, etc. However, these are in the development stage Per capita energy consumption is used as a and there are limitations on bulk energy production yardstick to measure the development of a country. through these sources as of now. India's per capita energy consumption is only about 4 % that of the United States of America and is about 20 In India, the per capita availability of resources of % of the average world values. India is one of the fast coal, oil, gas, hydro potential and uranium are limited developing, nations in the world, so its energy when compared to the per capita world reserves. consumption is bound to increase. However, a vast reserve of thorium is available in India to produce electric energy. This thorium has little other Amongst all modes of consumption of energy, use unlike coal and oil which are required for industries electric energy utilization is preferred over the other and transportation. In terms of energy output potential, forms of energy due to its convenience of use. Hence the Thorium reserves in India are at least six times that share of electric energy utilization is increasing steadily. of the coal reserves. India has formulated a three stage For example during 1950, in India, 9 % of energy nuclear power program for optimum use of its nuclear consumed was electrical, whereas it became 25 % of fuel resources. This paper discusses this program which the total energy consumed in 1991. Electric energy is is aimed at tapping the large energy potential of our mainly generated by the resources of coal, oil, gas, thorium resources. hydro potential and nuclear. It is also generated from SK01K0071

"SCRAPYARD CHALLENGE"

Adete Rollick

British Nuclear Fuels pic, United Kingdom

'Plutonium'. The word evokes deep reactions Statistics suggest that the quantities of military outside of the nuclear industry. Although the majority Plutonium stockpiles will slowly fall whilst that of Plutonium currently in existence is man-made and produced by civilian uses will rise. Is the fear associated therefore perceived as being unnatural, plutonium has with this potential rise justified and how does the been found as a product of the Oklo natural reactor in industry meet these concerns? Importantly, the Gabon. This paper seeks to challenge two concepts, differences between civil and military Plutonium are that of the Nuclear Control Institute that Plutonium is insignificant when used for nuclear fiiel, thus the overall unnatural, "fiendishly toxic" and one of the "substances quantities can be considered as one stockpile. This most hazardous to man" and the second image that a paper discusses the three current alternatives open to high security Plutonium store is merely a 'scrapyard* us in order to dispel these fears and which of these can containing a material which has little use. The nuclear be seen at the present moment to be the best one for industry has often been accused of treating Plutonium stockpile reduction. and its accumulation casually in proportion to the risks These alternatives demonstrate the opportunities perceived by those outside the industry. As a result this that Plutonium can provide, as compared to Uranium, paper seeks to demonstrate that the industry is aware which is currently used as the largest fuel component of the concerns of the public and is actively seeking in most nuclear reactors world-wide. This paper will viable solutions. The paper looks at Plutonium itself continue the debate about what we can do with and explores the issues surrounding military and civil Plutonium in general - to put it simply, do we bury it Plutonium in adding to the current stockpiles. It also or 'burn' it? suggests three possible alternatives for dealing with these Plutonium stockpiles and arrives at a conclusion The first alternative: Mixed Oxide Fuel (MOx) as to which solutions currently appear most viable. It is widely known that Plutonium can be used in Discussion large quantities in MOx (mixed oxide fuel) to replace current thermal fuel. MOx fuel has been used in the current Light Water Reactors (LWR's) as a partial In dealing with Plutonium, one must consider how loading and can produce more energy at higher burn- hazardous it is and how dangerous it is to handle. Like ups. The success of this venture is such that the many other more hazardous materials, such as intention is to develop the reactors into full Plutonium Hydrogen Fluoride (used extensively in the semi- burners. The ability of MOx to use stockpiles of conductor industry) Plutonium must be handled safely Plutonium and thus remove it as a perceived problem, and correctly. It is handled with care above its was the driver for the MOx plant in the UK and the radiological and chemical danger within current robust reason behind reprocessing within the THORP plant and comprehensive international security and going ahead. Further benefits can be seen world-wide safeguards requirements. This has and is being done when the reduction in Uranium required from mining in many institutions world-wide and this paper is taken into account. A discussion is carried out in discusses the myths propagated by pressure groups this paper of the amounts of Plutonium that can be used about plutonium and the real difficulties in handling by converting it into MOx fuel and the added benefits such a material. of doing so. The paper, as previously mentioned, seeks to The second alternative: Fast Reactors explore the differences between weapons grade Plutonium and civil Plutonium. It also discusses the An alternative well known within the industry is difficulties of converting between the two as well as to use the Plutonium from LWR reactors as the initial the justification for doing so. The perceived ease of charge fuel in fast reactors (FBR's). conversion has been raised by many groups as an issue These require large quantities of Plutonium in the of great concern to the general public. Many anxieties also relate to the quantities of Plutonium that are in fuel in contrast to Uranium fuelled LWR reactor and existence whatever their origin. are less sensitive to the Plutonium isotopic composition. The perception is that FBR's generate Plutonium One possible image that the public may visualise however once the reactor isiuelled and the fuel cycle is the world as a nuclear scrap-yard trying to find a reaches steady state, the Plutonium is effectively disposal route for both civil and military Plutonium. 'immobilised' within it. Furthermore the reactor 134 I ORAL HEHHmin requires an external feed of depleted Uranium of which Utilising these stockpiles however would provide there is a worldwide glut, and the reactor can be energy security for such nations making them more self- engineered to burn Plutonium. Whilst not advocating sufficient. Making nations more secure in their energy nuclear power for all countries to meet their energy supplies has been thought to lead to more stable shortfalls we must consider in those countries with political environments and increased wealth.1 If the stockpiles whether it is more environmentally friendly world were to depend greatly on Uranium in the current or socially responsible to bury it or use it more thermal nuclear reactors, then the world Uranium efficiently? reserves are currently estimated to be depleted someway through the next century. Using the stockpiles Development work on FBR's has shown that they gives a guarantee of sustaining energy supplies for a can burn the fuel for longer periods of time at higher longer period of time. power ratings thus utilising the fuel more efficiently in comparison to current thermal reactors which only use In terms of the more immediate environmental a small proportion of the natural Uranium in the fuel arguments, utilising the stockpiles to produce nuclear before requiring reprocessing. power can be seen within these particular countries to avoid carbon dioxide emissions that would otherwise There has to be an attempt to increase public occur by replacing nuclear requirements with fossil acceptance of seeing spent fuel as useful rather than fuels. This would enable these countries to achieve simply burying it. This means that with FBR's the fuel their Kyoto targets. cycle would be a closed one, containing nuclear power, fuel reprocessing and fabrication in one unit thus Currently several countries are takingsteps along reducing some of the risks perceived. FBR's have been the road of eliminating the use of nuclear power, if so developed to a demonstration stage but are not yet in it would appear that they are advocating a route known commercial operation. The paper shows the challenges as a 'NIMBY policy'. This acronym refers to an that are perceived with FBR's that affect their future attitude that can be defined as "Not In My Back Yard". development, when they might be in commercial Although nuclear power can be eliminated in one operation and the risks associated with FBR's. country, in order to deal with the energy shortfall they The third alternative: Immobilisation are prepared to pay for the import of power produced by nuclear means elsewhere. In the case of countries The third alternative available is to immobilise and such as Sweden, the closure of nuclear power stations bury the stockpiled material. An analysis of the means that the imported energy is more likely to come amounts that can be incorporated into an immobilised from more polluting fossil-fuel energy sources which matrix is also carried out. The reasons for this option simply moves the carbon dioxide burden to currently being the most favoured one are investigated. neighbouring countries. This makes such policies The message that must be conveyed is that once appear ill advised. Currently it is seen as imprudent to immobilisation is carried out, it is irreversible and produce all a country's energy from one source, for contrary to popular belief this may not be a positive example, wind or solar energy, and therefore there approach. In the future we may find that Plutonium is seems to be no reason why nuclear power should not vitally necessary for energy production and it will be used as a clean technology to produce electricity. unfortunately be inaccessible, because in response to current attitudes we have effectively closed an open Conclusions door with a possible route to a solution. Consequences of Disposal We must remember that currently all of the Irretrievable disposal of Plutonium stockpiles alternatives listed are open to us, both those inside and removes the possibility of using it to stabilise the those outside of the nuclear industry. By closing the varying costs of electricity generation due to the door on this opportunity we will have created a finite fluctuations in the world uranium price. Uranium is supply of fuel that like coal and gas will eventually run currently mined as a raw material and mining expansion out. Do we really want to eliminate our options one by could be slowed down whilst public acceptance of the one and leave ourselves with no alternatives? necessity of nuclear power is encouraged thus The paper concludes with the choice of alternatives preventing lengthy and costly debates similar to the that currently appear to be the best ones to deal with Jabiluka mine in Australia where environmental the opportunities and the challenges that Plutonium degradation was a big issue. presents.

FUEL CYCLE EHAUf NBES j 135 SK01K0072 PERFORMANCES OF ACTINIDE TRANSMUTATION BASED ACCELERATOR-DRIVEN SYSTEMS (ADS) AT CIEMAT

Miguel Embid, D. Cano, E. Gonzalez and D. Villamarfn

CIEMAT, Av. Complutense, 22 Edif. 17, 28040 Madrid, Spain

Abstract , , . , ,. The program has two main research lines. The _, p.™- , rtmoT • *. J • A first one is dedicated to the development of concepts, The FACET group at CIEMAT is studying the ,. ... ,. • i *•• .. " i-..K... , ,. .j . . , designs, operation model s andJ computet r simulation properties and potentialitie s off several liquid metal- . , , v. ... _.,. ,. . . ^. . i J .r,r J • /• x. -J J^. • • •• •• tools charactensticsofthis kind orf systems. The second cooled ADS designs for actimde and fission product ,. . , , ...... /,. , , , . - ... r:, •«.*•*• r . line includes the participation and the data analysis of transmutation. The mam characteristics of itthese ^, • . . . • ^ r- 1TJ J .; ,., * , ... . . the most advanced experiment As in the field and systems are the use of lead or lead-bismuth eutectic as .....,-., • . • , •• • , ^ , „ , . , international benchmarks, primary coolant, moderator and fuels made by. transuranics.

1351 HAL PBESEITAntllS iniiiiniiii SK01K0073 DYSNAI: FESTIVAL OF INTERNATIONAL YOUTH NUCLEAR ASSOCIATION

Alexander Bolgarov

Honored President, Taikos 72b-30, Visaginas, Lithuania, bolgar@dkd,ot.lt Abstract procedures) but also with establishing of human contacts, a kind of «human diplomacy)). It's a dialogue JJysnai is a tents camp on the Glade nearby of colleagues divided by state borders but having high wonderful Lithuanian lake, a kind of interesting and professional potential and united by the same job and funny show in the forest that takes place for seven fortune. Such, policy is supported by NPP' summer days each year. Beside the technical reports administration, welcomed by plant's personnel, press one can find a lot of causes for self-expression because and public opinion due to festival's spirit of friendship of spirit of freedom and friendship. Existing for 12 and coexistence. years, the festival provides contacts of youth, which have similar living and work conditions, interests and Discussion problems.

Introduction Participants of Dysnai movement looking for new colleagues may join to the festival. 12-years history For the first time the festival of young nuclear proved that it's useful and interesting. But the festival operators called «Dysnai» was held in 1988. Since that needs new experience and we're ready to discuss moment annually guys and girls that work at nuclear suggestions on events and procedures. power plants (NPP) or live in towns - satellites of NPP The International Youth Nuclear Association arrive at the same place in Lithuania in the first week includes legal (registered) participants from Lithuania, of July. They participate not only as individuals but as Russia and Ukraine but exist itself de-facto not de-jure. teams too-delegations from Leningrad, Kursk, Kolsk, May be it is useful to establish registered organization Beloyarsk, Smolensk, Rovno, Chernobyl, under authority of European Nuclear Society and Zaporozskaja and Ignalina NPP, Institute of Nuclear IAEA. Energy (Obninsk). Conclusions Methods In July 2000 the Xlllth international festival of Putting the festival's camp in order - we mean an young nuclear operators will take place. Any delegation arrangement of each delegation camp, a stage, or person can enjoy it. electrification, etc. - are fulfilled by the teams. The duty team prepares programme for a day - one team Acknoledgements after another. Programme includes technical and creative parts. Beforehand -usually fa the early April Thanks to Michael Laenko, Andrew Runov and - a workshop of youth forum co-ordinators held at one Grigory Zavarin for their assistance in making Dysnai's of the NPP-participant's place. They prepare a web-sites, containing good quality dysnai's songs and programme's package of the future festival, get to know pictures. ihe work of certain NPP, meet with nuclear enterprise and town administration. Such workshops have been References held in Visaginas (former Sniechkus), Desnogorsk, Sosnovy Bor, Kurchatov, Slavutich, Polyarnye Zori, Paul R. Josephson, Red Atom: Russia's Nuclear Zarechny, Kuznetsovsk Power Program from Stalin to Today, New York, Results W.H.Freeman and Company, p.36 (1999) Web-sites: That form of youth policy deals not only with www.iae.lt/mama professional purposes (there are some other www.dvsnai .narod.ru

Yfi SESSIII ] 141 SK01K0074

RADIATION LITERACY FOR FUTURE ECOLOGISTS

D. Belous

St.Petersburg State Institute of Technology, Russia Radiation, education, public understanding

One of the basic problems in the development of nuclear physics, the conception of natural and man- nuclear technologies is the hostility of the population made radiation, measurement methods and finishing and the resistance of some "Greens". Very often, such with the Nuclear Fuel Cycle, biological effects of a situation is the consequence of a lack of information radiation and legal rules. Labs and practical training and technical literacy and leads to fear. This fear acquainted the students with properties of different develops due to incomprehension of radiation nature types of nuclear radiation. They improved students' (no smell, no color, no taste and hypertrophy of the skills in working with dosimeters and radiometers of effects in disaster films) and ignorance of real danger various construction. During the: labs and practical level at the influence of radionuclides on human beings. training, methods of protection from ionized radiation were shown and comparison of measurement results This year the author of this report and the teaching with natural values were made. staff of the Engineering Radioecology and Radiochemical Technology department of The first cycle of studies carried out in the autumn- St.Petersburg State Institute of Technology, which has winter of 1999 demonstrated: a 50-year long experience in handling radionuclides, • students' high interest in learning and wish for has created the complete theoretical and practical knowledge; course for training the students of the Environment Protection' faculty (future ecologists). The author of • changing of students* attitudes toward the object the course had the goal of acquainting the students with of the study (from watchful and wary to reason- the basics of radioecology and radiation safety. The able and adequate). lectures elucidate a wide variety of questions which are necessary for understanding the influence of In the future it is planned to increase the number radiation on the biosphere and the culture of dealing of students that will take this course. with radioactive substances: starting with the basics of

HflfAfl EBBCAT1II All TMKfEl IF IIIW-HW1145 SK01K0075

TRAINING PROGRAM FOR OPERATORS AT THE BN600 POWER UNIT REACTOR DEPARTMENT D.VChichikin, OA.Potapov, A.EAminev

Beloyarsk NPP, Russian Federation At the Beloyarsk Nuclear Power Plant (NPP), • examinations; and operator training is carried out by step-by-step • backup. preparation, probation, and on-the-job backup in In the case of a Reactor Control Engineer, the accordance with standard training programs. duration of minimum training is 2.5 years (which Programs identify: includes no less than one month of independent work at each position). • training duration; • scope of knowledge on theoretical training and The most critical part of the training is allocated practical skills; and to tiie probation and the on-the-job backup. The terms of probation and backup for the control room engineers •; general procedure of reactor department duty op- are scheduled in a way that allows the operator to erators training. participate in startup or shutdown of the unit (loop), The duration of the operators training is defined having previously mastered the mode on the simulator. according to the following steps: Upon successful backup, the operator is permitted to •i probation aiming at the elaboration of practical start work. skills and the mastering of the work place;

14B | PISTK PKSEMTATIIHS SK01K0076 COMMUNICATIVE FOREIGN LANGUAGE EDUCATION FOR DEVELOPMENT OF INDIVIDUAL IN DIFFERENT CULTURES TSSolovkva

Obninsk Institute of Nuclear Power Engineering, Russia

Social changes in community and also in principles * conditions, technical means; of education, psychology of activity and philosophy • purpose of education. lead to changes in views on human as a whole, on place of individual in society and on definition of individual This model was tested on the example of itself, on essence of education and on foreign language collaboration with the Islam republic Iran. as business communication mean and separate Collaboration accordingly contract between institutes discipline. of higher education was accomplished by final testing in the as specialty language of narrow In this article the development of the individual in nuclear power profile (module) of students and the sphere of nuclear industry and technology is specialists. considered on the example of Obninsk Institute of Nuclear Power Engineering as the model of totality of Further in the article the full description of United all these factors. State Testing System in the Russian Language for Foreigners in context of European educational It is necessary to emphasize the key conceptions standards. In the article the experience of work of inter- of this model: region center of testing, which is Obninsk Institute of • matter of education; Nuclear Power Engineering, is given.

KHClfftB EDOCATill AM TMI5FEB IF IHW-BBW 1147 SK01K0077 IAEA ACTIVITIES IN NUCLEAR REACTOR SIMULATION FOR EDUCATIONAL PURPOSES

By A. Badulescu and R. B. Lyon

International Atomic Energy Agency, Austria Abstract code, developed in the United States. During the past seventeen years, the performance of The International Atomic Energy Agency (IAEA)has PCTRAN has been documented against known established a programme in nuclear reactor simulation data. The ARS runs on a typical PC and models the computer programs to assist its Member States in education reactor types: PWR, BWR and HWR in the 600MWe and training. The objective is to provide, for a variety of range. For the PWR models, plants with vertical inverted advanced reactor types, insight and practice in their U-bend steam generators of Western design, plants with operational characteristics and their response to horizontal steam generators as designed in the former perturbations and accident situations. To achieve mis, the Soviet Union, and a next-generation PWR with passive IAEA arranges for the supply or development of simulation safety features are included. The Simulator operates in programs and training material, sponsors training courses real or accelerated time and covers the nuclear steam and workshops, and distributes documentation and supply system, containment, control systems, and safety computer programs. Currently, the IAEA has simulation systems. Malfunctions and parameters can be selected programs available for distribution that simulate the to model normal and abnormal design basis conditions behaviour of BWR, PWR and HWR reactor types, only relevant to each reactor type. Conditions outside the two available that simulate all three reactors. design basis can also be simulated. Methods Discussion

This poster covers two simulation programs available The CARDs program simulates the neutronic for distribution: the Classroom-based Advanced Reactor behavior of the reactor core and the thermal-hydraulic Demonstrators package and the Advanced Reactor behavior of the primary heat transport system and all the Simulator. The two simulators model the same three types. • major surrounding systems such as feedwater, main steam, of reactors, BWR, PWR and HWR, however they are emergency core cooling, and the corresponding logic somewhat different in nature and may be of interest for associated with these models. The major digital control differing purposes. The Classroom-based Advanced computer programs emulated are: heat transport control; Reactor Demonstrators package contains modules from boiler level control; boiler pressure control; reactor an advanced full scope simulator, with an interaction regulating system; and moderator temperature control. capability reduced to only the necessary software required Available malfunctions are: steam generator malfunctions; to demonstrate the general behavior of the plant It is thus loss of coolant accident; loss of feedwater, liquid relief well suited to giving a rigorous demonstration of the valve failures; and other miscellaneous malfunctions. behaviour of the reactor systems for a selected set of The ARS program provides an on-screen "mimic"; a perturbations. The Advanced Reactor Simulator uses user-friendly interface that facilitates control actions and simplified models and provides ahands-on user interface diagnostics by the operator. Virtual control panels mat are that permits extensive user interaction with the simulation. simplified versions of those used in atypical nuclear power It is thus well suited for use as an educational tool. To plant of that type are included. Normal operation, power achieve this, the IAEA arranges for the supply or maneuvering, perturbations and accident situations can be development of simulation programs and training material, induced by selection of initial conditions and malfunction sponsors training courses and workshops, and distributes states, and by actual adjustment of control points, valve documentation and computer programs. positions, and equipment status via the mimic. Results Conclusions

The Classroom-based Advanced Reactor Nuclear reactor simulation computer programs are Demonstrators package (CARDs), is a suite of nuclear educational and training tools provided by the Agency power plant simulators developed by CAE Electronics to its Member States. The objective is to provide, for a Ltd., a Canadian company specializing in full scope variety of advanced reactor types, insight and practice flight, industrial, and nuclear plant simulators. The suite in their operational characteristics and their response to consists of PWR, BWR and HWR simulators and perturbations and accident situations. Currently, the operates on a typical PC. The simulators are modeled to IAEA has two simulation programs available for the discrete component level. Control logic is based on distribution: the Classroom-based Advanced Reactor the plant elementary diagrams and dynamic models of Demonstrators package and the Advanced Reactor plant hydraulic circuits are incorporated. The simulators Simulator, which model BWRs, PWRs and HWRs. are fully calibrated against both design and plant data However, they are somewhat different in nature and may but serves as a demonstrator and not a training simulator. be of interest for differing purposes. One is well suited The Advanced Reactor Simulator (ARS) was to giving a rigorous demonstration of the behaviour of developed by Microsimulation Technologies, based on the reactor systems for a selected set of perturbations PCTRAN, a PC-based FORTRAN transient analysis and the other well suited for use as an educational tool. 1411 nSIEI PRESEKTJITIIIS SK01K0078

ESTIMATION OF DOSES OF IONIZING RADIATION FROM PATIENTS TREATED WITH 131I

Y.Chaban, A.Roziev, O.Mileshin, A.Klyopov, N.Shishkanov, EMatusevich

Russia Introduction Materials and methods

Radioiodtherapy is recognized to be the most In order to change the duration of stay of a patient effective technique for the treatment of diffuse toxic under special conditions, we treated data onradiometry goiter and various types of differentiated thyroid cancer. of patients. Archived data that had been accumulated Adverse effects of irradiation following the therapy of in the MRRC RAMS for several years of application diffuse toxic goiter are considerably low in comparison of radioiodtherapy were served as the source of with complications that could be caused by surgery information. The rate of exposed dose to the surface Interventions. There is no method for the treatment of of a body and at a distance of 1 m from a patient was distant metastases of differentiated thyroid cancer other measured daily with a scintillation counter. than radioiodtherapy that is based on selective adsorption of 13II by thyroid cancer cells. The cohort under study consisted of more than 250 individuals (each of subjects had several entries to the A patient taking treatment with radionuclides can clinics), divided into 4 groups in relation to the disease be dangerous for medical personnel, relatives and other (diffuse toxic goiter and differentiated thyroid cancer) and individuals because after peroral administration of I3II ttierypeofappliedradionuclidetherapy(ablationofremnant he is both a source of ionizing radiation and a source thyroid tissue, treatment for metastases to lung or bones). of radiocontamination of the environment. In Consecutive treatment of available experimental radiological clinics these patients stay under special data allow one to build up averaged curves of temporal conditions; they are in special wards equipped with changes of exposed dose rate in each group of patients special ventilation and sewage, shields, telemonitors, with regard to the distance from a patient. telephone for contact with hospital personnel and other devices which allow one to minimize the exposure of At the next step of the study it will be possible to the staff to radiation. use the averaged temporal changes of the dose rate for Estimation of the potential dose to be received by people In accordance with standards acting in Russia, in contact with a patient in relation to the following: discharge of a patient from a hospital depends on the dose rate measured at a distance of 1 m from the patient; 1. The time that passed between the administration the limit is 3 mmSv/hour. That is why the patients must of the radionuclide and the start of the contact stay at clinics for 3-12 days on average. ' 2. Type of the disease (diffuse toxic goiter and From the experience of foreign clinics it is known differentiated thyroid cancer) and the number of that patients with diffuse toxic goiter do not need a administrations of the radionuclide long stay at radiological clinics if there is no special 3. Various scenarios concerning duration and prescription. As a rule, the duration of stay of apatient character of the contact of exposed and unexposed with differentiated thyroid cancer is no longer than 5 individuals days.

In order to make a decision on whether a patient Conclusions can come out from special facilities it is necessary to estimate the dose that can be received by individuals Patients will be asked to follow special instructions in contact with him and to make typical for public behavior, duration of contact with a spouse, recommendation for his behavior outside the facilities. children, adult relatives, colleagues, etc. It will allow one This will secure the unexposed people form receiving to minimize dose received by people around a patient. doses above the permissible limit that is under strict control by the National and International Commissions The results of the study will be given in the of Radiation Protection. presentation.

IBCLEAB TEEIKILIEY 11151 SK01K0079 THERMOHYDRAULIC RELATIONSHIPS FOR ADVANCED WATER COOLED REACTORS AND THE ROLE OF THE IAEA

A. Badulescu IAEA

andD. C. Groeneveld AECL, Austria

Abstract Results

Under the auspices of the International Atomic The final report of the CRP will provide: (1) a Energy Agency (IAEA) a Coordinated Research summary of important and relevant thermohydraulic Program (CRP) on Thermohydraulic Relationships for phenomena for advanced water cooled reactors on the Advanced Water-Cooled Reactors was carried out from basis of previous work by the international community, 1995-1998. It was included into the IAEA's (2) a state-of-the-art review and a recommended Programme following endorsement in 1995 by the prediction method for the critical heat flux, which has International Working Group on Advanced been established through international co-operation and Technologies for Water Cooled Reactors. The overall assessed within this CRP, (3) a state-of-the-art review goal was to promote international information exchange and three methods for predicting film boiling heat and cooperation in establishing a consistent set of transfer coefficients developed by institutes thermohydraulic relationships that are appropriate for participating in this CRP, (4) a compilation of relevant use in analyzing the performance and safety of pressure drop prediction methods and an assessment advanced water-cooled reactors. of these relations and the resulting recommendations, Methods (5) a discussion on a methodology to select the range of interest for parameters affecting CHF, film boiling and pressure drop in advanced water cooled reactors, The IAEA through CRPs provides a framework and (6) concluding remarks on the relationships for international collaboration among institutes in referred to in (2)-(4) above and comments on future industrialized and developing countries, typically 3 to research needs in thermohydraulics of advanced water 5 years in duration. The IAEA requires a significant cooled reactors. contribution from participants in a CRP, e.g.;, Discussion experimental results, analytical efforts, and assignment of guest researchers of team- . . .:: The. nuclear community has developed The CRP on Thermohydraulie Relationships for thermohydraulic codes for predicting the performance Advanced Water-Cooled Reactors collected and peer of water-cooled reactors under normal, transient and reviewed relationships for critical heat flux (CHF), post- accident conditions. These codes are used for plant CHF heat transfer, pressure drops under low flow and design, evaluation of safety margin, establishment of low-pressure conditions. The CRP participants have emergency procedures and operator training. The combined databases, where possible, to prepare performance of these codes is dependent on the relationships for use in predicting these phenomena; and accuracy and consistency of the thermohydraulic have provided information on the validation of these relationships and thermophysical properties data prediction methods. Organizations participating in the contained in the codes. Extensive validation CRP have provided their experimental data to augment programmes have been carried out to demonstrate the the database of the International Nuclear Safety Centre applicability of the codes to plants, e.g., experimental (INSC) at Argonne National Laboratories, which can data have been extensively compared with code be accessed at http://www.insc.anl.gov/thrmhydr/chf. predictions including International Standard Problems The database includes: (1) look-up table for Critical of the OECD Committee on the Safety of Nuclear Heat Flux (CHF) in 8-mm tubes, (2) CHF databank for Installations (CSNI) and IAEA standard problem VVER reactor applications, (3) look-up table for the exercises. post-dryout (PDO) heat transfer in tubes, (4) look-up table for CHF in WER rod bundles, (5) CHF data from The objectives of the CRP were (1) to low power and low flow experiments. systematically list the requirements for thermohydraulic relationships in support of advanced water cooled 1521PBSTEB PBESEHTAnillS reactors during normal and accident conditions, and provide details of their data base where possible and (2) results of this international collaboration in an IAEA to recommend and document a consistent set of TECDOC has facilitated the transfer of knowledge of thermohydraulic relationships for selected leading scientists in thermohydraulics to the next thermohydraulic phenomena such as CHF and post-CHF generation. heat transfer and pressure drop. Hence this expanded the previous knowledge of these thermohydraulic Conclusions phenomena by providing prediction methods having a wider range of validity including geometries being Evaluation of reactor performance under normal considered for Advanced Water-Cooled reactors. operation, accident and severe accident conditions require accurate representations of thermohydraulic Key collaborative activities of the participating relationships and thermophysical properties data. The institutes within the CRP include: (1) preparation of results of a 4-year CRP have been documented and internationally peer reviewed and accepted prediction will be issued shortly as an IAEA TECDOC. The data methods for CHF, post CHF heat transfer and pressure and relevant thermohydraulic prediction methods drop, and (2) establishment of a base of non-proprietary contributed by the organizations participating in the data and prediction methods available on the Internet. IAEA's CRP has been stored in the INSC database, This activity that includes documentation of the which is maintained by ANL.

HICIEAB TECmiLICV 11153 SK01K0080 GRADIENT AND STEPWISE INFRARED (IR) FIBER LIGHT GUIDES BASED ON SILVER AND THALLIUM HALOGENIDES

VD.Fedorov, S.E.Orlov, AASarmakova

A-RICT, Moscow

Abstract very important for fabrication technology of IR light guides with high optico-mechanical properties as well A method of heat vacuum extrusion in solid solutions as the stage of «tube-rod» interface formation.The of silver and thallium halogenides is developed. Optical crystal surface becomes free from organic and inorganic losses values are 0,2 -1,2 dB/m. Those are 20-30% lower impurities, chemosorbated layers and extractions of than those of ordinary double layer light guides. colloid silver. Crystal vacuuming at fabrication of the double layer preform provides evacuation of air from Methods the core and cladding crystal interface and results in absence of gas inclusions on the inner surface. The Crystal light guides are made by vacuum extru- «tube-rod» interface becomes practically invisible at sion method at the 150-250° C temperature range. Ex- Visual examination. Quasi-steady condition of double truding single or double layer preforms of thallium or layer optical fiber flow takes place after the initial silver halogenide crystals are put into a vacuum press 2 period of fiber extrusion and obtaining a steady chamber. Residual pressure was less than 10" mmHg. extrusion profile. Distribution of refractive index (DIR) Extrusion of optical fiber was performed in the desic- fits adequately to both: double layer preform and cated inert atmosphere. The design of the press-tool extruding fiber. «Camebax» microprobing TlBr-KRS- ensures the process of reverse extrusion. The vacuum 5 also proves the inheritance of radial profile of atom press-chamber was placed at a lower plate of the hy- composition (Br, J) measurement during the extrusion draulic testing machine with maximum capacity 1000. of a preform into optical fiber. KN, equipped with an electronic pressure control sys- tem. The special feature of the technology is the op- Disturbance of extrusion in double layer optical portunity to extrude optical fiber from the bottom up- fiber can cause geometric inhomogeneity, such as wards. The rate of light guide extrusion was 5-20 cm/min deviation from concentricity or some curving of cylinder ft specific pressures of 50-100 KN/cm2. The extrusion «core- cladding» interface. Light guides with optical rate was kept constant due to control of crystal defor- losses less than 1 dB/m were used for transmission of mation degree, which is approximately propor- .10-30 wt radiation of COj laser. There were no volume tional to shrinkage of extruding fiber. Crystals of sil- failures on the «core-cladding» interface, which usually ver and thallium halogenides used for fabrication of led to local overheating at weak points of light guides. core and cladding of IR light guides had the absorp- Homogeneity of light guides also remained high for light tion factor less than 2-5«10-4cnrl. Evaluation of the guides, with a gradient profile which had optical losses absorption factor was made by adiabatic laser of 20-30 % less than light guides with stepwise profile of refractive index (at the similar values of n^). calorimetry on A.= 10,6 (CO2 laser). Grain sizes of polycrystal light guides was measured with micro- scope statistic analyses by secant method. Mechanical Conclusions strength was measured in conditions of dynamically stretching axial loading with a constant rate of stretch- The method of heat vacuum extrusion in solid ing loading of 5 mm/min. The results of measurements solutions of silver and thallium halogenides makes it of breaking strength values were processed with weibular possible to produce: statistics based on the model of weak links. Optical losses D Single and double layer light guides; were determined on X= 10,6{im (CO2 laser) on optical - double layer light guides with a gradient profile fibers of more than 10 m length by a destruction check. of refractive index distribution. Optical homogeneity of IR light guides was evaluated by local thermoprobing of the total light guide length at Optical loss values are 0,2 - 1,2 dB/m. Optical losses in gradient light guides are decreased by 20- passing a 10 wt radiation CO2 laser. 30% as compared to ordinary double layer light guides.

Results and Discussion The method of heat vacuum extrusion helps to produce double layer light guides, which are Stages of chemical modifications of surfaces and homogeneous along the entire length of optical fiber and structural modifications of surfaces, tubes and rods are have no volume failures on the interface «core-cladding». SK01K0081 LOW-WASTE FABRICATION OF BARIUM, LITHIUM AND CALCIUM FLUORIDES FOR THERMOLUMINESCENSE DETECTORS AND HIGH DENSITY SCINTILLATORS

AA.Barmakova, S.E.Orlov

A-RICT, Moscow Abstract mother liquor in 4,0 -4,5 times. Acid consumption decreases 3-5 times as compared with previously used Especially pure fluorides of barium, lithium and technology. calcium are very promising materials for growing monocrystals and fabrication of ionizing radiation To achieve a deep deoxygenation, a heat treatment detectors. To make a high quality detector, the content of received fluorides is carried out in several steps: of colouring matter impurity in fluorides should be argon drying at 150°C and die following treatment with about 10-4 -10-6 mass %, oxygen and OH-group - not hydrofluoric acid and fluoride at 400 - 450°C. The 2 more than lxlO- mass.%. process provides fabrication of BaF2, LiF, CaF2 with content of impurities (mass.%): Fe*£l*10"4, Methods Ni,Co,Cr,Cu,Mo,V- l0*..l0r7, C-210-3,0-H0-2.

Universal, low-waste technology of fabrication of Results barium, lithium and calcium fluorides implies precipitation of the above mentioned fluorides from The technology and instrumentation for reception pure carbonates by means of 45% hydrofluoric acid of of high-purity halides is developed. The base on syn- specially pure grade and following fluoride treatment thesis of experimental batches of fluorides of barium, of dewatered fluorides of metals. lithium and calcium with impurity level 10-5 - The distinction of the process is a two-stage 10-6 wt.% is created. precipitation of fluorides: at the first stage carbonates Glass forming compositions of aluminium - fluo- react with mother liquor of the second stage and at the ride glasses, transparent in distant UV area (0,18-0,25 second stage, a product of the first stage istreated with microns) are found, the ways of reception of ingots hydrofluoric acid with 10-15 % excess. It provides a and cores with high homogeneity are developed, and lull utilization of the acid and reduction of a spent the resistance to colouring is stabilized.

NUCLEAR eUUKY 11155 SK01K0082 BN600 REACTIVITY DEFINITION

V.Zheltyshev and A. Ivanov

Beloyarsk NPP, Russian Federation

Since 1980, the fast BN600 reactor with sodium If the sensor efficiency is expressed by the justi- coolant has been operated at Beloyarsk Nuclear Power fied analytic function with one or two unknowns, there Plant. The periodic monitoring of the reactivity modi- can be simultaneously defined the reactivity, the pa- fications should be implemented in compliance with rameters of sensor efficiency modifications and the the standards and regulations applied in nuclear power effective outer sensor, basing on the requirement of engineering. The reactivity measurements are carried the constancy of reactivity to be measured after the out in order to confirm the basic rieutronic features of rod movement and the method of least squares. a BN600 reactor. The reactivity measurements are This method of defining the efficiency of the con- aimed to justify that nuclear safety is provided in course trol rod operating components provides for the error of the in-reactor installation of the experimental core decrease down to 2-3%. However, the unfeasibility of components. on-line reactivity meter application can be considered The intrinsic feature of the BN600 reactor neu- as one of the essential shortcomings of the method. tron kinetics is the very noticeable impact of the spac- Thus two reactivity meters are to be used on BN600 ing effects expressed in terms of the real neutron flux operation: time behavior, a difference from how it is described by 1. Digital on-line reactivity calculated under the kinetics point-wise model. The spacing effects stationary reactor operation on power (approximation coming from the in-reactor neutron field modifications of the point-wise kinetics is applied). during control rod movement results in changing of the efficiency of neutron records by sensors as reactiv- 2. Second reactivity meter used to define the ity changes. The immediate measurements on BN600 reactor control rod operating components efficiency have revealed that ihe application of the traditionally under reactor startup and take account of the changing accepted methodology of kinetic equation inverted efficiency of the sensor, however, this is more time- solution kinetics point-wise model may cause rather • consumptive than the on-line reactivity meter. serious systematic errors. In case of reactivities inputs The application of two reactivity meters allows within the range often fractions p^ the difference in outputs may reach ~ 10% with further increase as re- for the monitoring of the reactor reactivity under ev- activity grows. ery operating mode.

tiB | PI51DIPKSEMTATIIIS SK01K0083 LIFETIME EVALUATION OF BOHUNICE NPP COMPONENTS

LudovitKupca

Bohunice NPP, the Slovak Republic

Introduction c)jstraight pressuriser pipeline close to the pressuriser The conaept of major components lifetime evalu- jd) pressuriser cold injection piping connection to MCP ation is to be Understood as continuous monitoring and e)hot elbow under the steam generator. lifetime assessment of these components. It is essen- NPP V-2 piping lines parts tial to acknowledge the stage of component damage case by case, not only from the nuclear safety point of a)jtriple T-junction on the main circulation pipeline view, but alsc by an argument of the reactor lifetime, 3K 5OO/3K 200 the possibility of operation extension. b)low pressure emergency core cooling system T-junction - hot loop Lifetime! evaluation of selected components of the primary and secondary system have been taken c)low pressure emergency core cooling system Into consideration for many years in Bohunice NPP. T- junction - cold loop Since the equipment performance conditions are Id) high pressure emergency core cooling system ^characterised by the coolant fluid operating overpres- T-junction sure and temperature changes that are linked with e)jpressuriser hot injection piping connection to the power level changes, the equipment lifetime evalua- MCP tion is performed from the high-deformation (low f) hot elbow under the SG. cycle) fatigud point of view. This is the limiting deg- radation mecjianism based on the international expe- rience for that equipment range. The corrosion- NPP V-1 and V-2 primary erosion failure mechanism is negligible for austenitic circuit components structural materials, and it is monitored in a sufficient time period [for the pressuriser, steam generator, a)reactor pressure vessel (RPV) feedwater and steam piping system carbon and fer- b) hot RPV nozzle including the safe-end ritic steels through wall measurement and defect c)tightness parts of the RPV cover monitoring ijn potential critical locations. These measurement results are analysed and considered in jd) reactor coolant (main circulation) pump lifetime analyses. e)loop isolating valve f) pressuriser Analyses concerning periodic lifetime monitoring bf major components, exposed primary piping system - upper lid body and related systems, are carried out based on a contract - cold injection nozzle between NPPRI (V_JE) and NPP Bohunice utility (SE - bottom nozzles of the surge lines EBO). On subjected utility activities, EG_ Bratislava, v e - nozzle connection on the safety valve NRI (J ) _ _> _AM Vitkovice, V_Z Bratislava, ENERGOV_SKUM Brno etc. had participated, except g)steam generator (SG) mentioned NPPRI Trnava. - feedwater piping nozzle For each particular component and its parts, de- - primary colllector termination of failure mechanisms can be sorted based - SG body close to the water level on theoretical analysis of operational conditions and - steam collector. operational feedback from the present experience. Failure detection caused by particular mechanisms Within the scope of NPP lifetime assessment, the full and their cooperation can be made using straight calculation and failure analysis is performed for these monitoring quantification techniques. No particular parts and components: approach mentioned gives reliable information sufficiently. Hence, evaluation of the damage level NPP V-1 piping lines parts arising due to cooperation of particular failure mechanisms has to be derived from the combination a)jtriple T-junction on the main circulation pipeline both of approaches. The introduced task can be divided (MCP))K500/5K200 in the following issues: b) straight pressuriser pipeline close to the T-junction a)| monitoring of operating loads HDClfAl TECHMLIGY 111S7 b) monitoring of defect size and location Conclusions c) monitoring of state and changes of microstruc- Operational experience shows that interaction of ture, mechanical and physical properties particular damage mechanisms occurs. Based on d) computational quantification of damage. worldwide NPP operating experience it can be stated that the fatigue damage is the determining degradation Presentation of the results mechanism for assessed components (except RPV in the core area and SG parts which are underlying to The results of mentioned calculations arid analyses corrosion affected by the water of the secondary are presented in two forms. The first approach consists circuit). Fatigue damage is affected by the following of the figure with a description of component exposed mechanism influence: locations or junctions. The figure is completed with a - operating pressure level fluctuation (fatigue dam- table of damage cumulation values for exposed age is affected by maximum (minimum) value of locations for the period from unit startup to the last transitionsj not the effect performance (pressure reactor fuel campaign (expiry). The second form of increase/decrease)) the results presentation consists of two graphic relations: - deformation effects by non-stationary thermal fields which occurs by unexpected fluid tempera- - damage-campaign histogram ture changes. - total damage-campaign histogram In addition, both of the mechanisms perform in existing residual stress, additional load dilations and dynamic effects environment. SK01K0084

CHILEAN EXPERIENCE IN PRODUCTION OF 18F-FDG FROM 18F IN A REACTOR

M. Chandia*,N. Godoy*,X. Errazu*, Hernandez*, M. Figols, G. Firnau andF. Troncoso*

Chilean Nuclear Energy Commission, Av. Nueva Bilbao 12501, Santiago, Chile

"F-FDG (fiuorine-deoxy-D-glucose) is an impor- carbonate, and kriptofix and hydrolysed to form 18F tant and useful radiopharmaceutical for imaging and fluoride. The nucleophilic complex reacts with 1,3,4,6, study of miocardial viability. tetra-O-acetyl- 2-0-trifluormethanesulfonyl- 0-D- mannopyranose. The acetiled carbohydrate by acid Usually cyclotron-produced "F is used to label hydrolysis produces "FDG. The final product was "F-FDG. The availability of a 5MW Nuclear Reactor purified using an ion retarding resin (AGl 1-A8) and a in Chile and the absence of a quality cyclotron to pro- system two Sep Pak Plus: Alumina and C-l 8 cartridge duce "F required that we developed a method in order and sterilised by Millipore 0.22 urn filter. to obtain suitable "F to label "F-FDG using the facili- ties we have at the Nuclear Center of La Reina, Chil- The "F-FDG was obtained in an apyrogenic and ean Nuclear Energy Commission. sterile solution. The "F Radionuclide Purity was higher 18 6 3 than 99.9% and the Radiochemical Purity of the F- The nuclear reactions involved are: Li(n,oc) H and FDG obtained was over than 99%. Residual 3H con- 16 3 l8 6 3 I8 O( H,n) F. Enriched Li2CO, ( Li =95%) was irradi- tent was as low as 20 (Bq H/MBq F-FDG.). The ated in a 5 MW swimming pool type nuclear reactor yield of the process of "F-FDG was 13.2 %. with a neutron flux of 5.7 x 10l3 n cnr2sec' for 4 hours. We carried out biological distribution in mice by The irradiated Li2CO3 was dissolved in H2SO4 (1:1) dissection of animals to 5, 30, and 60 min post injec- and distilled as trimethylsilyl (" F) fluoride tion. ("F-TMS). Studies in patients with recent myocardial disease The labelling of the sugar was carried out using have shown the success of the I8F-FDG. The C.CH.E.N the method described by Hamacker. The "F-TMS was supports an active diagnostic nuclear cardiology pro- trapped in a solution of acetonitrile, water, potassium gram.

lUClfAB TECHHB1IBY11159 SK01K0085

CURRENT PROBLEMS OF WER-1000 REACTOR CORE OPERATION IN UKRAINE

A. Bykov

National Atomic Powergenerating Company "Energoatom", Reactor core operation and fuel usage department

Planned control rod drop time registration was and the central tube with the same structural ma- passed two times a year per reactor unit. In 1992-1993 terial to avoid different temperature lengthening. some control rods at almost all VVER-10.00 units ex- 2. To reduce hydraulic resistance, perforation of the ceeded the prescribed 4 second time limit. control rod drive shroud tubes was made. There were different versions to consider: 3.The control rod drive weight was increased on 1. deposition on inner side of fuel assembly (FA) some reactor units (Rovno-3)! guide tubes; 4. Operation of the new control rod type with in- 2. corrosion product drift into FA guide tubes; creased weight now is in progress. It is planned 3. mechanical bowing of FA spring blocks and con- that all the old types of control rods will be re- trol rod drives; placed by the new type, as was done in Russia. 4. mechanical bowing of FA guide tubes; To be sure that the drop time is under regulatory 5. radiation Creep of the structural materials. conditions through unit operation, the test period was reduced up to 90 days. One 50%.power capacity test Tp determine the real cause of that phenomenon, per fuel cycle is demanded. It is about 4 tests per fuel some spent FAs were tested at the hot cell. As a result cycle. Now almost all correction measures are made. of testing versions 1 -3 were not confirmed. The fourth A large number of time measurements were made. version seems to be the primary one. The last version Evaluation of the test results shows that the drop time cannot be proven due to limited data, and cannot be of control rods is the same for both the 50% power the main cause. capacity test and the test in the hot shutdown state. It is Large amounts of statistics were obtained in necessary to reduce the test quantity per fuel cycle be- Russia, Ukraine and Bulgaria VVER-1000 [FA cause of negative impact of transients. The general curvature, inter-assembly gap, control rod weight dia- constructor recommendations were obtained. Techni- grams, etc.]. FA guide tube bowing is the same as bow- cal solutions and corresponding documentation were ing of fuel rods in the FA. FA bowing has a dollar-type sent to the regulatory authority. That technical solu- complex form. FA bowing is the consequence of dis- tion is now in expert organization. proportionate FA lengthening and the possible spring More than 7000 individual control rod time tests block compression value. The bowing has a form of were made from the main correction measures. The collective movement and cannot be removed following conclusions have been made from the current momentarily because of FA operation history. statistics data: There are some ways to reduce the control rod • The main role of the vibration factor is proven in drop time: the FA bowing process. The greatest drop times • remove or reduce the main cause (FA bowing); and maximum of bowing values are concentrated • reduce hydraulic resistance of the control rod - at the vibration zone (2-4 FA rows from the reactor drive system; partition). The first FA row seems to be stable due |to the interaction with the reactor partition; • increase the weight of control rod - drive system. • Bowing relaxation will proceed during several fuel cycles (estimated value is 4-6), and depends on Correction measures to reduce control rod drop previous FA use history. It seems to be proven that time in WER-1000 were: previously bowed FAs effect the new FA, so pre-. viously bowed FAs are straightened until the l.To reduce FA bowing the reactor lid level was middle of the fuel cycle. At some reactor units measured and increased if needed by reactor de- signer. Also, FA spring block roughness was re- small drop time reduction is observed up to half duced. From 1999, all FAs have reduced spring of the fuel cycle from the start time values; block roughness. From our viewpoint, it is also • Control rod drop medium time (f) has almost linear important to construct FA fuel rods, FA guide tubes dependence on operation time (T) [t=kxx+b]. Es- 160 |P0STEfi PRESENTATIONS timated by the method of least squares, values of heutronics consequences. Core power distribution will k and b differ from unit to unit and from cycle to change due to local water-uranium ratio changing. So cycle. Values of A and b are in following ranges: new core calculation methods are needed to avoid DNB 4 *=2.0-T-2.6 seconds, JH5-50X10" seconds per ef- (departure from nucleate boiling). One of the possible fective operation day, ways is to recalculate the maximal available Kv values • Control rod drop time distribution changes through using previously measured inter-assembly gap values operation time. The position of maximum starts {Kv - non-uniformity coefficient of volume power pro- to shift after 240 eff.days, and the form of the dis- duction}. To avoid annual gap measurements, a gen- tribution start to change at the same time. Before era! contractor now prepares recommendations on the 240 eff.days, the distribution essentially does not generalised result basement. This way seems to be very change. effective because there is no need to revise the core calculation code. To guarantee that the control rod system reliability is now within prescribed limits, we should continue Now it is evident that currently proposed phe- testing. Additional analysis is needed. Test frequency nomenon models do not explain all collected facts. To can be reduced to avoid additional unreasonable tran- get the precise theoretical explanation of all facts it is sients. necessary to carry out a great deal of various types unplanned experiments. If we have a proven theoreti- The FA bowing that was discovered also has cal model, we can predict zone behaviour. SK01K0086

PIN WISE CALCULATION BY MONTE CARLO CODES

I.Popova Institute for Nuclear Research and Nuclear Energy, Bulgaria,

J.Bucholz Dak Ridge National Laboratory, USA

Abstract multigroupe neutron cross section library from SCALE4.4 system: 44GROUPNDF5. Pin wise calculations of core power distribution have been performed for a criticality mock up installation that models the WER-1000 reactor. The Results and Discussion Monte Carlo method is well suited to solving these Pin wise calculations of core power distribution complicated three-dimensional problems. Two Monte have been performed for criticality installation bases Carlo codes have been applied: the MCNP4B code and on the «Czech VVER-1000 Mock-Up Core» used in the KENOVI code from SCALE4.4 system. The codes the dosimetry experiment. The installation includes 32 use different kinds of neutron cross section data: VVER-1000 assemblies with different enrichment. pointwise continuous-energy and multigroupe. The Each assembly consists of 312 fuel rods (pins), 18 comparison of calculated results shows thattheKENO6 guiding rods and 1 central rod. The original KENO- and MCNP results are in good agreement. VI model of criticality installation was based on a 6- cm-thick horizontal slice through the midplane, with Methods perfect reflection applied on the top and bottom surfaces. The 3-D MCNP model has been enhanced to The codes that use the Monte Carlo method have include the finite length of the active fuel (including the been applied. This method is particularly useful for materials directly above and below the fuel), as well as complex geometry that cannot be modeled exactly by the height of the borated water in the actual experimental computer codes using deterministic methods and allows configuration. In addition, the effect of the grid spacers us to obtain answers by simulating individual particles was represented by applying the equivalent amount of and recording aspects of their behavior. steel as a thin layer of secondary cladding along the full Two worlds widely known for Monte Carlo codes length of each fuel pin and guide tube. Total numbers of have been applied with different neutron cross section histories run are respectively 66,253,550 and 52,500,000 data. The first code is the MCNP4B code. The code of MCNP4B and KEN VI calculations. treats arbitrary three-dimensional configuration of materials in geometric cells bounded by first and second Conclusions degree surfaces and some special fourth-degree surfaces. Pointwise continuous-energy neutron cross 260 pins have been chosen for the comparison of section data ENDF60 are used. The second one is the power distribution. The KENO6 and MCNP results KENOVI code from SCALE4.4 modular system. The . were compared. The difference between calculated three-dimensional volumes with first and second degree results is about 3%. Based on these comparisons, both surfaces describe the geometry. The code uses the the KENO6 and MCNP results appear quite credible.

162 [ P0S1ER PBESEKTAniNS SK01K0087 NUMERICAL EFFECTS IN THE NEUTRON FLUX CALCULATIONS INTO WWER-TYPE REACTOR VESSELS USING THE MONTE CARLO METHOD

F. Garcia Yip, CM. Alvarez Cardona, M. Rodriguez Gual andS. Hernandez Valle

Nuclear Technology Center Institute of Nuclear Sciences and Technology Ave. Salvador Allende y Luaces, Havana, Cuba. Tel: +(537)575663 ext. 38, e-mail: [email protected]

Abstract fluence. The effect of this irradiation on mechanical properties becomes more acute because of the The calculation of neutron fluxes and fluence into impurities measured in the Russian steel alloys. a reactor pressure vessel is a regulatory requirement in the stages of the design, operation, and plant lifetime In the present paper, a PC version of the Monte extension. Since the reactor vessel is part of the primary Carlo 3-D HEXANN-EVALU system is used for the circuit, its integrity should be preserved under all estimation of the WWER reactor pressure vessel operational regimes. The reactor vessel is considered irradiation. It was selected on the basis of its flexible to be a unique and non-substitutable part of the NPP options that, however, need to be quantified in that undergoes degradation. The main source of aging connection with the desired magnitudes. comes from fast neutron damage induced in the steel The parameters that control the random walk of crystalline lattice. neutrons, as well as the efficiency increasing options In the case of the WWER-type reactors, vessel included in the code, are studied in order to identify embrittlement has been identified as one of the main their impact on the final results for fluxes and fluence problems concerning the safety of NPP's. Due to the in the reactor pressure vessel. As a result, an optimal proximity of the core edge to the vessel inner surface, set of parameters is suggested. the steel in the vessel is exposed to high fast neutron

WCifH TtailllEY 111S3 SK01K0088

SPECTRAL NODAL METHOD FOR SOLVING THE NEUTRON DIFFUSION EQUATION

D. Sdnchez*, R. C de Barros2, C. R. Garcia1, andD. E. Milian3

'Instituto Superior de Ciencias y Tecnologfa Nuclear, Salvador Allende y Luaces Quinta de los Molinos, Ciudad de la Habana 2Instituto de Energfa Nuclear, Rio de Janeiro, Brasil 3Centro de Tecnologfa Nuclear

Abstract then considering flat approximations for the current. „ ... . , ,, . . These flat approximations are the only approximations Presented here is a new numerical nodal method ...... „ ...... ,. . . that are considered in this method. A s a result, the forsolvmg the static muln-dimensional neu ron

d.femequat.onmrectangulargeometry^Ourmethod ^^ We show ^^ ^ ^lustra* the is based on a spectral analysis of the nodal diffusion -.^, ^ . . , , .. • . ~ - . . J ,...... ,. accuracy of the metho dJx for coarse mesh calculations. equationst . These equation t s are obtained by integrating the diffusion equation in the x and y directions, and

1641POSIER PRESENTATIONS SK01K0089 CASCADE ENERGY AMPLIFIER

A.P. Banilov. A. V. Gulevich, and O.F. Kukharchuk

State Scientific Center of the Russian Federation Institute for Physics and Power Engineering I, Bondarenko Sq., Obninsk 249020 Russia [email protected]; [email protected]; [email protected]

The technical problem of long-life fission product The coupled blanket system utilizes the fast and and minor actinide incineration and production of the thermal cores in one reactor unit, which might be plutonium fuel in the prospective nuclear systems will effectively used for energy production and burning/ arise at significant scales of nuclear power industry transmutation of long-life fission products, minor development. Subcritical nuclear reactors driven by actinides and weapon graded plutonium. Waste stream external neutron sources («energy amplifiers))) are of this system might be separated into fast and thermal considered as incinerators of toxicity of complete regions and have an optimum effect for the nuclear Industry. transmutation rates. For example, one of important features of this dual-spectrum system is reduction of In the frames of this concept, the subcritical reactor inventory of neptunium and its impact to the long-term part consisting of two coupled blanket regions (inner radiological hazard in comparison with the hard fast neutron spectrum core and outer thermal core) spectrum core. driven by external neutron source is discussed. Comparative analysis of several types of the Two types of source are studied: spallation target coupled subcritical system with different thermal output and 14-Mev fusion burn of micropellets. Liquid metal was performed. The power parameters and neutronics pb-Bi is considered as target material and coolant of of systems, which make it possible to provide the jnner fast core. Thermal core is a heavy-water multiplication of source neutrons by 100, are studied subcritical reactor of the CANDU-type. The fast core in detail for both the steady-state and pulse-periodical is protected from thermal neutrons influence with the operation modes using the methodology and codes for boron shield. All reactor technologies used in this computation of neutronics (Monte Carlo method) and concept are tested during years of operation and dynamic parameters, which were elaborated and commercially available. experimentally tested at the IPPE on coupled reactor The system operates in steady-state or pulse- installations UKS-1M, BARS-6 and Stand B [3]. periodical modes as a two-cascade energy amplifier Thus, the cascade energy amplifiers have a set of for neutrons of spallation source located inside inner advantages in comparison with traditional concepts: core [1]. Because of special shield of the fast core, in energy production, in transmutation efficiency, and neutron coupling between these two subcritical regions in economics. has one-directional behavior. Using this "cascade amplifier" principle, it is possible to achieve in deeply subcritical (1^=0.95-0.96) coupled system more high power output (up to 10 times [2]) in comparison with References traditional «single» core under the same multiplication 1. A.P. Barzilov, A.V. Gulevich, O.F. Kukharchuk, factor, then the source requirements could be essentially et al, «Concept of a Coupled Blanket System for the reduced. It is very important feature for the feasibility Hybrid Fission-Fusion Reactor,» in IEEE/NPSS 16th of high voltage charge particle accelerators for Symposium Fusion Engineering, eds. G.H. MHey and accelerator driven systems and lasers of required super CM. Elliott, IEEE, Pescataway, NJ (1996). high beam energy for inertial confinement fusion 2. A.P. Barzilov, A.V. Gulevich, O.F. Kukharchuk, systems (from economical point of view, too). et al, «Hybrid Fission-Fusion Reactor Initiated by a Laser,» The coupled cores have a high level of inherent in 8th Int. Conf. on Emerging Nuclear Energy Systems, safety: the system is always deeply subcritical, neutron ed. A.V. Zrodnikov, IPPE, Obninsk, Russia (1997). leakage from the system is very low, and it is not 3. A.P. Barzilov, et al, «Neutron Problems of necessary to provide complex control and shielding Reactor-Pumped Laser Systems: Theory and systems. Further, the heavy-water core makes it possible Experiment,)) in 9th Int. Conf. on Emerging Nuclear to use uranium of natural enrichment. Energy Systems, Tel-Aviv, Israel (1998).

TECWfOieEY 11165 SK01K0090 DIRECT UTILIZATION OF INFORMATION FROM NUCLEAR DATA FILES IN MONTE CARLO SIMULATION OF NEUTRON AND PHOTON TRANSPORT

R Androsenko, D. Zholudov, A. Kompaniyets and O. Smirnova

Institute of Nuclear Power Engineering, Obninsk, Russia

Abstract computational efforts. However, it introduces additional unvalued inaccuracy into the final results. In order to improve both the economics of Nuclear The direct utilization of data obtained from libraries is Power Plants (NPPs) as well as their safety, data and the most accurate technique, but it is time-consuming computer codes that perform benchmark calculations and expensive in CPU. The above-mentioned fact while simulating N?P parameters must be utilized. This indicates that direct utilization of information by Monte work is mainly concerned with application of computer Carlo methods poses an important and challenging codes using the Monte Carlo method, which provides problem and is the subject of this paper. advanced accuracy of equations to be calculated.

Methods Results An approach to solving neutron and photon The development of nuclear power engineering transport problems by the Monte Carlo method based imposes higher requirements on both the economics on detailed energy conversion has been developed in and the safely of NPPs. In order to provide maximum this work. The developed algorithms allow recovering safety and achieve economic efficiency, an NPP needs, features of neutron and photon interactions using to improve the precision of radiation shielding nuclear data libraries of ENDF-6 format. For example, calculations. By using real 3-D geometry to describe the algorithms were developed to reconstruct point- the details of an object being studied, the necessary wise cross sections for a given neutron energy in the accuracy can be achieved when solving these problems. resonance energy region, to perform the most accurate The Monte Carlo method is characterized by its modeling of 2-D energy-angle interaction densities, and adaptability to solve multi-dimensional tasks while to model coherent and incoherent photon scattering accounting for the total information of simulated characteristics. processes. Although the accumulated data for various substance interactions are sufficient, thus far, the A set of routines was programmed based on application of such data is a very difficult problem. algorithms mentioned above. The resulting new The demands for improving the accuracy of segments were included into Constant Unit of the calculations necessitate the employment of the most BRAND Monte Carlo Code System and then some advanced information. This information is involved computational experiments with different ENDF-6 data in files of nuclear data libraries, such as ENDF/B-VI, libraries were performed. The experiments consisted JENDL-3, FEND^-2, and BROND-3. of measuring the neutron and photon leakage spectra emitted from spherical samples. Within the code systems, which use Monte Carlo methods for solving neutron and photon transport The results obtained were then compared with the problems, the process of simulating a particle's experimental benchmarks and with the results obtained interaction with substances is performed by routines from other programs (ANISN, MCNP and ROZ) and of Constant Unit. It is characteristic of Constant Units showed good agreement with both. to derive parameters directly from data files without any simplification. The term "direct utilization" means extracting necessary values from storage immediately Conclusions in program's run-time. But usually, preliminary The precision software tools were developed and processing of information from data files is followed included into Constant Unit of the BRAND Monte by its utilization in Monte Carlo calculations. This Carlo Code System to validate other programs and technique is essentially more efficient for required information of nuclear data libraries.

IBB} POSTER PfiESENrATflHS SK01K0091 PHOTONUCLEAR REACTIONS DATABASE FOR FUNDAMENTAL RESEARCH AND APPLICATIONS

Vedernikov Sergei

Institute of Nuclear Power Engineering, Obninsk, Russia

Abstract cross section, and peak width. For the strongly deformed nuclei the splitting of a peak on two ones is The database of photonuclear reactions is well known, that accordingly doubles the number of constructed on the base of the EXFOR library. Some parameters needed for the description of experimental files with experimental data contain mistakes and are data. There are numerical methods for these parameter incorrect. extractions and they were used in the creation of parameter compilations [1,2]. The main differences Methods of our parameter compilation from [1,2] are: the presence of data for nuclei with A<51 and splitting of The situation often arises when data from resonance for non-magic nuclei. experiments carried by several different authors differs so greatly that such distinctions cannot be explained by errors of experiment. Therefore the process of data Conclusions inclusion was preceded by checking accuracy and From the analysis of the dependence of the comparing it with other data. Sometimes data is parameters on quadrupole phonon energies and on incorrect due to multiplying by some factor, or because quadrupole deformations, the essential correlation was of a mistake in units of measurement. established. It is confirmed, that the peak splitting is absent only in magic and near magic nuclei. The critical Result values of quadrupole deformations and phonon energies are extracted. After the selection of correct experimental data, the database was used to obtain the theoretical description of photoabsorption cross sections with [1] S.S. Dietrich, B.L. Berman. Atlas of Lorentz dependence. Photoneutron Cross Section Obtained with Monoenergetic Photons. Atomic Data and Nuclear Data Tables 38 (1988) 199. Discussion [2] A.V. Varlamov, V.V. Varlamov, D.S. Rudenko, The curve depends on three parameters (the M.E. Stepanov. Atlas of Giant Dipole Resonances. parameters of giant dipole resonance) energy, peak INDC (NDS) -394. Vienna, 1999.

NDCIEAB TECHKHLIGY11167 SK01K0092 EXACT AND NUMERICAL SOLUTIONS OF NONLINEAR KINETIC EQUATIONS

MA. Zaboudko

Obninsk Institute of Nuclear Power Engineering, Obninsk Abstract Mathematical correctness of this model was proven. A new method of numerical simulation was The methods to solve a nonlinear kinetic equation built on the bases of the Galerkin method and the are considered in this paper. This type of equations convergence of this method was proven with the use arises at simulation of pores spreading in metals under of the theory of functional solutions introduced in [1] the influence of a fast neutrons source. Some exact by Galkin. solutions are received for this type of equation. A numerical regularization is made on the bases of the Exact solutions of problem (1), (2) were received Galerkin method and convergence problems are also for cases <&(y,m)=0 and investigated. q(m,t)=0, q(tn,t) - AS^ (dm), where SXo (dm)- Dirak function. Methods This paper investigates the nonlinear kinetic equation (1) with initial condition (2). This model, as Results a hypothesis, describes the process of pores spreading Several numerical experiments were conducted in metals under the influence of stationary fast neutrons with different functions

dt dt (1) Conclusions w. The new numerical method for solving nonlinear 0yb(m,y)f(y,t)dy+q(m,t),t > 0 kinetic equation (1) is true and reliable, o (2) References where f(m, ^-particles density distribution; 1. Galkin V. A. Global correctness of Cauchy problem for nonlinear conservation laws systems and Ofm^-collision intensity, one example for the gas dynamics // International se- : ®(y,m) <&> and <5>(m,y) is smooth ries of numerical mathematics, 1999. Vol. 129, p. 361- function; 367. source, 0

1SB j P6STEB PDESEHTATIQHS SK01K0093 MEASUREMENT AND CALCULATION OF ACTIVITY IN VANADIUM AND V-Tf-CR ALLOYS IRRADIATED IN BR-tO REACTOR

Yr.R. Kottdrashechkin, Gji. Birzhevoy, A.I. Blokhin and M.I. Zakharova

State Scientific Center of Russian Federation Institute of Physics and Power Engineering, Obninsk, Russia

Low activation is one of the important param- the FISPACT-3.0(5) code and nuclear cross sec- eters which determine the choice of structural tion library FENDL-1, tiie isotope composition, materials for fusion reactors. V-Ti-Cr alloy is the activity, heat production and contact dose rate favored concept for an advanced breeding blan- have been calculated taking into account the ket for the HER. chemical composition of samples, irradiation con- ditions and time. The comparison of experimen- Specimens of V, V-4Ti-4Cr and V-9Ti-5Cr al- tal and calculated data was made and the contri- loys were prepared by methods of electron beam bution of the isotopes to total activity of the sam- and arc melting and then irradiated in the BR-10 ples was determined. Nuclear reactions in which (Obninsk, Russia) at 400°C with two total neu- M 2 long-living isotopes are produced were consid- tron fluence of 2.55xl(F and 5.15xlO n/m . The ered. corresponding fast neutron fluences were 25 2 210X10 and 4.24xl(F n/m (En>0.1 MeV), re- It was shown that V, Ti, Cr nuclei do not pro- spectively. duce long-living isotopes except for Sc46 which is After extracting the specimens from the re- generated by Ti. It was established that the ma- actor, the activity of isotopes Scw, Mn54, Co58, Co<°, jor isotopes, which contribute to the total activ- Nb94 and Tamwas measured for 3 years. Using ity, are produced by technological impurities: Ni, Nb, W, Re, Co, Fe, Mn, Cu, and Mo.

mm lEUHiuev 1I iss SK01K0094 THE POLARIZATION OF FAST NEUTRONS

KV. Talov

Scientific Center of Russian Federation - Institute of Physics and Power Engineering Obninsk, Kaluga region, Russia

Abstract is called polarized in this direction. It is essential to know polarization of neutrons for characteristics of a It is insufficient to know coordmates and momentum to describe a state of a neutron. It is a so !arkationof 'fastneutron sc Je upm ,9 Vs. necessary to define the spin orientation. As far as it is known from quantum mechanics, a half spin has a The present work is the review of polarization of projection in the positive direction or in the negative fast neutrons and methods of polarization analysis. This direction. The probability of both projections in an also includes information about polarization of fast unpolarized beam is equal. If a direction exists, in neutrons from first papers, which described polarization which the projection is more probably, then the beam in die D(d,n)JHe, ^(p^'Be, and T(p^)» He reactions. SK01K0095

DOSIMETRY METHOD FOR MEASUREMENTS OF ENERGY OUTCOME FROM GADOLINIUM AFTER NEUTRON CAPTURE

SA. Kfykov, YuA. Kuratchenko, E.S. Matusevitch,

Institute of Nuclear Power Engineering, Obninsk, Russia, 249020, Studgorodok 1

S.P. Kaptchigashev, V.I. Potetnya, S.E. Oulianenko

Medical Radiological Research Center RAMS, Obninsk, Russia

A very important task now is to measure a real measurement or direct registration is a very involved absorbed dose for Gadolinium Neutron Capture problem due their low mean energy - less then 10 keV. Therapy. A method for direct measurement of Gd Substances with gadolinium of Gd(NO ) -?6H O reaction dose was offered and worked through in the 3 3 2 and GdCl -?6H O, and two different dosimeters - FBX study. 3 2 and Fricke were investigated. The absorbed dose A chemical dosimeter for measurement of ab- measurements were carried out with the Fricke dosim- sorbed dose from neutron capture Gd(n,y) reaction eter for different gadolinium concentrations in the products - gamma-radiation, internal conversion elec- solution; thus, this simulated probable treatment trons, and Auger electrons, that are very essential for condition for thermal and fast neutron beams. The ex- neutron capture therapy, is described in the paper. Au- perimental results were compared with calculations ger electrons are of interest, since their dose made using the MCNP Monte Carlo code.

KUCIEAB TECflMELflGY 11171 SK01K0096

CSI(TL) DETECTOR APPLICATION FOR LOW ENERGY MIXED RADIATION FIELD SPECTROSCOPY

V.A. Khriachkov, M.KDunaev, N.N. Semenova, I.V.Dunaeva

Institute of Physics and Power Engineering, Obninsk, Russia

Abstract tion. The spectrometric "'Ra layer served as a source of source of The CsI(Tl) crystal was used for charged particle registration. Photomultiplier signals were digitized by a-particles, P-particlesandy-rays. A polyethylene quad waveform digitizer. The particle energy was esti- film, irradiated by fast neutrons from a Pu-Be source, mated with accuracy ~180 keV. Each signal was de- served as protons. scribed as a sum of three exponential curves using the least squares method. This approach allowed us to Results create a method for pulse shape identification of the charged particles. In addition, this method may be used The method using the fast component area has the for pull-up rejection. The main luminescence proper- better resolution by a factor of 1.5 - 2. Energy ties (light yield, rise time, decay time) for various par- resolution of 180 keV was obtained. ticles were investigated. It shows that electrons, pro- tons and the alpha-particles with energy up to 1 MeV Discussion can be effectively separated. The comparison of the resolution of these two particle identification methods for aa-particles and Methods protons of different energy is shown in this work. It Photomultiplier signals were digitized by the 2262 shows that the method using the fast component area Quad Waveform Digitizer . The LeCroy model 2262 has better resolution by a factor of 1.5 - 2. This method is a high-speed, waveform digitizer providing high reso- showed that aa-particles, p-particles , protons and lution, multichannel solutions to waveform recording y-rays can be effectively separated. requirements. Designed in a modular standard for ease of system configuration, the 2262 offers four digitizers Conclusions per module. The 2262 is compatible with CAMAC, and works with an IBM PC. In the experiment a crys- The main luminescence properties of CsI(Tl) (light tal with a diameter of 13 mm and a thickness of 3 mm yield, rise time, decay time) for various particles were was used with a PM-118 photomultiplier. The purpose investigated. It was shown that electrons, protons and of this work was to study CsI(Tl) scintillator proper- alpha-particles with energy up to 1 MeV can be effec- ties for charged particles of small energy identifica- tively separated.

17? | P1STEIPDESEWTAT1IH5 SK01K0097 APPLICATION OF DIGITAL SIGNAL PROCESSING FOR RADIATION SPECTROSCOPY

V.A.Khriachkov, I. V.Dunaeva, M.KDunaev,N.N.Semenova

Scientific Center of Russian Federation - Institute of Physics and Power Engineering, Obninsk, Kaluga region, Russia.

Abstract Results

In our work the waveform digitizer (LeCroy 2262) The advanced methods of digital signal processing was used. The advanced methods of digital signal were applied for signal processing of the pulsed processing were applied for operation with the pulsed ionization chamber and scintillator. ionization chamber and scintillator. It was shown that For fission fragments it was possible to carry out the numerical processing did not yield a traditional measurements of masses, energies and emission angles analogue method of signal processing. Usage of unique from a target; to take into account processes occurring opportunities of numerical signal processing allows in the detector; to carry out a high-performance us to use not only information on amplitude and rejection of pile-up. When processing impulses from a duration of a signal but also that part which contains scintillator Csl (Tl) it is shown, that electrons, protons the waveform from the detector. In particular, for fission and the alpha-particles with energy up to 1 Mev can be fragments it is possible to carry out multidimensional effectively separated. (mass - energy - a emission angle from a target) measurements based on the analysis of the shape of anodi impulses; to take into account processes Discussion occurring in the detector; to carry out a high- performance rejection of pile-up. The developed , It is shown that the numerical signal processing approach in a combination to detectors, traditional for does not yield, and in series of cases exceeds analogue. nuclear physics, gives their new properties, to It is becoming possible to analyse a waveform, more raise(increase) precision and reliability of precisely to carry out a rejection, to take into account measurements. influence of the measuring equipment on a pulse shape, to estimate the contribution of various processes to an aggregate impulse. Methods

In our work the waveform digitizer with a sampling Conclusions rate of 80 MHZ was used. The signals from the detector were transduced to a numerical view and were The usage of unique opportunities of numerical maintained in a computer memory. During processing signal processing allows us to take into account the least squares method and numerical filtration were processes occurring in the detector, and it is more correct widely used. to interpret the experimental information. As a matter of fact, we have received a new, more perfect detector.

WIELEAB TEC8NILIGV J1173 SK01K0098 CALCULATIONS OF ACCELERATOR-BASED NEUTRON SOURCES' CHARACTERISTICS R.G.Tertytchnyi, V.S.Shorin

State Scientific Center of Russian Federation jlnstitute of Physics and Power Engineering, Obninsk, Russia

Abstract spectrum because of energy losses of incident ions in the target materials and angular scattering (cinematic Accelerator-based quasi-mortoenergetic neutron broadening) connected with the geometry of the sources (T(p,n), D(d,n),T(d,n) and Li (p,n)~reactions) experiment. The practice has shown that simulation of are widely used in experiments on measuring the neutrons histories from their birth in the accelerator interaction cross-sections of fast neutrons with nuclei. target up to absorption is an effective enough method The spectrum of outgoing neutrons in such reactions of the problem's solution. depends on target characteristics (chemical structure and form - solid or gas) as well as target design and The present work represents the code for experimental setup around it being a source of scattered calculation of the yields and spectra of neutrons neutrons noticeably distorting the shape of the primary generated in (p,n)~ and (d, n)- reactions on some targets neutron spectrum and causing a low-energy "tail." of light nuclei (D, T, '£/). The peculiarities of the stopping processes of charged particles (with incident It is very important to have a correct estimation of energy up to 15 MeV) in multilayer and a neutron spectrum at the researched sample in multicomponent targets are taken into account. The experiments with a small amount of investigated code version is made in terms of the "SOURCE," a material when samples are placed in immediate subroutine for the well-known MCNP code. Some proximity from the target generating neutrons. Such calculation results for the most popular accelerator- conditions result to the broadening of a neutron based neutron sources are given. SK01K0099 RBMK-1500 REACTOR PUMP TRIP EVENTS MODELLED USING THERMAL-HYDRAULIC CODE CATHARE2 V1.3L

Audrius Jasiulevicius

Helsinki University of Technology, Department of Engineering Physics and Mathematics, Finland

Two of the seventeen operating RBMK (Russian in Ignalina in 1996. The results of the calculations were acronym for "Channel Type large Power Reactor") type compared to the values of the actual transients. reactors in the world are located in Lithuania at Ignalina Values that were predicted using the CATHARE Nuclear Power Plant (NPP). The Ignalina NPP was model show good agreement with real NPP data. Data designed by Soviet engineers to be the first of the was compared in several groups of parameters, such RBMK-1500 type power plants. The first unit of as pressures, temperatures and flow rates in different Ignalina NPP was commissioned in December 1983, parts of the main circulation loops. Flow rates through and the second unit in August 1987. Following the main circulation pumps (MCP), feed-water inlet, and accident at the Chernobyl RBMK-1000 type NPP in steam output rates were in good tolerance with the real April 1986, several actions have been taken by RBMK transient values. Temperatures in the fuel center, specialists in order to increase safety at the Ignalina cladding and fuel channels did not exceed maximum NPP (Almenas et al., 1998). Among other actions, design values during all transient times. The pressure investigation and safety assessments of such reactor - in the most sensitive part of the system during this type types using best estimate thermal-hydraulic computer of transient, the drum separators, remained in the safe codes is being implemented. The western computer range during transients. bodes already used in safety studies of nuclear power plants, are now available to be used in the study of The studies presented have shown the capability RBMK thermal-hydraulic behavior. of the CATHARE code to treat thermal-hydraulic transients with a reactor scram in the RBMK in case of In this paper, a computer simulation model of a pump trip and establishment of subsequent natural RBMK-1500 main circulation loops is presented. The circulation. modeling was performed using the CATHARE2 V1.3L code. Initially, nominal operational conditions, so- called steady state, were established with the model. References Temperatures, pressures and other parameters, ./•I.Almenas K., Kaliatka A., Uspuras E., 1998, calculated using CATHARE, were compared with Ignalina RBMK-1500. A Source Book, Ignalina safety known values under nominal conditions in the real analysis group. Lithuanian Energy Institute, Kaunas, nuclear reactor. This comparison to plant data allows ISBN 9986-492-35-1. a person to make a judgement about CATHARE 2.Jasiulevicius A., Kouhia V, Sarrette Ch., 1999, accuracy to simulate such an operational mode. Values Trip OF Main circulation pumps of RBMK-1500 cal- that were calculated using the CATHARE code were culated with thermal-hydraulic code CATHARE2 in good agreement with the plant data. VI. 3L. Submitted for ICONE 8 conference. After establishment of the steady state, transient S.Eurasto T., Sandberg J., Marttila J., 1993,2%

NBCLEAB TECUULIBY11175 SK01K0100 OPTIMIZATION OF lONIZATION CHAMBERS ARRANGEMENT IN WER-1000 TYPE REACTOR FACILITIES

Stanislav PhilippqffandAlexei Soldatov

MEPhI, YDRNS

Abstract becomes possible. It necessitates frequent verifications of IC readings at nominal power when changing axial The current work proposes a mathematical model offset. for the optimization of ionization chamber arrangement. Taking into account specific features of neutron flux instrumentation set at WER-1000, one of the ways Text to increase reading accuracy could be the optimization ilt is known that power level monitoring at the NPP of IC arrangement relative to the core. WER-1000 reactor facility is performed by means of For the first time, at Novovoronezh NPP 5 Unit, neutron flux instrumentation. The principle of neutron examinations of neutron field distribution in channels flux instrumentation functioning is based on the of reactor B-1000 IC were performed in order to define measurement of leakage neutron flux density in optimal IC arrangement Neutron fluxes were measured channels of ionization chambers (IC), which are located by activated methods during the xenon oscillation in biological shielding. process. The recommendations were given to increase The change of coolant moderating and absorbing the distance between two power generation ranges of features during temperature and pressure changes leads ionization chambers 100 mm compared with the design to change in moderating neutrons leakage, and this fact, arrangement. Subsequently, the task for optimal in turn, distorts IC information about core power value. arrangement of power generation range ionization The bigger the thickness of iron-water shielding and chambers was being solved at different Units during annular gap between reactor cavity and vessel, the more commissioning activities. The shortcoming of visible this dependence is. That is why this effect examinations conducted is the absence of methods for becomes much more apparent at VVER-1000 than at mathematical treatment of the experimental results. In VVER-440. consequence of that aspect, the recommendations to change chamber arrangement are given on the basis of The change in height distribution of neutron flux speculative analysis of diagrams that are not density as a result of xenon oscillation appearance or summarized sufficiently. This approach has led to control rod relocation, as well as change in physical different results for the Units with the same reactor features during the fuel burn-up process, leads to type. inadequate readings of neutron flux instrumentation concerning reactor power. As a result, the decrease of The current work proposes a mathematical model margin for reactor scram and alarm protection actuation for optimization of ionization chamber arrangement.

17i f PaSTER PRESEHTATieilS SK01K0101

THE EPR LAYOUT DESIGN

Vwe Mast

Siemens AG / Power Generation Group (KWU) P.O. Box 3220 D-91050 Erlange.n, Germany http://www.siemens.de/kwu

Uwe.Mast(5),erl 11 .siemens.de

General the mitigation of such accident on the plant area. For that, a spreading area for molten corium, a channel from The European Pressurised Water Reactor (EPR) the reactor pit to the spreading area and the In is a French - German development for the next Containment Eeruelling Water Storage Tank (IRWST) generation of Pressurised Water Reactor. The new for flooding and initial cooling of the corium, were reactor is based on the experiences of operation and implemented in the design of the Reactor Building. design of nuclear power plants in both countries. The EPR fulfils enhanced safety standards, higher availability and a longer service life. Layout results Following buildings are arranged on a common Boundary conditions raft due to design earthquake: Reactor Building (KB), Safeguard Buildings (SAB) division 1 - 4 and Fuel Utilities aspects Building (FB). The other Nuclear Buildings have For the utilities one significant requirement is the separate rafts. The Safeguard Buildings are strictly reduction of personnel exposure during maintenance separated into 4 divisions: the SAB division 1 on the and in-service inspection. west side, the SAB DIV 2 and 3 on the north and Safeguard Buildings division 4 on the east side of the The other important requirement is of economic Reactor Building. The Safeguard Buildings are nature. The main points influencing costs, which have separated into a cold and hot mechanical, an electrical also impact on the layout are: outage times, accessibility and a ventilation part. The main control room is located of the reactor building and the available maintenance in the middle of the Safeguard Building division 2 and and set down area. 3. The Fuel Building is arranged southern of the Reactor The utilities have also required to load the spent Building. In this building the spent and the new fuel fuel assemblies into the shipping cask from the bottom assemblies are stored. The equipment for the fuel of the fuel pool, because of the exclusion of the drop assemblies and cask handling and a part of the set down of the cask and in order to avoid contamination at the area for the refuelling outage are also located here. The outer cask shell. Nuclear Auxiliary Building (NAB) is arranged at the Layout and safety aspects south side of Safeguard Building division 4 and at the east side of the Fuel Building. In the NAB the All Nuclear Island (NI) buildings are designed operational systems and also a part of the set down against design earthquake as well as explosion pressure area are implemented. The Diesel Buildings (DB) are wave. The protection against Airplane Crash (APC) arranged close to the SAB DIV 1 and DIV 4. Each of can be realised through civil and layout possibilities. the DB contains two diesel generator and one station The Reactor Building, the Safeguard Buildings division blackout diesel, which are separated according its 2 and 3 and the Fuel Building are protected by concrete redundancy requirements. structures. The other nuclear buildings are protected by geographical separation. AH NI buildings with their main routings were designed in a 3D model. The utility design reviews Important safety requirements are the further were performed with this model. reduction of the probability of severe accidents and

KUClfflR lENHfOLIDV11177 SK01K0102 STUDY OF CORROSION PRODUCTS FROM SECONDARY CIRCUIT NUCLEAR POWER PLANT V-1 BOHUNICE

A. Zeman, V. Slugen, J. Lipka

(Department of Nuclear Physics and Technology, Slovak Technical University Bratislava, lkovicova 3, 812 19 Bratislava, Slovak Republic, e-mail: [email protected]

Abstract Results and Discussions The properties and the composition of the corrosion More than 20 specimens were collected from the productsofthe stainlessCrNi andmildsteels independence NPP Bohunice secondary circuit. The investigation was on the conditions (temperature, acidity, etc.) in the secondary focused mainly on the steam generator SG35 where the circuit of tiie nuclear power plant (NPP) is of such range feed water pipeline was changed in 1994. During in-situ that, in practice, it is impossible to determine the properties inspection 3 specimens were scrapped from different of the corrosion products for an actual case from the places in steam generator NPP VVER 440. One feed theoretical data only. Since the decontamination processes water dispersion box was removed and analysed in for the materials ofthe water-cooled reactor (VVER-440) laboratory conditions. Magnetite was identified as a secondary circuits are in the progress of development, it is dominant component in all studied specimens. necessary to draw the necessary information by the Results confirmed that during operation time, a very measurement and analysis ofthe real specimens taken from weak oxidation surrounding was in the observed steam ablockofWER-440[l]. generator and the corrosion specimens were fully without base material particles. During visual inspection ofthe removed feed water dispersion box, 2 disturbing unde- Experimental fined particles, fixed in one of outlet nozzle, were found. The transmission MOssbauer spectroscopy was Difference in size of particles is significantly observable used in the investigation of magnetic phases of corro- in the portion of ct-Fe component by larger size. It has sion products obtained from NPP V-1 Bohunice steam been shown mat these highly corroded particles (found in generator and secondary circuit pipelines. The corro- outlet nozzle) do not originate from the 172 47 steel. sion layers were separated by scraping the rust off the surface and the powder samples were studied by trans- mission Mflssbauer spectroscopy. It should be noted Conclusions that the gamma spectroscopic measurements give no All measured specimens comprise iron in mag- evidence of the presence of low-energy gamma radia- netic as well as paramagnetic phases. Magnetic phases tion emitted from the samples. The scrapped specimen are in the form of hematite (

CVD DIAMOND DETECTORS OF IONIZING RADIATION

Andrea Perdochovd andVladimir Necas

Department of Nuclear Physics and Technology, Faculty of Electrical Engineering and Information Technology, Slovak University of Technology, 812 19 Bratislava, Slovakia Diamond possesses many attractive properties covered by a gold layer to produce the ohmic contact which make it an ideal material for particle detection in J3]. Strip detectors are used to track particle paths. An physics experiments as well as for medical dosimetry. incident particle is recorded by a strip which generates the signal. Strip detectors were used in CERN to study In 1981, CVD (chemical vapour deposition) dia- the mechanism of charge collection and to measure the mond manufacturing technology was developed. This spatial resolution of diamond detectors by a pion beam. technology has the potential of allowing low cost pro- The first diamond strip detectors with a 2*4 cm2 area duction of diamond in large sheets and with higher have been produced. These can be used in real physics purity than natural diamonds [1 ], [2]. Diamond has very experiments. The signal-to-noise ratio was 23/1 and the high band gap, 5.5 eV, therefore it is an excellent spatial resolution was 14 um. insulator with high resistivity, more then 10 (£2cm). The high breakdown electric field of diamond (10 VI The first 16*16 matrix of 150*150 um2 diamond cm) allows application of high electric field and speeds pixel detector was tested in 1996 [3]. The pixels were up the charge collection. The electron and hole mobil- wire-bonded to a fan-out on glass substrate and read-out ity is very high (higher than at GaAs and Si). The satu- using VA3 chip. The measured signal-to-noise ratio was ration velocity of diamond is high and allows high elec- 27/1 and the position resolution was consistent with the tron collection. The dielectric constant of diamond is digital resolution of the pixels. The development of dia- low (5.7); it represents low detector capacitance at the mond pixel detectors has made good progress in the dif- input of the spectroscopic read-out amplifier. Therefore ficult area of detector metallization for connection with a diamond detector causes less noise than a geometri- the bump-bonding technique. The first results from a pixel cally identical Si detector [3]. Energy to form electron- detector (ATLAS3) in a test beam demonstrated 98% ef- hole pair is for diamond 13 eV. This is the only poor ficiency for the bump-bonding of the pixels and spatial diamond property. The low atomic number of carbon resolution for the track measurement in both dimension. can be an advantage in two ways. The first: reducing high-energy cascades and multiple scattering in physics Recently, diamond layer technology has been suc- experiments. The second: absorption characteristics are cessfully applied at the Slovak University of Technol- well matched to human soft tissue, so that diamond (Z=6) ogy [6], In collaboration With our DNPT and Depart- can be considered soft-tissue-equivalent (Z=7.42) [4], ment of Microelectronics, diamond detectors for do- High radiation hardness (lMGy) makes diamond an simetry applications are currently being prepared. ideal material, especially in the high radiation References: environments of future colliders like the LHC in CERN. (I] Spitsyn, V. et al.: J. Cryst. Grow*, 52,1981, p. CVD diamond coaxial X-ray microdetectors are pre- 219 pared by CVD technology. CVD diamond film is depos- \2\ Matsumoto, S. et al.: Jpn. J. Appl. Phys., 21, ited by a hot filament CVD technique on metallic(W, Mo) 1982, p. 183 wires and tips. The diamond film (10 um thick) covers [3] Adam, W. et al.: Review of the development of dia- uniformly the W (tungsten) tip and is used as an active mond radiation sensors, Nucl. Instr. Meth. A 434,1999, p. Semiconductor region for X-ray detection. Agraphite layer 131 covers the diamond film and W substrate wire acts as an j[4] Manfredotti, C. et al.:. CVD diamond detec- electrode. Coaxial detectors are stable and relatively fast. tors, Nucl. Instr. Meth. A 410, 1998, p. 96 Their signal-to-noise ratio is approximately 1000 [5]. (5J Manfredotti, C. et al.: CVD diamond coaxial Coaxial detectors can be used for monitoring X-rays and X-ray microdetectors, 11* International Workshop on y-rays in medical and health physics fields. Coaxial ge- Room Temperature Semiconductor X- and Gamma- ometry is more suitable for probe miniaturization. Probes Ray Detectors and Associated Electronics, (Book of can have diameter up to 50 urn. Abstracts), IAEA, Vienna, 1999, p. 30 (to be published in Nucl. Instr. and Meth. A) CVD diamond micro-strip detectors have a pat- {6] Bederka, §.: Ciele a v^sledky rieSenia tern of strips with 50 urn pitch on the irradiated side of vedeckotechnicke"ho projektu Diamantove" a diamond and a guide ring around this structure. The semidiamantove" vrstvy, -In: Dubravcova V. (ed.): other surface of diamond is covered by a uniform me- Fyzika a technika rastu diamantovych vrstiev, tallic contact. For contacts, chromium is used as a rule, Bratislava, STU 1999, s. 6 (in Slovak) NUdEAH TECHNOLOGY 11178 SK01K0104 NEUTRON DIFFRACTION MEASUREMENTS OF RESIDUAL STRESSES IN NPP CONSTRUCTION MATERIALS

Robert Hinca1 and Gizo Bokuchava2

Department of Nuclear Physics and Technology, STU Bratislava, Slovakia Frank Laboratory of Neutron Physics, JINR Dubna, Russia

Neutron diffraction is one of the most powerful able thermal gradients are present. Surfacing high- methods for condensed matter studies. The principal nickel filler on an austenitic base metal is one of tech- advantage of using neutrons rather than the more con- niques in repair of primary collector in the primary ventional X-rays is the fact that neutrons can penetrate circuit of nuclear power plant type VVER. The repair deeply (2-4 cm for steel and more than 10 cm for alu- technology was developed at Welding Research Insti- minium) into metals to determine internal parameters tute Bratislava. Measurements of residual stresses in within the bulk of materials. the weld overlay and in the base metal are necessary for approving the mechanical analysis and verifying Material structure information is obtained from of residual stresses determination on welded material measured neutron intensity spectra. A regular crystal by numerical welding computer simulation. Investiga- lattice gives rise to sharp Bragg reflections (Fig. 1). tions of residual stresses are important for developing Disorder, like interstitials or vacancies, reflects itself optimal welding techniques. in diffuse scattering between Bragg peaks. Large ob- jects, like clusters in alloa, polymer conglomerates, or At the current state of investigation are estimated biological macromolecules, are imaged at small neu- residual strain components ex and ey(Fig. 2). Further tron scattering angles. measurements are planned for January 2000.

0.001

800 1000 1500 2000 JSOO Channtl. It Fig. 1 Diffraction spectrum of an austenitic steel

Fig. 2. Residual strain components ex and ey (from Stress investigations using neutron diffraction oc- profile refinement) versus gauge volume location depth cupy a special position in non-destructive testing. The d within thei sample. principle used is similar to that of the well-known X- ray technique in which the internal lattice stress present in a material is obtained from the measured elastic lat- References tice strain it produces in the crystallites of which it is \X\ Bacon G. - Neutron diffraction, Oxford composed. Lattice strains for all available crystal planes University Press, 1975 (hkl) were obtained from relative shift of diffraction {2) Noyan I. C, Cohen J. B.: Residual Stress, peaks. Springer- Verlag New York, 1987 £3] Allen A., Hutchings T., Windsor C. G., Residual stresses measurements were performed Andreani C: Neutron diffraction methods for the study with HRFD at the IBR-2 pulsed reactor in Dubna (Rus- of residual stress fields, Advances in Physics, 1985, sia). HRFD is TOF diffractometer equipped with a vol. 34, No.4,445-473 Fourier chopper. Slovak republic is a member state of [4] Berky R., Magula V., Tomasovic P.: Numerical JINR Dubna. simulation of welding processes, Research report, VUZ The aim of the investigation is to study the level Bratislava, June 1998 of residual stresses induced by the surfacing in the weld deposit zone and in the base metal, where consider-

1801P9S1ER PRESEHTATIBIIS SK01K0105 INCREASING OF UNIT CAPACITY WHILE CONTINUING CAMPAIGN

Slapakov Aleksandr, MironovaAnastasia, and A. Ivanov Valerie

Branch of St. Petersburg State Technical University, Sosnovy Bor, Russia

Abstract In the process of decreasing pressure in the sec- This report considers features of regulation by ond '<*™*>*» *™&&t of the main turbine de- "sliding" pressure on reactors of a VVER type with "J" *nd "superfluous" steams arise, which can be prolongation in terms ofthe working company and to P^uced by the reactor, but camot be used by the bas.c ^ c • C •«. r vmo turbine. One of the ways to use it would be the usethemformcreaseofcapaatyofaNPP. .^.^ ofm ^^ ^^ ^.^ on g The thermal effect of reactivity in WER-1000 line leading around the HPT ofthe basic turbine, type reactors allows an increase in fuel burn-upby de- basing the temperature of the coolant. The coolant „ i^rt temperature in the primary circuit depends on the sec- ^ for VVER-1000 wjh ondarycircuitpressure-Thisiscalledslidingpressure 65/1500 1S also considered, regulation.

I9U£A11EEUHIBY 11 111 SK01K0106 THE ROLE OF CHANNEL TYPE REACTORS IN RUSSIAN NUCLEAR ENGINEERING

Alexander A. Karev, Anatoly P. Eperin, Nickolay N. Erentin

[Branch of Saint Petersburg State Technical University, Sosnovy Bor, Russia

Abstract double protection from radiation dose to personnel for all the periods of NPP operation. The potential for the This report considers the advantages of the unit to release radioactive contamination to the MKER-1000 unit over RBMK-1000 in terms of safety, environment and to expose people to radiation is greatly based on the Leningrad NPP as an example. Technical diminished. characteristics of these units and facilities are referred to and are presented and analyzed here. In order to protect people, the area around the unit is comprised of a protective zone and an observation Theses zone. Analysis of possible radiation exposure to residents while the unit is in normal operation and in The power unit with an MKER-1000 reactor is case of top rated or abnormal accidents determines the being developed as an automated technological zone boundaries. Evacuation planning and protective complex, destined for safe and economically effective arrangements for the affected people are planned. production of electrical and calorific power and isotope The protective zone radius is 250 m (zone borders production. The Reactor MKER-1000 (multi-loop are not beyond the edge of the manufacturing grounds). boiling water reactor) is being designed as a channel The zone of people in the evacuation areas requiring water-graphite thermal neutron reactor of "increased" protective arrangement planning has a radius of 220 m safety performance. The term "increased" refers to (since this measure is less than the protective zone the existence of systems with a high degree of safety, radius, where there are no residents, planning for which enhance channel type reactor properties of self- population evacuation arrangements drops out). The defensibility, representative of channel-type units. zone of protective arrangements planning has a radius These systems include multi-loop modular construction of 2 km (for an NPP with an RBMK-1000 unit, the of the unit; ability of transferring heat through the protective zone radius is 3 km, therefore there should graphite setting; passive principles; a emergency reactor be no living areas or social buildings within 2 km from cooling system consisting of two subsystems; reactor NPP and there is the possibility to evaluate economic control, operation and protection system; automated advantages or losses from settling this territory). technological processes control system; emergency Finally, the observation zone radius is 4.5 km. steam derivation system; containment existence, etc. Decreasing the observation zone radius from 30 Technical solutions make the annual limits of km (for RBMK) to 4.5 km (for MKER) will open about individual and collective radiation dose not to be 2750 square km of federal land for non-restricted exceeded while in normal operation. Technological activities. Socio-economic compensations for processes are performed while the unit is in operation. additional factors of risk and emergency planning This preventive work provides extra protection from arrangements will not be applied to this area. internal and external radiation to personnel and acts as SK01K0107

EXPLOITATION QUESTIONS REGARDING CHANNEL TYPE REACTORS: WATER GRAPHITE CHANNEL REACTOR -1000 (OPERATION, RECONSTRUCTION, ADVANTAGES, AND DISADVANTAGES)

Chichindaev Dmitry Alexandrovich

Saint Petersburg Technical University, Russia

This is a short informative review about channel reactivity effect caused the accident in Chernobyl. type reactors. Electricity is necessary for a country's The reputation of channel type reactors came to developing economy. Today there is 5-10 times less harm as a consequence. electricity per capita than in developed Europe and f- The large amount of separated assembly channels jLFSA. In our region, the Leningrad Nuclear Power Plant for the 1st loop produced cumbersome partition- ipLNPP) provides 60% of all electricity. It is a Water ing of reactor equipment systems and supporting Graphite Channel Reactor-1000 (RBMK-1000). systems. At first, the project took into considera- tion the availability of the safety shield contain- Introduction ment, but later it was decided that containments would be built on the 1st and 2nd unit of LNPP to There are two ways of developing nuclear energy protect it from external effects. channel type reactors and pressure vessel reactors - Arrangements on the reconstruction before and (advantages, and disadvantages). Channel type reactors after the accident in Chernobyl are still taking are one of these ways. place. They are essential in increasing the safety of this type of reactor. ADVANTAGES 1. Nuclear safety, reactor Installation - The first and main advantage of these reactors is metal control, and operation and security simplicity and low cost of manufacturing separate systems of the reactor assembly channels and bundles for the reactor, and their transportation and mounting. The first stage of reconstruction of the LNPP was - jRefuelling on the capacity allows us to raise reac- completed on units No. 1, 2, and 3. The 4th one is tor operation and economic performance, efficient under construction. The main goal is to keep prolonged use of the nuclear fuel and the possibility of re- endurance of the plant. shuffling it in the core prolong operation. 2. Necessary justification of NEW reactor - One assembly channel failure does not induce shut (MKR-1000) flown (if operation/exploitation limits are not bro- ken, even 9 assembly channel failure do not in- Energy consumption and energy users are volve damage of the reactor). constantly increasing so we need more powerful sources - Maintainability of the reactor and its bundles al- of energy. We must not prolong the lifetime of current lows fixing or changing any bundles inside the plants by way of infinite, reconstruction and modernisation, reactor (on LNPPs unit No. 3 was conducted not because it causes an increase in expenditure and a decrease only a change of 50% of all the assembly channels jin achieved safety performance - it is necessary to build (but even recovering of graphite units of one assem- new power plant units. bly channel, destroyed in an accident in 1992, by way of reshuffling of the graphite pillars). MKR-1000 is a new generation reactor that has high safety performances, advantages of the channe} j- Usage of the reactor ion source for receiving indus- type reactors, but without some disadvantages" (Of trial radiation materials like Co, doped silicon, etc. RBMK reactors. - The process of reconstruction and bundles renewal yields prolonged operation and endurance of the plant Jn our country we have gained great experience in building and operating channel type reactors. It is less DISADVANTAGES expensive under our conditions. Our town of Sosnovy Bor is a powerful building base, and our region allows - The physical performances of the reactor do not us to build new units of MKR-iOOOs quickly and provide its safety, for example, a positive steam cheaply, and partially using LNPP's infrastructure. KICIEM 1ECIIILMY 11113 SK01K0108 EXAMINATION OF CONTROL ROD EJECTION IN WWER-440 TYPE REACTORS USING THE CODE DYN3D

G. Petofi, A. Aszddi

Institute of Nuclear Techniques, Technical University of Budapest, Hungary

Abstract 3. Control rod ejection at low power

For nuclear reactors it is very important to examine The control rod ejection event is also relevant at the reactivity initiated transients caused by the ejection of a hot zero power. In computer simulations the definition control rod. The event is found to be dependent on different of this power level can be different from type to type thermalandneutronicparatrieters. Inthispapertheemphasis of events. For this reason, besides the safety related is laid on the effect of me power level at which the transient points, the location of the hot zero power level is also began and on the effect of the heat ftansfer coefficient investigated. Calculations are made at 10%, 1% and measured in the gap between the fiiel and Hie cladding. 0.1% of the nominal power. The transient induced less reactivity change and weaker feedback effects than it 1. Determination of the most effective did at nominal power. For this latter reason, for a certain control rod time the core had a weak positive feedback. The importance of the examination of this occurrence is The most significant transients can be established coming from the possibility of the appearance of by the ejection of the most effective control rod. So another safety relevant event during this time. For the first step is to determine the position of this rod. It example the loss of one or more main coolant pumps was done by steady state calculations using the following could lead to additional reactivity growth. scenario. A calculation was carried out with all the rods inserted to the half level of the core, criticality was 4. Control rod ejection at different reached by adjusting the power level. Seven other values of heat transfer coefficients calculations were made for each control rod at withdrawn in the gas gap position while the other six rods were inserted to the half plane of the core. The most effective rod is the one The mostproblematic part of the heat transfer process that could cause the most significant difference compared is the one in the gas gap. In WWER-440 type reactors, to the power level determined in the first calculation. this value at the beginning of the cycle is about 2500 W/ m2/K. In order to take into account the uncertainty of this All the rod ejections are carried out in 0,1. s. value, calculations were made with the constant values of Criticality is reached by dividing the cross-sections by 1500 W/mVK and 3500 W/mVK. Additionally, a kejr During the calculations it is supposed that by some calculation was made to check the implemented heat reason scram is not initiated. Only the negative feed- transfer coefficient calculating procedure of DYN3D. backs are available for the reactor. The primary effect of the modification of the 2. Examination of rod ejection coefficient is the change of the radial temperature at full power distribution. As the examinations showed, this difference and the change of the transient process lose their effect After the ejection the feedback brings back the at the outer surface of the cladding, because the cladding reactivity very quickly as a consequence of the high has a high heat conduction coefficient Further inquiries temperature change in the fuel. showed that the DNB did not occur even at the calculations with the highest coefficient. The heat transfer The temperature increase occurs in the first few coefficient calculating procedure of DYN3D gave seconds. It was found that the highest temperatures appropriate results in this range of parameters. appeared near the withdrawn control element in the highly enriched assemblies just below the half level of Conclusions the core. The risk of melting of the core is not significant. Another interest is whether DNB occurs or not, By summarising the results, it can be said that the By all the three correlations of DYN3D, it is assessed modification of the heat transfer coefficient between . that the state of the first type of boiling crisis, even in real limits at either power level could not cause the most loaded part of the core, has not come into significant problems from the safety point of view. The existence. The ejection of the most effective control small change in the absolute power level induces less element at this power level from this point of view does important effects. The possibility of a boiling crisis has not cause an important risk. According to further not been approached as much as at the higher power results, there has not been oxidation of the cladding, level. A significant difference was that at lower power butat the hottest part a void-fraction of 12% has arisen. levels, void-fraction have not been formed. 184J PfiSIER PRESEIfrATieifS SK01K0109 MATHEMATICAL SIMULATION OF GAS DYNAMICS ON EULERIAN MESHES: A REALIZATION OF YOUNG'S METHOD FOR INTERFACE TRACKING IN A THREE-DIMENSIONAL CASE ON ORTHOGONAL MESHES

Veselov Roman, AM Bykov and B.L. Voronin

Russian federal nuclear center - VNIIEF, Russia

improvements in nuclear technology provide of a mesh. This is why additional algorithms of strong incentives towards development in other areas restoration of the interfaces are required. of science, particularly for mathematical simulation of Several materials can be present in cells in Eulerian physical processes. Some tasks, for example a study computations. Information about those materials in a Of nuclear reactor safety, generally can be only explored mixed mesh is usually stored in computer memory, for by means of mathematical modeling. Nowadays it has example: the number of materials and their volume become possible to use more accomplished fraction and thermodynamic magnitudes (density, formulations, including algorithms that require large temperature, energy). Any additional information about computing expenditures for simulation of multiple spatial disposition of materials in a cell is not stored. processes of gas dynamics, thermal conduction, The elementary algorithms of restoration of the contact transposition of neutrons, etc., with the emergence of boundaries assume that the interface is parallel to the new high-efficiency multiprocessor computers. lines of a mesh. The plane interface has an arbitrary The mathematical simulation of gas dynamic position in a more complicated algorithm known as processes using difference methods requires a Young's Method. construction of the initial difference mesh covering a This report contains the exposition of Young's given area. This simulation can be described and interface tracking method and includes determination performed by two methods: Lagrangian meshes (when of the flux across a cell side, as well as modifications the mesh is joined together with substance) and Eulerian Of a method for cases when a mixed cell contains more meshes (where the mesh is motionless in space and the than three materials. substance is irrelevant to the mesh). The Lagrangian computation is more economic, but often results in a The outcomes of simulations for three sample tangled mesh and necessitates a local reorganization calculations: "Jack and Balls," "Spinning Disk," and of the mesh, which is often necessary to perform f Spuming Jack," are also presented. The simulations manually. The Eulerian computation is smoother, but represent linear motion and/or rotation of geometric requires large computing expenditures. In the planes on rectangular meshes, which provide an Lagrangian computation, the interface is traced estimate of the resolution of the interface-tracking automatically as a line in the mesh. In the Eulerian algorithms. computation, the interfaces do not coincide with lines

mieiEAR immaioGY 1 SK01K0110 CONTAINMENT LJEAK41GfiTNE$S ENHANCEMENT AT WEB440

Milan Prandorfy

VtJEZ, Levice, Slovakia

Abstract and hidden leaks (by removal of a part of cover The hermetic compartments of VVER 440 NPPs concrete or using injection) fulfil the function of the containment used at NPPs all - local, individual and periodical integrated leak rate over the world. The purpose of the containment is to tests to prove the efficiency of sealing works. protect the NPP personnel against radioactive impact With regard to the design of the individual as well as to prevent radioactive leakage to the technological nodes, different methods of leak testing environment during a lost of coolant accident. are applied to hermetic boundary components. Leak-tightness enhancement in NPPs with WWER With regard to this improved leak tightness of 440/213 and WWER 440/230 reactors is an important containment, new methods for leak rate evaluation and safely issue. New procedures, measures and methods for verification of leakage-test accuracy have been were adopted at NPPs in Mochovce, J. Bohunice, adopted. Dukovany and Paks for leak identification and sealing works performed by yUEZ Levice. Results

Methods The integrated leakage rate tests performed at the end of refueling outages have proven the efficiency of Mochovce units 1 and 2: sealing works conducted during refueling outages. Containment leakage rate is still lower than that During the years 1997 to 1999 (before startup obtained during the last refueling outages. operation) a lot of leak tests were performed according to improved programs prepared in compliance with (Charts of leakage rate will be presented) standards and requirements contained in internationally To obtain better results of integrated leak rate tests, recognized regulations e.g. KTA 3405 or ANSI/ANS periodical resealing as well as local and individual leak 56.8. More than 1000 local leak tests (super-checking) testing should be repeated. Resealing of structural have been performed per unit. After leak resealing the nodes (hermetic doors, hatches, ...) should be individual test of hermetic compartments and the integral periodically repeated. The procedure adopted for the leak rate test of containment have been performed. performance of sealing works proved itself. J. Bohunice units 1-4 and Dukovanvunits 1 and 3: Leakage rate reached at Mochovce NPP is Enhancement leak rate measures were started for comparable with those in western Europe NPPs the period beginning with 1990 at the V- 1NPP and with containments. Components on the external hermetic 1995 at the V-2 NPPs and Dukovany NPPs (during boundary preserve proper integrity during lost of operation of the units). To increase the containment coolant accidents. leak-tightness, new procedures of leak identification and sealing works have been adopted: Conclusions - application of sealing compounds, Containment properties would deteriorate without - application of adhesives, periodical tests of individual hermetic nodes on the -vapplication of coating, hermetic boundary and sealing works. - application of foams and injection. The excellent leak-tightness of the Mochovce NPP The process of leak rate reduction has been containment comparable with that of western Europe performed in stages: containments and the step-wise enhancement of leak- tightness in the Bohunice and Dukovany NPPs enable - leak detection (follow-up integrated leak rate test), to rank the WWER 440 NPPs among those meeting - repair of leaks detected on the hermetic boundary international nuclear safety standards.

1U f NS1E8 PSESEITATtllS SK01K0111

THE GRADUAL DEVELOPMENT STEPS OF THE EXTERNAL COUPLED RELAP5- DYN3D CODE

Ctibor STRMENSKt

VtiJETrnavaa.s. Okru4na5 918 64TRNAVA tel:+421 805 599 1317 email: [email protected] Introducion The first goal was correct installation of RELAP5 code under the operating system. The test of the This paper describes the on-going and finished installed program was done successfully. The pilot parts of project: «The external coupled RELAP5- calculation of nuclear power plant model was DYN3D code». The RELAPS is thermo-hydraulics performed and its correctness and accuracy were tested. code used for analysis of the thermohydraulics Steady state of model was evaluated. problems in the nuclear facilities. The DYN3D is three- dimensional dynamic code used to calculate the As listed as the second and third points, the dynamics processes in the nuclear core. modification of die thermohydraulics data is being done. At the present time, the calculation and the modification of input data into KELAPS code and Process of the external coupled RELAP5- DYN3D code are being done. In this phase of DYN3D code modification the values are set manually. The individual The project was divided into four development values of selected parameters are obtained from output- phases (Fig. 1). files (Tab. 2). The calculation in the RELAP5-code is stopped when all values of parameters are stabilized. Fig. 1 The scheme of the exchange data Fourth, the RELDYN-code should manage the modification of input data into both codes and synchronize the time in transient processes automatically.

Table 2 The selected parameters for exchange Code RELAPS DYN3D Parameter coolant temp coolant temp at output of core To* at input of core mass flow above mass flow belov. core Go* core Gu

Table 1. The development phases Conclusion Name of phase Flag The progress of external coupling is continued. 1.Install and test The implementation of RELAP5-code and the first tests the RELAP5 code was performed of nuclear power plant model were performed without 2.Modify input data into RELAP5 code and DYN3D-code is performed any substantial difficulties. The modification of selected 3.New tests of the modified input parameters is being exchanged manually. The data for both codes is performed exchanged parameters are not selected definitely. For 4.Program the RELDYN code - this group of selected parameters, it is possible to adjust automatic exchange of input data or upgrade their number. The RELDYN-code will between RELAP5 and DYN3D replace manual manner of input data modification. codes is planned

MHClfftR TECHNBLIGVI I 187 SK01K0112

THE ANALYSIS OF PHYSICAL PERFORMANCES OF THE REACTOR RBMK-1500

Valdas Ledzinskas

Engineer of Reactor Control, Ignalina NPP, The Trainee of the Kaunas Technological University, Kosmoso 28-411, Visaginas, 4761, Lithuania, [email protected]

outcomes have shown a satisfactory likeness with actual Abstract data. Main direction of work is construction of mathematical models of severe accidents and models of different transient regimes in the reactor RBMK- 1500 and in all complex of the NPP equipment as well. Discussion The analysis of serviceability of the important There are some difficulties in model constructing technological schemes in emergency operation also is and accuracy of calculating in such data as graphite carried out. temperature, damage of fuel channels, fuel melting and its transfer from active zone into under reactor space because of difference of channel and case reactors in MELCOR codes. Introduction Despite the Designer and Main Constructor of RBMK made their own calculations it was useful to make independent calculations using worldwide Conclusions experience to ensure safety. In other hand, the work This work is in the context of safety analysis of has certain applied significance. Ignalina NPP with Chernobyl type of reactor. Calculation results show the resistanse of these reactors and self-protection from after-effects of severe accidents. And, therefore, possibility of safe operation Methods the RBMK-1500. For these purpose mainly the program MELCOR is used, which author is Sandia National Laboratories (USA). The adaptation of this computer code (initially intended for NPPs with case reactors) was conducted Acknoledgements in Argonne National Laboratories (USA) during 10 Thanks to the scientific chief of this issue professor 1998-04 1999. Jonas Gylys (Department of Thermal and Atomic Engineering), for his authority and wise care over this Because RBMK have essential distinctions with work as well as to A. Adomavichjus (Kaunas case reactors, an original method in program modeling Technological University) and J. Sienski (Argonne was necessary to use. For example, to simulate one National Laboratories). loop of main coolant circus and divide it into conditional "control volumes", "imagine" graphite Special thanks to A. Bolgarov (Leading Engineer moderator as the material of absorber rods and of Reactor Control Ignalina NPP) who helped in design artificially change their thermo physical properties. and translation of the report.

Results References The mathematical model of the reactor, main K.Almenas, A.Kaliatka, E.Ushpuras Ignalina coolant circus, other systems of Ignalina NPP was RBMK-1500: a Source Book Kaunas, Ignalina Safety constructed and the number of accounts, which Analysis Group, Lithuanian Energy Institute (1998).

Ill [ PISTE! SK01K0113 PERSPECTIVE PRODUCTION OF RADIOISOTOPES AND RADIOPHARMACEUTICALS IN DIVISIONS OF IPPE

G.O.Tkrentyev ?

Institute of Physics and Power Engineering, Russia

The first reason for the production ofradioisotopes The reactors that are being used for irradiation of in IPPE was for the generation of Tc needed for targets are: domestic medicine in Russia. Currently, 48 1. A reactor AM (the first NPP): thermal rating radioisotopes are can be manufactured. The profile of up to 10 MW, stream of thermal neutrons up to 4*10" production is based on the possibilities of die irradiation base and the need of medicine; the sale volume of the n/cm\s*in a.period ofmaintenance until theyear 2004. raw isotopes constitute about 20% (mostly abroad) and . 2. A reactor BRilO: thermalrating up to 6 MW, the remaining 80% is composed generally of isotopes stream-.:df fast neutrons up to 8,6* 10" n/crir\s, in a that are ready to be used. period of maintenance until the year 2002. In IPPE, much research concerning these issues is [The questions of why the construction of these being conducted. These are based on the technology replacements for the existing reactors and the of 1-125 and Pd-103 (which are obtained from alpha possibility, of its usage for the. irradiation of targets of emitters); Th-229; generation of Th*229 and Ac-225; indirect reactors are being considered.. creation of Ac-225 - Bi-213; and production of the closed gamma source.

®CANDU is a registered trademark of Atomic Energy of Canada Limited (AECL). HUCIEAB eUHDLIGV II1183 SK01K0114 UTILIZING HORIZONTAL REACTOR CHANNELS FOR NEUTRON THERAPY

E.Yu. Stankovsky and YILA. Kurachenko

Institute of Nuclear Power Engineering, Obninsk, Russia, 249020, Studgorodok 1

Abstract Results Two experimental heterogeneous reactors have The comparison of calculated and experimental been considered. The reactors may be applied in neu- data inside the core allows one to obtain the absolute tron capture therapy and in a conventional manner. The values of neutron flux at any point in the channel. The channel out of the core serves as the neutron source. normalizing relation is At each of these facilities, both fast and epithermal 9* 1.15'10 18 xN, (1) neutron fluxes for BNCT research, human clinical tri- where cp is experimental flux density, als, and characterized common computational tech- N is calculated neutron flux. niques have been evaluated. [The expression (1) corresponds to 10 MW of re- actor power. All results of the calculations inside the channel are normalized to one neutron per fission, since Methods this tally is the standard of MCNP code. Also obtained Longstanding investigations proved that neutron is the neutron spectrum at different points of a chan- beams out of reactor cores are effective for medical nel, the ratio between thermal and fast neutrons along applications. In the Obninsk IPPE BR-10 reactor the the channel, and angular distribution of neutrons in therapeutic room for conventional neutron therapy has different points of channel. been created and 200 patients have passed the course of treatment. It is assumed that the characteristics of the WWR-c reactor at the Obninsk branch of State Discussion Scientific Center "Karpov Institute of Physical Chem- The calculated data for WWR-c reactor beam port istry" are advantageous in comparison with the ones correspond to real neutron flux densities at 10 MW of of BR-10. Monte Carlo analyses were performed to reactor power as jtotal=9.12'10 9 ± 2.27'10 9 (neu- evaluate the relative significance of the epithermal neu- trons/cm3 s), jthermal=2.17'10 9 ± 0.713'10 9 (neu- tron source and boron compound characteristics on a trons/cm2 s). theoretical tumor control using a standard model. Cal- Such values demonstrated the applicability for culations were performed for both existing facilities applying this reactor in the neutron capture therapy by using 3D models. To get the angular distribution of goals, and made it possible to optimize the neutron the neutron beams the adjoint problems were solved. field characteristics by filters. Utilizing boron com- This data was used as a source for subsequent neutron pounds in fast neutron therapy has advantages if a pre- transport treatment in water phantoms. The dose rate liminary slowing-down moderator is used. distributions of neutrons, gammas, and boron captur- ing component have been calculated. Various mod- erators for fast neutron beams that upgrade spectral Conclusions characteristics have been studied. To select the proper Determination of energy and angular characteris- arrangement and dimensions of the moderators a tics of neutron sources for treatment is very difficult. number of preliminary problems must be solved. All If the core of the reactor serves as a neutron source for these studies demonstrate the applicability of fast neu- therapy it is useful to apply the adjoint problem. Only tron beams for combined "boost" therapy. Neverthe- Monte-Carlo based algorithms with appropriate vari- less, it is too difficult to achieve an epithermal beam ance reduction techniques can guarantee good agree- for boron neutron capture therapy. Preliminary stud- ment of experimental data with those that are calcu- ies of WWR-c beam port were based on "in-core" ex- lated. As in all radiotherapies, it is important to per- periments. The geometric model of this facility has been form extensive computational dosimetry studies on a used in MCNP calculations. The cross sections from patient-specific basis in order to deliver an optimum ENDF/B-VI nuclear data library have been used. Ex- treatment. The field size, number of fields, angles of perimental data for channels placed inside the core were incidence, and the spectrum of the therapy beam are available. To verify the model, the comparison between all adjustable parameters that must be determined for calculated thermal neutron flux distribution and experi- optimal treatment. mental data has been performed.

1941 PflSTEfi PSSEMTAT1BMS SK01K0115

188RE-MICROSPHERES OF ALBUMINE - THE POTENTIAL PREPERATION FOR RADIOTHERAPY

D.N.Dyomin, V.M.Petriev

Medical Radiological Research Centre RAMS, Obninsk, Russia

An important direction in radiopharmacy is the intrapleural administration at metastases covering a search, the development and estimation of particular cavity. properties of radiopharmaceuticals for radiotherapy of Microspheres, 188Re with sizes 40-60 micron for tumoral and non-tumoral diseases, where the main thing treatment of disseminated kidney cancer (intraarterial, is the choice of optimal carriers and radionuclides. selectively), intratumoral administration to damaged The microparticles have a much higher degree of nodules less than 2-3 cm. I88 selective accumulation in tissues of patients than solu- Microspheres, Re with sizes 80-100 micron for ble preparations. And, this process can be programmed large neoplasms and metastases of liver (intraarterial, by change of physico-chemical properties of selectively), intratumoral administration to damaged microparticles and variation of methods of injection to nodules with sizes over 3 cm. patient. In developing preparations for radiotherapy the preparation of albumine microspheres is carried (various microparticles can be used. The colloid parti- out by thermal denaturation of protein in vegetable oil. cles uptake in organs of the reticuloendothelial sys- Microspheres are obtained with the necessary range of tem. The microaggregates and macroaggregates are un- sizes by ultrasonic fractionation. At our laboratory the stable, liposomes are quickly exposed to decompos- method of preparation of albumine microspheres with ing. It results in the higher irradiating of organs and any sizes of particles (from 5 - 10 up to 800 - 1000 tissues that do not need to be treated. microns) has been developed. For preparation of microspheres, various materi- Labelling efficiency is increasing with the increase als can be used: glass, ceramics, synthetic polymers, of the amount of microspheres. The experiments have inorganic salts, starch, casein, gelatin, albumine. The shown, that the direct introduction of 188Re results in albumine microspheres have of advantage: binding 1S8Re with an efficiency of about 70 %. Though physiologicality and biodegradability; technologically and in the low efficiency of labelling of albumin simplicity of method of preparing of microspheres; pos- microspheres, stability of a labelled preparation is sibility of preparation of microspheres with given sizes. rather high. The study of kinetics of eliminating 188Re Among traditionally used radionuclides in from albumine microspheres at room temperature has radiotherapy in recent years, significant attention has shown, that l88Re practically is not eliminated from been given to the radionuclide of 188-Re. The devel- microspheres and remains (stays) in the linked state at opment of preparations on the basis of l!8Re is a per- a level of 70 % within 24 hours. spective direction of radiopharmacy. 18BRe has the op- ;To elevate the percent of binding 188Re with timal nuclear-physical characteristics in its use for microspheres, we have decided to carry out modifica- radiotherapy. tion of microspheres by incorporating various The energy of particles (E = 0,728 MeV (25 %), complexes or function groups with the surface of E = 0,716 MeV (79 %) MeV) create a good dose for microspheres. Such complexes can be DTPA, treatment. The presence of a gamma-component with diphosphonic acids derivative, mercapto- derivative energy 155 KeV - allows observation of behavior of and so on. the radionuclide in an organism by use of a gamma - (We have carried out the study of kinetics of bind- camera. ing !88Re with this complex, which is composed of phos- The short half-life of mRe (17 hours) allows to phorus-containing groups. keep the patients in special conditions short time. The binding of 188Re with the complex was stud- 188Re is a generating radionuclide Therefore kits ied in dependence from pH, concentrations of the to the generator of 188Re, similar to kits to the genera- carrier in eluate 188Re, time of carrying out of the tor of MmTc, can be designed. reaction and storage of a labelled preparation at room Being grounded on what is stated above, we un- temperature. dertake an attempt to develop kits to the generator of Optimal conditions for preparation of a 188Re- l88Re on the basis of albumine microspheres for complex follow: radiotherapy of both oncological and nononcological • The time of labelling the complex is 40 - 50 diseases. minutes. Microspheres, l88Re with sizes 10-20 micron for • Volume of eluate l88Re for labelling complex is treatment of rheumatoid arthritis (damage of large and 1,5 ml. intermediate joints), intraperitoneal administration and • The eluate 188-Re should be with the carrier.

HBGIEAB TECHNDIIEYII | IBS SK01K0116 RADIOIODTHERAPY: DOSIMETRY PLANNING

Ajipyan, A.Roziev, O.Mileshin, A.Klyopov, N.Shishkanov, E,Matusevich

Russia Introduction (the distance between a body and a radiometer was 1 m). In order to build up a data curve on accumulation of Radioiodtherapy of diffuse toxic goiter and differ- radioiodine in the thyroid, measurements were made at 2 entiated thyroid cancer is the conventional method for and 24 hours after the administration of diagnostic activity treatment due to the selective accumulation of radioac- for diffuse toxic goiter and calculations made for the thyroid tive iodine to the thyroid tissue and cells of the differen- cancer were used. Data extracted from dosimetric records tiated thyroid cancer regardless of their localization. of annual measurements of remaining activity of radioiodine The therapeutic effect ofl311 is based on the break- in the thyroid following the administration of the gamma down of the thyroid tissue by p-particles. So, due to rays was measured with a distant scintillation counter (DSU- radioiodtherapy of the diffuse toxic goiter the number of 61). The device consisted of the crystal of Na(Tl) and functional follicles of the thyroid is reduced, it leads to photornuMplier FEU-13. The sensor was shielded. Radi- the normalization of its functioning and to the reduction ometry of patients with dosimeter DRG-05Mwasmadedaify. of the thyroid size. Post-surgery iodine ablation for the The device was designed for the measurement of exposed differentiated thyroid cancer leads to the considerable dose of X- and y-irradiation, as well as for quantitative esti- reduction of the risk of relapse and to the enhancement of. mation of the existence of p-particles. the efficacy of radioiodtherapy of metastases. The size of the thyroid or its remnants was meas- jHoweyer, analysis of Russian and foreign data shows ured with the use of sonograms. The volume of the that there are considerable distinctions between recom- thyroid is calculated with the formula of an ellipsoid. mended optimal therapeutic absorbed doses (from 40 to 120 Absorbed dose to the thyroid following the adminis- Gy) for the diffuse toxic goiter. As for the differentiated thy- tration of a therapeutic amount of radioiodine was calcu- roid cancer, of empirical radioactivity on the outcome of lated with the use of curves of accumulation --elimination radioiodtherapy has been carefully studied for a long time; of iodine, size of the thyroid or its remnants in the bed of however, absorbed dose to the remaining tissue has not been the thyroid calculated on the basis of ultrasound exami- taken into consideration. Few authors who tried to investi- nation, and data on administered activity. gate the dose-response relationship proposed to form the absorbed dose in the range between 300-700 Gy on aver- Results age; the doses used varied from 200 to 1300 Gy. There is no consensus among those who recognized If the administered therapeutic amount was l-10mCi the method of administration of standard activity. For the absorbed dose in the thyroid of patients with diffuse radioiodablation there are two tendencies: to deliver low toxic goiter was within the range of 4-185 Gy. Patients activity (30-40 mCi) and to deliver high activity (100- with thyroid cancer received standard amount of l3lI (70 150mCi) of !31I. Administered therapeutic activity for mCi) for radioiodablation, the absorbed doses in the rem-. diffuse toxic goiter is varied from 2-3 to 4- mCi ofl311. nants of the thyroid varied from 70 to 7000 Gy. Calculations showed that if standard activity had been The relationship between the biological period of administered the true dose to the thyroid of patients var- half-elimination of radioiodine from the thyroid and grav- ied considerably (from 40 to 1221 Gy if administered ity of the disease of interest (diffuse toxic goiter and thy- standard activity was 15 mCi). The fluctuations of the roid cancer) was detected. There were considerable dis- absorbed thyroid dose in patients with diffuse toxic goiter tinctions between biological periods of half-elimination led to relapses and the development of hypothyreosis. of radioiodine in patients with diffuse toxic goiter with regard to the amount of administered radioiodine (for the Materials and methods purposes of diagnosis or for the therapy). So, if calcula- . tion of absorbed dose in the thyroid of patients with dif- The results of treatment of 142 case histories of 125 fuse toxic goiter is based on the biological period of half- patients who had been treated with radioactive iodine at elimination of radioiodine administered for diagnosis, the Medical Radiological Research Center of RAMS from the modeled absorbed dose in the thyroid will be reliably 1983 to 1999 are given in the presentation. Among the higher than the real dose. In order to estimate the absorbed patients, 3 5 cases of diffuse toxic goiter with signs of thy- dose more accurately a relevant coefficient must be used. rotoxicosis of a mild degree, 25 cases of Diffuse toxic goiter with severe thyrotoxicosis, 6 cases of differenti- Conclusions ated thyroid cancer with metastases to lymphnodes of the neck, 30 cases of thyroid cancer with metastases to Accounting for our experience and data fromothe r inves- lymphnodes of the rieck and lung and 1 case of thyroid tigators, we think that it is necessary to cany out individual cancer with metastases to bones were diagnosed. . dostaetry planningfcriadfoiodinerapy ofpatients. Inthemefhod Curves for accumulation and elimination of the based onadministration of standard activity of 13IIboth mass of radioiodine from the thyroid were built up for every pa- the thyroid or its remnants, its functional activity and kinetics of tient. The curves show the true speed of accumulation of the radioiodine are not taken into consideration, this consider- the iodine in the thyroid tissue and its elimination from the ably effects the abscabed close in the target site. thyroid gland or its remnants (a radiometer was positioned Recommendations for individual dosimetry plan- close to the target organ), as well as from the whole body ning for radioiodtherapy are given in the presentation. SK01K0117 NON-PROLIFERATION ISSUES WITH WEAPONS-USABLE PLUTONIUM

Leonard W. Gray

Lawrence Livermore National Laboratory Livermore CA 94551 United States of America

ABSTRACT to load 50 tonnes of Pu into the reactors and 20 years to irradiate those 50 tonnes. After 13 years the first All Pu isotopes produced in either a Pu production weapons-grade-spent fuel will have aged 10 years. reactor or a nuclear-power reactor have a critical mass. Therefore at the end of the 20 years irradiation cycle, Theoretically, then, any isotopic content of Pu can be nearly a third of the 50 tonnes of Pu is back to being a used to fabricate a nuclear weapon. However, there maj or terrorist proliferation threat. Even at the greater must be sufficient mass of Pu and that mass must be rate of 5 tonnes per year, it will take 10 years to load assembled fast enough. While fabrication of a weapon land 13 years to complete the irradiation. Therefore, may be above the resources of any particular terrorist about 25% of the Pu will again be a nuclear weapons group, it certainly is within the capabilities and proliferation risk. The US answer to counter this risk resources of many countries of the world. is to entomb the weapons-grade MOX spent fuel in a Consequently, both so-called "weapons-grade Pu" and repository. The repository then becomes the "commercial power-reactor-grade Pu" can be made into proliferation deterrent — not the radiation field. a usable weapon. Therefore, the stockpile of Pu being returned from the Russian and American militaries When Pu is immobilized by mineralizing in a (About 100 tonnes of plutonium will be removed from ceramic matrix and then placed into a canister and the US and Russian stockpiles by the year 2005) will surrounded by HLW glass, the package will have remain a proliferation problem if burned in a once- approximately the same radiation field as 30 year old through MOX-reactor cycle. For comparison, in the spent fuel. However, the mechanical disassembly of year 2000, approximately 1200 to 1600 tonnes of the canister is much more difficult than chopping spent weapons-usable plutonium exist in spent fuel from fuel assemblies. Simplistic time-and-motion studies worldwide power reactors. (This number is increasing indicate that the radiation dose that would be absorbed at a rate of 60 to 70 tonnes per year.) by a terrorist group would be much higher while trying to dismantle the can-in-canister immobilized Pu than The Russian proposal is to burn all of the excess She dose absorbed to chop-up spent fuel that has been weapons-grade Pu in power reactors as MOX fuel. The aged only ten years. Whereas the fissile materials are US proposes to burn only a portion of its excess Pu in very easily leached from chopped spent fuel, that is existing reactors. The remainder of the Pu, the US hot the case with dissolution of the Pu ceramics. proposes to immobilize in a ceramic matrix and then Although several recovery methods are possible, each surround that matrix in a canister filled with high-level method for the recovery of mineralized Pu from the Waste (HLW) glass. This is referred to as the can-in- ceramic immobilized form is much more vigorous than canister concept. that required for the recovery of Pu from the simple The radiation field from spent fuel (both uranium solid-solution ceramic form of spent fuel. oxide and MOX fuel) will give a measure of protection With increasing age, the radiation dose decreases against a terrorist group but that field decays with time. for both the spent fuel and the immobilized form. The Between discharge and one year, the gamma radiation ease of dissolution of the solid solution of Pu-U oxides dose falls by about a factor of 10, another factor of of spent fuel become easier as the radiation dose to the about 10 between years 1 and 10, and then falls roughly 137 matrix increases. Therefore with aging, spend fuel with the half-live of Cs for years 10 to about 300. become easier to process. Although the radiation field Whereas fuel which has recently been discharged from will diminish with age, the actual mechanical difficulty the reactor exhibits a lethal radiation dose, 10 year old of removing the cans of plutonium from the canister of spent fuel does not. With the easy of which spent fuel glass does not diminish with time. Neither does the can be chopped using water shielding and then leached difficulty of dissolution of the mineralized plutonium with hot nitric acid, spent fuel aged for ten years no ceramic. While it is true that the reduction in radiation longer has a sufficient radiation field to be a significant field allows longer human contact with both forms, this deterrent to any knowledgeable group. reduction in dose foes not translate into a equal At the proposed irradiation rate of 3 tonnes reduction in the difficulty of Pu recover for the two weapons-grade plutonium per year, it will take 17 years forms. The difference in the difficulty of recovery of

PflUTICAL ASPECTS I IBS Pu from the immobilized form over once-through United States and Russia are at about their midpoints. burning of weapons-grade plutonium and over uranium The total amount of Pu will be 15 to 20 times as much. oxide power reactor fuel increases with age. At the present processing rates, there will also be twice as much separated weapons-useable Pu in the From the host country standpoint, one must commercial sector, as the United States and Russia have remember that the United States and Russia will be declared excess. holding an undisclosed amount of plutonium in reserve. Since neither country has the equipment neither Conclusions installed nor developed necessary to recover plutonium from the ceramic immobilized form, it makes no sense Excess Pu, mineralized in a ceramic matrix and for either the United States or Russia to do the R&D incased in HLW glass, is a less attractive target for and construction necessary to recover the Pu from the terrorist groups than either aged, irradiated weapons- immobilized form. grade MOX foel, or aged, U oxide spent fuel. This is From a terrorist group standpoint, there is more especially true after the Russian and United States' Pu than 10 times as much weapons-usable plutonium in Disposition Programs have been completed, until the spent fuel that will be approaching an average age of material (spent MOX fuel or the immobilized form) is 30 years when the Pu Disposition Programs of the stored in a sealed, repository* SK01K0118 RUSSIAN FEDERAL NUCLEAR CENTER VNIIEF - POSSIBILITIES OF INTERNATIONAL COOPERATION

The lecturer - Shaburov Vasilii M. Co-author - Mozharov R, V.

Russian federal nuclear center - VNIIEF The Russian Federation Nuclear Center - the A11- - high-temperature plasma physics; Russian Experimental Physics Research Institute i- development of physical models of complex physi- (RFNC-AREPRI) is a major scientific-technical center cal processes and the creation of mathematical of Russia capable of solving the most difficult problems methodologies and software based on these mod- in the interests of defense, science and the national els; economy. There was a time when the RFNC-AREPRI - energy; played a decisive role in liquidating the U.S. monopoly on nuclear weapons and ensuring half a century of - medicine; world civilization without global political and military - ecology; conflicts. Today, RFNC-AREPRI specialists are - progressive technologies for various sectors of the entrusted with the mission of maintaining and perfecting economy. Russia's nuclear shield that ensures its security and independence. As well as defense-oriented projects, International cooperation includes: (he Institute is busy developing and implementing a number of projects in the most diverse fields of science - fundamental research; and technology. - research and experiments in super-powerful ex- [At present, the Institute possesses an experimental plosive magnetic energy sources; and testing base that includes: a gas dynamic complex - arrangement of j oint production of fuel cell power for testing manufactured products and explosives, plants; irradiation facilities, nuclear reactors, laser systems, - contacts on conversion projects, etc; complexes for mechanical, temperature and climatic testing of specific manufactured products and The Institute's foreign contacts are boosted by the instruments, and an aeroballistic testing complex. The activities of the International Scientific-Technical Institute's material base, with its mathematical support, Center. is one of the most powerful in Russia. The Institute is currently concerned with its The RFNC-AREPRI employs about 20,000 conversion-connected activities, based on fundamental workers, including 9,500 scientists and engineers. and applied research including: The 50 years of the RFNC-AREPRI experience, - work on inertial thermonuclear fusion for which its scientific-experimental, mathematical and the Institute has an adequate experimental base, production-technological base, as well as its highly with unique capabilities in Russia; professional specialists presently enable the Institute - research into high-energy physics where Institute to tackle the most complicated scientific-technical scientists are world leaders; problems and carry out fundamental and applied - research on the problems of transmutation of ra- research work in various fields. dioactive waste and development of sound and en- Today, the RFNC-AREPRI is engaged in activities vironmentally pure atomic energetics; in the following principal directions: - development of a supercomputer capable of par- allel computing of complex problems of physics -properties of material under extreme pressure and at a speed of more than a billion operations per temperature; second. - gas dynamics; ;The Institute is being transformed from a top - nuclear physics; secret facility into a more or less open scientific - radiation physics; establishment ready and eager for wide-range in- - laser physics and equipment; ternational scientific-technical contacts and coop- - superpowerful magnetic fields; eration.

PflUTICAL ASPECTS 1 SK01K0119 PREVENTION OF ACCIDENT CONSEQUENCES IN PLUTONIUM

Oleg LTetchko

Duhov str.6, SarFTI,607190 Sarov (Arzamas - 16), VNIIEF, Region, Russian Federation

1. Introduction emergencies, the primary means to accident localization is' the presence of fire extinguishing On the basis of analysis of causes and equipment in FMS facilities and thermostable filters consequences of possible radiation accidents in fission in their ventilation systems; these systems should shut materials storage (FMS), there are proposed measures off automatically in the event of an accident. Storage and facilities to prevent accidents and their facilities must be resistant, without hermetic loss, to Consequences. Accident prevention systems and control explosive shock waves for maximum possible explosive of radiation in FMS have been developed and technical power. The parameters for defining safety equipment facilities for prevention of accidents are presented. requirements based on possible causes and effects of The fissile materials storage project undertaken accidents art shown in Fig. 1. by Russian and American specialists presents unique challenges concerning the quantity of plutonium to be 4. Conclusions stored and the duration of operation (100 years). Possible accident consequences associated with FMS the complexity, duration, and safety concerns of radiation hazard factors for personnel and the general the ongoing FMS project demands further investigation population can be compared with those formerly and experimentation involving equipment and associated with nuclear armaments (NA). Proposals on techniques for radiation accident control and response. measures and facilities to localize and prevent accidents Future work should be concentrated in the areas of have been drawn from NA experience in the US. and equipment reliability and possible accident scenarios. Russia and VNIIEF experience in minimizing accident The following equipment should be tested further: consequences at Chernobyl in 1986 - 88. • Protective containers; . 2. Accident Hazard Factors In FMS • Ventilation system; . • Thermostable filters; and Explosions and fires are the primary accidents • Autoblocking systems (falling locks, autolocks). leading to plutonium dispersion and its entry into the Additionally, experimental test of modefe of the environment Also, self-action chain reactions of fission following accident scenarios and their effects on can occur, which are accompanied by momentary facilities and fissile materials containers would prove dispersion of gamma-neutron radiation and beta- useful: . . gamma active fission products.' ' • Shock-wave impact and gas composition resulting Radiation accidents are also accompanied by from explosions; •.••-•••• chemical Hazard factors, which can increase accident • Temperature and gas composition in fires; consequences. Experiments with explosions (45-240 kg) in closed volumes (30 m 3 ) in the presence • Aerosol release in facilities and the atmosphere ' of organic substances demonstrated the dispersion of during explosions, fires, and self-acting chain re- the following gasses (in volume). actions; and' ••:;•"• • The influence of explosions', fires, and self-acting 3. Technical Equipment for Prevention - chain reactions on automatic control systems. Accidents and-Minimizing Their Conae- • quences '

In addition to an effective response system, FMS facilities must also be well equipped to deal with SK01K0120

THE CHARACTERISTICS OF INTAKING CS-137 AND SR-90 WITH FOOD BY OZYORSK'S INHABITANTS

Mariya Dronova

Ozyorsk Technological Institute of Moscow Physical Engineering Institute, Russia

The problem of the intake of radioactive accumulation and concentration of toxic and radioac- substances into humans through the consumption of tive materials, they are at a higher level on the ecologi- food is currently of interest to specialists. Man-made cal pyramid and, thus, the material concentration of radiation from nuclear weapons tests and nuclear power increases by approximately so many times. accidents has actually increased the global radiation As a result of the analysis of food composition, average. 14 components were marked: milk, salt, water, veg- It is known that radionuclides contained in food etables, potatoes, greens, sugar, tea, fish, bread, fruits, are a source of additional exposure to people. macaroni, and groats. For these components, the quan- Radionuclides can move in biological systems. tities of radionuclides were measured. Composition of food was determined by eating standards. The tech- The greatest contribution to radioactive nological considerations were also taken into account. contamination in food is from 1-131, Cs-137, and Sr- For example, butter, cheese, sour cream, yoghurt and 90. The following are classifications for the others were transferred to milk. Sausage and meat prod- methodology of studying food intake: ucts were divided into their different ingredients and 1. The incidence of food use. transferred to meat. By knowing the specific activity 2. The direct use of food: of the main ingredients and the quantity of consump- a. for groups of people; tion, we defined the activity of products and the over- b. in the family; all activity of Cs-137 and Sr-90 in people's rations per c. for an individual; year. In the statistical calculations, age and gender were This work is based on a questionnaire of PA accounted for. "MAYAK" workers eating habits. Cs-137 and Sr-90 Constitutive gender and age differences have not were chosen as experimental radionuclides. Cs-137 is been established. Evidently, it is related to the fact distributed in organisms equally. It is considered to be that the quality and quantity of consumed products are a source of genetic damage after entry into the human the same. Children represent a special case. Because body. This radionuclide is almost completely absorbed of the peculiarity of their eating habit and food intake, as it passes into the bowels after peroral entry. In the they can be exposed to intensive irradiation from ra- blood Cs-137 is distributed almost evenly between dioactive isotopes that enter with their foodstuffs. The organs and tissues. composition of a child's diet, especially at the suck- It was established that Cs-137 accumulates in ling age, can change considerably with age and also muscle, kidney, heart, spleen, lungs, and liver to a higher depends on local conditions. degree. Although the half-life of Cs-137 exceeds Sr- Milk plays a more important role in the nourish- 90, Cs-137 leaves the body more rapidly. The radia- ment of children than adults. The irradiation of chil- tion dose of Cs-137 per unit of intake is less than for dren by Cs-137 and Sr-90 can be effectively estimated Sr-90. In addition, Cs-137 is absorbed by plants to a by the contamination of milk alone because the radio- lesser degree than Sr-90. Taking into account that a considerable part of the daily diet consists of vegeta- active substances come via the milk in the largest bles, it should be noted that different radionuclides get amounts. into plants in different amounts and different ways. Cs- So, we can conclude: 137 concentrates in cereal, beans, oilplants, potatoes, beets and tomatoes. Meat and milk are also a main :1. For the first time, research on the uptake of Cs- source of Cs-137 in a person's diet. 137 and Sr-90 through food was appraised for the popu- lation of the Ozyorsk on the bases of individual diets Sr-90 contributes to radiation dose for bone and estimated by a variety of foodstuffs and individual eat- marrow, almost all Sr-90 is taken up by organisms ing habits. through milk, greens, and cereal. 2. The results of this work will be used for esti- Since human beings head the list on the trophic mating the contribution of food to the uptake of radio- food chain, in conformity with the law of biological active substances and to effective equivalent dose.

ENVIRIHMEIIT & SAFETY 12B5 SK01K0121 PROBABILISTIC SAFETY ASSESSMENT AS A STANDPOINT FOR DECISION-MAKING

M. Cepin

«Josef Stefan» Institute, Reactor Engineering Division, Jamova 39, Ljubljana, Slovenia marko. cepin@ijs. si, http://www2. ijs. si/~cepin Introduction 3. Definition of implementation and monitoring pro- gram. Probabilistic safety assessment (PSA) is a tool for assessment and improvement of safety systems 4. Submittal for regulatory acceptance. reliability and nuclear power plants safety in the sense of evaluating and reducing risk. It is increasingly used D. Examples of Use of PSA in decision-making [1,2,3]. The principal role of PSA in Decision-Making in decision-making is to provide a risk evaluation of impact of issue under investigation. iSeveral proposed technical specifications (TS) changes were evaluated. The summary of proposed Role of PSA in Decision-Making changes, their acceptance and extent of use of probabilistic safety assessment (PSA) for decision- A. Features of PSA making is reported in ref. [4]. Several prerequisites for use of PSA in decision- Conclusion making exist: 1. All operating modes of nuclear power plant and It has been recognised that probabilistic safety as- all initiating events should be addressed. sessment could be increasingly used in decision-mak- ing on questions about the nuclear power plants safety. 2. The level of detail in psa should be sufficient to model the impact of addressed change. Probabilistic safety assessment does not replace the defence-in-depth principle but supports it. Even- 3. Psa should represent the modelled entity in ac- tually, there could exist a rule or regulation in which cordance with accepted practices. Independent peer probabilistic safety assessment would be the dominant review is strongly recommended. basis for decision-making, while there could exist other case in which the defence-in-depth principles would 6. Acceptance Guidelines remain dominant.

The risk acceptance guidelines are well established References in NRC Regulatory Guide 1.174, summarised in ref. [2]. Regions are established in two planes generated 1. W. E. Vesely, Principles of Resource-Effective- by a measure of the baseline risk metric along the x ness and Regulatory-Effectiveness for Risk-Informed axis and the change in those metrics along the y axis. Applications: Reducting Burdens by Improving Effec- Risk metrics that are used are the following: core dam- tiveness, Reliability Engineering and System Safety, age frequency (CDF) and large early release frequency 1999, Vol. 63, pp. 283-292 (LERF). Acceptance guidelines are established for each 2. M. A. Caruso, M. C. Cheok, M. A. Cunningham, region. E.g. if the issue under investigation results in G. M. Holahan, T. L. King, G. W. Parry, A. M. Ramey- the decrease of CDF, it will be considered, that this Smith, M. P. Rubin, A. C. Thadani, An Approach for Using issue has satisfied the relevant principle of risk-in- Risk Assessment in Risk-Informed Decisions on Plant- formed regulation with respect to CDF. Specific Changes to the Licensing Basis, Reliability En- gineering and System Safety, 1999, Vol. 63, pp. 231-242 C. Decision-Making Process 3. M. Cepin, B. Mavko, Probabilistic Safety As- sessment Improves Surveillance Requirements in Tech- Decision-making process is an iterative process nical Specifications, Reliability Engineering and Sys- consisting of number of steps: tems Safety, 1997, Vol. 56, pp. 69-77 1. Definition of issue under investigation. 4. L. Fabjan, R. Jordan-Cizelj, M. Cepin, Assess- ment of Krjbko NPP - Technical Specification Changes 2. Performing the engineering analyses, which in- Related to the Surveillance Requirements, IJS-DP- clude risk-based methods (e.g. Psa). 7737, 1998 ZflB | POSTER PttSEKTATIflNS SK01K0122 SAFETY PREDICTION FOR BASIC COMPONENTS OF SAFETY-CRITICAL SOFTWARE BASED ON STATIC TESTING

Han Seong Son and Poong Hyun Seong

Dept. of Nuclear Engineering, KAIST 373-1 Kusong-dong, Yusong-gu, Taejon 305-701, Korea

Abstract combines the major factor with the quality factor for The purpose of this work is to develop a safety ** comPonent3> j"«*.» ***** *•* «» *e prediction method, with which we can predict the risk ^^sures proposed m this work. TJe appl.cat.on to a ofsoftwarecomponentsbasedonstatictestingresults f^?? *t "TO™ TT TT"^ ** at the early devetopment stage. The predictive model feasiblllty of *• "^ Predlctl0n method

EtfVIfffllfM£HT & SAFETY 1207 SK01K0123 INFLUENCE OF HEAT LOSSES AND ACCUMULATED HEAT UPON THE ACCIDENT PROCESS EVOLUTION

Ilya V. Gashenko

Electrogorsk Research and Engineering Center (EREC) on Nuclear Power Plants Safety Russia, 142530, Moscow Region, Electrogorsk, Bezymyannaya str. 6 Abstract Methods

Studies of accident processes in the primary circuit Using the RELAP5 thermal-hydraulic computer of nuclear power plants are limited to experiments at code, the author performed the study on the influence scaled mockups - test facilities. Presently, it is the • of the aspects of ambient heat losses and accumulated common understanding that only the qualitative heat upon the accident process evolution. reproduction of the phenomena being studied is This study comprised performing several possible when simulating the accident processes at calculations (the heat losses and the accumulated heat these test facilities. More or less, correct modeling of cases respectively) for two ISB-WWER tests. The tests the reference nuclear power plant is possible only being considered are two similar experiments involving through use of the computer codes based on the intermediate break in the upper plenum with the high- mathematical models of the two-phase flows using pressure pumps of the Emergency Core Cooling System numerous empirical correlations. Test facilities are (ECCS) activation and hydro-accumulators being used for validation of these computer codes. unavailable. The difference between the two tests is In the case of an accident process, it is crucial that the location of the ECCS cooling water injection. simulations at the test facilities correctly account for ambient heat losses and accumulated heat because both Results under- and over-estimation of these factors might lead to the significant distortion of the whole accident The important results on the influence of the process being simulated. ambient heat losses and the accumulated heat upon the accident process evolution were obtained in the course The EREC operates two integral test facilities of the study performed. simulating the primary circuit of the Russian designed pressurized water reactor WWER - 1000. This paper deals with two experiments performed at the ISB- Discussion WWER Integral Test Facility which is the ERE's pilot test facility. The results obtained could be used as guidelines for issuing the corrective measures on potential The ISB-WWER simulates the reference reactor experiment scenario change at the stage of planning WWER-1000 with a volumetric power scale of 1:3000. over the future experiments at the EREC's test facilities. All the elevations are kept to a 1:1 scale. The primary circuit of the ISB-WWER consists of intact and broken Conclusions loops that are connected to the reactor model. These two loops are simulating four loops of the reference The results of the study are presented in the paper. NPP with WWER-1000 reactor. The loops' mass flow The influence of the mentioned factors was illustrated rates ratio is 1:3. The reactor model consists of the by the example of two similar ISB-WWER experiments downcomer, the core simulator, the by-pass section and using the RELAP5 thermal-hydraulic computer code the upper plenum. The facility is designed for as an instrument. operations at a maximum pressure of 25 MPa and a core simulator power of 1.8 MW.

2Q81 POSTER PflESEtfTfllfQNS SK01K0124 POST-TEST CALCULATION OF THE LOOP SEAL 1% LEAK ON THE PSB-WER TEST FACILITY WITH THE HELP OF THE CODE RELAP5/MOD3.2

A. V. Kapustin

Electrogorsk Research and Engineering Center, Electrogorsk, Moscow region, Russia Abstract PSB-VVER is operated in Electrogorsk Research and Engineering center on NPP safety to study With the help of the RELAP5/MOD3.2 code a thermal hydraulic processes under VVER LOCA con- post-test calculation was done of the 1% leak from cold ditions. The PSB-VVER main goal is to form an ex- leg horizontal part at 2/3 HPIS failure perfomed on perimental data base to validate computer codes. PSB-VVER test facility in 1999. PSB-VVER is a four loop installation, which is Methods structurally similar to VVER-1000 with a power-volu- metric scale 1:300. The circulation loop consists of a RELAP5/MOD3.2, post-test calculation steam generator, pump, cold and hot pipe-line. All loops are connected to the reactor vessel model. PSB-VVER Results includes also a pressurizer and ECCS. The calculation model contains all basic equip- The nuclear power plants potentially endanger an ment of PSB-VVER, except for pumps (experiment environment with pollution by fuel fission radioactive was performed at natural circulation through pump products. Today, NPP evolution is based on new in- bypasses). The nodalization scheme consists of 420 creased safety designs. NPP designers should guaran- control volumes, 438 junctions and 361 heat control tee safety. To justify NPP safety, one needs clear rep- volumes (metal structures and thermal insulation). The resentation concerning accident relevant thermal hy- leak unit was simulated as a diaphragm. draulic processes, that results in the necessity to simu- late them. Simulation is both possible, experimental Good agreement between calculation and experi- and numeric. Today it is arguable, in what extent the ment was achieved. The initial stage of the regime is test facilities reproduce NPP accident transient regimes. specified by subcooled water release and fast depres- They reproduce partially accident transients. Thus, NPP surization, which results in high pressure inj ection sys- simulation can be implemented with computer soft- tem operation. At two-phase mixture discharge the pres- ware, such as best estimate thermal hydraulic codes: sure decrease diminishes. Coolant phases separate, loop TRAC, RELAP, ATHLET, CATHARE, TRAP. Nu- seals form, and core water level depletes with core heat- meric simulation adequacy is validated on experimen- up before hydroaccumulator actuation. tal data. For that closure relations are obtained with separate effect tests, integral effect tests being used to Discussion verify nodalization schemes and codes capability to quantify accident processes. To increase simulation quality, boundary conditions were tuned: thermal insulation heat Best estimate thermal hydraulic code is a power- conductivity (heat losses variation), leak critical mass ful means to simulate numerically vapor-liquid systems. flow rate coefficients (for subcooled water and two- However, results obtained depend on calculation model phase mixture). (nodalization scheme), which the code user develops. So, the model adequacy relates to a user understand- Conclusions ing of the simulated processes. With the help of the RELAP5/MOD3.2 code a The post-test calculation was done of the post-test calculation was done of the 1% leak from the experiment on the PSB-VVER test facility. Good cold leg horizontal part at 2/3 HPIS failure performed agreement between calculation and experiment was on PSB-VVER test facility in 1999. achieved.

ENV1S0HMEHT & SAFETY 1209 SK01K0125 THERMAL-HYDRAULIC CODES VALIDATION FOR SAFETY ANALYSIS OF NPPS WITH RBMK

Brus N.A., Ioussoupov O.E.

Electrogorsk Research and Engineering Center for NPP Safety, Russia Abstract Discussion This work is devoted to validation of western The phenomena of counter current flow and coun- thermal-hydraulic codes (RELAP5/MOD3.2 and ter current flow restriction were analyzed for different ATHLET 1.1 Cycle C) in application to Russian heat powers during analysis of the first standard prob- designed light water reactors. Such validation is needed lem. Counter current flow restriction was observed in due to features of RBMK reactor design and thermal- different cross sections of the channel depending on hydraulics in comparison with PWR and B WR reactors, the heat power. for which these codes were developed and validated. the thermal hydraulic instability is characterized These validational studies are concluded with a by flow rate oscillations. The flow rate oscillations are comparison of calculation results of modeling with the associated with variations of other thermal hydraulic thermal-hydraulics codes with the experiments parameters. Such oscillations may lead to critical heat performed earlier using the thermal-hydraulics test flux conditions. To ensure safe operation of rbmk re- facilities with the experimental data. actors it is necessary to know the range of parameters within which no thermal hydraulic instability occurs Methods as well as methods to widen these ranges. Thermal- hydraulics instability boundaries for different geom- Leading experts of russian organizations erec, rrc ki, etry and mode parameters were obtained as a result of entec and relapS code developers from ineel (usa) have the analysis of the second problem disscussed. developed a validation procedure and prepared general Comparison of calculated results with experimental guidelines. A list of processes/phenomena was developed data was performed for the standard problems analyzed. and each entry was assessed for two criteria: importance It is suggested to discuss the following questions for safety of rbmk and degree ofknowledge. Previous relap5 as well: validation work was assessed for sufficiency. Asa result of 1. Capabilities of large-scale integral thermal-hy- analysis of experimental data available in russia a prioritized draulics test facilities; list of potential standard problems was defined, 2. Capabilities of up-to-date thermal-hydraulics codes; Results 3. Computer thermal-hydraulics codes validation Validational studies have been performed for two procedure; standard problems from the list of potential standard prob- 4. Numerical methods of adequacy assessment of lems. Independent review of the results was performed in modeling with the code of particular phenomena or russian and in the usa. The modeled tests are: process. 1. Countercurrent flow investigation with rrc ki ex- Conclusions periment "stop of flow rate at the inlet to rbmk fuel channel when removing residual heat" performed at From the analyses 1he following conclusions were drawn: ks test facility. 1. Thermal-hydraulics codes relap5 and athlet in The objective of the experiment was investigation general are capable of modeling the processes and phe- of fuel assembly model cooling under condition of nomena occurring in the selected tests; phases counter current flow. 2. For some tests under investigation, consider- 2. Thermal hydraulic instability in parallel heated able disagreement between experimental data and cal- channels performed at erec 108 experimental test facility. culational results is observed; The objective of the experiments was to investi- 3. Experimental data available are of low quality, gate the boundary of thermal hydraulics instability in therefore for complete validation of thermal-hydraulics parallel channels. codes in application to rbmk new experiments with up- As a result of analysis of the first experiment, time to-date thermal-hydraulics tests facilities are needed; periods between complete stop of flow and deterio- 4. Numeric methods of adequacy assessment are rated heat exchange onset were determined. required to make a final conclusion on adequacy of Boundaries of thermal-hydraulics instability in modeling with the code of the process or phenomena dimensionless coordinates were obtained as a result of under investigation. These methods should be accepted the second experiments. by leading organizations in the field of npp safety. ?1D! POSTER PRESEPTTATIBMS SK01K0126 PROPAGATION OF A WAVE THERMAL DETONATION IN VIEW OF MICROINTERACTION MODEL

O.LMelikhov, V.I.Melikhov, A. V.Sokolin

Electrogorsk Research & Engineering Center , Electrogorsk, Moscow region, Russia

Abstract - sop.

'• -' • A new mathematical model of thermal detonation 400 • in the melt-water mixture is proposed. The model uses : 1 300 Eulerian multiphase flow equations. The micro- y / --/• e interaction concept is used for description of melt : 3 200 fragments-coolant interaction. Coolant (steam-water 7 • mixture) is divided into two phases. The first phase is [2. the part of the coolant, which immediately interacts : with generated fragments of melt drops. The other part | is the second phase. The model has been applied to DliUncs

Methods The calculations were made under the following parameters: We have considered transient one-dimensional Initial pressure 0.1 MPa planar propagation of the thermal detonation wave in the system "coolant-water". The initial mixture ahead Initial melt temperature 3000 K of the detonation front is assumed to consist of melt Initial melt volume fraction 0.1 droplets, water and steam. The melt droplets are under Initial void fraction 0.5 film boiling. Behind the detonation front the droplets Initial droplet size 0.005 m are fragmented and water is heated by the fragments. To describe this situation we have used four different The solution domain represents a closed vessel. components: melt droplets, fine melt fragments with Thus, the only boundary condition is zero velocity at surrounding coolant (microinteraction phase) and the vessel walls. To simulate triggering, a high pressure coolant liquid. Each component has its own velocity (10 MPa) was determined in the first numerical cell. and temperature fields. Fig. 1 shows calculated profiles of pressure for subsequent moments (every 0.5 ms). Length of the Mathematical description of the process is based unsteady part of wave propagation is about 4 m. on the multiphase flow equations, where the presence Maximum pressure is about 470 MPa, velocity is 450 of each component is specified by a volume fraction m/s. and all of the components are at a common pressure. The melt droplets and fragments are assumed to be incompressible. Discussion The partial differential equations are solved by a Comparison of results obtained with the proposed finite difference method which employs the usual model and previously used model (without staggered grid arrangement. All convective terms are microinteraction phase) have been carried out. The approximated using upwind differencing. All source main discrepancy concerns the length of the unsteady terms are evaluated using new time values. Global part of wave propagation. The new model gives ~ 4 m iterations are used for implicit consideration of in comparison with ~ 20 m obtained by the old model. convective terms. Conclusions Results Preliminary results of the task about non-stationary Using the numerical solution above we have propagation of thermal detonation in system water-drop implemented a simulation of transient thermal of melt corium were obtained, using the micro- detonation propagation in the corium-water mixture. interaction model.

EHVIBONMEHT & SAFETY 1211 SK01K0127 DETAILED SPECTRA DATA FOR THE INTERNATIONAL HANDBOOK OF EVALUATED CRITICALITY SAFETY BENCHMARK EXPERIMENTS

Yevgeniy Rozhikhin

Institute of Physics and Power Engineering (IPPE) 1 Bondarenko Square 249020 Obninsk Russian Federation Abstract Metal, compound, and miscellaneous systems are subdivided into fast, intermediate, thermal, and mixed Critical configurations described in the spectra systems, depending upon where the majority "International Handbook of Evaluated Criticality of the fissions occur. However, subdivisions are not Safety Benchmark Experiments" [1] are being applicable to solution systems. recalculated and detailed spectra data are being collected. These spectra data represent a significant In this handbook, fast, intermediate, and thermal enhancement to the handbook and will enable criticality systems are defined as systems in which over 50% of safety practitioners to better understand the the fissions occur at energies over 100 keV, from applicability of each configuration in the handbook. 0.625 eV to 100 keV, and less than 0.625 eV, respec- These data are published in a document that is separate tively. Systems for which over 50% of the fissions do from the handbook, but are linked to the appropriate not occur in any one of these three energy ranges are evaluations in the handbook. Eventually, spectra data classified as "mixed" spectra systems. for every configuration that appears in the handbook will be available. These data were beginning to appear Recently, the Idaho National Engineering and Envi- in the 1999 version of the Handbook. A description of ronmental Laboratory (INEEL) and the Institute of Phys- the spectra data is included in this paper. ics and Power Engineering (IPPE) in Obninsk, Russian Federation, began a collaborative effort to recalculate Results and Discussion every configuration in the Handbook and collect the spec- tral characteristics of each experiment. Included in these The 1999 Edition of the "International Handbook data are the energy corresponding to the average lethargy of Evaluated Criticality Safety Benchmark Experiments" of neutrons causing fission; the average energy of neu- contains 268 evaluations with benchmark specifications trons causing fission; the percentage of the flux, fissions, for nearly 2240 critical or near critical configurations. and captures that occur in the fast, intermediate, and ther- The evaluations are divided into seven volumes repre- mal energy ranges; the percentage of fissions and cap- senting seven general types of fissile systems: tures by isotope over the core region, and vE/S,. A plot of the neutron spectrum is also provided for bounding Volume I: Plutonium Systems configurations in each evaluation. Volume II: Highly Enriched Uranium Systems (wt.% 2J5lte60) Conclusions Volume III: Intermediate and Mixed While small amounts of spectra data are provided Enrichment Uranium Systems in the International Handbook, much more can be done (10

2121 POSTER PRESENTATIONS SK01K0128 VERIFICATION OF THE CODES RELAP5MOD3.2 AND CATHARE 2 V1.4 BY MODELING THE LONG TERM COOLING IN THE WER-640 POWER PLANT AFTER A LARGE BREAK LOCA ON THE PACTEL FACILITY

Alexander Alexeev Obninsk Institute of Nuclear Power Engineering, Russia Jdzsef Bdndti, Eero Virtanen Lappeenranta University of Technology

One way to increase the safety of a nuclear power MOD3.2 and CATHARE 2V1.4 are used to calculate station is to use passive safety systems in them that do the behavior of the PACTEL facility during the Long not require any intervention on the part of the operator Term Cooling after a Large Break LOCA test. in the event of potentially dangerous design basis RELAP5/MOD3.2 is an American computer code accidents. This approach, which is in line with present that has been developed for the best-estimate transient day global trends in the development of nuclear energy, simulation of the light water reactor coolant system is realized in the concept of safety of the new generation during a severe accident. The code models the coupled of nuclear power stations of moderate capacity with behavior of the reactor coolant system and the core WER-640 reactors that are now in their designed phase. during a severe accident transient, as well as large and A fundamental element of this concept is equipping the small break loss of coolant accidents and operational VVER-640 nuclear power station with a set of passive transients, such as anticipated transients without scram, systems that can handle the situation during shutdown loss of feedwater, and loss of flow. or cooldown of the reactor, and also with lengthy removal CATHARE 2VI.4 is a French computer code, of the residual heat (in the case of a hermetically sealed developed by CEA, EDF and Framatome for PWR safety coolant circuit, and also in the event of depressurisation analysis. It simulates the thermalhydraulics of primary and of this circuit). These systems do not require any actions secondary circuits with the two-fluid model, and the fuel on the part of the operator and do not require power thermomechanics. 0-D, 1 -D and 2-D modules are available supply from an external source for at least 24-h. with walls, heat exchangers, and transport of two In nuclear power plants a sufficient cooling of the honcondensable gases, transport of boron and activity. It reactor core must be confirmed in all operational may simulate all accidental transients including the Large transients. In order to gab more detailed information Break LOCA using a specific Reflooding model. about the complicated physical phenomena taking place It was established that, depending on performance in nuclear power plants sophisticated computer codes parameters and the hydraulic characteristics of the and test facilities have been designed. circuit under consideration, three different modes of In cooperation of Lappeenranta University of natural circulation are possible: (a) the evaporation Technology (LUT) and the Technical Research Center of mode with obstruction of the natural circulation in the Finland (VTT), a test facility for V VER-440 reactors has circuit; (b) the oscillation mode with low-frequency been constructed. PACTEL (Parallel Channel Test Loop) pulsation of the parameters; (c) the mode of steady is the largest facility to model pressurized water reactors natural circulation. A comparison of experimental data (PWR) with horizontal steam generators. The research is with the results of calculation conducted using the codes focused on different loss-of-coolant accidents (LOCA), RELAP5/MOD3.2 and CATHARE 2VI.4 allows us natural circulation and operational transients. For carrying to assert the following: (1) these codes can simulate all out experiments connected with WER-640 PACTEL three stages of the transients that were observed in the (Facility was reconstructed. The main goal of these experiment; (2) acceptable quantitative agreement was experiments is to study the thermal hydraulics that will obtained between calculations and the experimental occur during decay heat removal to the emergency cooling data on the amplitude of oscillations of most of the and fuel storage pools in the VVER-640 should a hot or measured parameters; (3) calculations gave the same cold leg of the primary coolant loop break. frequency range of oscillations of the parameters as The reasonable way to predict the thermal- that observed in the experiments. hydraulic behavior of a reference power plant is to use A comparison of experimental data and calculations thermal hydraulic computer codes. Experiments in test showed that the codes RELAP5/MOD3.2 and facilities provide data to verify the code calculation. CATHARE V1.4 could simulate with a good correlation Pata from experiments examining decay heat removal behavior of the most parameters and characteristics, to emergency cooling and fuel storage pools in the which were obtained at the PACTEL Facility. VVER-640 using the PACTEL could verify thermal- Thus, verification of codes RELAP5/MOD3.2 and hydraulic codes at low pressures and low flow rates. CATHARE 2V1.4 that was done in the present stage of RELAP5/MOD3.2 and CATHARE 2VI.4 investigations shows thatthese codes describe sufficiently, thermal-hydraulic codes for simulating accidental and with acceptable accuracy, the thermal and hydraulic transients in reactor systems. In this work RELAP5/ processes that take place during these experiments. BlVIBBNMEirr & SAFETY i 213 SK01K0129 CREATION OF CADASTRE OF GROUND CONCENTRATION OF CHEMICAL POLLUTING SUBSTANCES IN THE CITY OF OBNINSK TAKING INTO ACCOUNT AERODYNAMIC SHADOWS OF BUILDINGS

O.Malik

Obninsk Institute of Nuclear Power Engineering, Obninsk, Russia

The summary volume «Protection of atmosphere of contamination, belonging to the enterprise) are and extreme allowable contamination (EAC) for the described. city of Obninsk « (hereinafter Summary volume) was developed in 1996 under aegis of State committee on The data bases function in a specialized program protection of the environment of the Kaluga region and complex; The database structure includes the programs Administration of. Obninsk. Now 96 industrial accounting ground concentration, converting of data enterprises and research institutes of the city are bases, and creation of a topology base. The list of the included in the Summary volume. PC which can be used is annually affirmed by the State Committee of Protection of the environment. When the The summary volume is used for Summary volume was originally developed, the PC a) Solving town-planning problems; first of all for «Ecologist 1.13» was used; it was upgraded up to the development of a General plan, for choice of places version «Ecologist 2.20» in 1998. Now, the PC is for construction of new objects of industrial and social upgraded to account for aerodynamic shadows of assignments. . buildings. To maintain the Summary volume by taking into account buildings, the PC «Prism - region» by b) Receiving information about expected ground «LOGUS» firm is selected. concentration of harmful substances in the city for prevention of a threat to the population at unattended During the work, the heights of 1263 buildings accommodation of the enterprises. • were determined, the data bases for the new PC were c) Fulfillment of requests of the State normative modified and calculations of ground concentration of documents in a sphere of Protection of an environment. 150 polluting substances from 1248 sources were made. The experts of NTP «ROSEKO» periodically Auxiliary software used in the work include: calculate ground concentrations of polluting substances - Operational system Windows 98/NT; in the city and analyze obtained results with the use of tabulated and graphic materials. - The package Autocad-14 in which the map of Obninsk is supported. It is necessary for point The summary volume represents a set of positioning of sources of contamination; specialized data bases in which the parameters of sources of contamination (coordinates, heights, -MS Excel and MS Word forprecomputations and diameters, temperature, velocity of gas streams, power maintenance of document circulation.

2141 PISTE! FttSEITATlIK SK01K0130 DOSE LOAD AT ORAL ENTRANCE AND INJECTION OF RADIONUCLIDES IN HUMAN ORGANISM

S.Korkoshko

Institute of Physics and Power Engineering, Russia

The norms of radiation safely NRB-96 are applied With the purpose of perfecting usage of sources to the following kinds of effects of ionizing radiation of ionizing radiation in medicine and the decrease of on humans: levels of irradiation of the patients, the federal organs of public health services with Goskomsanepidem- • irradiation of staff and population in conditions nadzor of Russia establish objective levels of medical of normal operation of plants which have sources irradiation in rontgenology, radionuclide diagnostics of ionizing radiation; and therapies based on the best world standards. The • irradiation of staff and population in conditions levels should also serve as a base of development and of radiation failure; perfection of a methodology of radiological medical • irradiation of workers of the industrial plants and procedures, designing and producing the equipment, population by natural sources of ionizing radia- radiopharmaceuticals etc. tion; There are several paths of entrance of radionuc- • medical irradiation of the population; lides in an human organism: inhalement (through A primary purpose of radiation safety is to protect respiratory paths), oral (with food and water) and the health of the people from the harmful effects of medical (introduction of a preparation immediately in ionizing radiation by observance of main principles and blood). In medicine prepared radiopharmaceuticals norms of radiation safety without unreasonably arrive in human organism with injections and tablets. restricting useful activity with radiation in various areas At the present time, quantitative valuations of favour of economy, science and medicine. and harm in medical practice do not exist. The principles of monitoring and restricting The main purpose of the given work is the deter- radiation effects in medicine are based on derivation mination of probability evaluation of risk of damage of diagnostic information, necessary and useful to the to a human organism from application of patient, or therapeutic effect at minimum possible levels radiopharmiaceuticals in diagnostics and therapy. The of irradiation. The maximum limiting dose is level of negligible risk divides the area of optimiza- determined and principles of the substantiation with tion of risk and the area of unconditional acceptable the indications of radiological procedures and risk and equals to (1 .OE-06) for one year. optimization of safety measures are used.

EKVIflllMEIT & SAFETY | 215 . . SK01K0131 DEVELOPMENT OF NEW EXTRACTANT UTILISED BY COORDINATION PROPERTIES OF DIGLYCOL AMIDE (DGA) TO TRIVALENT CURIUM AND LANTHANIDES

T. Yalta, H. Narita, S. Tachimori andN.M. Edelstein1

Japan Atomic Energy Research Institute, Lawrence Berkeley National Laboratory1

Abstract carbonyl group and Cm were about 2.43 E, and the Cm-ether oxygen distance was, therefore, not observed Coordination properties of diglycolamide (DGA) in the primary coordination sphere of Cm complex. to trivalent curium and lanthanides were studied by EXAFS and X-ray crystallography. EXAFS and XRD Discussion data suggested that DGA would behave as tridentate and semi-tridentate ligand in coordination to lanthanides and DGA is good ligand for trivalent lanthanide and curium, respectively. This difference in the coordination actinide extractions compared to the other amides, and properties would be very useful for development of new furthermore, has an advantage in the disposal of used extracting for trivalent lanthanide-actihide separation. extractant because of its incinerablity. The structural determination by both single crystal XRD and EXAFS Methods showed that DGA essentially coordinated to lanthanides in a tridentate fashion in both solution and • EXAFS measurement solid state. This result suggested that the high extractability of DGA in lanthanide extraction would The sample solutions were prepared by dissolution be due to complex stabilization by coordination of eHier of metal (Nd, Er, Cm)-DGA complex into ethanol. The oxygen. In contrast, the EXAFS data for Cm complex curium sample was prepared in the glove box. EXAFS suggested that DGA would coordinate to curium with spectra at the L^-edge of metals were measured at the • only semi-tridentate mode in the coordination to BL 4-1 of SSRL in Stanford University, U.S.A. and curium. Actually, the distribution ratios of curium were BL 27B of PF in KEK, Japan. The experimental data slightly lower than that of lanthanides in our previous was analyzed by using a WinXAS97 program. The Work. Accordingly, the coordination of ether oxygen theoretical phase shift and back scattering amplitude to metals would be very important for their extraction. parameter was calculated by FEFF 7 code. X-ray crystallography o/Er-DGA complex Phenyl The single crystal of Er-DGA complex: group ^^jj^jjj was prepared in the mixture of ethanol and 2-pentanol. The obtained crystal was nwasuredby aWgalmAFC7Rdiffractometer equipped Carbonyl with graphic monochromated Mo-Kcc radiation. The data oxygen Erbium analysis was performed by the TEXAN crystallographic package of Molecular Structure Corporation.

Results Oxygen of water Figure 1 shows the crystal structure of Er-DGA complex. From the results, DGA coordinated to Er3+ Figure 1 View of the structure of Er-DGA with tridentate fashion. The XRD and EXAFS data complex. Chemical formula: showed that the bond distances between oxygen of the [Er(C,,H20NIO})J(HJO)J]CV6HIO carbonyl group and Er are about 2.30 E in the crystal, Crystal system: Orthohombic; Lattice Parameters: and 2.33 E in solution. The ether oxygen distances are a=25.351(4)A, b=10.252(3)A, c=18.114(2)A; about 2.47 E in both crystal and solution state. Two Space Group: Pcca DGA were observed in a complex inborn crystal and solution. Accordingly, both crystal and solution Conclusions structures were very similar in the structural parameters XRD and EXAFS determined the detailed structure and stoichiometry. The EXAFS data for Cm-DGA of lanthanides and Cm complexes. The high extractability complex showed that the first coordination sphere of trivalent lanthanides and actinides would be due to the consists of only oxygen of water and carbonyl group. coordination of ether oxygen to these metals. The bond distances between oxygen of water and 711 i PflSTII PESEITAT1I1S SK01K0132

A NEW SYSTEM FOR THE MEASUREMENT OF THE SPACE RADIATION

T. Pdzmdndi1,!. Apathy1, R. Beaujean2, S.Deme1

1 KFKI Atomic Energy Research Institute 2 Kiel University, Hungary

Abstract of the earlier applied components of the system, i.e., the Pille system and the single-axis silicon telescope Radiation from space mainly consists of charged DOSTEL, were used in earlier flights. The small size, heavy particles (protons and heavier particles). Due extremely wide range system - consisting of a set of to this fact, the effective dose significantly differs from bulb dosimeters and an on-^board TLD reader - was jthe physical dose. Current measuring equipment is not used mostly for space flight on different space stations fully suitable to measure both of the quantities and space shuttles as well. The DOSTEL had its simultaneously. A combined device for measurement maiden flight on STS76 in 1996. The instrument is of the mentioned values consists of an on-board part of the dosimetry experiment E094 in mission 6A thermohuninescence (TL) dosimeter reader and a three- on the International Space Station. axis silicon detector linear energy transfer (LET) spectrometer. This paper deals with the main characteristics of the. new system. This system can be Discussion , applied for dosimetry of aircrew as well. The earlier Pille dosimeter could measure only the absorbed dose, and the silicon telescope was able to Methods determine the LET spectra in only one direction. The tasks in this project are the development and The physical dose (Gy) is measured with a TXD manufacturing of the three-axis silicon detector system named Pille. This system applies different TL connected to the automatic qn-board TLD Reader and materials (CaSO :Dy, 7LiF,LiF and AljOjtC) and these 4 the development of software for TLD data evaluation dosimeters are read out by a lightweight (1.5 kg) on- using gated and ungated LET spectra measured by the board reader. The reader is able to work in manual and particle telescope. in automatic mode. In the latter case, the reader meas- ures the dosimeter left in the device. This TLD system The combined device in automatic readout mode was developed in, the Space Dosimetry Laboratory of will be able to determine the physical dose and the the Atomic Energy Research Institute, Budapest. dose equivalent in the direction of the three axes. This complex system will be applicable for on-board The radiation weighing factor (RWF) is used to % calibration and the high temperature method can be convert the physical dose to the dose equivalent (Sv). accomplished by using the complex analysis of the The RWF is a function of the LET of the concerned heating curve. radiation. The LET value of the space radiation shows a wide range distribution between 0.3 to 200 keV/um. Conclusions The DOSimetry TELescope (DOSTEL) is based on two Si detectors, ft was developed with the Monitoring exposure of astronauts to space cooperation of DLR and the University of Kiel in 1995. radiation is very important because of the high dose This distributionis measured by a three-orthogbnal-axis rate due mostly to high-energy protons and alpha silicon detector telescope and the evaluation software particles. It will be possible to simultaneously converts the LET spectrum to an averaged RWF. determine the absorbed dose and the dose equivalent with the complex DOSTEL-Pille system. The given Results system can be applied for aircrew and high-energy laboratory personal dosimetry as well. The new combined system is under development and the prototype will be ready in the year 2000. Both

EHVfBBHMEIIT & SAFETY I 217 SK01K0133

ANALYSIS OF EFFECTS FILLING OF STEAM LINES WITH WATER AT IGNALINA NPP

Rolandas Urbonas, Algirdas Kaliatka

Laboratory of Nuclear Installation Safety Lithuania Energy Institute Breslaujos 3, LT-3035 Kaunas, LITHUANIA Phone: +370 7 45 46 83 Fax: + 370 7 35 12 71 E-mail: [email protected] Abstract Results

Accident analysis of LOCA's performed at Ignalina During the analysis, the Ignalina NPP behaviour NPP Safety Analysis Report (SAR) [I] and was calculated for the first 1000 seconds from the independent calculations [2] showed that due to accident initiation. The calculation showed that there Emergency Core Cooling System (ECCS) water supply, are two dynamic loading peaks: at the very beginning the intact Main Circulation Circuit (MCC) side's Drum of the accident and approximately 700 seconds after Separators (DS) can be overfilled and coolant is going the accident initiation. The maximum dynamic loading to flow to the steam lines. Two-phase mixture and water appears to be at the very beginning of the accident, supply to the vertical and horizontal parts of steam Jines when the turbine isolating and regulating valve is will cause dynamic loading on the piping. closing. Hydraulic testing of steam lines is performed during the planned preventive maintenance. During the Discussion testing steam lines are filled with water. However, Filling with the coolant of DS 15 minutes after the dynamic loading on steam lines during the transients accident initiation was obtained only in the case of when the piping is filled with coolant will differ guillotine rupture of one LWC pipe. Due to the coolant considerably from the hydraulic testing. Thus, the flow into the steam lines, the piping has dynamic separate analysis should be performed as recommended loading. Calculations showed that dynamic loading due in reference [2]. to steam lines filling with the coolant is considerably Methods lower than dynamic loading due to steam supply to the turbines because of closure of the turbine control valve (TCV). A number of accident scenarios were reviewed. The worst scenarios were selected from the steam lines filling with coolant point of view: Conclusions • Partial rupture of Group Distribution Header In the paper, results of the two accidents when (GDH) with flow are of 139 cm2; steam lines are filled with coolant are shown. Analysis • Guillotine rupture of one Lower Water Commu- revealed that the maximum dynamic loading occurs at nication (LWC) piping with loss of site power. the very beginning of the accident due to steam supply to the turbines because of closure of the turbine control For the analysis of the earlier stated problem, the valve (TCV). The dynamic loading due to the steam best estimate system analysis code RELAP5/MOD3.2 lines filling with the coolant is considerably lower. was employed. 1. Ignalina NPP Safety Analysis Report, 1996. 2. Review oflgnalina NPP Safety AnalysisReport, 1997.

2181 POSTER PSESEKTAnONS SK01K0134

THERMAL-HYDRAULIC ANALYSIS OF IGNALINA NPP COMPARTMENTS RESPONSE TO GROUP DISTRIBUTION HEADER RUPTURE USING RALOC4 CODE

Egidijus Urbonavicius

Lithuanian Energy Institute, Breslaujos 3, LT-3035 Kaunas, Lithuania Introduction Results and discussion

The Accident Localisation System (ALS) of The maximum pressure in break node occurs at Ignalina NPP is a containment of pressure suppression about 15 s and reaches 142 kPa (abs.). After reaching type designed to protect the environment from the the maximum value the pressure reduces due to: the dangerous impact of the radioactivity. The failure of ALS decrease of energy release from the break, condensation could lead to contamination of the environment and of steam in the condensing pools and the release of prescribed public radiation doses could be exceeded. clean air into the environment in the initial phase of the accident. The purpose of the presented analysis is to perform long term thermal-hydraulic analysis of compartments The closure of the flaps on connections to the response to Group Distribution Header (GDH) rupture environment after 300 s leads to pressure increase in and verify if design pressure values are not exceeded. the whole ALS. Later on the accident-generated energy exceeds Short description of ALS and RAL0C4 the capacity of the energy sinks (mainly the active model CTCS and heat transfer by walls) and pressure increases again. After about 2000 s the pressure in compartments The condensing pools form a barrier dividing the before the condensing trays equalises as well as the whole ALS into two distinct parts. The compartments pressure in the compartments behind it. before condensing pools are designed to withstand high pressure. The steam generated by the accident After about 3.5 hours the energy discharged by condenses while bubbling through the water resulting the break and the energy removed by active systems in the lower pressure beyond condensing pools. A part and structures equalises and a further pressure peak of of clean air is released to the environment in the initial 139 kPa (abs.) occurs. phase of the accident. This is achieved by the discharge Later on the thermal energy of the graphite stack pipes that are closed in 5 min after accident start to has been largely dissipated and decay heat has prevent radioactivity release to the environment. The diminished. The capacity of the energy sinks in the ALS release of clean air helps to reduce the peak pressure exceeds the release from the break and, consequently, in the ALS compartments. pressures in compartments before condensing trays The condensing pools are cooled by the Condenser decrease. The pressure in compartments behind the Tray Cooling System (CTCS). During long-term condensing traysdecreases slower. transients CTCS becomes a major energy sink. The The maximum pressures calculated for the CTCS pumps deliver water to the condensing pools compartments of ALS are below the design pressure and sprays. RALOC4 code applied for thermal- values and the failure of ALS is not expected. hydraulic analysis of compartments response to the break. The RALOC calculation model of ALS include Conclusions all the accident affected compartments, the structures for consideration of heat sinks/source and condensation, The behaviour of the Accident Localisation System simple junctions between compartments for gas and of the Ignalina NPP, unit 2, in the case of Group water flows, special junctions with specified pressure Distribution Header break was investigated for a time difference for opening and other features. The RALOC interval of 24 hours applying the RALOC4 code. Special model consists of 26 nodes, 84 junctions of different attention was paid to the detailed and realistic simulation type, 10 pump systems and 89 structures (heat slabs). of engineering systems and heat transfer structures. The correct simulation of the structures is of great The RALOC model was set up in correspondence to importance because they play an important role in the the assumptions applied in the determination of the mass long-term analysis of the accident. The orientation and energy release rates, calculated with RELAP5 code. (wall, floor, ceiling) and shape (plate, cylinder) of structures are considered in the model. All pressures calculated for the ALS compartments are below the maximum allowed design pressures. Condenser tray cooling and spray systems are modelled in detail because these systems have big influence to the energy balance in long term. The make-up of condensing pools from deaereator considered as well. ENVIROHMEHT & SAFETY I 219 SK01K0135 ACTIVATION OF BWR COOLANT IN TRANSIENTS

Sergey Grachev

MPEI, Russia

Abstract shown that its activity is not a result of penetrating fission products to coolant. After a number of This paper will discuss the elaborate activation of experiments were performed at heat power stations and BWR coolant in transients. Will will also make VK-50, it was noticed that when the power is recommendations solve the, problem of decreasing decreasing, the quantity of admixtures is increasing in coolant activity. the coolant at several degfeases and vice versa. This feature was documented in literature as "Hide-out". At Summary the Nuclear Power Plant department of the Moscow Power Engineering Institute, they have elaborately Two years ago, increasing activity of coolant was modeled this process and suggested variants of solving noticed when the boiling reactor VK-50 stopped. At this problem. One of these suggestions is optimization that time, it was suggested that this was the result of of the water regime. We have done a number of non-hermeticai'ly sealed fuel. But coolant tests have experiments, on which we will elaborate.

221 j POSe FfiESEIUTIBIS SK01K0136

SAFETY CULTURE IN FUTURE

Evgeny Barishev

MPEI, Russia

Abstract Safety culture concerning the work of NPPs has 4 aspects: studying the principles of safety culture, The ways of evolution df safety culture in future methodical basts of the performance, practical will be described here. realization of the principles, analysis of the level of safety culture in organization. Summary The research andthe practical work about all these aspects is being carried on now. Further developing The concept of safety culture is defined by the and deeper understanding of safety culture is an International Nuclear Safety Advisory Group as junportant goal for the future. It is necessary to improve follows: Studying facilities. Special training systems and new Safety culture is that assembly of characteristics methods of analysis of the safety culture levels should and attitudes in organizations and individuals which be made. National peculiarity should be considered establishes that, as an overridingpriority, nuclear plant there. That's why, in the future, further research and safety Issues receive the attention warranted by their development of safety culture should be carried on with significance. international cooperation.

EHVIBUMENT & SAFETY SK01K0137 MINERAL-LIKE URANYLSILICATES OF ONE-VALENCE METALS OF STRUCTURE A'[UO2SIO3OHJxNH2O

V.E. Kortikov and KG, Chernorukov

Nizhny Novgorod State University, Russia

Abstract of the synthesized uranylsilicates. For all analyzed The mineral-like uranylsilicates of one-valence substances in the field of valence numbers 1100-700cnv 1 1 metals belonging to a morphotropic line and 640-400 cnr ! there are strips of absorption of A'[U02Si0j0H]-nH20 are synthesized and investigated. valence and deformation fluctuations of a silicate tet- + The methods of a roentgenography, IR-spectroscopy and rahedron (SiO4). The valence fluctuations UO* are thermography were used to investigate compound struc- answered by strips 925 cm1,940 cm'1 for uranylsilicates tures and their thermal degradation. of an ammonium and lithium accordingly. For 1 1 One of the main tasks of nuclear power is the im- K[UO2SiO3OH]H2O v =938 cm , v3=883 cm . Wa- mobilization of spent nuclear fuel. Research on the ter in structure of compounds keeps molecular indi- iowest forms of uranium linkage and also perfection viduality and is shown by strips of valence and defor- of reprocessing and reuse technologies are currently mation fluctuations. Fluctuations of OH-grbup in struc- underway. ture SiO3OH, for Li[UO2SiO,OH]-1.5H20, K[UO SiO OH]-H O the strips are found in the field This paper presents the results of synthesis of 2 3 2 of 3180 cnr1 and 1380 cnr1:. In the structure mineral-like compounds of uranium and complex NH [UO SiO OH]0.5H O the width of the strip with physicochemical research* 4 2 3 2 a maximum at 3179 cm*1 and the presence of excesses on a strip with a maximum at 1415 CM"1 supposes im- Methods posing fluctuations of ammonium and OH-groups. The data of the thermal analysis show that at a The technique for synthesis of uranylsilicates with temperature of 100cC there is a dehydration of all the general formula A'fUOjSiOjOHl-nHjO (where A1 - + + uranylsilicates. The uranylsilicates of ammonium and NH4 , Li*, K ) was developed according to natural con- lithium are metastable and upon cooling to room tem- ditions of apotassiumboltwooditeK[U02Si03OH]-H:0. perature turn to initial crystal hydrates. It was also Results of the chemical analysis indicate the synthesized necessary to expect removal of water molecules in compounds have the following chemical structure: uranylsilicates of lithium and potassium results in re- NH4tUO2SiO,OH]0.5H2O, LitUO^SiOjO^l.SHjO, duction of interlayer parameter (Table 1). K[UO2SiO3OH]-H2O. The thermal stability of the synthesized com- pounds depends on the nature of the cation located in Results interlayer space. The least thermal stability was ob- served for the uranyisilicates of lithium; at 300-380°C From the roentgenographic measurements data, the the given compound is decomposed. For uranylsilicates uranylsilicates of one-valence metals are determined of ammonium and potassium, the removal of the con- to be crystallographic analogues with close parameters stitutional water from a crystal lattice occurs in tem- of elementary cells (Table 1). perature intervals 340-410°C and 500-580°C, respec- tively. Thermal degradation of uranylsilicates of Table 1. Parameters of elementary cells of lithium and potassium produced the uranates of the uranylsilicates of one-valence metals appropriate alkaline metals and SiO2; NH4[UO2SiO3OH]0.5H2O is decomposed with forma- Compounds a,A b,A c,A 6 P, tion U3OS and SiO2.

NH4[HSiUOJ0.5HjO 7.01(2) 7.03(8) 6.65(9) 05.6(0) Li[HSiUOJ1.5HjO 6.95(0) 7.19(4) 6.54(3) 06.7(8) K[HSiUO6]HjO 7.03(3) 7.05(3) 6.62(8) 05.5(5) Conclusions Li[HSiUO6] 6.89(2) 6.89(7) 6.30(3) 04.6(9) K[HSiUOJ 7.17(7) 7.02(8) 6.61(0) 05.2(5) Thus, the mineral-like uranylsilicates one-valence metals are synthesized and the features of their struc- Discussion ture and thermal stability are investigated.

The IR-spectroscopy research gives information The investigations are maintained by the Russian about functional structure and features of the structure Fund of Basic Research. ?271 PflSTIH PBESEVTAT10NS SK01K0138

SYNTHESIS AND INVESTIGATION OF URANYLPHOSPHATES URANYLARSENANES AND URANYLVANADATES OF LANTHANIDES

N.G. Chernorukov, A.V. Knyazev, S.V.Barch, O.V. Feoktistova, E.V. Suleymanov

Nizhny Novgorod State University, Russia

Abstract (U0j)2Vj07 with water solution of anitric acid (pH=3) at temperature 180°C. Currently compounds Ak(BvUO ) 6 k ; The general features of the crystal structure of (OH) nH O (Ak - Y, La-Lu, Bv -P, As, V) are of m m 2 discussed compounds are the layers of structure interest as the possible mineral-like forms of link- V k [B UO ]"-, between which the ions of metal A and age in an environment of uranium and radioac- 6 molecular H O are located. tive isotopes of lanthanides of an artificial origin, 2 Infrared spectroscopy research of dry samples of as modeling systems for study of similar com- uranylphosphates and uranylarsenates lanthanides testi- pounds of six-valence neptunium and plutonium, fies to the presence in their structure of uranyl groups and as objects, suitable for allocation and divi- (band about 925 cnvl).and orthophosphate (orthoarsenate) sion lanthanides and actinides. In order to deter- tetrahedrons (band about 1100-1000 (1000-900) cm"1). mine a role of water molecules in the formation of In the structure of uranylvanadates as individual oscilla- the crystal structure of inorganic hydrates and proc- tory systems it is expedient to allocate square pyramids esses of dehydration uranium minerals with the gen- VO :, pentahedral bipyramid UO and molecular water. eral formula Ak(BvUOXJOH) -nH O and their 5 7 m 2 Vibrations of H O molecules in IR spectra of all obtained synthetic analogues (Ak - one-, two- and three- 2 compounds are observed as bands, v (H O) = valent elements, Bv- elements of the fifth group of ju 2 3650^-3150cm'1 and 8(H O) =1625 cm1. Periodic system) are rather convenient objects. 2 The study of these compounds by methods of a Methods roentgenography, infrared spectroscopy and thermogravimetry has shown that their thermal degra- The authors have synthesized themosthydrated dation proceeds in stages. Down to complete dehy- k v dration in structure of compounds the layered motif is compounds of the series A (B UO6)^(OH)mnHjO, (Ak-Y, Ln, Al; Bv- P, As, V). The common feature of kept. According to features of the structure, it is pos- k v their crystal structure are layers [B^OJ, ions of metal sible to divide all compounds in systems A (B UO6)k k m(OH)m-nH2Oj into four groups: the maximum hy- A and molecular H2O are located between the lay- ers. Processes of dehydration and intermediate hy- drated compounds (interlayer cations cooperate mainly drates are investigated by X-ray diffraction, infrared with molecular H2O), average hydrated crystals (the spectroscopy, and thermogravimetry methods. number of molecules of H2O and interlayer atoms of oxygen in an environment of an ion Ak is comparable), Results the lowest hydrated crystals and waterless compounds (the coordination sphere of an ion Ak mainly and com- The data of roentgenographic measurements, in- pletely consists of atoms of oxygen). frared spectra, thermograms and thermodynamic func- Atomic radii and electronic structures of interlayer tions (enthalpies, entropies and Gibbs's functions) of cations consistently affect the structural parameters of given compounds were obtained. The solubilities in the investigated compounds with the retention of lay- various conditions were calculated. ered motif. For example, decreasing of the interlayer distance because of lanthanoid contraction, tetrad ef- Discussion jfect, and splitting of molecules of water on cations with The uranylphosphates and uranylarsenates are high ionic potential. obtained by synthesis from solutions on the following k+ 2+ 3 Conclusions reaction A (s) + (k-m)UO2 (s) + (k-m)P(As)O4 (s) + (n+m)H2O(l) -> A'CBVUO^OHVnHjOCc) + Thus, some uranium mineral-like compounds were + mH (s). The synthesis of uranylvanadates synthesized and the features of their structure and ther- v A (VUO(i)1,-nH2P carried out by interaction of crystal mal stability were investigated. compound HVUO6-2H2O with 0.1M solution nitrate k of the appropriate metals A in the closed ampoule at The investigations are maintained by the Russian temperature 80°C. Synthesis of HVUO6-2H2O car- Foundation of Basic Research. ried out in hydrothermal conditions by the reaction of

EHViRgNMEHT & SAFETY (223 SK01K0139 SOLID SOLUTIONS OF SOME URANYLPHOSPHATES- URANYLARSENATES. STRUCTURE AND THERMODYNAMICS

N. G. Chemorukov, E. V Suleymanov, SA. Ermonin

Nizhny Novgorod State University, Nizhny Novgorod, Russia

Abstract asymmetric stretching mode and splitting of phosphate and arsenate asymmetric stretching mode are observed. Uranylphosphates and uranylarsenates of hydro- Enthalpies of mixing for solid solutions A'P, gen and alkali metals are minerals and are analogues. jjASjjUCynHjO are positive for all compositions. The In nature they can occur both in individual phases and lowest values for enthalpy of mixing is observed for in the form of solid solutions, and can be formed dur- the hydrogen family of solid solutions. For the lithium ing migration of uranium in nature. This paper is dis- family, enthalpies of mixing are a little higher. The high- cusses the synthesis, structure, and thermodynamics of est values (in the range of error) are observed for the solid solutions having a general formula: 1 sodium, potassium, rubidium, and cesium families. A'P,.xAsxUO6-nH2O (A - H, Li, Na, K, Rb, Cs). Partial excess Gibbs's functions of mixing of HP,, xAsxUO6-4Hjp solid solutions have been calculated Methods on data of the dependence of the distribution coeffi- cients of phosphate and arsenate components between Samples of HPlxAsxUO4-4HsO have been pre- solid and liquid phases. Full Gibbs's function of mix- pared by precipitation reactions of solutions of uranyl ing of HP1JCAsxUO6-4H2O have been calculated. It is nitrate and mixtures of phosphoric and arsenic acids. less than zero for all compositions. Substituted samples have been obtained by ion ex- change reactions. Powder X-ray diffraction data have been collected Discussion using Cu Kcc -radiation. Patterns have been indexed Data of X-ray diffraction method and IR- and the cell constants have been refined by using least- spectroscopy prove the formation of solid solutions of squares analysis. Diffraction patterns were indexed the series A'P^ASxUO^Hp. Splitting of the IR bands on the basis of tetragonal unit cells for all the salts. argue the presence of strained chemical bonds, whose Infrared spectra have been recorded on a SPECORD appearance can be explained by the partial substitution of 71 IR using KBr -pellets. phosphate groups to arsenate. The magnitude of this bond Enthalpies of the solution of samples in effort can be assigned with the value of enthalpy of mix- hydrofluoric acid have been measured on adiabatlc ing. This means that the least effort is observed in the Calvet calorimeters. Enthalpies of mixing have been hydrogen family of solid solutions. This fact can be ex- calculated as residual enthalpies of solution of physi- plained by the influence of interlayer water molecules to cal mixture of individual phases and of solid solution. oxygen atoms of phosphate groups. In the structure of Gibbs's functions of mixing of solid solutions lithium solid solutions, all water molecules are coordi- HPlxAsxUCy4H,O having distribution coefficients of nated to lithium atoms, this makes the effect of bond ef- phosphate and arsenate components between solid and forts increase. The greatest efforts are observed for so- liquid phases have been determined. dium, potassium, rubidium and cesium families, where alkali metals substitute for one of four water molecules. Results Excess Gibbs's functions of mixing of solid solu- tions HP,.xAsxUO6-4H2O are equal to enthalpies of Elemental analyses have indicated that investigated mixing in the range of error. It means that excess en- samples have compositions of X: 0 (pure phosphate), tropies of mixing are equal to zero, and solution can 0.19,0.44,0.72,1 (pure arsenate). It should be noted be considered as regular. The same conclusion can be that the X value does not change during the ion ex- made for all the rest families of solid solutions because change reaction. Hydration value n depends only on type of similar layer structure. A1, and is equal to 4 (A1 - H, Li), 3 (A1 - Na, K, Rb), and 2.5 (Cs). Conclusions The X-ray-diffraction data obtained for intermedi- ate solids indicate them to be in single phases. Corre- Solid solutions A'P^As^Oj-nHjO have been sponding lattice constants derived from patterns are in- synthesized, their structural properties have been in- creased with X in good agreement with Vegard"s law. vestigated. Thermodynamics functions of mixing have Spectra of intermediate solids differ from a physi- been determined, and the possibility of applying regu- cal mixture of individual phases. Shifting of uranyl lar model of the solution have been established. 2241 POSTER PRESENTATIONS SK01K0140 MODELLING THE IMPACT OF CERNAVODA NPP ON THE ENVIRONMENT

Angela -Ioana Constantin Cernavoda NPP - CNE - PROD; Romania Abstract Results In connection with a stay as a visiting researcher The results of modeling the dispersion and radiation at VTT Energy (Technical Research Centre of Finland), exposure caused by releases from the Cernavoda nuclear a study was conducted related to the dispersion and power plant were obtained for two different cases: for dose assessment for releases from nuclear power plants. normal function and for major accidents. These results The most important aspects of this study are described show that the dose rate grows as height is decreased be- in this paper. These include: the development of a spe- cause ata lower height there is less air giving attenuation cial program INTEGRATION for the calculation of between the layer and the target, which is 1.0 m above cloud gamma dose rate of 72 nuclides and the applica- ground surface (a standing person). tion of the TRADOS dispersion and dose assessment The results from the program INTEGRATION rep- model in the case of the Cernavoda NPP. resent dose rates in air,, whereas the results of Kocher refer to dose rates applicable to the total human body Methods and thus include attenuation within the human tissue. The program is written in FORTRAN 77 language The TRADOS (TRAjectoryDOsemodel), system has and runs on a UNDC operating system, on HP workstation. been developed in co-operation between VTT and the Finn- With the program INTEGRATION several dia- ish Meteorological Institute (FMI). TRADOS determines grams were produced for different radionuclides (for the air parcel trajectories and the dispersion conditions along example -1-131) and figures covering simultaneously them. Air transport of successive periods of the release is all the 72 nuclides covered in the Kocher publication. described by 3D trajectories that are usually based on data from the Nordic HIRLAM weatherprediction model. Dis- Discussion persion of the cloud around its centerpoint is computed separately. In the vertical direction, the K theory of turbu- In this study the potential radiological effects in Ro- lent diffusion is used, and the cloud is divided into three mania due to some hypothetical accidental airborne releases layers according to the possible values of the mixing height. at the Cemavoda nuclear power plant site were analyzed. Besides doses, the output quantities of TRADOS are con- The trajectory-dose model TRADOS was employed. Tra- centrations at breathing height, deposited amounts and ex- jectories are started at 3-hour time intervals and the conse- ternal dose rates from fallout and the cloud itself. quence code utilizes the same time increment. This ena- The program called INTEGRATION is intended for bles the consideration of long duration releases. Further- the calculation of external cloud gamma dose rates for more, it is possible to consider wind field variations as well different nuclides, atmospheric layers and 'rings' with con- as changing dispersion conditions. Dose conversion fac- stant distance to the target. The code theoretically pro- tors are based on data published by Kocher. vides the basis for using an infinite number of possible vertical concentration profiles, because the dose rate from Conclusions every possible profile can be decomposed into layers. To calculate dose, one starts with the so-called en- The programs called INTEGRATION and srgy fluence rate, which is defined as *F = energy flue- TRADOS application for the Cernavoda NPP demon- ice rate (J/(m2 s)= the amount of energy per unit time strate the permanent preoccupation for radiation pro- that has entered an infinitesimally small sphere-shaped tection. This preoccupation has been also proved at volume around a point divided by the area of a circle with Cernavoda NPP by the lowest individual radiation the same radius as the sphere. The program INTEGRA- doses registered in 1998 for the atomic radiation work TION is further elaborated for consideration of a fi- (ARW) (0.77 mSv/year, in comparison with the legal nite cloud. The finite cloud model involves simulating limit of 18 mSv/year) and by the critical population the plume by a series of small volume sources and in- group dose collected from the nuclear power plant ra- tegrating over these sources. There are two stages in dioactive emissions (5.58 uSv/year in comparison with' the calculation, the evaluation of the photon flux at the 2200 nSv/year, the environmental dose). point of interest and the conversion of the photon flux These positive results represent a convincing ar- :o absorbed dose in air. In general a number of pho- gument for the continuation and end of the work at the :ons of differing energy and intensity are associated CernavodaNPP second unit which will bring great ben- with the decay of a particular radionuclide. The proce- efits for my country's power: the reduction of the or- dure for estimating the dose for photons of a discrete ganic fuel import; the essential reduction of atmos- decay energy is described; the evaluation of the dose pheric pollution of the thermo-electric power stations' from the decay of any radionuclide is obtained by sum- on the inferior coal, in present in majority; the cost mation over the photon decay energy spectrum. In pro- reduction of electric energy, and the possibility of a gram INTEGRATION, this is accomplished by the \arge number of -work places and energy export in the function all_photons. neighboring countries. SK01K0141 ALARA AND RADIATION PROTECTION OPTIMIZATION STATUS OF CERNAVODA NPP

Dragos lone! HA U

Cernavoda NPP, Romania

ALARA in Radiation Protection is specifically • Radioactive work planning applied as a tool in loweringB the risk of high expo- Higher approvallevelsforhigher hazards, HealthPhys- sure when performing necessary activities by reducing ics Department work planning assistance, Use of ALARA the doses as much and as reasonably possible. tools, Individual dose restraints (Dose Checkpoint); Is ALARA also a matter of IMAGE ? ALARA is • Reduction of exposure time one of the main aspects that contributes to forming the Detailed and comprehensive work procedures, image of a certain facility. Many experiences in the world Use of the least number of experienced workers, Use show a decrease in the collective dose per station after of specially designed tools, Prefabrication of materi- proper ALARA implementation. This trend is more als in workshops, Removing components from systems, relevant for "adult" nuclear sites. At Cernavoda NPP Ensurance of preliminary work completion, Use of the authorization process itself included implementation Personal Alarm Dosimeter to control stay time; of an effective ALARA program. • Effective use of shielding Employing portable shielding, Shielding cabinets L Objectives of Health Physics Program for works at reactor face, Shielding by water filling of • Orgaruzationandmanagementforradiationcontrol appropriate vessels/tanks; • Training and qualification for radiation control • Distance from radioactive sources • Reduction of occupational exposure Use of remote tools, Use of long handled tools (Ex. radiation detectors), Employing robotics; Source control, Protective barriers and • Decay of radioactive sources equipment, Administrative control, Surveillance, Delaying entrance in contaminated areas; Personnel Dosimetry Program • Use of protection equipment • Contamination control Respiratory Protection, Training in correct selec- • Radioactive solid waste management tion and use of equipment; • Radiation work control • Decontamination of areas and equipment • Improvements of radiation control tasks Removing radioactive materials from equipment • Public Radiation Control Effluent Monitoring Program, Environmental proir to maintenance, Large area decontamination, Monitoring Program Minimizing the number of contaminated areas; 2, Focus items of Radiation Protection Optimi- • Ventilation systems and equipment zation Program High capacity ventilation systems, Ventilationflow • Review of operating experience of other NPPs toward areas with high contamination potential, Pass- (especially CANDU 6 Plants) ing of ventilation air through vapour driers; • Identifying design changes • ALARA goals and targets • Identifying changes to operational procedures Staff acknowledgement, Establishing realistic and • Revision and implementation of ALARA Program achievable Annual Dose Targets, Defining outage and 3. Current ALARA elements of Radiation specific job Dose Targets, Managerial commitment; Protection Optimization Proeram • Other elements Program documentation Investigation of Radiation Incidents, Providing a high Guideline documents for Radiation Safety Program; level ofRP services and expertise, Accessing the latest • Protection by facility design information by memerbership in international organizations; Employment of highly qualified radiation safety 4. Future Plans for reflnine the Health Physics design specialists,Access Control Systems, Fixed Area Proeram Monitoring, Dedicated leak collecting systems; • Improvement of work management and work • Elimination of radioactive sources dose management Suitable construction materials, Early detection Improvement of job dose monitoring systems, of failed fuel (GFP System), Tritiated water vapour Instrumentation optimization, Improvement of recovery system (dryers), Reactor shutdown/power Radiation Protection Training system; reduction prior to accessing restricted areas, Close • posimetry Program optimization reactor water systems chemistry control; Dosimetry Instrumentation optimization, Dose • Training of personnel Records System optimization; Extensive radiation protection training, Special- • Refining and improvement of ALARA elements ized radiation protection training for radiation pro- ALARA Database development in connection with tection assistants, Planned training refreshments/ Dose Records System; requalifications, Work planning training, Use of mock- ups and practice areas; ZZt | PiSIER PKESENTJITI1HS SK01K0142

INCREASE NUCLEAR SAFETY OF WWER-440

Teodor Nochev, Sabin Sabinov

NPP, Kozlodui, Bulgaria

ac horizontal steam generations, primary loops isolation _ ' . nmrnn AAn <* *i valves and two turbine-generators. Twenty-six WWER-440 units are currently e> working in Eastern Europe. In NPP Kozloduy, there A complete program for increasing nuclear safety are four operating units of this type. WWER —440 has been made at NPP Kozloduy with the participation has some positive qualities from a safety perspective, of German, French, Russian and American specialists. These include low thermal density, low specific thermal This effort cost greater than 100 mil $. This report loading and a large coolant inventory. Some of the includes the methods of increasing nuclear safety. The design solutions benefit the safety characteristics, such style of management in NPP Kozloduy has been as multi-loop configuration of the primary circuit, changed for the last seven years.

EHVIBBHUENT & SAFETY 1227 SK01K0143 NEW INTERMEDIATE HEAT EXCHANGER FOR LOOP TYPE LMFBR

K. Miyazaki, N. Uda, J.Toyooka andH, Horiike

Osaka University, Graduate School of Engineering 2-1 Yamadaoka Suita, 5650871 Osaka, Japan Abstract Thermal inter- action experiment Secondary sodium loop elimination is proposed was conducted for Pb for the loop type LMFBR with using Advanced Inter- and Pb-Bl, which are mediate Heat Exchanger (AIHX) for reduction in size candidates of an in- and cost[l]. This heat exchanger contains primary termediate medium sodium tubes and tertiary water tubes in a tank filled of AIHX. In case of with intermediate heat transfer media. Figure 1 shows tertiary tube failure, a comparison thermal interaction with a may be occurred be- conventional tween water and hot loop type sys- intermediate. The tem. A concept experiment was PutteqWd verifying experi- conducted in two ment has been cases, hot metal drop performed with into water and water Figure 2 Experimental setup using Ga as the drop into hot metal pool with a temperature range oi intermediate 200-600C. In both cases, Pb-Bi has a strong interac- medium in natu- tion, like the Sn case in atmospheric pressure. On tto ral convective region to low ve- other hand, Pb showed a milder interaction. No inten> locity forced cir- sive interaction was observed under the pressurizec condition up to several bars. Figurel Comaprison with culation. From conventional loop system the experimental correlation, AIHX - steam generator was conceptually designed. In order to use Pb or Pb-Bi for the Intermediate me- dium, the thermal interaction of Pb and Pb-Bi with water was studied experimentally. Interaction was sup- pressed under the pressurized condition of several bars, and the possibility of vapor explosion will be ruled out. Experimental result and discussion

A basic AIHX experiment has been conducted for tube bundle of primary straight - tertiary helical configuration. Experimental setup is shown in Fig.2. Ga was circulated with electro-magnetic pump. The experiment covered a range of U=0.5-18 cm/sec at Figure 3: Size comparison of typical AIHX an< heater surface, q=20-240kW/m2 at the inlet of cooling Monju water temperature of 40C. Overall heat transport Jl] K. Miyazaki et al., Advanced IHX-S< coefficient \ was estimated by signals from 0.5mm Combined FBR System Design and Basic Exp., Pro< thermocouples, and hm=q (In AT^-lnAT^AT^-AT^). 10-PBNC, Kobe, 1996 The data were arranged into non-dimensional correlation form of Nusselt number as a function of Peclet number, i.e. Nu=4.38+0.27Pe067. Figure3 shows a typical size comparison of AIHX and Monju.

221 i PBSTEt PBESEITATieiiS SK01K0144 THE TEQHA RIVER: 50 YEARS OF RADIATION PROBLEMS

Dmitriy Evlanov

Ozyorsk Technological Institute of Moscow Physical Engineering Institute, Russia

In 1948, the first industrial complex that obtained y-background near the river and internal radiation from Plutonium was in the Chelyabinsk region. Later PA radioactive isotopes that enter organisms through wa- "Mayak" was formed on the basis of this complex. At ter and food. External radiation was determined to be first, the plant's work led to extensive radioactive pol- from Cs-137, Ru-106, and. Zr-95. The largest doses lution of the Ural region. There were several steps to (50 to 100 cSv/yr) were received by the people of the production of plutonium. Some formed great vol- Metlino village. Some people were taken ill because umes of radioactive waste of medium- and low-level of external radiation. Internal irradiation of elderly activity. Our scientists decided to dispose of radioac- people amounted to about 4.6 mCi from the river. Sr-90 tive waste in the Techa River. and Cs-137 were the major radioactive.isotopes. ' The Techa River rises from the Irtyash Lake and In the autumn of 1951, an effort to stop thesources empties into the' Iset River that joins the Tobol River. of pollution from PA "Mayak" to the Techa River The total length of this river system is approximately started. In 1952, the radiochemical plant did not dis- 1000 km. The length of the Techa River is 240 km. pose technological waste in the Techa River. Activity Before the disposal of radioactive pollution, there were of liquid waste contained not more than 20-30 Ci/day. 38 villages with a total population about 28,000 on the In 1956, radioactive waste disposal in the river riversides of the Techa. stopped. In 1956 and In 1963, dams were built on the In 1949, the first radioactive waste was disposed Techa River. These dams insulated technical objects in the Techa. About 80 percent of the water volume of the plant from contaminating the bottomland at the had low-level activity from the Kyzyltyach Lake to source of the river. Cascades of ponds were built to Techa and 20 percent was of medium-level activity prevent leakage of radianuclides. In addition, about from the plant. From 1949 to 1956, 76 million cubic 8000 hectares of bottomlands that were polluted be- meters of liquid waste with a total activity of about cause of floods were drawn from land settlement. 2.075 million curies was dumped in the Techa River. These steps led to an improved radiation situation At first, the disposed liquid contained about 10'MO"4 Ci/ in die region. But the idea to evacuate people from I. After an accident, the liquid contained lO^-lO4 CM. some villages was accepted. From 1955 to 1960, about From March 1950 to November 1951,95 percent 7500 people from 19 villages were evacuated. But of all activity was dumped in the Techa. During this evacuation of people was untimely and this step was period, 4300 Ci/1 was dumped daily. The water con- not effective because the people had already received tained: Sr-89, Sr-90 - 20.4%; Cs-137 -12.2%; Zr-95, a large dose of external and internal radiation. Some Nb-95 - 13.6%; Ru-103, 104, 105, 106 - 25.9%. of Metlino's people received the biological equivalent During the next five years, waste in the Techa River of about 200 roentgens per year. Cluster sampling was greatly decreased. There were 9500 Ci dumped in showed about 935 cases of radiation sickness. 1952 and 500-2000 Ci in 1953-1956. Creating Techa's cascade ponds posed several The radioactive waste that was dumped in the problems such as rehabilitation (cleaning water, reduc- Techa was diluted by 5-10 times by the water of tion and utilization of radioisotopes) and stabilization KyzyltyashLake. During the flowing ofthis dirty water of closed ponds. about 25 percent of all the activity was embedded in lake bottom sediments. The activity that was accumu- The PA "Mayak" solves the complex problems of lated in animals living in and around the river exceeded the industrial ponds on the Techa River using the fol- the limited concentration by 75 to 100 times. The spe- lowing: cific activity in the mud was about 20 mCi/kg. Hydrological and hydro-geological studies of the Techa cascade ponds, and In 1955, radioactive water polluted the bottomland of the Techa because of a flood. The inhabitants were pevelopment of methods and technologies for using this place as grassland. cleaning the bottom of the ponds. For inhabitants of Techa's riverside, the river was The final solution to Techa's problems is possible the basic source of water for food and irrigation. The only with the collaboration of the South Ural Nuclear riverside population incurred external radiation from Power Plant.

EKVIIIKM£IIT&SftFm!22! SK01K0145 DYNAMIC OF AGE STRUCTURE AND THE NUMBER OF POPULATION IN OZYORSK AND AFFECTING FACTORS

Olga Panchenko and Marina Rtischeva

Ozyorsk Technological Institute of Moscow Physical-Engineering Institute (Technical University) Ozyorsk, Chelyabinsk region, Russia

The analysis of demographic processes in cities was included in the aim of the study. In considerin; having atomic industry is of great interest. The aim of this question, a natural growth was analyzed; it wa this work was an evaluation of the dynamics of age defined as a difference between birth rate and deatl structure and population for the city of Ozyorsk, based rate factors and migration growth, showing the growtl in connection with creation of the nuclear plant of population per 1000 people due to migration. "Mayak," the "first-born" of the Russian atomic It should be noted that by January 1998, for Xh industry. first time in the history of town, the number of th> As it was shown in that the highest percent of population reduced by 150 people in comparison witl children (32%) was in 1959, but in following years it 1997. These data point to negative dynamics. Sue] began gradually to fall and in 1989 is was 23%. The processes such as birth rate, death rate, and migratioi growth of IAS indicates the increasing process of affect the changing of the number of population demographic aging of population. The load of children Analyses have shown that since 1950, as a result of tht on parents decreased with each year, but load of birth rate factor, there had been falling natural decreasi grandparents increased. The later in conditions of among the number of young families in the course o reducing of the load of children presents an time from 1965 to 1989. It stabilized on average to J inauspicious tendency. The results of the determination level of 16.4 per 1000 people, but since 1990 it ha: factor of aging have shown that in a period from 1959 fallen to values of 8.5 per 1000 people. In its turn to 1979 their values did not exceed 8%, that since 1950 the death rate factor had been gradually corresponds to a demographic youth, from 1989 the increasing, but the rate of its increase stayed at i values of these factors have exceeded 12%, that points constant level until 1989, then it sharply increased uf to demographic aging of children in a total population to 12 events per 1000 people. In this connection, is less than 25%, demographic aging is more natural growth increased and took a negative value ol denominated, i.e. of second degree. Rates of aging of -3.5. During the whole period considered in this study, population from 1989 exceed the rates of rejuvenation, a factor of arrival constantly exceeded a departure as a result, the IDA that characterizes their correlation factor. However, the values of both factors decreased exceeded a critical value (>1.0). Subscripted factorial that reflected a reduction of the intensity of migration analysis has shown that increasing the aging factor is processes. As a result, by 1997 the difference between affected by not only natural processes of aging, as well these factors became small (0.3 per 1000 people). as by reduction of part of the young, which depends on birthrate. The obtained results indicate that since 1950 demographic processes in Ozyorsk were more Thus, as a result of studying the age structure of favorable, in spite of fact that it was in this period the Ozyorsk population, it was established that from workers PA «Mayak» and population as a whole, got 1959 it carried a stationary nature, but from 1989 it comparatively greater radiation doses than in the was characterized by a regressive type of development, following years. However, dynamics the number of in which rates are enlarged at later years. The observed population has an unfavorable trend to reduce, picture is characterized by the approach of the second connected with sharp worsening of social-economic degree of demographic aging, caused by both the situation in the town as a whole, as a result of the raising of the rates of aging, and reducing the rates of economic reforms in the country. Reduction of the rejuvenation of the population at later years. This number of population in the town is expressed by the indicates the inauspicious demographic changes, negative natural growth and by reducing migration defining the age structure of the Ozyorsk population. processes, which resulted in sharp decrease of the Besides, a correlation of the nature of demographic general growth of population, and in its stopping in processes with social-economic and ecological 1998. conditions of lifestyles of the population in this town

2311PISTER PRESENTATIONS SK01K0146

ANALYSIS OF MODES OF HEAT TRANSFER BY FUEL DAMAGED BY BREAKING A FUNCTIONING SIDEBAR OF A WATER-COOLED REACTOR

A.V Golubinsky

Russian Federal Nuclear Center - VNIIEF, Russia

Damage connected with breaking a functioning In the study, a new way to measure the length of sidebar of a water-cooled reactor is accompanied by film boiling on surfaces of heating particles with non-stationary heat and the hydrodynamic processes. reference to the fault condition (interaction of fuel with In particular: a quick increase in power; a decline or coolant) in a NPP was considered. Presented is a cessation of coolant; a drop in pressure as a description of the experimental installation, strategies, consequence of an unhermetic sidebar; and the onset and results of length measurements of film boiling of of reactor emergency cooling systems. water on spheres of different diameters from honeys and aluminum under different temperatures of Emergencies can bring about melting in a interaction. These results can be used to test models reactor core (RC) and melted fragments of fuel in the and codes for the evaluation of heating efficiency during water coolant. The interaction of fuel and water creates boiling under corresponding conditions. vaporization. Under certain conditions, vaporization The primary use of the present work is to verify can result in a steam explosion as a result of energy and calibrate heat transfer correlations using the results transfer from fuel fragments in the mode of detonation of experiments. Using correlations to cot uluct analyses waves. Such damage was observed in the reactor SL- of heat transfer during the boiling on spheres from 1. In order to evaluate the probability of development honeys and aluminum and with provision for results of and the scale of the steam explosion in a NPP, it is the experience, boiling curves are produced describing necessary to determine the transient features of the heat dependency of flow of heat (Q) on fuel temperature transfer process in the fuel up to the period of boiling {To), size (R), ball material, and temperature of when it is possible to have a steam explosion. surrounding water (TL).

ENVIBINMEKT & SAFETY 1231 SK01K0147

DYNAMIC ANALYSIS AND SAFETY MEASURES OF WER-440 NPP PAKS 1 TO 4 DUE TO SEISMIC AND OPERATIONAL LOADING CONDITIONS

T. Burjan PAKS Nuclear Power Plant LTD, PAKS, Hungary

B. Schwemin, U. Erben, A. Halbritter, W. Trubnikov Siemens AG, Power Generation (KWU), Offenbach, FRG

Abstract and the final dynamic response and bearing capacity calculations performed. In the course of the gradual safety upgrading of the VVER-440 NPP PAKS, the loading capacity of The excitation was defined by a site-specific design the civil structures due to earthquake lo- spectrum, with a maximum acceleration of 0.25g for ading superposed by operational conditions was the horizontal and 0.23g for the vertical direction. analyzed. Furthermore, a wide range of soil parameters and the liquefaction effect was considered. Based on a three-dimensional detailed finite- element model, several dynamic analyses were The maximum forces and stresses were calculated performed in the frequency domain for the main for the load cases, dead weight, snow loads (NOL), building complex of each of the four nuclear units. and seismic excitation. In accordance to the specified criteria made for NPP PAKS, the extraordinary loading At first, the bearing capacity of the as-built of earthquake was super-positioned with the other load structure was evaluated, followed by additional cases in a realistic manner. calculations considering required and optimized upgrading measures. An additional requirement was The stress results were evaluated by the acceptance that implementation of the upgrading steel elements criteria defined for the seismic re-qualification of the had to be executed during operation. PAKS NPP and AISC. The effectiveness of different upgrading options As a result of the analysis and the provided for the structures and systems, as well as the upgrading measures, the civil structure is now capable implementation feasibility of the upgrading measures to withstand a horizontal and vertical seismic loading and seismic safety, were investigated. For the essential equivalent to the above excitation level. retrofit concept, the mathematical model was updated

232 [ P3STEB PRESENTATIONS SK01K0148 CONFINEMENT STRENGTH AND INTEGRITY, NUCLEAR POWER PLANT V-1 BOHUNICE SLOVAKIA Magnar Berge Siemens AG / Power Generation Group (KWU), Offenbach, Germany Magnar.Berge@off 1 .siemens.de

JozefJanik EBO, Bohunice, Slovakia

In the course of the gradual safety upgrading of Beyond Design Basis Accident (BDBA) is the Bohunice V-1 Plant, the confinement strength and initiated by a double-ended rupture of a main integrity at accidental conditions has to be circulation piping with the diameter of 500 mm demonstrated. Concerning the material properties the long-term The purpose of the present paper is to show the phenomena creep and shrinkage as well as the influence specification of design loads and load combinations of concrete cracking are considered. which are considered in the evaluation of the Proceeding from the original design format and confinement structure, and to explain the applied following the development of the European standards, acceptance criteria and used evaluation methods. the concept for the re-evaluation of the confinement- Besides dead loads, live loads, and thermal loads compartments is based on design in limit states. The the structure is subjected to postulated accidents which partial safety factors for design loads and design are defined by the regulatory body tJD SR as follows: resistance are adopted in compliance with the recommendations of IAEA. Design Basis Accident (DBA), which is initiated by one of the following events: Regarding the general shape complexity of the confinement a three-dimensional finite element model - a double-ended rupture of a pressurizer surge line of the whole concrete structure is developed to perform with the diameter of 200 mm the design checks. - a partial rupture of a primary piping with the equivalent diameter of 200 mm The applied partial safety factors comply with international requirements and result in an adequate - a double-ended rupture of a main steam line and/ safety level. The required total safety margin is about or feedwater line with the discharge into the SG 1,5 for load case DBA. In load case BDBA the strength compartment capacity can reach it's theoretical limit.

EiiVII!ONMENT& SAFETY 1233 SK01K0149

PROVISIONS OF COMUNICATION BETWEEN NPP OPERATIONAL PERSONNEL: DRIFTS OF DEVELOPMENT

Sergey A. Piskarev and Valeriy R. Aksenov

Branch of St. Petersburg State Technical University, Sosnovy Bor, Russia

With the increase of NPP unit capacity there is a • Hierarchical performance of the information about substantial growth of information and, as a NPP technological process and safety. consequence, information loading of operators in • Provision of fast access to a required format or emergency operation and the necessity of perfection automatic performance of a required format in of their activity has increased. The decision of this emergencies. task in many respects depends on accounting for human factor at the creation of die control room system (CRS). • Hierarchical organization of the signal system; a The accident on Three Mile Island (TMI-2) NPP uniform format of display organization. revealed an essential lack of communication between • Information support of the operative personnel at the operator and the control system. Methods of normal operation and in accident conditions. granting of the information did not give the operator The paramount role in information provision an exact picture of general condition of the NPP during systems belongs to screens of multiple access, the the accident. This accident has laid the foundation for personal displays of the operators are a level below. intensive research in different countries on the creation Form and volume of given information is also of automatic systems, information maintenance, and important. It should be clear to the operator, and its informational support of operators. volume should not interfere with concentration of The central problem regarding the perfection of operator attention on the most important messages. operative personnel activity is the reduction of information provided to an operator, granting to him Therefore one method to perfect the operator- only that information which is necessary for safety system interaction is the conventional reduction of monitoring, and control and management of information volume on display screens in abnormal technological process. The solution to this problem is situations, so that the operator has an opportunity to possible through the introduction of a third generation concentrate their attention on the most important safety of CRSs, which are characterized by a high degree of messages. The following algorithms can be applied to computerization utilizing both integration of software the reduction of the display messages volume: and management. algorithm using a time delay; algorithm of association of the equivalent messages; number reduction Approaches to the creation of CRSs in foreign algorithms of interdependent event messages. countries are based on support of operator actions. The basis is the provision of compressed information about It is possible to allocate three basic tasks of support safety and technological processes with performance systems: a rating of the current status of technological on screens of monitors and by paying special attention process and trends of its development; recognition of to human factors. In contrast to the traditional, the abnormal and emergency conditions and definition of advantage of such boards consists in granting to the its reasons; a choice of necessary actions on elimination operator a flexible format with the integrated or mitigation of consequences of this situation and their information, with graphic, generalized, etc., culmination. The following operator support systems representation of the data. now take place: systems of granting of safety When organizing information concerning the parameters; automation systems of emergency operative personnel, the following principles are fixed: instructions provision; and expert systems.

2341 POSIES PKSEffTATIIIIS SK01K0150 A SIMPLE EVALUATION OF CONTAINMENT INTEGRITY AGAINST EX-VESSEL STEAM EXPLOSION ON ADVANCED PWR

Kenji MVRAYAMA

The Kansai Electric Power Co., Inc., Japan 1. Abstract The energy which would act on the reactor cavity wall is assumed to be equivalent to 80 percent of the The guideline for consideration of severe accidents mechanical energy which would be generated by the on containment design for the next-generation LWR was steam explosion. This rate represents an area of the published in 1999 in Japan. The guideline specifies the side wall surface as occupied in the explosion region. probabilistic performance targets for the frequencyo f events Moreover, four different levels of mechanical energy in which the capability of containment to retain fission conversion ratio - 0.01%, 0.1%, 1.0% and 5.0% - have productscannotsatisfythe requirements. Also the guideline been applied in the calculation. This is because, although specifies that assessment should be made from the no results of experiments on a full-scale facility are deterministic point of view in order to ensure that the available, general observations are that the mechanical containment has a proper safety margin against possible energy conversion ratio tends to decrease as the amount loads that may be produced during a severe accident. of melted products increases and it is considered to stay Pressurized water reactors (PWRs) in operation at 1% at a maximum. In addition, ALPHA experiments in Japan adopt the wet cavity formula as a severe conducted by the Japan Atomic Energy Research Institute accident management, which focuses on the control of (JAERI) show that the conversion ratio ranges from 0.6% molten core concrete interaction (MCCI) during a to 5.7%, though the results of these experiments are based severe accident. Therefore, there may be a possibility on about 20 kg of melted Alumina. bf an ex-vessel steam explosion in case of the reactor The strength of the reactor cavity wall is evaluated by vessel failure. Seeing that the wet cavity formula will the strain energy which can be absorbed by reinforcing be adopted in the next-generation of PWRs, we have elements. In this study, the total strain energy which decided to develop a method of assessing the containment accumulate by the time one of the reinforcing elements first integrity against ex-vessel steam explosion in order to reaches its failure strain in a simulation analysis using finite check the validity of future containment designs for the element method is regarded as the strength of the wall. guideline. First, we conducted a simple evaluation on an The analytical region is the range filled with the APWR, the latest designed PWR in Japan, though it is water injected in the reactor cavity wall (the primary not categorized as a next-generation LWR. shielding wall). The explosion energy was assumed to be a function of the mass of the dropping melted core and the 3. Results conversion ratio. The strength of the reactor cavity wall was assumed When assuming the mechanical energy conversion to be equivalent to the total strain energy which would ratio of 1 %, it is calculated that the energy which would accumulate by the time one reinforcing bar element be generated by ex-vessel steam explosion would amount would first reach the failure strain in FEM analyses. }o about 1 MJ if the mass of the dropping melted core As a result of this study, the explosion energy corresponds to the break of one instrumentation guide would be less than the wall strength; thus, the tube penetration, and about 40 MJ if the mass containment integrity would be maintained even if an corresponds to the simultaneous break of all penetrations. ex-vessel steam explosion were to occur. In the structural model used in the analysis of the strength of the reactor cavity wall, the upper part of the 2. Methods structure suffers larger deformations. When the strain in the uppermost node has reached 18%, which is the rate The scenario of the ex-vessel steam explosion we guaranteed forthe steel bar JIS G3312 SD390 forreinforced assume is that the reactor cavity wall would fail and concrete, the internal pressure is found to be 6.3 MPa. The would become unable to support the reactor vessel and total strain energy accumulated in the reinforcing elements then pipes connected with hot and cold legs would be by this point of time is evaluated about 72 MJ, the energy pulled strongly, which would result in a break of the which can be absorbed by the reactor cavity wall. In this pipe penetrations through the containment. evaluation, the effect of concrete is ignored because many In this study, while referring to the evaluation cracks are presumed to have developed in concrete. methods used by foreign researchers, we have evaluated ihe energy which would be generated by steam explosion 4. Conclusion and the strength of the reactor cavity wall separately. The thermal energy of the melted core which would It is expected that the energy which would be be dropping in the water is converted into mechanical generated by ex-vessel steam explosion would be less energy at the mechanical energy conversion ratio. By than the strength of the cavity wall; thus, the calculating this mechanical energy, the energy which containment integrity of APWR would be maintained would be generated by the steam explosion is evaluated. even if an ex-vessel steam explosion were to occur. ENVIBBHMEHT & SAFHY j 235 SK01K0151 AN ANALYSIS OF THE INTENT OF ENVIRONMENTAL STANDARDS IN THE UNITED STATES THAT APPLY TO WASTE DISPOSED AT THE NEVADA TEST SITE by Anthony E. Hechanova and Brett T. Mattingly

Harry Reid Center for Environmental Studies, University of Nevada, Las Vegas, USA This paper addresses the disposal of transuranic disposal site. EPA assumed that any radionuclide waste at the Nevada Test Site (NTS), the intention of migration away from an undisturbed site and into the the environmental standards under which the disposal is accessible environment would occur via a ground water completed, and some lingering controversy surrounding pathway, which would then cause a widespread the U.S. nuclear weapons complex remediation effort. population exposure. A goal of this paper besides the informational value is The containment requirements were derived to to provide points of discussion regarding this very costly assure that this initial release into the ground water is and large-scale program in the U.S. and provide a below a certain limit. Since the assumptions which platform for the exchange of ideas regarding remediation designate this pathway as the most limiting case for activities in other countries. study are not valid for the GCD boreholes at the NTS, For the transuranic waste (TRU) in the Greater a different population exposure pathway should be Confinement Disposal (GCD) facility at the NTS, analyzed. This pathway most likely will be tied directly application of current compliance requirements and to the deposition of drill cuttings on the surface as well regulatory guidance defined for a generic disposal as the migration of radionuclides in the vadose zone to system, although satisfying the "letter of the law," is the accessible environment. shown to be incompatible with the "intent of the law" Furthermore, because the exposure pathways are based on a thorough review of the preamble and altogether different, the release limits from the. background documents supporting the regulation. regulation are not directly applicable. Separate release The standards that apply to transuranic waste limits should be calculated based on a new population disposal were derived assuming deep geologic disposal exposure scenario that is consistent with the intent of and much larger and more hazardous waste forms: EPA in protecting public health and the environment irradiated nuclear reactor fuel and high-level and these metrics should be scrutinized against the DOE radioactive waste. Therefore, key assumptions that and the NRC standards for radiation protection. underpin the analyses used to justify the standards (e.g., The authors believe the first step to determine the the ground water pathway being considered the only level of required protection would be a hazard assessment major release mechanism) are inconsistent with the focused on the GCD TRU facility that will investigate nature Of the radionuclide inventory and the intermediate whether or not the excess cancer burden (or other limiting depth of waste emplacement in the GCD boreholes at criteria) to a future population from releases (e.g., from the NTS. The authors recommend that site specific drill cuttings) at the GCD is below what the EPA considers performance metrics be determined to foster an analysis an acceptable risk. Simply scaling down release limits is which is transparent and consistent with U.S. not the best scientific method to show compliance and Environmental Protection Agency (EPA) intent in instill public confidence that the site is safe. developing the standards for a generic disposal system. The goal of such work would be two-fold. First, By adhering to the release limits listed in EPA review of site-specific risk characteristics will identify regulations, the U.S. Department of Energy (DOE) is site-specific data gaps and uncertainties that may abiding by the "letter of the law." However, in this case, significantly affect the ability to assure that the simply adhering to the rule does not constitute significant performance of the GCD disposal facility is in scientific finding mat the intent of the regulations is being compliance with the intent of the EPA rules. Second, fulfilled. Instead, a more appropriate technical analysis a hazard assessment focused on the actual facility is a should be performed which accounts for the specific much more scientifically defensible, and credible, characteristics of the facility. Then a more transparent and means of assessing an appropriate protection standard conclusive scientific argument could be made that the. than simply showing that the site calculations fall below facility does meet the intent of EPA in promulgating their a certain release limit, which is not intuitive. Such an regulation - specifically, that the GCD facility adequately assessment would allow transparency and flexibility protects the public and the environment from exposure to so that a variety of stakeholder concerns could be harmful quantities of the disposed radionuclides. addressed through model and scenario adaptation. The The intent of the containment requirements is to authors believe this to be the most robust approach to protect a future population from a low but widespread setting site standards in light of the special cases dose from radionuclides which are released from the encountered at the NTS. . ?3B | POSTER PBESEItTATJBHS SK01K0152 INVESTIGATION OF ELECTRO-KINETIC METHODS FOR SOIL DECONTAMINATION

A.N. Shabanova

Urals State Technological University-UPI, Ekaterinburg, Russia The choices of effective methods for ecological sys- periment during 800-1100 hours. To study the dynam- tem decontamination, their perfection and introduction ics of radionuclide accumulation in the adsorbing into practical use have been actual tasks for the Urals element membrane, the latter was taken off daily and region. The objective of this work has been to study the sent for spectrometry measurements. The intensity of potentials of electrical kinetics method of ISOTRON the decontamination process by the volt-ampere Corporation (US) for decontamination of the Urals' soils. characteristics and activity level of the membrane (Fig. 1). K=(Am/Ap)*100%, Object studied where Am - activity level of membrane; Beloyarsk NPP site was used as a test site, with Ap - activity of soil at the moment of loading two land plots selected that had different soil types. into the installation. The plots were contaminated over a decade ago, as a result of a leakage of liquid radioactive wastes. Results Experimental techniques Major contaminants of soil include: Cs-134, Cs-137, Co-60, Sr-90, Pu-238, Pu-239 (240). Comparative Preliminary radio-ecological investigations have evaluation of content of various forms of radionuclides been carried out at the plots selected: y-radiation dose in soil has shown Cs-137 to be present mostly in the rate and p-particle flux density; y-spectrometry assess- fixed form: Co-60 (in acid-soluble) and Sr-90 (in ex- ment of the composition and specific activities of changeable forms). All isotopes under study proved to contaminants at various depths, radiochemical assess- be fixedin the soil characterizedby a low content ofhumus ment of specific activity levels oP'Sr at various depths, and loamy sand soils. The radionuclides were ranged radiochemical assessment of plutonium isotope according to the decrease of effectiveness of loamy sand composition and specific activity, neutron activation decontamination as follows: analysis of soil micro-element composition and, physic- • Strontium (95±5)% chemical parameters of soils have been defined. • Plutonium (70±30)% Analysis has been given to the forms of y-,p~ and a-active radionuclides present at various depths in soil. • Cobalt (19±3)% In this respect, the following forms have been assessed: • Cesium (0.43±0.05)% water-soluble and fixed. The effectiveness of the electro- The decontamination effectiveness appears to be kinetic method has been evaluated in two types of directly related to the ratio of radionuclide forms laboratory installations: vertical and horizontal layout. present in soils. Radionuclides with a weaker fixation The vertical variant, the "NIKIET (Moscow) de- on soil particles are removed easier. Humus and loamy sign, was a cylindrical facility with graphite plate serv- soils manifest a worse decontamination effect. ing as a basement that functioned as an anode. A cathode polymer membrane, manufactured by Conclusions ISOTRON Corporation, was inserted into the top cover of the polyacryl vessel. For the vertical installation Results bbtained have shown the method proposed loading, a sample averaged for all soil horizons was to be usable for decontaminating some types of soils used taken from one of the plots, where the soil was from strontium and plutonium; it is low effective for uniform loamy sand without humus layers. The soil decontamination in the area of South - Urals sample mass was 2.153 kg. radioactive plume. Thus, a low effectiveness can be The horizontal installation of ISOTRON type de- expected in podzolic and leached chernozems sign is a parallelepiped of 36cm length, 19cm width, characterized by a high content of loamy sand and sandy and 14cm height. The cathode and anode polymer soils, as well as for sobby-podzolic ones. membranes manufactured by ISOTRON were placed at The method can be promising for decontamina- the end faces of the polyacryl vessel. The installation tion of soils and wastes from chemical contaminants, was loaded with two layers. The upper layer is an aver- such as Zn, Ni, Cu, Pb, Hg, and others. Important ad- age sample of the humus horizon, and the lower one is vantages of this method compared to others have been an average sample of loamy sand horizon. its simplicity, small amount of wastes, and feasibility Volt-ampere characteristics were recorded in the ex- of decontamination in areas difficult to access.

EHVIRflKMENT & SAFETY j 237 SK01K0153 STATIONARY EQUIPMENT FOR DETECTING RADIOACTIVE MATERIAL ON PASSING PEDESTRIANS DEVELOPED AND MADE IN RFYC - VNIIEF

D.S.Kapustin

CYRY RFYC-VNIIEF, Russian Federal Nuclear Center, Russia

The atomic industry has accumulated significant Post KPRM is a device that detects gamma and experience in the development and fabrication of neutron radiation of controlled material from external radiation monitors which, unlike radiation measurement radiation over background. Features of the post include and monitor check-out instruments, are not confined the following: to the facilities, but are threshold devices that • self-diagnostics of equipment; registering gamma and neutron radiation from NM at a rate of external background radiation plus an excess • background radiation calibrations; of threshold values which give a signal. • detection of wear out radioactive material; At VNIIEF, radiation monitors have been designed • continuous radiation monitoring (checking "pos- that measure for transported or pedestrian carried session" of radioactive object across space of post radioactive material and allow a quick determination and around it); of an excess gamma and/or neutron background on • checking an unauthorized access to equipment; natural or fixed levels. For this problem, neither the and, type of the material or its amount need to be determined, • an "Alert" indication under prescribed conditions. but other parameters are measured to detect highly sensitive br priority materials, defined by the threshold In the event of finding radioactive material, of detection, operational monitoring efficiency, anomalous changing of the background or breaking an simplicity of use and demonstrative imaging of results. operation of blocks is triggered and sound, light, and All these parameters were considered at VNIIEF information signal "ALERT" with an indication of the in the design of pedestrian radiation monitor KPRM- type of event that caused operating the alarm signalizing P1. Post KPRM-P1 is an automatic pedestrian radiation (for example: "wear out," "possession," "DAMAGE" monitor, installed on communicating and KPP (fault of detector), or "BACKGROUND") and the enterprises, and intended for checking for authorized detector(s) number registering the event. or unauthorized radioactive material possessed by Information about changing a mode or any change pedestrians crossing a controlled space. of a condition of the post is fixed and can be sent to the The post can be used in physical protection central computer by request or with specified systems, accounting and monitoring for radioactive periodicity. material on enterprises, used for specified material in If needed, the central computer can also change its technological cycle, and on other objects as well, parameters, defining sensitivity of post and possible where monitoring is required for any movement of false alarm level. radioactive material. The post can be operated in the local mode or as part of a complex safety system.

2381P8S7EU PffSEMTflTIIWS SK01K0154 THERMO-HYDRAULIC ANALYSES OF BOHUNICE NPP WER-440

TomdS Kliment,

VUJE Trnava Inc., Slovakia

This contribution presents results of thermo- > The radially averaged fuel pellet enthalpy does not hydraulic calculations, which were performed using the exceed 963 J/g at any axial location of a rod (Maxi- code RELAP5/Mod3.2.2. The calculations were mum calculated value: less than initial value of performed for Bohunice V-l NPP, which is operated 499.5 J/g). with the VVER-440/230 type. > There shall be no melting of the fuel pellets, even One of the aims of gradual upgrading, demanded locally (melting point 2639 °C for fresh fuel, 2605 by the Slovak Nuclear Regulatory Authority Decision °C for burned fuel) (Maximum calculated value: 1/94, is reconstruction of ECCS to cope with following less than initial value of 1958 °C). extended DBA: > Calculated changes in geometry are such that the core remains amenable to cooling • double ended rupture of one of the pressurizer (PRZ) surge line (2 x 200 mm) (Number of leaked fuel rods equals 0.423 % and rod flow blockade is less than middle empirical • rupture of main coolant pipe with equivalent di- value of 20 %) ameter 200 mm at the worst place > Effective dose at the boundary of the protective For evaluation of the effective changes, a 6-loop zone (3 km from NPP) does not exceed 50 mSv design model for Bohunice V-l NPP has been for whole body and 500 mSv for the thyroid (Maxi- developed for RELAP5/Mod3.2.2 arid validated using mum calculated value: 2.91 mSv for whole body, a large data base of operational measurements. The 6- 4.80 mSv for the thyroid). loop model describes in detail the primary and secondary circuits, as well as all safety-important LOCA 200 mm and PRZ surge line break analyses systems and most control systems. The nodding were also performed with conservative assumptions scheme reflects both the general needs of thermo- from the point of view of mass and energy release to hydraulic simulation (dimensions, elevations, volumes, the hermetic zone. The purpose of these calculations areas) and the extent of safety analysis (simulation of was to show whether the acceptance criteria related to large scope of scenarios). the hermetic zone are met. The RELAP5 calculation results (mass flow rate and enthalpy flow at the break, The basic analyses (double ended rupture of one ECCS flow rate) were used for hermetic zone load of the PRZ surge lines 2x200 mm and rupture of the calculations, which were calculated using the code main coolant pipe with equivalent diameter 200 mm) TRACO. The following results were obtained: were performed from the point of view of core cooling as well as mass and energy release into the hermetical Acceptance criteria Maximum calculated value zone. PRZ surge LOCA line 200 mm For the LOCA 200 mm analysis, the conservative assumptions from the point of view of core cooling Maximum overpresure were assumed. The initial and boundary conditions inHZ<60.0kPa 52.9 kPa 52.9 kPa were chosen to show whether all acceptance criteria Maximum temperature in HZ< 120.0 °C 111.1 °C 111.0 °C are met. In the brackets, results of analysis are shown. Maximum sub-pressure > The fuel rod cladding temperature does not exceed inHZ<15.0kPa 6.3 kPa 6.3 kPa 1200 °C (Maximum calculated value: 800 °C). > The total local oxidation of the cladding does not exceed 17 % of the initial thickness before oxida- Conclusions tion (Maximum calculated value: 0.5 %). DBA analyses indicate that none of the acceptance > The total amount of hydrogen generated from the criteria for these events are violated. No danger of loss chemical reaction of cladding with water or steam of the core coolability is predicted, the criteria for the does not exceed 1 % of the hypothetical amount that hermetic zone are met and dose equivalents are deeply would be generated if all of the cladding in the core under the limits. were to react (Maximum calculated value: 0.046 %).

EHVISaNMEHT & SAFETY 1239 SK01K0155

RESEARCH OF CHEMICAL STRUCTURE OF ATMOSPHERIC PRECIPITATION

Author: Korenyak DA. Instructor: Korenyak T.K.

Russia

In the paper " Research of chemical structure of The relationship SO^/Cl1" > 1 in the warm season atmospheric precipitation," features and conditions of testifies to the continental origin of the salt component formation of the chemical structure of atmospheric of atmospheric precipitation in this period. precipitation under the influence of natural and The change of concentration of hydrogen ions of anthropogenic factors are stated. In this paper the author precipitation is well explained by the joint presence of presents data on the chemical structure of precipitation calcium and magnesium in water solutions of sulfete-ions. in the region for the first time. So anthropogenic changes in structure of precipitation During preparations for research, we studied the had a temporary local nature since they were observed processes of forming the structure of atmospheric during the short period, for the reason of unfavorable precipitation water and we defined the range of the weather conditions for dispersion of ejection. components that require investigation in the structure of atmospheric precipitation. • Mineralization of precipitation and the sum of concentrations of its main components are having During the period of observation, 35 samples were good correspondence. taken. The most mineralized precipitation was falling in the warm season. The main quantity of calcium and • While choosing the location of the town, the dis- magnesium falls with precipitation during the warm itance was taken into account, ensuring the disper- 2 sion of ejection from the plant. season. The relationship of concentrations SO4 7C1'~ < 1 in the cold season is explained by slackening of • The influence of ejection from Nizhny Tagil and eolian processes and also by introduction of chloride- Serov cities was not revealed on the structure of ions from condensation nuclei of marine origin. atmospheric precipitation.

24RI POSTER PBESENTATIBNS SK01K0156 INTRODUCTION OF REGULATORY AND LICENSING PROCEDURES OF SOME OECD COUNTRIES

Piter Zagyvai, Szabolcs Czifrus, 6rs Benedekfi, Piter Ormai, Gyula Dankd

Technical University of Budapest, Institute of Nuclear Techniques, Hungary

Abstract exemption and clearance, etc. Then we analyzed the regulatory and licensing practice of the following In the OECD countries more than 50 nuclear power countries: United States of America, France, United plants will have to be closed in the beginning of the Kingdom, Germany, Belgium, , Spain, Sweden, next century since their licenses expire. For this reason Canada, Japan. It turned out that the regulations are it is very important to establish reasonable regulations very different in the OECD countries from the detailed, in the field of decommissioning. In this poster, first we prescribing regulatory concept (like in the U.S) to the define the basic principles related to decommissioning. mostly sight related regulations (like in the U.K.) that Then we account on our survey of the situation of the provide more opportunities for the licensee. regulatory and licensing procedures in some OECD countries. Finally, we show some case studies for It is very interesting, that there is no intention to decommissioning. standardize neither the regulations nor the licensing procedures (not even in the EC countries). We looked Methods through the policies of the following international organizations: IAEA, OECD-NEA, EURATOM. In the We established contacts with professionals of first appendix of the report we publish an analysis about institutions of some OECD countries and read through the Greifswald, Rheinsberg and WAGR all the literature we could reach by INIS database, decommissioning projects* In this part we introduce Internet, etc. the strategy, the licensing, the waste management plan and the structure of the whole project. In the second Results appendix a bibliography can be found about decommissioning in which we classified the items to We compiled a report for the Hungarian Atomic the following categories: general, case studies, research Energy Authority (HAEA) about the practice of projects, strategy and planning, regulatory and licensing, regulatory and licensing procedures of safety, costs, radiation protection, decontamination, decommissioning in some OECD countries. The report waste management, dismantling technologies, includes a study about the Greifswald and Rheinsberg decommissioning of research reactors, other. decommissioning projects in Germany and the WAGR project in the United Kingdom. We also compiled a Conclusions bibliography about all available books, articles, reports and proceeding related to decommissioning from 1990 On the basis of this study, some research topics of to June 1999 (approximately 800 items). particular importance are recommended for HAEA: isotope inventory, examination of dose conversion Discussion factors, metabolic pathways of interesting isotopes, classification, volume reduction, conditioning, placing Combining of the terminology of IAEA and NEA of radioactive wastes, collecting and working up of we defined the following terms: stages, strategies and experiences related to decommissioning of VVER phases of decommissioning, decommissioning plan, reactors, PR management of decommissioning.

EMVIHflKMEMI & SftfETY | ?41 SK01K0157 DEVELOPMENT OF VENDOR INDEPENDENT SAFETY ANALYSIS CAPABILITY FOR NUCLEAR POWER PLANTS IN TAIWAN

Jan-Ru Tang

Institute of Nuclear Energy Research P.O. Box 3-3, Lungtan, Taoyuan 325, Taiwan, Republic of China E-mail: [email protected]. FAX-.886-3-4711404

Abstract on these improvements. A vendor independent safety analysis capability is therefore significant to meet the This paper presents the development of vendor needs for supporting current and future operation of independent safety analysis capability in response to the nuclear plants. On the other hand, with years of the needs for supporting nuclear power plants in the operation of the plants, many systems require design 21^ century. One of the major needs is the need of changes or modifications to at least keep or improve analysis, especially safety analysis such as for reload the plant efficiency. This also requires vendor core design and for plant system design changes. independent analysis capability to support the plant Usually, the Nuclear Steam Supply System (NSSS) either to review vendor's analysis or to perform the vendors such as General Electric (GE) and analysis directly. Westinghouse, and the fuel suppliers such as Seimens Power Company (SPC) provide all the analyses; As a country lack of natural resource such as however, more and more needs from the plants are petroleum oil or coal, Taiwan started nuclear power raised for the vendor independent analysis capability. plant construction in 1973. The first and the second The significance of this capability for the current and nuclear plants, Chinshan and Kuosheng were designed the future needs is discussed. The analysis by NSSS vendor of GE, each with two units of BWR/ methodology is described and the experience of Taiwan 4 and BWR/6, respectively. The third one, Maanshan, is presented. has two units of Westinghous designed three-loop Introduction PWRs. They started their commercial operation in a consecutive way of 1978,1981, and 1984, which totally In recent years, one of the major factors of the provide about one-third of electrical power of Taiwan. continuation of nuclear power plants is their economic Due to ardent needs of electricity, the fourth nuclear competitiveness, which will also be true for the next power plant, Lungmen with two units of ABWR generation. While there have been many aspects designed by GE is now under construction. In order to including improving management to enhance ensure operational safety, to improve plant efficiency competitiveness, the major development in the and to review the new design as well, a heavy load of technology area is the high-energy core design. This vendor independent safety analysis capability is on the includes high burnup fuel, extended fuel cycle length, nuclear society of Taiwan. The cooperation between advanced fuel design, uprating, and more flexible Institute of Nuclear Energy (INER) and Taiwan Power operating strategies such as operation under high Company (TPC) play significant role on the peaking factors. All of these demand for much more development of this analysis capability. The capability complicated design and analysis. While most of these has been applied to support the plants in many events analyses were performed by the fuel vendors, utilities and is under further development projects to enhance and their supporting organizations are responsible to its applications. ensure safely and to gain efficiency of their plants based

2421 POSTER PRESEHTATIDHS SK01K0158 THE INFLUENCE OF NUCLEAR RISK SENSE TO ATOMIC INDUSTRY'S PUBLIC RELATION

Polina Biryukova and Yuliya Kazakevich

Ozyorsk Technological Institute of Moscow Physical Engineering Institute, Russia

The formation of adequate perception depends on jA third attitude of people to danger is defined by the right information about the origin of risks. how well they are familiar with it. On one hand, factors Perception of the risks from dangerous events presents affecting risks are about information that people do a complex problem. People are prone to easily accept not even know. On the other hand, the well-known factors connected with relatively large risks while they factors attract a lot of effort. Nuclear technology was should be associating the factors with vastly less risk kept a secret for a long time. to people. It is possible to judge this on a South Ural example From the public's point of view, the most about how important it is to inform people on the dangerous factors threatening their health and life are authentic risk connected with nuclear energy notalways far from them. There are great differences production. between supposed and real dangers. This is represented Because of the lack of information, the catastrophe by the inadequate sense of risk by the people. These of PA "MAYAK" resulted in the steadfast radio-phobia differences are caused by a number of reasons. of the majority of inhabitants living in the polluted At first, a big difference exists between a voluntary territory and the formation of radio-phobia in some risk and a constrained one. Many people willingly go inhabitants living in the clean territory. The syndromes with the risk for the sake of one's inclination. They of radio-phobia, including an increase in fear and stress, think that pleasure without danger is less satisfying. have lowered the physical and mental capacity for work AH that it presents are factors of voluntary risks, so in juvenile aged children already. that people find it the wholly acceptable. Many people The population is sufficiently informed about the are convinced that freedom of risks to their own health current radiation situation, the degree of contaminated and life are essential rights of their own. Public opinion territory around their residencies, and how relocating perceives a voluntary risk with less hostility than a can help in reducing radiation damage and risk. In constrained risk. The nuclear risk is perceived as radioactive contaminated territories, the population negative because it is necessary. The people are expresses social pessimism and an increase in critical subjected to nuclear risks, and hence it becomes a attitude towards authorities and organizations. negative perception. The public role of offering risk evaluation must People are more prone to accept the risks from be increased. Without it, the people will express their natural phenomena's than accept a stipulated risk from unwillingness to be exposed to any force, including nuclear factors. For this reason, the ratings of risk the risks from a nucleus, because it will not provide factors for natural processes are always understated. reliable and objective! information.

CBMMUHICAT1SII & PUBLIC PEflCEPTIOH 1245 SK01K0159

DEMONSTRATION-INFORMATIVE CENTER BASED ON RESEARCH REACTOR IR-50 IN HEAT REGIME

Phaina Krupenina

Research and Development Institute of Power Engineering (ENTEK), Moscow, Russia

Many problems exist in the nuclear field, but the Demonstration of the safe functioning of the most significant one is the public's mistrust of Nuclear heating reactor IR-50 in the center of the city is a basic Energy. Strong downfalls of the radiological culture goal of the Demonstration Department of DIC. affect public perception, the main paradox being the The Information Department will be the second situation after Chernobyl. The task of creating a part of the Center. It will include the following groups: Demonstration-Informative Center (Minatom RF) on reactor IR-50 research is conducted by Research and • information-analytical group; Development Institute of Power Engineering • heat supply NPP information group; (ENTEK). The IR-50 is situated on the grounds of the • atomic energy information group. institute. It will be a unique event when the functional reactor is situated in the center of the city. The technical support group will be the third part of DIC. It will ensure technical activity of the Center. Purposes of the Demonstration- The projects of ENTEK successfully passed an Informative Center examination in Gosatomnadsor, Russia. Such a project was done in Canada (Pinava, Manitoba, Whitshell The Purposes of the Demonstration-Informative Nuclear Research Establishment) with positive results. Center (DIC) are as follows: • Demonstration of the safe functioning of the reactor Concept of IR-50 conversion to situated in a densely populated region of Moscow; heat-regime • Organization of lectures, seminars, laboratory and Design of the IR-50 research reactor project and practical work based on IR-50 for students; its construction was completed more than 30 years ago. • Overcoming the "Chernobyl syndrome" and However, the accepted method and technology of radiophobia, meeting with residents, advertising, reactor safety has not lost its urgency in the present lectures in schools, universities and other institutions; Currently, in this society, as all over the world, • Demonstration of the prospect of NPP construction there is a counteraction to development and even for heat supply (heat supply NPP) with pool-type maintenance of Nuclear Energy. Some people, reactors; pursuing their own mercenary interests, actively • Demonstration of positive qualities of Nuclear introduce such attitudes in the public consciousness. Energy and comparative features to other sources Many of them are united under the "green" flag. They of energy; were widespread and greedy for sensational mass media • Demonstration ofmodem level and prospect ofNuclear in the struggle with nuclear technologies. Energy development in Russia and in the world; Thus, it is necessary to lead educational activity • Raising ofNuclear Energy prestige and drawing for the public. The most effective way is through so- young specialists to this branch. called Visitors' Centers (only in Great Britain there The main base of the Center will be the research are more than 10). But Russia offers another kind of reactor IR-50 in a heat-regime with exploratory center, a Demonstration-Informative Center based on functions for the reactor. the research reactor IR-50 in the heat regime.

24B} PDSTES PSESEKTATIONS SK01K0160 PUBLIC OPINION SHIFTS TO THE FAVOUR OF NUCLEAR ENERGY DUE TO STEAM GENERATOR TRANSPORT

/. Lengar, T. Nemec

Josef Stefan Institute, Jamova 39, 1000 Ljubljana, Slovenia

Abstract Early morning on the third day of the transport 13 members of Greenpeace chained themselves to the In late August and early September of 1999, steam generator structure, and thus stopped the nuclear energy topics occupied a central place in the transport for three hours. Activists were all Austrian Slovenian media because of the transport of two new citizens; there were no Slovenian sympathisers isteam generators to the Kroko nuclear power plant, supporting this protest action. Later Austrian and also due to the protest action of an Austrian Greenpeace also demanded a counter move from the Greenpeace group. Before these events, the public Austrian government, as a response to the «brutality» opinion in Slovenia was not in favour or nuclear energy of Slovenian police action during the arrest and land Greenpeace had a good reputation. In September unchaining of activists from the steam generator. it has lost much credibility because of their clumsy action of protest, and in just one month this caused a Before this nonsensical protest action by shift of public opinion in Slovenia towards support of Greenpeace, this organisation had a good reputation Slovenia's only nuclear power plant. The Greenpeace in Slovenia. But in this action only foreign citizens protest action occurred during the transport of the two participated and used an offensive approach to the NPP new steam generators to Krjtko. By replacement of modernisation. This caused the Slovenian public and the old steam generators the operation of the Rntko the media to oppose and criticise this action. NPP will be extended until 2023. The transport envoy The shift toward a favourable opinion of nuclear travelled during the first half of September '99 across power was enhanced in the same week, because of the a considerable part of Slovene territory, passing by the statements made by Austrian politicians. Seeking for capital of Ljubljana. votes in the approaching elections, they put pressure on Slovenia to close the Kr^bko NPP, and threatened Discussion counter measures. The Austrian minister of foreign affairs expressed his concern due to «the fact», that the Over the last few years, the public opinion in «eastern technology NPP» (made by Westinghouse, an Slovenia has not much favoured the use of nuclear American company) does not meet the safety standards. energy, though many people were ignorant. The Slovenian media gave a hostile response to these In September '99 the replacement of the steam unprofessional antinuclear statements. generators for the Krj&ko NPP took place. They had to Coincidence of the steam generator transport, the be transported in a huge envoy from the Koper port to failure of the Greenpeace protest action, and the Krjbko, across more than half of Slovenia and the political demands for the closure of the Krjtko NPP as attention of the media started to focus on the event. On one of the central issues in the Austrian elections, was the first day of transport, the front pages of most a big promotion for nuclear power in Slovenia. These newspapers were covered with pictures of the steam combined events shifted the public opinion in favour generator, and large numbers of spectators by the of the extension of Slovenia's NPP operation. The latest roadsides to see the colossus. The media coverage of poll showed that only 26.7 % of Slovenians still support this event certainly helped to de-mystify nuclear energy. the end of NPP operation, compared to about 50 % People started to lose their fear of nuclear energy, and before these events took place. to inform themselves about the high technology used in a nuclear power plant. Another push in popularisation was provided by Conclusions the Greenpeace organisation and politicians from The Slovenian example illustrates the effect of a Austria, a neighbouring country. Austrian politicians failed protest action by Greenpeace on the public have demanded immediate closure of the Krjtko NPP, opinion of nuclear energy. Greenpeace activists were which makes them very unpopular in Slovenia. The not local people, but were foreigners, and this protest Austrian national standpoint gave strong incentive for action was seen by the public and the media as an the Austrian Greenpeace to start protest actions against interference in the Slovenian internal affairs rather than the steam generator transport, and eventual Krjtko NPP a true protest. upgrade and modernisation. COMMUNICATION & PUBLIC PERCEPTION 1247 SK01K0161 ANALYSIS OF ACTIVITY OF INFORMATION INQUIRED GROUP ON RADIOECOLOGY AND PUBLIC COMMUNICATION IN OZYORSK (THE TOWN OF NUCLEAR INDUSTRY) Elena Govyrina

Ozyorsk Technological Institute of Moscow Physical Engineering Institute, Russia The Information Inquiry Group on Radioecology 9. Participation in preparing scripts and in survey- and Public Communication is a branch of the Depart- ing popular science education and documentaries on ment of Production Association "Mayak." "Mayak" radioecology, technology of processing of spent nuclear was formed in1989. fuel, and waste management. The main tasks of the group are: 10. Organization of plant visitations for foreign 1. Formation of objective public opinion about the delegations, delegation of Russian specialists, and activity of the plant through: representatives of mass media. 1.1. Organization and presentation of 11. Providing answers to public questions on is- informational lectures, sues dealing with the ecological situation in the plant's il .2. Implementation of excursions to the plant and region. the museum of PA "Mayak." 12. Informing plant management about group's 1.3. Informing the mass media. activity. 1.4. Organization of fairs, "round table" discus- 13. Making visual aids. sions, and press conferences. ;14. Wide interaction with another branches of the , 1.5, Production of educational documentary vid- plant on issues dealing with providing efficient publicity eos. about incidents, publications, transport, inventory, etc. 1.6. Distribution of informative materials. Important insights on public opinion and public 1.7. Creation of a database on nuclear problems. understanding with regard to the nuclear field: 1.8. Keeping of an archive at the PA "Mayak" 1. Necessity of addressing both interest and museum. contradictions of different social groups. 2. Improving the structure of control on labor 2. Systematically informing the public about organizations. achievements in the field of nuclear power safety and 3. Provision of high-level labor discipline. utilization of radioactive wastes on federal and regional 4. Organization and improvement of qualified levels. personnel. . 3. Examine the alternative of no nuclear industry The functions of the group are: with the public exposed by questioning of public opin- 1. Efficient delivery of information to the mass ion, challenge nuclear socialists on objective shortages media about events that took place on the plant, and in the nuclear field. present public interest issues. 4. Outstripping exposition of main problems of 2. Collection and study of plant information pub- nuclear power, perturbing the public. lished in mass media, private contacts with public 5. Efficient informing of the public about radiation representatives. incidents with an estimation of its consequences on 3. Organization of visitations by public humans and the environment. representatives to the plant and museum of PA 6. Development of a system of radioecological "Mayak." education of the public on issues dealing with biologi- 4. Organization of continuous business contacts cal action of radiation, natural radioactive background, with journalists, scientists, specialists, and public out- and the history of nuclear power. reach workers. 7. Wide publications of modern medical and 5. Participation in the organization of press radioecological data by analysis of consequences of conferences and discussions of issues dealing with Chernobyl and another radiation incidents. radioecology and nuclear power. The main quality of a prepared specialist is not 6. Preparation of informative and objective mate- only knowledge of the technical and scientific side of rials on issues dealing with the plant!s activity. a problem, but the ability to clearly and accurately ex- 7. Distribution of informative materials developed plain its content in a form understandable by non-spe- by the group. cialists. Nobody must abandon doubts in our free will 8. Organization of speakers in schools, medical and in our aspiration to be the best. It may be reached organizations, and labor collectives on issues dealing with only in the future if our reports and information will the plant's activity, its technologies, nuclear and radiation subdue listeners by being frank, honest and trustwor- safety, and implementation of established norms and rules. thy. 248! POSTER PRESEKTAT1BHS SK01K0162 A RISK COMMUNICATION CASE STUDY: THE NEVADA RISK ASSESSMENT/MANAGEMENT PROGRAM

by Anthony E. Hechanova

Harry Reid Center for Environmental Studies, University of Nevada, Las Vegas, USA

The Nevada Risk Assessment/Management The process of risk assessment for the DOE sites Program (NRAMP) is part of a national effort by the in Nevada is complicated by many contaminant types, U.S. Department of Energy (DOE) to develop new potential land uses, exposure pathways, and public sources of information and approaches to risk interests. In addition, few definitive conclusions can assessment, risk management, risk communication and be made about the risks because characterizations are public outreach as these objectives relate to the not complete as a result of the size, complexity, and ecological and human health effects of radioactive and limited funds expended to date. hazardous material management and site remediation The NRAMP approach to risk assessment taken activities. This paper reviews the innovation behind during the 1995-1996 effort was designed to assist all the NRAMP project and presents a synopsis of the involved parties (the NRAMP Stakeholder Working NRAMP effort which occurred from 1995 to 2000. Group and technical team, the Community Advisory Board for Nevada Test Site Programs, and the general The primary goals of the DOE in awarding the public) to participate in the development of a Preliminary cooperative agreement establishing NRAMP were to Risk Assessment for DOE sites in Nevada (PRA). (1) use a risk-based approach to evaluate the consequences of alternative actions in DOE's For the sake of consistency in dealing with the Environmental Remediation Programs at sites in complex and varied technical issues of DOE sites in Nevada, the PRA focus was limited to Maximally Nevada and (2) use a neutral and credible institution Exposed Hypothetical Individual (MEHI) risk from outside the DOE to perform the risk assessments and specific land use scenarios and various contaminant Contribute to public education about environmental source categories. Critical to this approach was the management issues at the Nevada Test Site. development and use of consistent assumptions and These goals are action-oriented interpretations of parameters. Therefore, the NRAMP technical team, the U.S. National Academy of Sciences advise to DOE with the input of the Stakeholder Working Group and on how risk-based decisions could be incorporated into the Scientific Peer Review Panel, determined five the Environmental Management Program. The source categories and five land use scenarios which resulting report by the U.S. National Research Council formed the basis for the NRAMP technical approach (NRC), identified certain obstacles to a risk-based to a preliminary risk assessment. The risk management approach and confirmed the importance . communication effort that resulted in the underlying of stakeholder involvement in performing the risk assumptions for the PRA is highlighted in this paper. evaluations. One of the primary NRC recommendations The PRA provides preliminary qualitative and was that the credibility of an evaluation of site-wide risks quantitative information about current and future public would be greatly enhanced if the evaluator were other health risks from the NTS. This information is intended than DOE, and the NRC identified six criteria which an to be applied to the development of refined risk estimates, institution should satisfy in order to establish credibility: stakeholder comments on site restoration activities, and (1) it should be perceived as being neutral; (2) it should prioritization of environmental activities by the DOE. have management capability; (3) it should have the Quantification of risks in this PRA is limited to Maximally ability to conduct scientifically valid and responsible Exposed Hypothetical Individual (MEHI) risk scenarios risk assessments; (4) its assessments should be subject developed in response to stakeholder safety and future to independent, external review by technical experts; (5) land use concerns. Results of risk assessment for these it should have the ability to plan, organize, manage, and land use scenarios are intended to provide insights on the facilitate public participation; and, (6) it should have location, timing, and severity of hazards at the NTS. This the ability to effectively communicate complicated MEHI risk approach is initially appropriate at a screening scientific information on potential risks and uncertainties. level because of the current land use that precludes public The implementation of DOE's Cooperative access to contaminated areas and the lack of Agreement with the University of Nevada, Las Vegas comprehensive information on all contaminants that would is believed to be the first site-specific application of be needed to support a more detailed risk assessment. the NRC-recommended risk assessment process for The PRA accomplishes two important objectives. supporting site-specific decision-making. As originally First, the PRA identifies gaps in technical knowledge proposed, NRAMP developed an integrated that will be useful to prioritize future risk assessment stakeholder, scientific peer review and risk assessment activities such as gathering additional needed data.. process that tracked the goals enunciated in the DOE Secondly, the procedure has been a valuable experiment Notice of Program Interest. in stakeholder involvement in scientific risk assessment. COMMUNICATION & PUBLIC PEBCEPTIOK 124B SK01K0163 PUBLIC ACCEPTANCE/PUBLIC UNDERSTANDING: NEW APPROACHES

Sergey Ryabtsun

Russia

Abstract education of a population. A crushing flow of horrible information, not always correct, and an absence of This paper will address ecological education of a systematization have brought about the prevailing population, with the example of Lesnoy city. thought on ecological problems of the city. Those valued by inhabitants include soiling water, the Summary atmosphere, and radiation harm, based on inconsistent opinions and judgements, most of which do not comply In planning territorial development, clear and with reality. objective factors must be considered, characterizing life ambience of a person. For one person - biological In the report, absolute data are given on the essence, with other - social-public, its life ambience quality of drinking water in the city, medical statistics, has two interconnected parts: natural and social- statistics from the center of employment of population economic. In accordance with the law of equivalence and other statistical data. This work carefully analyzes of ecological factors, each separate contribution plays the level of ecological education of a people and brings an equivalent role. concrete recommendations on how to work with the population, Results of sociological questioning in Lesnoy city ; have revealed a defect in the system of ecological

251 f POSTER PRESEHTATIIHS SK01K0164 PUBLIC INF< NORTH WfcC'l

Anna Saiapina

State Educational Center St. Petersburg, Russia

The objectives of the CPI are: Center of Public Information (CPI) in North West - conducting the all-round informational, region of Russian Federation is a part of the State enlightenmental and educational activities for Regional Educational Center of Ministry of the.Russiori forming the positive attitude and treatment to Federation on atomic energy. atomic energy and nuclear technologies; The premises of the Centre (about 500 sq. rn.) -^ providing the population and the means of public include the exposition hall, video hall, the hall for press information with the reliable information about the confrences, rooms for meetings, conferences. objects of potential risk; - organizing active exchange of the information with CP.I provides the visitor with the wide range of the enterprises, using nuclear tecnologies in the information dealing with the nuclear power, It was region. opened in the structure of State Regional Educational Centre in 1997. In out exposition the following sections are presented: Producing electricity. Why nuclear power? Regional Centre of Public Information of Nuclear reactors in Russia. Nuclear safety. The new MINATOM of Russia in Saint Petersburg was created Russian reactors. Nuclear power and environment. according to the agreement with the Eruropean Union Radioactivity. About Saint Petersburg. Nuclear power: Cornission in the framework of the TACIS program who does what? with the participation of French companies EDF, COGEMA, STEPFER. The work of the North West Center of Public Information is described in the presentation.

Oiiffifflfifl SK01K0165 THE STRATEGY OF IMPLEMENTATION OF AN INFORMATIONAL MANAGEMENT SYSTEM FOR A POWER STATION

Carmen Isar

CNE-PROD CERNAVODA, Maintenance Support Section, Romania Abstract permits and material control, to the preparation, .-»... revision, and control of documentation. For many years now, the use of computerized data has been spreading at an increasing pace, and as a result, The implementation of an Information now covers every imaginable application. Power Management System requires modification of work stations have been no exception to this process, The processes. The cost to acquire and implement this aim of data processing applications used in power system is high. However, many utilities have found stations has been the direct support of the organizational that these costs are justified for many reasons including and operational processes that are normally performed safety, efficiency and cost aspects, and strategic by people. This has resulted in a very wide field of reasons, applications, ranging from the processing of work

ECONOMICS 1257 SK01K0166

DEVELOPMENT OF THE NUCLEAR ENERGY PRODUCING AS A MAIN FACTOR OF PRESERVING THE ECOLOGICAL BALANCE ON THE EARTH

M. Yanusova

Russia

The main objective of the paper is to show by «Brundt!and Commission)) («Our common futures, comparison and to justify the potential of atomic energy 1987) and recommendations of Conference «RIO-92». development as a single way of producing In brief, all recommendations could be reduced to environmentally-safe energy in a scope required for selection of a strategy of civilization development mankind. defined as «Sustainable Development)) (SD). For the ' first time this term appeared in the paper «A11-World At the end of the XX century, on the threshold of Strategy of Environment Protection)) presented by the the third millennium, civilization finds itself at a historic International Union for Environment Protection and point which is often named differently («moments» by Natural Resources. Tan; «nodes» by A. Solzhenitsyn, «infraction» by A. Tointy, etc.). This point determines dynamics and This paper considers in detail one of the SD directions of civilization evolution for a long-term components - energy. It is power availability that perspective. determines living standards (life quality) and capability to influence the environment in a proper manner. Today, It becomes obvious at the threshold of the XXI people spend energy not only to meet their needs but century that traditional stereotypes, if preserved, will more and more to compensate negative consequences not ensure a constructive life of civilization and will of human activities (including reprocessing of wastes). pose a threat to the balance and stability of the historically-formed ecological structures. The paper considers in detail advantages and disadvantages of different types of energy, including This problem is a multidisciplinary one, and it alternative ones, and analyzes prospects of their depends on a large number of interrelated factors. A development in the XXI century. simplified scheme can be presented as «resources - energy + mankind's activity - wastes (ecology). In the paper special attention is given to different Contradictions between the earth population growth types of atomic energy, including both conventional and necessity to meet energy needs of mankind, on the methods of atomic energy production and altenative one hand, and limited capabilities of environment to ones, such as deuterium energy. assimilate products of mankind activities, on the other hand, become antagonistic. References: Study of various aspects of future development of civilization begins with basic research by Darvin 1. "Explosion deuterium energy", G.Ivanov, and Maltus, the theory of overpopulation. N.Voloshin; Snezhinsk, 1996 Most complex research was performed by the 2. "Growth limits", "Roman Club", 1972 «Roman Club» («Growth Limits», 1972). Currently of 3. "Ecology and sustainable development", V.Los, special interest are the materials prepared by O.Draer; Moscow, 1997

NUCLEAR PROGRAMS & TECHNICAL COOPERATION 1201 SK01K0167 THE KEY FOR COMPETITIVE NUCLEAR POWER, A VIEW FROM TAIWAN

James Lin

Nuclear Safety Department, Taiwan Power Company, Taiwan Abstract Taiwan. However, the capital cost for constructing the new power units, the two GE-ABWRs, will have an The article mainly deals with the current situation adverse effect on its financial status as a whole. Further of nuclear power generation in Taiwan. The cutting down the 0 & M costs is a must and continued development of nuclear power has been long and to be the most challenging work for nuclear people in punctilious, whereas the contribution to meet the power Taiwan. demand in the lean resouce country is prodigious. This article delineate the structure of power generation costs with coal, petroleum oil, and nulear reactors in Taiwan Discussion in the recent 20 years, which highlights the superiority of nuclear application. However, as we see it from Although safety is the limiting condition for Taiwan, the nuclear power could have been better if nuclear power plant to operate, economic we can simplify the design and regulation of the reactor. competitiveness is the only reason to run or to shut a nuclear power plant. This is why this article put Methods emphasis on reducing the costs for generation of nuclear power. In recent years, changing attitude of This article scrutinizes the electricity generation the nuclear people help to simplify the work process costs with coal, petrolium oil, and nuclear reactors in and delete non-value added steps in nuclear power Taiwan in the recent 20 years. The costs are broken- plants to reduce the O&M cost. Although the reduction down, trended, compared, and discussed. From the of the production cost has being reduced, production operational experiences and from the stand point of cost is only part of the total cost and is not easy to view of risk, this article also questions the effectiveness . depress. For more competitive, simpler designed, faster of the safety systems in atypical nuclear power plant. constructed nuclear units shall be developed,

Results Conclusion

The capital cost for constructing a nuclear power From the cost comparison and the safety system plant turned out to be the most influential factor for review, This article proposes that the nuclear family the development of the nuclear industry. Attributed should scrutinize the effectiveness of the current nuclear to cutting down the operational costs, the nuclear power hardwares and softwares, and pursue simpler work is still the most competitive electrical power source in processes and simpler reactor design.

2B21 POSe PRESENTATIONS SK01K0168 ASPECTS REGARDING THE FUEL MANAGEMENT FOR PHWR NUCLEAR REACTOR

Bobolea Alin, Voicu Alexandru, Dragusin Octavian

"POLITEHNICA" University of Bucharest, Romania

Abstract 1. Group irradiation and poison management Fuel management for the PHWR nuclear reactor 1.1 Uniform removal of poison 1.2 Removal of poison from an irradiated zone (natural uranium - D2O ) is completely different from the P WR reactor fuel management. PHWR reactor fuel 1.3 Removal of poison in order to achieve a loading procedures are repeated after an interval of constant power distribution time, as defined and specified in the project 1.4 Constant poisoning documentation, using a fuel machine that can be 2. Discontinue irradiation due the unloading of the attached to the terminal fittings of horizontal pressure reactor tubes while the reactor is at full power. Another aspect 3. B i-directional continual irradiation of fuel management policy is related to the possibility Input data for the model were PHWR - GANDU of bi-directional loading of the reactor, with the primary 600 characteristics and fuel properties during the burn- advantage of uniform and symmetrical characteristics. up process.

Methods Results This paper presents the dynamic programming method used to solve the problems related to the fuel Topics studied: management for several loading phases during a period in which the reactor is functioning at full power. • 'axial variation of relative power density for con- stant axial speed displacement of fuel Optimal methods of manipulating the fuel inside • power density radial distribution the core of the reactor has to fulfill three main objectives: • dependence between maximum burn-up of fuel, enrichment and fuel displacement techniques - minimizing power peak - maximizing the fuel burn-up • medium relative power density variation - minimizing stopping time Conclusions Reactor models, used for the dynamic programming method, simulate the fuel burn-up in the reactor through results obtained by resolving mono- The fuel manipulation process of a PHWR reactor energetic diffusion equation. can be optimized using the dynamic simulation program. To know and minimize the Xenon V-DVO -O-V -0 = concentration consequences when the reactor is shut down, a calculation code was built. The results were The calculation is only for half of the reactor's used to determine the optimal enrichment of the fuel, core because the neutron flux is symmetric over the the techniques of poison control and the core minimum reactor's center. The models analyze the management volume in order to get maximum power. procedures that can be used for a PHWR reactor, which include:

FUEL CYCLE IMLLffiES 1265 SK01K0169 CALCULATIONS OF NON-STATIONARY PROCESSES AND SPATIAL DISTURBANCES IN THE BN-600 REACTOR CORE

A.M. Tuchkov, A.L Karpenkoy Yu.A. Blinov Beloyarsk NPP, Russia

E.F. Seleznev VNIIAES, Russia

The results of subassembly characteristic and The algorithm pertaining to the disconnection of neutronic and physical calculations during local one of three heat transfer loops without a drop of control disturbances at power density field distribution by rod (reducing the reactor power by 33% against the initial movement of reactor BN600 control rods are presented level) has also been presented. The design algorithm of in this report. a loop disconnection transient is compared. The calculations were performed using the The proposed algorithm prevents a reduction in integrated GEFEST software designed by VNIIAES. fuel power to lower than the present value and total The calculation by GEFEST code performs by diffusion distortion of the core power density field. approximation in 3-D hexagonal geometry. The reactor The possibility of substitution of boron carbide BN600 model contains about 18,000 points. safety control rods by boron carbide with reduced The calculational results of BN-600 reactor control enrichment of boron-10 can be performed without rod effectiveness and coarse control rod interference reducing the BN-600 reactor control rods system's coefficients are presented. effectiveness is also illustrated.

2661 POSe PRESENTATIONS SK01K0170

APPARENT STABILITY CONSTANTS OF THE COMPLEXES OF AM(III) AND CM(III) WITH HUMIC ACID AS A FUNCTION OF PH BY THE SCHUBERT METHOD

Tomofumi Sakuragi1, Shusuke Sawa1, Seichi Sato1, Hiroshi OhashV, Toshiaki Mitsugasira2, Mitsuo Hara2 and Yoshimitsu Suzuki2

'Graduate School of Engineering Hokkaido University, Japan institute for Materials Tohoku University, Japan

Abstract creased with de- 10 creasing concen- TRU elements released from nuclear waste after trations of Eu(III), disposal may form complexes with humic acid occur- Am(in)andCm(III) ring in natural aquifers. Their migration behavior may in pH 5.0-5.5. In be affected by the complex formation. Apparent sta- the Schubert me- bility constants of the complexes of Am(III) and thod, ion exchange Cm(III) with humic acid, pw, have been determined resins strongly ad- Dialysis method by a Schubert method in the pH range from 4 to 8. sorb Am(III) and (previous work"*) The results compared the present data with those by a 0 1 1 Cm(III), whose dialysis method '. It was found that P " remarkably concentration in 6 8 changed, depending both on pH and its concentration. aqueous phase pH (Am(III)+Am-humate) decreases a few orders of mag- Experiment nitude lower than that by the previous dialysis method. app M2 The difference of log P between two methods can Batch solutions containing "'AmCIII), Cm(III) be elucidated by the difference of concentration of and Dowex 50W-X4 cation exchange resin were pre- Am(III) and Cm(III). pared at solid/liquid ratio of 0.10 g/dm3, ionic strength of 0.1, HA concentrations of 0, 1, 5, 10, 20 ppm and The increase of log papp from pH 4 to 5 would be pH from 4 to 8 under a CCyfree atmosphere, using an caused by dissociation of humic acid with the rise of w N: glove box. After the resin was equilibrated with pH. On the other hand, the decrease of log P from the solution at 298 K, the specific activities of Am(III) pH 5 to 8 can be ascribed to decrease of stability of and Cm(III) distributed between solid and aqueous AmJ* and Cm3+ with the pH-rise. In the previous work, phase were measured by a-spectroscopy(2). pw have been determined in the pH range from 3.0 to 6.0. Beyond pH 6.0 the dialysis method was not ap- To determine Pw, the fitting curves for distribu- plicable because Am(III) and Cm(III) could not per- tion coefficients at the 0 (K/), I-20 (K,) ppm of hu- meate through the membrane which strongly suggests mic acid were obtained as a function of pH. Then, ]T 3+ hydrolysis of Am and Cm with the rise of pH. P»pp was determined by plotting the distribution coefficient as a function of humic acid concentration Conclusions ([HA]) at a constant pH, based on the equation: Apparent stability constants p*"? of the complexes of Am(III) and Cm(III) with humic acid have been de- termined in the pH range from 4 to 8. Logarithm of $*& where i is the number of humate ligand bound to a increased from pH 4 to 5 and decreased from 5 to 8. metal ion. References Results and Discussion (1) M. Samadfam et al.: Effect of pH on Stability Apparent stability constants Pipp obtained are Constants of Am(III)- and Cm(III)-Humate shown in Fig. 1, together with those by previous di- Complex, Proc. 2nd NUCEF International Sym- alysis method. The difference between P3*1" for Am(III) posium (1998). and Cm(III) was indistinguishably small. Therefore, (2) T. Mitsugashira et al.: 1996 SERNIA Symp. the results are described using a single curve both for Envirn. Radioactive Nucl. Impact in Asia 291- Am(III) and Cm(III). 296 (1996). w The present log P by the Schubert method is (3) W. Hummel et al.: Complexation of Radionuclides remarkably greater than the previous one. Hummel et <3) with Humic Substances: The Metal Concentration al. reported that the stability constant remarkably in- effect, Radiochim. Acta 84,111-114 (1999). FUH CYCLE CHALLENGES 1267 SK01K0171 AN UNEXPLORED FRONTIER: FLUORINATION OF URANIUM OXIDE BY CF4/O2 R.F. PLASMA

Yong-soo Kim, Yong-daeSeo and Sang-hwan Jeon Nuclear Engineering Department Hanyang University, Seoul 133-791, Korea

Byung-oke Cho Nuclear Power Technology Development Section Division Nuclear Power Department of Nuclear Energy IAEA, Vienna, Austria

Abstract experimental findings, the overall dominant reaction of UOj in CF/O plasma is determined as: Research on the direct dry fluorination process of 2 uranium oxides in CF /O r,f, plasma is being carried 4 2 UO + 3/2CF, + 3/802 -> UF + 3/2(CO or C0 ) out for the future replacement of the current multi-step 2 6 2 uranium fluorination processes which contain the solvent extraction, reduction, and fluorination. In the Surface analysis reveals that UO2F2 is formed on present study, the apparatus for the plasma reaction is the surface as a primary intermediate species in the designed and manufactured, and the fluorination fluorination reaction. reaction of the uranium dioxide, UOj, is investigated. The reaction rate is expressed in terms of the etching rate in this study and it is found that the highest First of all, the reaction rates are investigated as etching rate of 0.4/min is obtained at the optimum 20% functions of the ratio CF/O , plasma power, and 2 0 mole fraction, regardless of r.f. power and substrate substrate temperature. It is found that there is an 2 temperature. It is also found that as the substrate optimum CF /O ratio for the effective fluorination, and 4 2 temperature and r.f. power increases, the etching rate the ratio is four. The optimum ratio is confirmed by is enhanced linearly. XPS and XRD analysis. Even though several uranium

fluorides are produced in the reaction, the mass Research on the dry fluorination of U30, is under spectrometry analysis identifies that the major reaction examination. In the conference, the results of U}0, product is uranium hexa-fluoride, UF^ Based on the fluorination will be presented.

2B01 POSTER PRESENTATIONS SK01K0172 OPERATIONAL BENCHMARK FOR WER-1000, KOZLODUY NPP UNIT 6

T. Apostolov, B. Petrov

Institute for Nuclear Research and Nuclear Energy Bulgarian Academy of Sciences Abstract Results

Benchmark calculations have been carried out Global neutron-physics characteristics of the using the 3 D nodal code TRAPEZ. The neutron-physics VVER-1000 core, Kozloduy NPP Unit 6, have been characteristics of the VVER-1000 core, Kozloduy NPP determined taking into account the real loading patterns Unit 6, have been determined taking into account the and operational history of the first three cycles. In the real loading patterns and operational history of the first axial direction the reactor core is divided into 20 layers, three cycles, The code TRLOAD has been used to each with a thickness of 17.75 cm. perform the fuel reloading between any two cycles. The reactor and component descriptions, as well Discussion as material compositions, are given. The results The calculated and the measured values of the presented in the paper include the critical boric acid critical boric acid concentration are compared. In all concentration, the radial power distribution, the axial three cycles the results from the calculations give good power distribution for the maximum overloaded agreement with the operational data. assembly, and the burnup distribution at three different moments during each cycle. Comparison is made between the calculated and the measured normalized radial power distributions at Calculated values have been compared with different time steps. Mean relative errors of the measured data. It is shown that the results obtained by calculated values are given. The analysis shows that the TRAPEZ code are in good agreement with the the differences are smaller than the inaccuracy of the experimental data. Thus the information presented measured data. The axial power distribution for the could serve as a test case for validation of code maximum overloaded assembly at some time steps is packages designed for analyzing the steady-state presented. The direct comparison here is not possible, operation of VVERs. because the axial positions of the detectors are different Methods from the ones used in the calculations. In any case, there is very good agreement between the experimental and the calculated values of the power distributions. The experimental data have been obtained by means of self-powered neutron detectors located in 64 The calculated burnup distribution at various time positions in the radial plane and in 7 positions in the steps is given. Although there is no corresponding axial plane. These data have been used to reconstruct measured data for this characteristic of the reactor the power field which will be referred to as measured operation, the fact that the power distributions continue data. The inaccuracy range of the measured data is ±5%. to give good agreement with experimental data for cycles 2 and 3 indicates that these burnup data should The calculations have been carried out using the be valid. 3D nodal code TRAPEZ. It solves the 2-group diffusion equation for hexagonal fuel assemblies, divided into one internal hexagon and six outer trapezoids in the Conclusions radial direction. The 3D problem is transformed by The comparison of the calculated values of basic separation into a 2D (x-y geometry) and a ID (axial) neutron-physics characteristics of the VVER-1000 solution. In the x-y direction partial solutions for the core, Kozloduy NPP Unit 6, with the measured ones, neutron fluxes within the nodes are approximated by and the subsequent analysis demonstrated the suitability exponential and trigonometric functions. For inner of this operational benchmark to serve as a test case iterations an overrelaxation method is applied, and for for the validation of VVER fuel management code the outer iteration a Chebyshev method is used. The packages. code TRLOAD has been used to perform the fuel reloading between any two cycles.

FUEL CYCLE CHALLENGES I 2EB SK01K0173 LEARNING FEATURES OF FISSION FRAGMENT INTERACTIONS WITH POLYMERIC COMPOUNDS

S. Kosarev The Institute of Nuclear Power Engineering, Obninsk, Russian Federation Abstract and also their distribution in the film thickness have been calculated. The main purpose of this work is to learn features Experimental research on the structure of fission of the fission fragment interactions with polymeric fragment latent tracks is planned with the help of gas- compounds. The calculation performed has shown that dynamics methods and electron microscopy. the available calculational research methods of energy losses and range in the substances, allows prediction Results of the results of the experiments with good accuracy. The obtained calculation results have been Methods compared with experimental data, including data obtained in SSC RF IPPE, The installation has been The calculation of energy losses and the heavy calibrated for the passage of air through a polymeric ion range through the substance was carried out with film. The number of irradiations of PET-films by the use of the theories of a Bethe-BIoch and Lindhard. fission fragments of U-235 is conducted. The The theory of a Bethe-BIoch is used at enough high experimental methods have been defined and also the velocities of incoming ion, as at small ion velocities. density of tracks of fission fragments on depth of a It captures the electrons of the target atoms that result polymeric compound, which is composed of several in the decrease of a charge of an ion. Atthe ion velocity thin films. less than voz (v0 - Bohr velocity of an electron; z - the nuclear charge of incoming ion), decrease of a charge Discussion of an ion occurs. When the charge of an ion differs very much yet not from the nuclear charge, the concept Calculational results have been compared with of an effective charge of an ion zeff (a parameter which experimental data and they have shown a good agrees the experimental and calculational values of agreement. The data, obtained in SSC RF IPPE and energy losses in the best way) is used. The effective calculational dependence of residual energy of the charge of a heavy ion inhibited in a substance is defined fission fragment on the path length are in good by the Brown-Moak formula. In the region of velocities agreement. The calculational dependence of residual M (vo

CALCULATION RESEARCH OF NEUTRONICS PARAMETERS OF THE WER REACTOR WITH Th AND ™U DIOXIDE FUEL

A.A.Belov, V.M.Decusar, A. G.Kalashnikov

Institute of Physical and Power Engineering, Obninsk, Russia

In this work some variants of uranium - thorium the point of view of neutron balance) of recycling fuel cycle of the VVER reactor are considered. The uranium in the thorium fuel cycle with adding and WIMS-ABBN code was used for variant calculations. without adding plutonium. Recycling uranium is These include recycling of uranium - thorium fuel for «good» from the point of view of nonproliferation of the VVER-1000 standard assembly with 11 pins on each divided materials. On the basis of obtained data, the side and for the 12 pin assembly with low power, and also conclusion is made that it is impossible to remain within one variant for these cycles with heavy water as both the the framework of primary design solutions to achieve heat-carrier and moderator. The variant of recycling of a reproduction factor close to one. uranium - thorium fuel for the 12 pin assembly is also calculated for each side of the reactor with reduced power Methods to increase the reproduction factor, such and adding weapons grade plutonium. as magnification of a multiplicity of reloading and magnification of burn-out, for a design of the VVER The calculations have shown a possibility (from reactor with reduced power are shown.

FUEL CYCLE CHALLENGES!?.' SK01K0175

RELATIVE ABUNDANCES AND PERIODS OF DELAYED NEUTRONS FROM FISSION OF 239PU BY FAST NEUTRONS

V.M.Piksaikin, L.E.Kazakov, S.G.Isaev, G.CKorolev, V.A.Roshenko, M.Z. Tarasko, R. G. Tertytchnyi

State Scientific Center of Russian Federation Institute of Physics and Power Engineering, Obninsk, Russia

Abstract processing of experimental data consists of the summation of DN time distributions from different The fundamental role of delayed neutrons (DN) measurement cycles. To obtain the estimations of the in the kinetic behaviour, control, and safety of nuclear DN relative abundances and decay constants the reactors is well known today. Delayed neutrons are of analysis of the total DN distributions was performed interest not only in the field of reactor physics, but also using the iterative least-squares fitting procedure. In in the field of nuclear physics and astrophysics. Delayed the process of the least-squares analysis the information neutrons also serve as a basic tool for studying the on the correlations between DN parameters was yields of various fission products as a function of obtained. This information was used in deriving the incident neutron energy. final set of the DN parameters averaged over different In spite of great efforts devoted to the investigation measurement runs related to the same energy of incident of delayed-neutron physics, the fundamental delayed- neutrons. neutron parameters of the most fissionable isotopes encountered in reactor systems are still poorly known. Results For example, delayed-neutron parameters used in most reactor physics applications are usually associated with Experimentally determined DN group constants either thermal or fast fission spectrum. It is difficult for the incident neutron energies from 0.37 up to 5 MeV for the reactor designer to select the most applicable are presented as a traditional 6-group model. The delayed neutron data set for an intermediate-spectrum comparison of the obtained energy-dependent DN system. group parameters with data of other authors was performed in the terms of the average half-life of the The primary objective of the. present work was to DN precursors. measure the energy-dependent delayed-neutron group parameters for 239Pu for incident-neutron energies It is shown, that there is a significant energy ranging from 0.37 up to 5 MeV. dependence of DN group parameters in the energy range of primary neutrons from thermal up to 5 MeV, Methods which appears as 10 % decrease of an average half- life of DN precursors in the studied energy range of The method used to perform these measurements primary neutrons. It is obvious, that such modification was based on a periodic irradiation technique in which of DN group parameters must be taken into account in a sample of 2}9Pu was irradiated by mono-energetic reactor calculations. The data obtained in present work neutrons from T(p,n) and D(d,n) reactions produced can be used in deriving of group constants for reactors at the KG-2.5 electrostatic accelerator at IPPE with an intermediate neutron spectrum, which differs (Obninsk). The delayed neutron decay curves were from both thermal and fast. measured after the irradiation of 239Pu sample. The

FUEL CYCLE CHALLENGES I 2/3 SK01K0176 TRIGA FUEL ELEMENT BURNUP DETERMINATION BY MEASUREMENT AND CALCULATION

TomazZagar, Matjaz Ravnik, Andreja Persie, Robert Jeraj

Institute Joief Stefan, Jamova 39, 1000 Ljubljana, Slovenija

Introduction The accuracy of the calculations was estimated also by comparing the calculated results to the results This analysis of fuel element burnup was done for of the experiments. Fuel element burnup was measured the spent fuel element shipment to the United States by the reactivity method [4], The reactivity method is for final disposal [1] and within the scope of the TRIGA based on the assumption that the reactivity worth of benchmark project [2] accepted by the International the fuel element is a known function of burnup. For Criticality Safety Benchmark Evaluation Project the measurements we established a practically critical (ICSBEP) working group. However, the results are core configuration with fuel elements to be measured acceptable to all TRIGA fuel element burnup at the selected measurement position. The digital calculations. Since experimental methods, even reactivity meter was used to measure core excess nondestructive gamma scanning, are normally too reactivity. Comparison shows agreement within ±1 complicated for determining burnup of large numbers burned 2JSU burnup for most of the fuel elements. of fuel elements, reactor calculations are the most Discrepancies for some fuel elements, which were common and practical method for fuel element burnup positioned near transient control rod air-fol lower, could determination. Good knowledge of calculated fuel be explained with poor fuel element surrounding element burnup accuracy is also needed for correct descriptions in the unit-cell calculation. interpretations of all research reactor experiments, when spent fuel elements are present in the reactor core. Conclusions Methods The results show that fuel element burnup estimates can be significantly wrong if one uses The most commonly used methods for non- calculation models that are too simplified. The error destructive fuel element burn-up determination at can be over 50% in mixed cores containing low and research reactors are reactor calculations, reactivity highly enriched fuel elements. The effect of new measurements and gamma ray spectrometry. Reactor WIMSD library on burnup calculations is not so calculations are normally used for continual monitoring pronounced. Results of experiments showed that one of the fuel element burn-up history because they are of the largest uncertainties of the calculated burnup easy to apply and do not influence reactor operation. arise from inaccurately known fresh fuel composition However, their accuracy was sometimes questionable, and from rather large inaccuracies in reactor thermal particularly in cases of complicated mixed cores and power calibration. mixed rings, when a one dimensional model was used.

To estimate the accuracy of the fuel element burnup References calculation different factors influencing the calculation were studied. To cover different aspects of burnup [1] T. ZAGAR, «Fuel Elements for Shipment - Basic calculations, two in-house developed computer codes Data», IJS-DP-8094, Ljubljana, Institute Jo2ef were used in calculations. The first (TRIGAP) is based Stefan (1999). on a one-dimensional two-group diffusion [2] M. RAVNIK, R. JERAJ, «TR1GA Mark II Bench- approximation, and the second (TRIG LAV) is based mark Critical Experiments)), International Hand- on a two-dimensional four-group diffusion equation. book of Evaluated Critical Safety Benchmark Ex- Both codes use WIMSD program with different periments, IEU-COMP-THERM-003, OECDNEA libraries forunit-cell cross section data calculation. The (1999). burnup accumulated during the operating history of the [3] M. RAVNIK, T. ZAGAR, A. PERSlC, «Fuel Ele- TRIGA reactor at «Jonef Stefan» Institute was ment Burnup Determination in Mixed TRIGA Core calculated for all fuel elements [3]. Elements used in Using Reactor Calculations)), Nuclear Technology the core during this period were standard SS 8.5% fuel 128,1,35(1999). elements, standard SS 12% fuel elements and highly [4] M. RAVNIK, M. STREBL, H. BUCK, A. TRKOV, enriched FLIP fuel elements. During the considerable I. MELE, ((Determination of the Burn-Up of period of operational history, FLIP and standard fuel TRIGA Fuel Elements by Calculation and Reac- elements were used simultaneously in mixed cores. tivity Experiments)), Kerntechnik 57,5,291 (1992).

2741 POSe PRESENTATIONS SK01K0177 RADIOACTIVE WASTE TREATMENT IN SLOVAK REPUBLIK

Pavol Dubovsky

Head of Primary Circuit Department, NPP V-2, Jaslovske Bohunice, Slovak Republic

The Slovak Electric pic. - SE a.s. ... . • Nuclear power plants 51,7% The joint-stock company SEa.s. is the major entity * Eossil-fuel power plants 27,9% in Slovakia involved in generation and distribution of • Hydro power plants 20,4 %

electric power. It is providing about 70% of all electric There are t^ nuclear facilities in Slovakia: power consumption. • Jaslovske Bohunice Within SE a.s. the electric power generation is # Mochovc provided with three types of power plants:

FUEL CYCLE CHALLENGES 1275 SK01K0178 CURRENT RADIOACTIVE WASTE UTILIZATION AT PA "MAYAK"

Alexander Merkushkin

Ozyorsk Technological Institute of Moscow Physical Engineering Institute, Russia The Production Association "Mayak" (also known resulting solution is then concentrated by an evaporating as Chemical Plant "Mayak" in the early days) was apparatus, which results in salt concentrations of established 40 years ago for its work on military approximately 350-370 g/L. This product is then plutonium and utilization of nuclear materials. Later injected into an electric furnace (model EH-500) with "Mayak" established a Nuclear Fuel Cycle CNFC) for displacement of 500 liters per hour. The duration of the civic production, including manufacture and presence of the glass in the seethe zone is about 79-80 preparation of nuclear emitters. In 1-955, the Nuclear hours. Nascent glass-heaps are ladled into container Isotopes Plant went into production, and in 1976, the cans with volumetric capacity equal to 200 liters. Each Conversion Plant came into existence to process three-can unit is packed into special boxes and placed nuclear fuel from atomic stations, research reactors, in temporary storage. The specific activity of glass-heaps and propulsion reactors. is 2500 Ci/L; the specific activity of the nascent glass is The military program at PA "Mayak" generated 200-600 Ci/L. The nascent condensates are then about 19,000 cubic meters of High Activity Waste combined with Medium Activity Waste (MAW). Today, (HAW), including suspensions with a total activity of all HAW is processed according to this technology. At about 135 million Ci. The chemical and radiochemical the PA "Mayak," fractionation of HAW to separate Sr composition of these suspensions are very complex and Cs from the waste using an extraction system based because of different types of primary wastes (hydrates, on chloride decorbollide of cobalt has been created and sulphides and ferrocyanide dregs). A portion of the this technology is continuing to improve. waste (about 8000 cubic meters) from extractive In the near future, the creation of the IMCC utilization of Warm Emitting Units is stored in tanks (inductive melter with cold crucible) line is expected as active nitrogenous solutions with total activity of to be complete. Also, the medium-level activity liquid about 200 million Curies. Every year at the wastes are removed mostly into opened ponds. Radiochemical Plant, approximately 2000-3000 cubic Since 1951, there was about 2.5 million cubic meters of HAW, estimated total activity around 100 meters of waste that was routed to Karachay for deposal million Ci, is processed. Accumulated high-active with an accumulated activity of about 120 million nitrogeneous solutions are stored in special cylindrical Curies. Historically, the Karachay repository has tanks made of rust-proof steel. All tanks have an received approximately 20,000 cubic meters of waste autonomous cooling system and they are continually each year with a characteristic specific activity of about monitored for the accumulation of content-generated 0.01-0.02 Ci/1. After its creation in 1970, an methane and hydrogen gases. • Records are kept on evaporation line in pond # 17 has accepted tritium accumulation of nuclear materials; the HAW units condensate. The total volume of this condensate is include a complex gas cleaning system. about 800,000 cubic meters with specific activity of 1 Liquid suspensions are stored in 20 tanks that have microCi/1 for tritium. gauges to monitor solution levels, pressure, temperature Presently, medium-activity liquid waste is typically and gas emissions, including an air ventilation system composed of drainage-desorption solutions, solutions to dilute and remove accumulated radiolithic methane from disassembled equipment, solutions after and hydrogen gases. extractions, and finally solutions after deactivation of The Principal Technological Scheme for Handling special containers. Typically, the specific activity of of Liquid Radioactive Waste of Any Activity Level this waste is about 0.02 Ci/1, and the average salt was developed according to the concept of cyclic water- concentration is approximately 12-15 g/1, including supply and safe-handling of radioactive waste on the condensates from cleaning of special gases. PA "Mayak". This scheme consists of three The total annual volume of nascent waste is about technological strategies that are applicable to solutions 16-20 thousands cubic meters (less than 1 million Ci). of high-, medium-, and low-level activity. Based on The first line of a plant to process MAW is currently radiologic safety regulations regarding waste storage, under construction at Karachay. This plant will contain phosphate glass is used as a matrix for high-activity units that will prepare, concentrate, and bitumenize waste. For low activity waste, matrices of bitumen and solutions, including a storage facility to contain and cement are used. The phosphate glass scheme utilizes keep accumulated products of bitumen shape. In the a glass furnace that is electrically operated. A primary future, this piece-meal conversion from bitumen solution of HAW to be processed is held in an technology to the processes that consume more intermediate apparatus and a specified quantity of invariable regional raw materials (such as clay, phosphoric acid is injected (the flux process). The metallurgical cylinders) is projected. 27B} POSTES PRESENTATIONS SK01K0179

THE METHODS OF CONVERSION OF RADIOACTIVE WASTE: A LOOK AT THE PAST

Denis Rezchikov

Ozyorsk Technological Institute of Moscow Physical Engineering Institute, Russia

The discovery by German scientists in 1938 of the system's control temperature and level of the waste ithe division of Uranium atoms and the release of large were exceeded. quantities of energy made it possible to create atomic Repair was impossible because radiation fields arms. Scientists emigrated to the USA to continue their were high. Contents of the volume were released in work. The Government of the USA developed a large the plume and spread into the atmosphere. Around 90 program for the creation of an atomic bomb. After percent of the radioactivity fell near the place of the bombardment of Hiroshima and Nagasaki in 1945, the burst. The rest of the 2 million curies was spread by Soviet Government organized production of atomic wind and appeared as the West Ural Trace. arms using Pu-239 in a short time. Thus, opposition between USA and USSR for domination in world It was very important to utilize solid waste (S W). policy began the cold war. Around 200 special places were made for storage of SW. High-activity S W was kept in reinforced-concrete In 1948, the Government of the USSR made containers. Middle-activity waste was put in a trench. provisions to establish an industrial complex to produce The total activity of SW is approximately 2 million Pu-239 in the Chelyabinsk region. PA "MAYAK" was Curies. part of this complex. It is now addressing radioactive waste disposition in peacetime. Much attention is now placed on pollution of the atmosphere. Many operations resulted in highly Making Pu-239 in metal form led to the formation radioactive aerosols in the air. There were no measures of large quantities of liquid radioactive waste. Since it for the protection of health. High stacks were one of was impossible to clean the water at that time, the main methods to reduce air contamination. In 1954, radioactive waste was put in the Techa River until 1951. questions about regulation of contamination to the The liquid waste contained a mixture of radioactive atmosphere were considered. The practice was stopped Sr, Cs, Nb, Ru and other elements. Sr-90 and Cs-137 when wind blew to the town. accounted for about a quarter1 of the radioactivity. Many people lived near the Techa and received large doses Therefore, methods for conversion of waste were of radiation. About 2.7 million Curies was put into the absent during the long history of the radiochemical river before this practice stopped and waste was then plant's work. There were two reasons for this. First, put into Lake Karachay. attention was concentrated on designing and building industrial plants to produce nuclear arms. Second, High-activity waste was placed in concrete and people did not know anything about the behavior of steel for protection. The volume was cooled with water. radioactive substances in the environment and its affects When cooling was stopped, the dry mixture heated to on health. 300-350 degrees and exploded. According to official information, the reason for the incident was because

FUEL CYC1E CHALLENGES 127/ SK01K0180 MANAGEMENT OF RADIOACTIVE WASTE IN NUCLEAR POWER: HANDLING OF IRRADIATED GRAPHITE FROM WATER-COOLED GRAPHITE REACTORS

S.S. Anfimov Research and Development Institute of Power Engineering, Moscow, Russia

One of the topical issues of nuclear power today in the ground and in other places not protected from is decommissioning of power, production and research underground water. nuclear reactors with expired lifetime. In addition to Cementing technology for processing lower technical, economic and social matters, activity radwaste is popular in the countries where dry decommissioning involves ecologically safe solids and beds free from underground water are confininement of radioactive waste (radwaste) that is available. generated, in particular, during dismantling of highly In recent years, the investigations conducted in radioactive reactor components. some countries were focused on exploration of As a result of decommissioning of water-cooled synthetic resins and polymers as possible binders for graphite-moderated reactors, a large amount of graphite radwaste. radwaste in the form of graphite stack fragments is Incineration of IG has an important advantage over generated (on average 1500-2000 tons per reactor). any other methods of graphite handling after That is why it is essentially important, although dismantling, because the amount of radwaste intended complex from the technical point of view, to develop for disposal will be significantly reduced as a result of advanced technologies based on up-to-date remotely- incineration. controlled systems for unmanned dismantling of the As commonly accepted in Russia and abroad, the graphite stack containing highly-active long-lived graphite stacks of gas- and water-cooled graphite- radionuclides and for conditioning of irradiated moderated reactors planned for decommissioning shall graphite (1G) for the purposes of transportation and be preliminarily cooled during the 50 to 100 years in subsequent long term and ecologically safe storage the reactor cavity, where it is isolated from the either on NPP sites or in special-purpose geological environment. Priority of this concept relegates to the repositories. background the efforts on development of cost-efficient The main characteristics critical for radiation and technologies for safe handling of IG. As a result, nuclear hazards of the graphite stack are as follows: technology of IG handling mastered in the industrial • the graphite stack is contaminated with nuclear fuel scale is not available in Russia today. that has gotten there as a result of the accidents; In terms of IG handling technology two lines were • the graphite mass is 992 tons, total activity -67104 identified: long-term storage of conditioned IG and IG Ci (at the time of unit shutdown); disposal by means of incineration. • the fuel mass in the reactor stack amounts to 100- The specific cost of graphite immobilization in a 140 kg, as estimated by IPPE and RDIPE, radiation-resistant polymeric matrix amounts to ~ 2600 respectively; USD per 1 t of graphite, whereas the specific cost of • y-radiation dose rate in the stack cells varies from immobilization in slag-stone containers with an 4 to 4300 R/h, with the prevailing values being in inorganic binder (cement) is ~ 1400 USD per 1 t of the range from 50 to 100 R/h. graphite. On the other hand, volume of conditioned IG radwaste subject for disposal, if obtained by means of Traditional methods of radwaste handling the first technology, is 2-2.5 times less than the volume of radwaste generated by means of the second In general, radwaste conditioning means waste technology. Besides, the duration of radwaste processing, processing to achieve pre-determined properties, including IG conditioning, package set transportation, parameters, characteristics or a new state. Typical and disposal is much shorter, if performed in accordance requirements to radwaste conditioning include with the first technology. Thus, the final choice of reduction in volume, dehumidification (dry material), graphite immobilization technology shall be made in the formation of solid radwaste (SRW) from liquid course of a feasibility study on decommissioning of radwaste, fractionation, i.e. separation of highly-active water-cooled graphite-moderated reactors, accounting waste from lower activity waste, etc.. for the strategy of futher graphite radwaste handling in Bituminization technology for radwaste processing the region of NPP locations. is used in many countries with highly developed nuclear It can be concluded from the above that advanced power. This technology is especially useful in such methods for graphite radwaste handling are available countries where dry territories suitable for radwaste today. Implementation of these methods will allow storage (e.g., deserts, dry sands, etc.) are unavailable. enhancement of environmental safety of nuclear power This technology uses bitumen blocks that can be stored that will benefit its progress in the future. 2781 POSTER PRESENTATIONS SK01K0181 CALCULATED INVESTIGATION OF ACTINIDE TRANSMUTATION IN THE BOR-60 REACTOR

/.Yu.Zhentkov, O.V.Ishunina, I.V.Yakovleva

State Scientific Center of the Russian Federation Research Institute of Atomic Reactors, Dimitrovgrad, Russia

In the course of reactor operation the formation ous minor-actinide compositions, and their quan- of fission products and accumulation of minor-actinides titative and qualitative behavior in time; (MA) and plutonium take place in the nuclear fuel. - carrying out experiments on the definition of the These materials define the radiation hazard to a great main neutron-physical characteristics (flux den- extent. One possible way to lower the activity of sity, neutron spectrum, reaction rate, etc.), change irradiated nuclear fuel is transmutation of long-lived of nuclear fuel composition (burn-up, isotopic ki- radioactive isotopes in the stable or short-lived forms, netics, etc.), separate fuel assemblies, fuel ele- that allows facilitation of the problem of the highly- ments, capsules and reactor as a whole, compari- active waste and improvement of the efficiency of son of experimental and calculated investigation nuclear fuel use at the expense of increased recycling results; and burnup. - testing and adaptation of the BOR-60 program and As numerous investigations demonstrate, the constant provision in order to improve the calcu- minor-actinide transmutation is possible in the thermal lation accuracy of neutron-physical characteristics. and fast power reactors. However, the most effective The BOR-60 reactor has gained significant actinide burnup can be achieved at a special actinide experience on the operation with power-grade burner-reactor only. These reactors provide a plutonium (the reactor was converted to MOX fuel in purposeful realization of optimum characteristics for 1980). The standard fuel assemblies are irradiated to actinide transmutations, which are "hard' spectrum and 20% of h.a. The separate fuel assemblies and fuel high density of neutron flux, a small factor of nuclear elements demonstrate a record value of fuel burnup up fuel breeding as well as stable neutron-physical to 35% h.a. At present, a great number of fuel assemblies features, that allows loading of a considerable amount involving weapon-grade plutonium was irradiated. The of actinides to the core without an important influence experimental fuel assemblies containing different types on the reactor (primarily on safety). One of the of nuclear fuel and capsules incorporating practically prospective actinide burner reactor type is the fast all minor-actinides were irradiated and their irradiation reactor with a "hard" spectrum and small breeding is being continued in the reactor. factor, which is the BOR-60. The calculated investigations demonstrate that loading up to 40% of A regularly performed comparison between minor-actinides to the BOR-60 reactor did not lead to calculated and experimental data concerning the nuclear the considerable change of neutron-physical fuel burnup in the fuel assemblies and minor-actinides characteristics. in the capsules demonstrated good agreement in the results. Numerous calculated investigations defining the The BOR-60 reactor can be considered as opti- possibility and efficiency of the BOR-60 operation as a mum for demonstration of burnup efficiency of minor- minor-actinide burner were carried out. The various actinides and plutonium, because the fraction of neu- versions of the BOR-60 core with minor-actinides in tron flux density at energies above 0,1 MeV makes-up the fuel and separate assemblies were considered. The 73-86%, neutron flux density - up to 3,55* 1019nrV, influence of minor-actinides on the main neutron- mean neutron energy in the core - 200-390 keV, nega- physical characteristics and reactor safety was estimated. tive effect of sodium void reactivity (- 0,052-0,068 Ak), The performed calculations show that the BOR- small breeding factor (<0,28), and experimental op- 60 reactor possesses a high efficiency of the minor- portunities of reactor facility. During 30 years of op- actinide and plutonium burn-up (up to 37 kg/(TW*h)) eration of the BOR-60 reactor significant calculated- that is comparable with properties of the actinide experimental experience has been gained. The BOR- burner- reactors under design. The BOR-60 reactor can 60 reactor provides to investigate and evaluate the provide a homogeneous minor-actinide loading (minor- possibility and efficiency of minor-actinide and pluto- actinide addition to the standard fuel) to the core and heterogeneous loading (as separate assemblies-targets nium transmutation: with a high minor-actinide fraction) to the first rows of - calculated investigation of critical mass, neutron- a radial blanket that allows the optimum usage of the physical characteristics of reactor fuel with vari- reactor and its characteristics. FUEL CYCLE CHAUENGES {270 SK01K0182

NO HJGH-Tfc ZATION OF HARD Fi

/, Yu. Pashkeev,A.KSeiiiii, E.N. Ily'm, KV. Gerasimova

South-Ural State University, Russia ••Abstract equipment-heat is allocated directly in briquette during burning. The special equipment for realization A chemical immobilization technique is tested by ofthe process is not required; only the expendable metal the method of self-propagating high-temperature container and protective layer between the briquette synthesis (SPHTS). The SPHTS application for the and the container are necessary. After the process ends, formation of multi-layer blocks containing simulators the block, ready for earth storage, is formed. of high-level wastes (HLW) is considered. The conditions of SPHTS realization and the structure of formed materials in Ti-B, Ti-Si, and Ti-C systems are ts anc fSCtfSSiCMIJ investigated. The influence of various solutions on the '. The technique of chemical immobilization was linear burning speed of powder mixes and structure of tested using simulators. The method provides the final cerate materials are investigated. v • ; : formation of the cermet matrix during a thermal interaction in: the: Al-Fe^Oj-CrjOj-CeOj system : ; : iletfiotls . : ;. . '. ; ' ' : '''.]':• . Containing CeO2 as simulator of radioactive elements. Tliecermet block consisted of Al2Op spinel solutions, :. SPHTS. is offered in addition to the existing high- and cerium alitminates. All. were formed as a result of temperature immobilization methods (sintering, SPHTS. The/ceramic immobilizers were received in vitrification, and induction melting), SPHTS permits the increase of the immobilization process adaptability to ;• '• Conditionsof SPHTS realization and structure of manufacturing. .SPHTS represents the exothermic cover cermet materials in Ti-B,Ti-Si* and Ti-C systems chemical interaction between simple substances or : areiinvestigated. During burning, chemically steady chemical connections in the condensed phase. The initial titanium borides, silicides, and carbides are formed; substahces mixed in me exothermic reaction burns itva .•r:;,;.,Theiinfjuencerof:several factors on burning level-by-level regime, A high temperature is created as a: parameters:was;investigated::: \.••••.:• . result of heat emission dutihgcnemicalreactiohsJ SPHjS 1. form: of'fuel: components (titanium, ferrotitanium receives; blocks consisting df the core (matrix Iwitli :tfie '^yhVi l chemically connected HLWcorripqnerits) surrounded by 2... ratio of fuel components in initial structures; a protective:cover that:does not :cqrii:ain;radioactive• i'3; quantity and type of metal (fenxjchromium X65, Ni, Cu, Cr, Fe) and nonmetal (Sip2, TiOr initial cover mix that is distributed to a. mix>containing . ;' AijOj) additives entered for regulation of composition, HLVV and they form an entire cenmet block, Such materials •: • structure, and properties of final composte material as SYNROC. can be used as an. immobilizerfqr t3ie first The introduction of metal and oxide additives in barrier ceramic, and pieces of initial oxide-mix .are ' thermal mixes results' in a decrease of burning-speed, uniformly distributed on the matrix volume, Oxides are but promotes crushing and more, unifonn distribution melted due to effect of the reaction heat. of phases in the new cermet structure. Cermet material composed of titanium borides (ceramic component), nonferrous metals alloys and steels (metal component) have the densest structures." Among samples with nbnmetal additives, the densest structures were obtained after TiO, addition.: : .

The technological efficiency of SPHTS application for cermet materials manufacturing is shown. On the basis of research results, mix compositions for block Figure 1, Multilayer blocks for IILAV: foi mation are offered. The samples, containing a core 1 - a container, 2 - matiix-immobihVer; with chemically connected HLW simulators, were made 3 - cermet cover in laboratory .conditions'.- Core consisted of components The considered method has a number of Oj-SiOj-CaO-NajO-CeOj,: and AI- advantages; The process does hot need special heating : systems.: •:. iilPiiiiiiiiiiiiis SK01K0183

SHADOW CORROSION EVALUATION IN THE STUDSVIK R2 REACTOR

Charlotta Sanders & Gunnar Lysell

Studsvik Nuclear AB, 611 82Nykoping, Sweden Abstract Zrrod Zr tube Post-irradiation examination has shown that Inconel increased corrosion occurs when zirconium alloys are in contact with or in proximity to other metallic objects. The observations indicate an influence of irradiation from the adjacent component as the enhanced corrosion occurs as a 'shadow' of the metallic object on the Figure 1. Outline of the INCA model. Everything zirconium surface. This phenomenon could ultimately unmarked in the pictures is water. limit the lifetime of certain zirconium alloy components In the reactor. The Studsvik R2 materials test reactor Discussion & Results has an IN-Core Autoclave (INCA) test facility especially designed for water chemistry and materials The electron depositions and the corresponding research. The INCA facility has been evaluated and currents calculated by MCNP show that the spacer found suitable for shadow corrosion studies. The R2 becomes negative charged relative to the cladding. This will enhance the corrosion of the cladding in close reactor core containing the INCA facility was modeled vicinity of the spacer, since the electric field between with the Monte Carlo N-Particle (MCNP) code in order spacer and cladding will promote the diffusion of to evaluate the electron deposition in various materials oxygen ions through the zirconium oxide to the and to develop a hypothesis of the shadow corrosion zirconium metal. At the same time, positive hydrogen mechanism. ions will be pulled away from the oxide/metal interface thus decreasing the normally observed level of Method hydrogen uptake. The estimated electron current density at the Zircaloy surface is of the same order of magnitude The INCA facility consists of two major parts: the as the corrosion current density for the observed oxide external water supply and analysis system as well as thickness assuming that all the current had to be provided the in-core rig. In the external INCA system, degassed by an external source (in this case the electrons were and de-ionized high purity water is pumped into the redistributed by the gamma irradiation). rig at various flow rates. Injections of additives and impurities, which is called the injection flow, can be Conclusions fed into the rig either before the water enters the rig or just above the in-core section of the rig. The specimen In-pile testing of spacer, shadow corrosion using rod is installed in the rig tube to serve as a carrier for the INCA facility in the Studsvik R2 reactor showed the test specimens. that the same type of shadow corrosion was obtained in the INCA rig as in a real BWR-reactor. To further The INCA facility was modeled in the MCNP code develop a hypothesis regarding currents involved in in order to evaluate the electron deposition and to the process of shadow corrosion MCNP modelling was analyze a hypothesis of the shadow corrosion. MCNP used to obtain rates of electron deposition in the is a three-dimensional, continuous-energy, coupled components involved. The MCNP model shows that neutron-photon-electron transport code. The principle electrons are redistributed within the INCA facility. behind MCNP is to track a particle from birth to death. Voltage differences are built up between the various MCNP was used to calculate the electron deposition components. The resulting polarity of the electric field arising from neutron-photon-electron transport. The between the cladding and the spacer is such as to case modeled represents the cladding specimen promote oxidation of the cladding and to reduce surrounded by the Inconel spacer cell in the INCA hydrogen uptake as has been observed in many practical facility and is shown in Figure I. cases of shadow corrosion on fuel rods.

FUEL CYCLE CHALLENGES 1281 SK01K0184 SYSTEM STUDY ON PARTITIONING AND TRANSMUTATION OF LONG-LIVED ISOTOPES

M. Szieberth Budapest University of Technology and Economics, Institute of Nuclear Techniques, Hungary Abstract the ORIGEN-S code. The radiotoxicity of the nuclides is defined in relation to the radiotoxicity of the uranium The management of the long-lived isotopes - ore that is consumed during the energy production. For transuraniums (TRUs) and fission products (FPs) -pro- an improved evaluation we introduce a global parameter duced by nuclear reactors is a problem that basically af- called residual hazard: fects the public acceptance of nuclear energy, and may influence the long-term hazard caused by energy produc- tion. Partitioning and transmutation (P&T) of spent fuel materials offer a suitable solution to this problem. After i I the nuclear community had realised this fact, the number of publications on this topic significantly increased but toxu there is a lack of studies that include the analysis of not where T is the time needed to reach the uranium ore's only one instrument but also the whole nuclear energy radiotoxicity level (tox^ and i is the index of the iso- system. However, from the viewpoint of P&T's imple- topes. The index of residual hazard could be obtained mentation a substantial question is the cooperation of from the above: plants optimised for energy production and others for partitioning or transmutation. In order to analyse this prob- lem, the schemes of different systems are framed and their [*]. mathematical models are worked out. The systems are evaluated through the long-term risks caused by the waste where Kmix=KJ0) for the once-through fuel cycle. deposited in final storage, which are described by a newly defined quantity, the residual hazard index. Results and Discussion Methods In the case of an appropriate power ratio of LWRs and TTRs we find that the system described above in It is obvious that the implementation of P&T can the first stage is capable of transmuting efficiently 2J7Np be reached in several steps. Therefore, we examine and "'•"'Am, but at the expense of the accumulation three basic types of systems* which represent three of the Cm isotopes. Because of the low thermal cross- possible stages of the implementation, The energy sections, the transmutation of Cm is possible only in production is based on conventional LWRs in each fast burners. The cooperation of thermal and fast stage and recycling of U and Pu is assumed. The first transmutational reactors allows stopping the stage contains additional thermal transmutational accumulation or even decreasing the amount of MAs. reactors (TTRs), which are slightly modified LWRs As the outcome of the evaluation of the systems with some special pins containing minor actinides based on long-term risks, we find a surprising result that (MAs) in their core. In the second stage, fast burner even the first stage represents a breakthrough in reactors (FBuRs) appear in order to improve decreasing the long-term risks compared to the once- transmutation rates. The third and final stage is the through fuel cycle. Moreover, the second and third stage double strata fuel cycle, where a symbiotic nuclear allows a reasonable storage time. Another result is the energy system is in connection with a P&T cycle confirmation of the facts that the transmutation of TRUs containing accelerator driven subcritical systems must be done by fissioning and that long-lived FPs' (ADSs), or both thermal and fast transmutational transmutation gains importance only if this has happened. reactors. The model that we worked out can describe the material flows in these systems, even under Conclusions changing potential of the NPPs. Since in this case the retardations in the fuel cycle (e.g. partitioning, fuel We can conclude the followings: fabrication, etc.) play a very important role, we obtain • The examined three stages mark out a possible differential equations with retarded arguments, which way to. the realisation of so-called "clean" nuclear can be solved by computational methods. In stead of energy. carrying out detailed computations on the different types • The mathematical model and computational meth- ofNPPs we gain the necessary data about them from other ods are suitable for the examination of these systems. studies and from a code that is able to calculate the • The examination of the long-term risks through the transmutation of actinides in different reactors' spectras. calculation of the long-lived isotopes' radiotoxicity In order to evaluate long-term risks, the accumulated is a convenient method for the evaluation and com- amount of waste is calculated and the decay of the long- parison of different P&T systems, especially after lived isotopes is computed in a one million year term by the introduction of the concept of residual hazard. 2BZ! POSTER PRESENTATIONS SK01K0185

NOVEL TECHNIQUE FOR MANIPULATING MOX FUEL PARTICLES USING RADIATION PRESSURE OF A LASER LIGHT Ryota OMORI Department of Quantum Engineering and Systems Science, University of Tokyo, Japan

1. Summary Principle A. MnO2 particles were used as a sample, because MnO2 is an oxide that is not transparent in the An efficient tool for manipulating MOX fuel particles visible region, as are UO2 and PuO2. Typical diameter remotely seems to have a lot of applications in the MOX of the MnO2; particles is several hundred nanometers. fuel fabrication plant. It will enable simplification of A few of the particles (~0.0 lg) were put in an area of equipment for handling MOX fuel particles and reduce -100 mm2; on a glass plate that was fixed on the stage secondary wastes and personnel exposures, which seems of the microscope. To break the strong bond between to be one of the important factors for public acceptance of the particles and the plate, the acoustic vibration was the MOX industry. On the other hand, laser manipulation applied by cylindrical piezoelectric ceramics. Applying is known as a powerful tool for manipulating remotely the vibration, a large number of particles left the plate microscopic objects including dielectric or metallic and were levitated. The incident laser light of P=~%0 particles, liquid droplets, and biological cells. It utilizes mW was irradiated on the levitated particles. The radiation pressure of a laser light and enables acceleration incident visible laser beam was irradiated for 10 seconds. in the direction of propagation of an incident beam. After the irradiation, the aggregation of collected We have continued theoretical and experimental MnO2 particles was formed around the beam axis on studies on laser manipulation of nuclear fuel particles, such the plate. It was not observed without the beam as UO2, PuO2 and ThO2. In this paper, we investigate the irradiation, which means that the particles were applicability of the collection of MOX particles floating in collected by radiation pressure. The collected particles air using radiation pressure of a laser light; some preliminary were not diffused due to the air convection resulting results are shown. This technique will be useful for removal from the light absorption of the particles. and confinement of MOX particles being transported by air current or dispersed in a cell box. First, we propose two 4.Numerical analysis types of principles for collecting MOX particles. Second, we show some experimental results. Third, we show We performed numerical calculations in order to numerical results of radiation pressure exerted on verify both the principles A and B. Among some models submicrometer-sized UO2 particles using Generalized for calculating radiation pressure exerted on particles, a Lorentz-Mie theory (GLMT). Because optical constants geometrical optics model has been widely used. of UO2 are similar to those of MOX fuel particles, it seems However, it is valid only when the particle diameter is that calculation results obtained hold for MOX fuel particles. much larger than the wavelength of an incident laser light. For submicrometer-sized particles, one can use 2. Principles of collecting MOX fuel the Generalized Lorentz-Mie theory (GLMT). Because particles using radiation pressure GLMT is based on wave optics, it is applicable in principle for spheres of any size. Principles A and B are Light carries momentum. So, when a light is incident validated by the calculation results obtained. However, on an object, its momentum changes due to the reflection, it should be noted here that, to manipulate the particles the refraction, and the that absorption results in forces in a wide area, an incident laser beam with high intensity acting on the object. This is called radiation pressure. We is required especially for Principle B. propose two principles for collecting MOX fuel particles. Principle A 5.Conclusions When a single ray is incident on highly reflective or absorbing particles, such as MOX fuel particles, they are In the present study, we proposed two principles of accelerated in the direction of the ray due to the repulsive collecting MOX fuel particles floating in air. While force. By irradiating a focused visible laser light at MOX Principle A is based on the acceleration of the MOX particles floating in air, it is, therefore, possible to push the particles due to the radiation pressure of a visible laser particles forward and to collect them in the focus region. light, Principle B is based on the gradient forces exerted Principle B on the particles when an infrared laser light is incident. It is also reported that, when a laser light is incident Principle A was experimentally verified using MnO2 on transparent particles, they are simultaneously accelerated particles of which optical properties are similar to those in the direction of the light and drawn into the beam axis. of UO2 and PuO2. Moreover, we calculated radiation The absorption coefficients of UO2 and PuO2 are very low pressure exerted on submicrometer-sized UO2 particles in the infrared region. Therefore, it seems possible to draw using generalized Lorentz-Mie theory. Numerical results floating MOX particles into the beam axis and to collect showed that UO2 particles were simultaneously pushed them at a point on the axis using a CO2 laser light. forward and pulled into the beam axis for both principles. It seems, therefore, possible to collect MOX particles 3.Experiment based on these principle because optical constants of PuO2 are similar to those of UO2, though an incident laser beam We performed experiments in order to verify with high intensity is required, especially for Principle B. FUEL CYCLE CHALLENGES 1283 SK01K0186 THE HIGH ORDER APPROXIMATION OF THREE-DIMENSIONAL NEUTRON DIFFUSION EQUATION BASED ON COMBINATION OF FINITE ELEMENTS AND FINITE DIFFERENCES SCHEMES INKORAT3DCODE

Tatiana V. Shemyakina, O. Zvenigorodskaya, R. M.Shagaliev

Russian federal nuclear center - VNIIEF, Russia

In most cases of reactor process simulations, the is not assured even with grid refinement. It results from neutron kinetics are taken into account by means of nonlinearity of such schemes. the diffusion approximation. In order to solve such In this paper we suggest the linear scheme for the problems it is necessary to take into account the three- three-dimensional neutron diffusion equation dimensional geometry of the reactor and the complex approximation. In this scheme we use the finite element composition of materials. It results in specification of method with biquadratic test functions for x,y growth on the stage of problem statement (for RBMK approximation. For z approximation of the equation reactor it is on the order of 80000 points). we use the finite difference method. Theoretically, such In general, reactor problems have solutions with a scheme provides convergence with a high order of significant horizontal gradients and smooth vertical accuracy: the third order or higher for x,y variables distributions. In order to simulate the RBMK-1000 and the second order for z variable. The scheme reactor by means of the second order approximation provides simulations with space grid refinement. Our with sufficient accuracy, it is necessary to divide the computational investigations showed that accuracy of grid in the horizontal plane (by a factor of 3 or more) calculations is acceptable even at the base (the coarsest) for each of the directions. This results in a large point grid. It provides significant reduction of calculation quantity, so simulation of the full reactor zone becomes time compared to simulations based on the second impossible. accuracy order schemes. In order to increase the efficiency of such problem • Our scheme was implemented in the KORAT 3D simulations, a three-dimensional neutron diffusion code. The RBMK reactor was simulated as the test equation approximation based on node type nonlinear problem with a different detailed order. The results of schemes is now used. Such schemes allow significant the computational investigation of convergence were increases in the calculation accuracy for coarse grids. compared with the results obtained by the second At the same time, the numerical solution convergence accuracy order scheme.

2M|PDSIR PRESENTATIONS SK01K0187 STUDY OF MATERIAL STABILITY SURROUNDING WITH LOESS-CLAY-LOAM ROCKS ON AN EXAMPLE OF «OLVIYA» MONUMENT OF UKRAINIAN NORTHERN PRICHERNOMORYA

B.Zlobenko, V.Kadoshnikov, VManichev, L.Demchenko, T.Golovko, V. Krapivina1

State Scientific Center of Environmental Radiogeochemistry, Kyiv, Ukraine; 'Institute of Archaeology of NAS of Ukraine, Kyiv Disposal of long lived radioactive wastes requires south-east part of the Upper State on the territory of an isolation for times that exceed reasonable the Rome times. expectations of maintaining institutional controls. Prior Study with the facilitation of physical-chemical to implementation of the disposal concept, a thorough and physical methods of research of chemical structure understanding of the long-term behavior of the barriers and surface layers of materials. The various physical including materials and geological environment of methods of samples investigation were the following: chosen must be reached. investigation with half-quantitative spectral analysis There are difficulties in showing compliance with performed on spectrograph (CTE-1); X-ray powder safety criteria over long time-scales because of the diffraction with DRON-UM-1 diffractometers use, X- increase with time of the uncertainty associated with ray fluorescent analysis (spectrometer VPA-30) and the the results of predictive models. x-ray microanalysis (JXA-5). Time, or rate of change, can be an important and Carried out microscopic researches have revealed useful indicator of the overall potential of a repository the specific forms of metals and glasses; corrosion and or its individual components to isolate and contain long dissolution of the surface that character is determined lived radioactive materials. For example, time can be by structure and physic-chemical conditions of its burial used as a direct indicator: place. - to show how a natural or engineered barrier per- The geochemical analysis of exhumed soil sample forms by observing or describing how long it takes represented by clay loam is carried. It demonstrates for an isotope to pass through or the isolation po- the pollution connected with ancient metallurgy. Is was tential of the natural system; shown that the contents of the mobile forms of copper - to describe the rate of change of important in cultural ancient layer, the OlViyan epoch, makes up pararmeters of the natural system (pH, Eh, hy- 5-6 mg/1. That is compared to the contents of mentioned draulic gradients, etc.) or to describe the natural copper forms in modern soils of Ukraine (0,18-6 mg/l). evolution of minerals in the repository (bentonite Study of structure superficial weathering of glass to illite clay forms). layer and its internal undestroyed part using method The anthropogenic analogues include all the man of the phase spectral analysis demonstrated that large made systems which may give useful information on cation (Na, Ca, Mg) flow out the glass first. Study of rates of the processes potentially significant in assessing chemical composition of both glass and its altered layer the performance of a disposal system or its components. showed the ions of alkali and alkaline-earth metals to Perhaps the most important research objects in this diffuse from the glass into the environment under the category are the archaeological artefacts. The disposal conditions, aluminum accumulated in the anthropogenic analogues can be also useful while hydrated gel-like layer of the glass. Left components, studying the behavior of potential barrier materials probably, crystallize, that is shown in irrization appear under repository conditions on a surface of glass. The subsequent consolidation of this layer leads to a layered structure formation, while In this work we have examined the archaeological the gel layer deeper alterations cause crystalline phases material exhumed from the archaeology monument formation. «Olviya». The ancient State Olviya is situated on the territory of Ukraine (Nikolaev region) and it is The work was carried out partly under considered to be an integral part of the world historical Research Contract No. 10748 with the IAEA in frame legacy. The samples of glasses were collected from P- of Co-ordinated Research Project on "Anthropogenic 25 excavation. The collected glasses and metals dated Analogues for Geological Disposal of High Level and by I-II centuries of our era. They are situated in the Long Lived Radioactive Waste".

FUEL CYCLE CIIALLFKGES | 285 SK01K0188 NUCLIDES 2000: AN ELECTRONIC CHART OF THE NUCLSDES

/. Galy and J. Magill European Commission, Joint Research Centre Institute for Transuranium Elements, Postfach 2340, D-76125 Karlsruhe, Germany

Abstract Radiotoxicity Data, and Derived Quantities (specific activity, isotopic powers, spontaneous fission rate, Radionuclides have many applications in specific gamma dose rate at lm, annual limits of intake). agriculture, medicine, industry and research. For basic Elements up to atomic number 114 are included information on such radioactive materials, the Chart although only default data have been used for elements of the Nuclides has proved to be an indispensable tool 113 and 114. For any existing element, isotopes can be for obtaining data on radionuclides and working out added and their data edited qualitatively decay schemes and reaction paths. These Charts are, however, of limited use when one requires quantitative information on the decaying nuclide and 3. Decay Calculations its daughters. This was the motivation for the development of the NUCLIDES 2000 software Through the «decay engine», one can investigate package. The radioactive decay data used in the foil decay scheme of any radionuclide to obtain the NUCLIDES 2000 is based on the Joint Evaluated File concentrations, masses, activities etc. accounting for (JEF) version 2.2. The present version of the program all the daughters, starting from an initial mass or activity contains decay data on approximately 2700 of the parent nuclide. radionuclides. 4. Searching the Database 1. Nuefide cx There are over 2600 nuclides in the database with In the Nuclide Explorer, a powerful navigational more than 70000 associated gamma energies. The da- interface allows fast access to the nuclides through a tabase can be searched to find nuclides with particular periodic table and Segra chart. There are 112 known characteristics. Comprehensive Boolean search tasks chemical elements and recently evidence for the can be performed. Searches can also be made on the existence of element 114, 116 and 118 have been decay type i.e. (alpha, B-, fi-n etc). reported. Because of the large number of nuclides, navigation through a chart of the nuclides is more 5. Case Studies complicated that navigation through the periodic table of the elements. For this reason, the starting window To demonstrate the use of the CD, examples of shown in the Nuclide Explorer contains a periodic table. the radioactive decay of ^Co and 232U are given in the Once the element has been selected, a list of all isotopes Help files and in the User Guide. is given in the element list box. This particular way of arranging nuclide was first proposed by Segra. The colours refer to the fact that the nuclide is unstable (radioactive) and decays by a particular process or processes. The decay processes and their associated colours are shown at the bottom of fig. 1. In the present version of Nuclides 2000, the user can select the colour schemes used in the original Karlsruhe, Strasbourg, and General Electric Charts.

2. Data Sheets

The data on each nuclide is based on the JEF2.2 decay datafile. In each Data Sheet, the information Fig. 1. The navigational interface showing the given consists of Decay Data (half-life, atomic weight, location of the selected nuclide in the Segra number of decay modes, branching ratios, decay energy, chart (colour scheme selected is that used in daughter products), Mean Decay Energies (associated the General Electric chart) with the decay modes), Discrete Energies,

M | POSe PRESENTATIONS 6. Articles 7. Online Resources

In the Articles section, a selection of articles, on In the Online Resources the browser window opens various aspects of radionuclides, ranging from with theNuclides 2000 homepage. From an investigation radiocarbon dating to the formation of 44Ti in of internet related websites, documents have been supernova, are given. There are also articles covering classified under the following general headings: the history of radioactivity and radiochemistry and Glossary of Nuclear Science, Classical Scientific descriptions of all the chemical elements. The articles Papers, Historical, People, New Elements, Origin of database is fully indexed and can be searched for Nuclides, Introduction to Radiation and Radioactivity, keywords. Use of wildcards and jokers is allowed in Applications of Radionuclides, Archaeology, Radon, the keyword selection. The displayed topic contains Nuclear Data, Others, Organizations. the keyword highlighted wherever it appears in the text. Nuclides 2000 is available on CD for Windows 95, 98, NT operating systems.

FUEL CYCLE CHALLENGES 1287 SK01K0189 TAKING BURNUP CREDIT FOR INTERIM STORAGE AND TRANSPORTATION SYSTEM FOR BWR FUELS

Ken-ichi Yoshioka

TOSHIBA Nuclear Engineering Laboratory 4-1 Ukishima-cho Kawasaki-ku Kawasaki, 210-0862 JAPAN

In the back-end issues of nuclear fuel cycle, requirements for the above items 2 to 4. The selection of reprocessing or one-through is a big issue. measurement system is something similar to the well- For both of the cases, a reasonable interim storage and known FORK system, and consists of two types of transportation system is required. detectors, namely, a Cd-Te detector for gamma-ray spectrum measurement and a fission chamber for This study proposes an advanced practical neutron measurement. monitoring and evaluation system. The system features the followings: A Cd-Te detector is used for burnup measurement in which Csl37 to Csl34 ratio measurement method 1 .Storage racks and transportation casks taking credit is employed. The method is less sensitive to deviations for burnup. in assembly position than Csl37 measurement method. 2. A burnup estimation system using a compact moni- The fact is advantageous for a fuel storage pool in which tor with Cd-Te detectors and fission chambers. a fuel assembly is not easy to be precisely fixed. A 3.A neutron emission-rate evaluation methodology, CdTe detector have the following advantages: especially important for high burnup MOX fuels. a.Sufficient energy resolution for burnup measure- 4.A nuclear materials management system for safe- ments, e.g., Schottky type CdTe detector achieves guards. the resolution of 2% at 662keV gamma-ray emit- Current storage system and transport casks are ted from Csl37. designed on the basis of a fresh fuel assumption. The b.Compact system, CdTe has high efficiency for assumption is too conservative. Taking burnup credit gamma-ray because of the large atomic number, gives a reasonable design while keeping conservatism. which results in sufficient efficiency even with small crystal. Furthermore, a CdTe detector is In order to establish a reasonable burnup credit operated in room temperature without a liquid ni- design system, a calculation system has been developed trogen tank which a Hp-Ge detector requires. for determining isotope compositions, burnup, and criticality. The calculation system consists of some The application of a CdTe detector enables the modules such as TGBLA, ORIGEN, CITATION, measurement system smaller. MCNP and KENO. The TGBLA code is a fuel design Neutron measurement method is utilized for code for LWR fuels developed in TOSHIBA burnup and neutron emission rates measurements. Corporation. The code takes operational history such Neutron emission rates are quite important for spent as, power density, void fraction into account. This code MOX fuels. Neutron emission rates of spent MOX fuels is applied to the back-end issues for a more accurate are significantly higher than those of spent UO2 fuels, design of a storage and a transportation system. The therefore, a neutron shielding design is a more ORIGEN code is well-known one-point isotope depletion important problem for a transportation cask. code. In the calculation system, the code calculates isotope compositions using libraries generated from the TGBLA Neutron measurements are carried out with two code. The CITATION code, the MCNP code, and the fission chambers placed both the sides of an assembly. KENO code are three dimensional diffusion code, A fission chamber is not sensitive to gamma-rays, which continuous energy Monte Cairo code, discrete energy results in small system without shielding. The sum of Monte Cairo code, respectively. Those codes calculate k- counts from two chambers is less sensitive to deviations effective of the storage and transportation systems using in assembly position, which results in smaller isotope compositions generated from the ORIGEN code. uncertainty in measurement. The CITATION code and the KENO code are usually Nuclear materials management is carried out used for practical designs. The MCNP code is used for through burnup management. The amount of nuclear reference calculations because of the long calculation time, materials is estimated on the basis of burnup values, although the code gives the most accurate results among initial enrichments, and operational histories. Nuclear three codes. materials management is especially important for spent MOX fuels in view of safeguards. A compact measurement system for a fuel assembly has been being developed to meet Burnup, neutron emission rates, and the amount 2881 POSTER PHESEHTATIONS of nuclear materials are determined with measured intensities are quite different between fresh fuels and values and calculated values. irradiated fuels, therefore, only total gamma-ray measurement with ion chamber is carried out instead Burnup and neutron emission rate estimations of gamma-ray spectrum measurement. The design gives should be carried out at the time of shipping to more simple system instead of smaller merit of burnup transportation casks in a power plant. The measurement credit. The design is considered as a first step to system is placed on a wall of a fuel storage pool. introduce a burnup credit system. Shipping of spent fuels is determined by help from the results of the measurements . Some model calculations were carried out for the BWR fuels of the latest design called STEP-3 whose Nuclear materials managements should be assembly-averaged initial enrichment was 3.5% and periodically carried out in interim storage facilities. assembly-averaged burnup was 45G Wd/t. For a current Reliable methods and technique are required to satisfy transport cask which accommodates 38 assemblies, it transparency along with non-proliferation. was found that all assemblies could not be shipped to a The measurement system is made with suspension cask without taking credit for burnup. For the current at fuel transportation machine. A fuel assembly is raised storage rack, it was found that burnup credit enables a vertically from a storage position and neutron design without borated stainless steel. measurement is made. After the measurement, the fuel assembly is returned to the storage position, a set of The above-mentioned advanced storage and the nuclear material inventory is evaluated through the transportation system has many merit, yet, the system burnup value deduced from the measured value. is not sufficiently validated with measurements. For the validation of the calculation system and the As a simple case, a procedure taking credit for measurement system, sufficient post irradiated only one cycle burnup is also developed. In this case, experiments and benchmark calculations are needed measurement system consists of only a ion chamber with international-basis to meet material management for a fuel storage pool in a power plant. Gamma ray requirements.

FUEL CYCLE CHALLENGES I 2HB SK01K0190 RUSSIAN SYSTEM OF COMPUTERIZED ANALYSIS FOR LICENSING AT ATOMIC INDUSTRY (SCALA) AND ITS VALIDATION ON ICSBEP HANDBOOK DATA AND ON SOME BURNUP CALCULATIONS

Tatiana Ivanovo, Marc Nikolaev, Alexey Polyakov, Tatiana Saraeva, Alexandre Tsiboulia

Institute of Physics and Power Engineering (IPPE), Russian Federation

Abstract Conclusions

The System of Computerized Analysis for SCALA system allows one to predict the keff values Licensing at Atomic industry (SCALA) is a Russian of critical assemblies varying from a very hard spectrum analogue of the well-known SCALE system. For to the well thermalized, large solution assemblies with criticality evaluations the ABBN-93 system is used with an uncertainty, as a rule, not greater than 0.5% and TWODANT and with joined American KENO and better if a bias factor derived from the provided Russian MMK Monte-Carlo code MMKKENO. validation would be used. Using the same cross sections and input models, Conclusions based on the set of rather simple all these codes give results that coincide within the systems considered here were confirmed by comparison statistical uncertainties (for Monte-Carlo codes). with many more geometrically complicated benchmarks from the Handbook and other sources. Validation of criticality calculations using SCALA was performed using data presented in the International As for the burnup calculations, the comparisons Handbook of Evaluated Criticality Safety Benchmark show the good agreement of the SCALA results with Expesriments. other codes. However validation on experimental material is required. A new approach should be Another task of the work was to test the burnup considered for the calculations a burnable absorber. capability of SCALA system in complex geometry in compare with other codes. Benchmark models of VVER type reactor assemblies with UO2 and MOX fuel including the cases with burnable gadolinium absorbers'were calculated. KENO-VI and MMK codes were used for power distribution calculations, ORIGEN code was used for the isotopic kinetics calculations. This paper contains: - the results of statistical analysis of discrepancies in ke(fbetween calculated and experimental bench- mark-model and conclusions about uncertainties of criticality prediction for different types of mul- tiplying systems following from this analysis. - Comparison of Pin-by-pin power distributions and reactivity losses for VVER assembly benchmark model calculated by SCALA system and other codes.

2901 POSTEil PRESENTATIONS SK01K0191 FUTURE PERSPECTIVE OF THORIUM BASED NUCLEAR FUELS AND THORIUM POTENTIAL OF TURKEY

Turan UNAK* and Yeliz YILDIRIM

Ege University, Faculty of Science, Department of Chemistry, Division of Nuclear Chemistry, Bornova, Izmir 35100, Turkey *) Phone & fax : +90-232-388-8264; E-mail: [email protected]

Introduction technical studies on the use of thorium based nuclear fuels in commercial reactors were slowly progressed The discovery of the fission of uranium by Otto during the last 50 years. Nevertheless, some prototype Hahn at the end of 193 8 opened the gateway of nuclear reactors were operated to test the potential use of energy applications as a new and powerful energy thorium based nuclear fuels in some countries such as source. The first applications of this energy source were the USA, Germany, UK, Holland. Following the closing unfortunately shown as the atomic bombs. So, the first of cold war and the agreement on the non-proliferation nuclear energy applications were initiated by the use of nuclear weapons at the end of 1980"s, the excess of fissile U-235 and then continued being added an Pu-239 either in the nuclear weapons or in the stocks, artificial fissile material Pu-239. As a consequence of the increasing of the terrorist activities, and the global this initiative, today's nuclear technology has environmental problems have caused to focus the principally been based on the use of these fissile attention of nuclear community into the use of thorium materials. The existence of thorium in the nature and based nuclear fuels. In this context, India as a country its potential use in the nuclear technology were not having considerable thorium reserves and special unfortunately into account with a sufficient importance. experience on nuclear technology had earlier directed Today, the excess of Pu-239, the proliferation potential to their national attentions to the use of thorium fuels of nuclear weapons, and also the anxious of nuclear in their own reactors. This country is independently terrorism are very serious and basic factors for global continuing to these efforts. In parallel to India case, suffering. the special interests in the use of thorium fuels have been focused years by years since about last fifteen Global Distribution of Thorium and years in some other countries such as the Russian Uranium Federation, Japan, China, Holland, Canada, the USA, etc. The studies carried out on the thorium have clearly The global abundance of thorium is about three showed that the thorium based nuclear fuels have the folder high than that of uranium. This means that the potential easily use in most of reactor types actually isotopic natural abundance of Th-232 is approximately operated with the classical uranium based nuclear fuels 24,000 folder high than that of fissile isotope of without any considerable modification. It is also very uranium, U-235. The global distributions of thorium promising progress that a new thorium based nuclear and uranium reserves indicate that in general some fuel proposed by Radkowsky as being based on a «seed- developed countries such as the USA, Canada, blanket» fuel design is seeming to be an ideal approach Australia, France have considerable uranium reserves, to replace the uranium based nuclear fuels with that of Radkowsky thorium fuels. Starting from this new fuel and contrarily only some developing countries such as approach Radkowsy Thorium Power Corporation has Turkey, Brazil, India, Egypt have considerable thorium been established in the USA in 1992. This corporation reserves as being totally about 70 % of the global has eventually aimed to test and to use the Radkowsky reserve. Turkey is the first country having about 800 thorium fuels in some commercial reactor by 2005. kilo tones assured thorium reserve, and this corresponds to about 52 % of the total global reserve. Conclusion Thorium Based Nuclear Fuels Nowadays, the nuclear technology has been Th-232 is not a fissile material, but can finally be interrogated because of proliferation of nuclear transformed into a fissile U-233 nucleus following a weapons, long-lived wastes, technical problems of thermal neutron capture reaction. So, the natural Th- excess Pu-239 either in dismantled nuclear weapons 232 as a bred of U-233 has became an important nuclear or in stocks. On the other hand, the nuclear energy material having very large potential applications. While cannot be abandoned because of increasing global the first studies on the thorium fuel cycle was begun in energy demand year by year, and so, sustainable nuclear the USA as early as with the studies on uranium and energy production systems should be developed in long plutonium following the , the term. All technical parameters obtained from the studies f IIEl CYCLE CRAllENGES 1291 on thorium fuel cycle during the last 50 years indicate existence of thorium in the nature and its potential use that in the case of developing the technologies based in the nuclear technology were not unfortunately into on thorium fuel cycle systems, these serious problems account with a sufficient importance. The global will probably be resolved, and sustainable nuclear distributions of thorium and uranium reserves indicate energy production will be realized in the next new that in general some developed countries such as the century. This means that thorium will probably be a USA, Canada, Australia, France have considerable nuclear material much more valuable than uranium in uranium reserves, and contrarily only some developing the future. For this reason, all developing countries countries such as Turkey, Brazil, India, Egypt have having thorium reserves should direct their considerable thorium reserves. The studies carried out technological attentions to the evaluation of their on the thorium during the last 50 years have clearly national thorium deposits like in the case of India, and showed that the thorium based nuclear fuels have the cooperate each others in this field for combining their potential easily use in most of reactor types actually efforts. This can be realized under the coordination of operated with the classical uranium based nuclear fuels a technology leader country. Finally, thorium is shown without any considerable modification. In the case of in the horizon of the 21 st century as a brighten star of the use of thorium based nuclear fuels in future nuclear new nuclear technology era. energy production systems, the serious problems such as the excess of Pu-239, the proliferation potential of Summary nuclear weapons, and also the anxious of nuclear terrorism will probably be resolved, and sustainable Today's nuclear technology has principally been nuclear energy production will be realized in the next based on the use of fissile U-235 and Pu-239. he new century.

2B2 I POSTER PRESEKTATIDNS