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Liquid Metal Cooled Reactors: Experience in Design and Operation

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Liquid Metal Cooled Reactors: Experience in Design and Operation IAEA-TECDOC-1569 Liquid Metal Cooled Reactors: Experience in Design and Operation December 2007 IAEA-TECDOC-1569 Liquid Metal Cooled Reactors: Experience in Design and Operation December 2007 The originating Sections of this publication in the IAEA were: INIS and Nuclear Knowledge Management and Nuclear Power Technology Development Sections International Atomic Energy Agency Wagramer Strasse 5 P.O. Box 100 A-1400 Vienna, Austria LIQUID METAL COOLED REACTORS: EXPERIENCE IN DESIGN AND OPERATION IAEA, VIENNA, 2007 IAEA-TECDOC-1569 ISBN 978–92–0–107907–7 ISSN 1011–4289 © IAEA, 2007 Printed by the IAEA in Austria December 2007 FOREWORD In 2002, within the framework of the Department of Nuclear Energy’s Technical Working Group on Fast Reactors (TWG-FR), and according to the expressed needs of the TWG-FR Member States to maintain and increase the present knowledge and expertise in fast reactor science and technology, the IAEA established its initiative seeking to establish a comprehensive, international inventory of fast reactor data and knowledge. More generally, at the IAEA meeting of senior officials convened to address issues of nuclear knowledge management underlying the safe and economic use of nuclear science and technology (Vienna, 17–19 June 2002), there was widespread agreement that, for sustainability reasons for fissile sources and waste management, long-term development of nuclear power as a part of the world’s future energy mix will require the fast reactor technology. Furthermore, given the decline in fast reactor development projects, data retrieval and knowledge preservation efforts in this area are of particular importance. This consensus concluded from the recognition of immediate need gave support to the IAEA initiative for fast reactor data and knowledge presevation. To implement the IAEA initiative, the scope of fast reactor knowledge preservation activities and a road map for implementation have been developed. The IAEA supports and coordinates data retrieval and interpretation efforts in the Member States joining the initiative and ensures the collaboration with other international organizations (mainly OECD/NEA) and eventually establishes and maintains a portal for accessing the fast reactor knowledge base. The IAEA assists Member State activities by providing an umbrella for information exchange and collaborative R&D to pool resources and expertise within the framework of the TWG-FR and the Agency’s International Nuclear Information System (INIS) and Nuclear Knowledge Management Section (NKMS). The IAEA collects and summarizes the scientific and technical information on key fast reactor technology aspects in an integrative sense useful to engineers, scientists, managers, university students and professors. This publication has been prepared to contribute toward the IAEA activity to preserve the knowledge gained in the liquid metal cooled fast reactor (LMFR) technology development. This technology development and experience include aspects addressing not only experimental and demonstration reactors, but also all activities from reactor construction to decommissioning. This publication provides a survey of worldwide experience gained over the past five decades in LMFR development, design, operation and decommissioning, which has been accumulated through the IAEA programmes carried out within the framework of the TWG-FR and the Agency’s INIS and NKMS. The IAEA appreciate the advice and support of the IAEA’s TWG-FR members in the preparation of the publication. The draft of the report, compiled by A. Rineiskii (consultant), in cooperation with W. Mandl, A. Badulescu and Y.I. Kim of the IAEA, has been reviewed by the TWG-FR Members of China, France, India, Japan, Republic of Korea and the Russian Federation. The work was guided by the IAEA officers A. Stanculescu and Y. Yanev and they are responsible for this publication. EDITORIAL NOTE The use of particular designations of countries or territories does not imply any judgement by the publisher, the IAEA, as to the legal status of such countries or territories, of their authorities and institutions or of the delimitation of their boundaries. The mention of names of specific companies or products (whether or not indicated as registered) does not imply any intention to infringe proprietary rights, nor should it be construed as an endorsement or recommendation on the part of the IAEA. CONTENTS INTRODUCTION...................................................................................................................... 1 1. LIQUID METAL COOLED FAST REACTOR DEVELOPMENT AND IAEA ACTIVITIES ................................................................................................ 2 1.1. Activities on fast reactors design and technology................................................... 2 1.2. The joint research activities on LMFR.................................................................... 6 1.2.1. Sodium void effect....................................................................................... 6 1.2.2. Intercomparison of LMFR seismic analysis codes: comparison of experimental results with computer prediction ............ 9 1.2.3. Sodium mixing problems........................................................................... 12 1.2.4. Core structural and fuel materials assuring high fuel burnup .................... 17 1.2.5. Concept of fuel resources and waste management .................................... 18 1.2.6. Core disruptive accident ............................................................................ 19 1.2.7. Acoustic signal processing for the detection of sodium boiling in reactor cores or leak detection and location in steam generators ....... 20 1.3. Activities on advanced fast LMFR and technology .............................................. 20 1.4. Safety research issues for advanced LMFR development..................................... 25 1.4.1. Progress made since the past...................................................................... 26 References ...................................................................................................................... 27 2. PROTOTYPE FAST REACTOR................................................................................... 29 2.1. Design features and review of operating history................................................... 29 2.2. Review of the PFR steam generator design concept and operating history .......... 32 2.2.1. PFR steam generator: choice of the design concept .................................. 34 2.2.2. Review of steam generators operating history........................................... 37 2.2.3. The under-sodium leak in PFR superheater............................................... 40 2.3. Effects of sodium aerosol deposition in LMFRs................................................... 43 2.4. The primary circuit oil spill................................................................................... 46 2.5. Cracks in the PFR air heat exchangers.................................................................. 50 2.6. Sodium mixing problems and flow and mechanically induced vibrations ........... 52 2.7. The effect of neutron-induced distortion of core components .............................. 52 2.8. Fuel development .................................................................................................. 55 Bibliography ................................................................................................................... 56 3. PHENIX AND SUPER-PHENIX REACTORS............................................................. 57 3.1. Phénix reactor........................................................................................................ 57 3.1.1. Commissioning and design features .......................................................... 57 3.1.2. Intermediate heat exchangers operating experience .................................. 64 3.1.3. Steam generator operating experience ....................................................... 66 3.1.4. Secondary circuit operating experience ..................................................... 72 3.1.5. Pumps operating experience ...................................................................... 73 3.1.6. Negative reactivity shutdowns................................................................... 74 3.1.7. Sodium aerosol deposit .............................................................................. 75 3.1.8. Sodium circuits, reactor and equipment inspection and renovation .......... 76 3.1.9. Protection for fires ..................................................................................... 79 3.1.10. Plant statistics and conclusions.................................................................. 80 3.2. Super-Phénix reactor............................................................................................. 83 3.2.1. Design features and commissioning history .............................................. 83 3.2.2. SPX steam generator: design and operating experience ............................ 88 3.2.3. Storage drum operating experience............................................................ 91 3.2.4. Primary sodium contamination .................................................................. 96 3.2.5. Argon leak from the sealing bell of the intermediate heat exchangers...... 98 Bibliography ................................................................................................................. 100 4. BN-600
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