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Regulatory Technology Development Plan Sodium Fast Reactor Mechanistic Source Term – Metal Fuel Radionuclide Release

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Regulatory Technology Development Plan Sodium Fast Reactor Mechanistic Source Term – Metal Fuel Radionuclide Release ANL-ART-38 Regulatory Technology Development Plan Sodium Fast Reactor Mechanistic Source Term – Metal Fuel Radionuclide Release Nuclear Engineering Division About Argonne National Laboratory Argonne is a U.S. Department of Energy laboratory managed by UChicago Argonne, LLC under contract DE-AC02-06CH11357. The Laboratory’s main facility is outside Chicago, at 9700 South Cass Avenue, Argonne, Illinois 60439. For information about Argonne and its pioneering science and technology programs, see www.anl.gov. DOCUMENT AVAILABILITY Online Access: U.S. Department of Energy (DOE) reports produced after 1991 and a growing number of pre-1991 documents are available free via DOE’s SciTech Connect (http://www.osti.gov/scitech/) Reports not in digital format may be purchased by the public from the National Technical Information Service (NTIS): U.S. Department of Commerce National Technical Information Service 5301 Shawnee Rd Alexandria, VA 22312 www.ntis.gov Phone: (800) 553-NTIS (6847) or (703) 605-6000 Fax: (703) 605-6900 Email: [email protected] Reports not in digital format are available to DOE and DOE contractors from the Office of Scientific and Technical Information (OSTI): U.S. Department of Energy Office of Scientific and Technical Information P. O . B o x 6 2 Oak Ridge, TN 37831-0062 www.osti.gov Phone: (865) 576-8401 Fax: (865) 576-5728 Email: [email protected] Disclaimer This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor UChicago Argonne, LLC, nor any of their employees or officers, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of document authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof, Argonne National Laboratory, or UChicago Argonne, LLC. ANL-ART-38 Regulatory Technology Development Plan Sodium Fast Reactor Mechanistic Source Term – Metal Fuel Radionuclide Release Prepared by David Grabaskas, Matthew Bucknor, James Jerden Nuclear Engineering Division, Argonne National Laboratory February, 2016 Acknowledgements The authors would like to express their appreciation to the members of the Regulatory Technology Development Plan (RTDP) project at Idaho National Laboratory for their assistance with the project development and report creation. Wayne Moe Jim Kinsey Mark Holbrook The authors would also like to thank the technical reviewers of the report. Thomas King (Ret., Nuclear Regulatory Commission) Leonard Leibowitz (Senior Chemist, Argonne National Laboratory) Candido Pereira (Principal Chemical Engineer, Argonne National Laboratory) The authors would like to thank those technical experts who provided insight and assistance throughout the course of the project. Steve Hayes Steve Frank Francine Rice Mark Williamson Walid Mohamed Aydin Karahan Gerard Hofman i Executive Summary The development of an accurate and defensible mechanistic source term will be vital for the future licensing efforts of metal fuel, pool-type sodium fast reactors. To assist in the creation of a comprehensive mechanistic source term, the current effort sought to estimate the release fraction of radionuclides from metal fuel pins to the primary sodium coolant during fuel pin failures at a variety of temperature conditions. These release estimates were based on the findings of an extensive literature search, which reviewed past experimentation and reactor fuel damage accidents. Data sources for each radionuclide of interest were reviewed to establish release fractions, along with possible release dependencies, and the corresponding uncertainty levels. Although the current knowledgebase is substantial, and radionuclide release fractions were established for the elements deemed important for the determination of offsite consequences following a reactor accident, gaps were found pertaining to several radionuclides. First, there is uncertainty regarding the transport behavior of several radionuclides (iodine, barium, strontium, tellurium, and europium) during metal fuel irradiation to high burnup levels. The migration of these radionuclides within the fuel matrix and bond sodium region can greatly affect their release during pin failure incidents. Post-irradiation examination of existing high burnup metal fuel can likely resolve this knowledge gap. Second, data regarding the radionuclide release from molten high burnup metal fuel in sodium is sparse, which makes the assessment of radionuclide release from fuel melting accidents at high fuel burnup levels difficult. This gap could be addressed through fuel melting experimentation with samples from the existing high burnup metal fuel inventory. ii Table of Contents 1 INTRODUCTION .................................................................................................................................................. 1 1.1 SCOPE .............................................................................................................................................................................. 2 1.2 METHODOLOGY .............................................................................................................................................................. 4 2 LITERATURE REVIEW SUMMARY ...................................................................................................................... 5 2.1 EXPERIMENTS AND REPROCESSING ........................................................................................................................... 5 2.1.1 MELT REFINING EXPERIENCE ........................................................................................................................................................................ 5 2.1.1.1 PRE-OPERATIONAL FUEL MELT TESTS ........................................................................................................................................................................ 6 2.1.1.2 FCF MELT REFINING EXPERIENCE ............................................................................................................................................................................... 9 2.1.2 FUEL MELT EXPERIMENTS ............................................................................................................................................................................. 9 2.1.2.1 OAK RIDGE FUEL MELT TESTING .................................................................................................................................................................................. 9 2.1.2.2 HANFORD FUEL MELT TESTING .................................................................................................................................................................................. 11 2.1.2.3 ATOMICS INTERNATIONAL FUEL MELT TESTING ..................................................................................................................................................... 12 2.1.3 CLADDING BREACH EXPERIMENTS AT EBR-II ......................................................................................................................................... 12 2.1.3.1 LEAKER EXPERIMENTS .................................................................................................................................................................................................. 13 2.1.3.2 RUN-BEYOND-CLADDING-BREACH TESTS ................................................................................................................................................................ 13 2.1.4 TREAT M-SERIES ....................................................................................................................................................................................... 14 2.1.5 METAL FUEL POST-IRRADIATION EXAMINATION (PIE) ......................................................................................................................... 16 2.1.5.1 FISSION GAS RELEASE EXPERIMENTS ......................................................................................................................................................................... 16 2.1.5.2 EBR-II PIE ...................................................................................................................................................................................................................... 17 2.1.5.3 FFTF PIE ........................................................................................................................................................................................................................ 22 2.1.5.4 METAPHIX PIE ............................................................................................................................................................................................................ 23 2.1.5.5 AFC SERIES PIE ............................................................................................................................................................................................................
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