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Regulatory Technology Development Plan Sodium Fast Reactor Mechanistic Source Term Development

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Regulatory Technology Development Plan Sodium Fast Reactor Mechanistic Source Term Development ANL$ART$'3' Regulatory Technology Development Plan Sodium Fast Reactor Mechanistic Source Term Development 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 Alexandra, 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. Box 62 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$ART$'3' Regulatory Technology Development Plan Sodium Fast Reactor Mechanistic Source Term Development Prepared by David Grabaskas, Acacia Brunett, Matthew Bucknor, James Sienicki, Tanju Sofu Nuclear Engineering Division, Argonne National Laboratory February 28, 2015 RTDP%–%Sodium%Fast%Reactor% 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 (biographies are provided in Appendix B). Mark Cunningham (Nuclear Regulatory Commission ret.) Richard Denning (Battelle/Ohio State University ret.) The authors would like to thank our industry contacts for their involvement in the formulation of the project and input on the report. The perspective of industry was vital to the project and their time and effort were greatly appreciated. Advanced Reactor Concepts, LLC GE Hitachi TerraPower Toshiba . ! i% % % ! RTDP%4%%Sodium%Fast%Reactor% Executive.Summary. Construction and operation of a nuclear power installation in the U.S. requires licensing by the U.S. Nuclear Regulatory Commission (NRC). A vital part of this licensing process and integrated safety assessment entails the analysis of a source term (or source terms) that represents the release of radionuclides during normal operation and accident sequences. Historically, nuclear plant source term analyses have utilized deterministic, bounding assessments of the radionuclides released to the environment. Significant advancements in technical capabilities and the knowledge state have enabled the development of more realistic analyses such that a mechanistic source term (MST) assessment is now expected to be a requirement of advanced reactor licensing. This report focuses on the state of development of an MST for a sodium fast reactor (SFR), with the intent of aiding in the process of MST definition by qualitatively identifying and characterizing the major sources and transport processes of radionuclides. Due to common design characteristics among current U.S. SFR vendor designs, a metal-fuel, pool-type SFR has been selected as the reference design for this work, with all phenomenological discussions geared toward this specific reactor configuration. This works also aims to identify the key gaps and uncertainties in the current knowledge state that must be addressed for SFR MST development. It is anticipated that this knowledge state assessment can enable the coordination of technology and analysis tool development discussions such that any knowledge gaps may be addressed. Sources of radionuclides considered in this report include releases originating both in-vessel and ex-vessel, including in-core fuel, primary sodium and cover gas cleanup systems, and spent fuel movement and handling. Transport phenomena affecting various release groups are identified and qualitatively discussed, including fuel pin and primary coolant retention, and behavior in the cover gas and containment. Radionuclides released from a primary sodium fire are also considered as potential sources. Any available experimental data and pertinent results relevant to the aforementioned phenomena are discussed, and operating incidents at domestically operated facilities are also examined. Considering the extensive range of phenomena affecting the release of radionuclides, the existing state of knowledge generally appears to be substantial, and may be sufficient in most areas. For core damage accidents, high retention rates should be expected within the fuel matrix and primary sodium coolant for all radionuclides other than the noble gases. These factors greatly reduce the magnitude of possible radionuclide release to the environment. Several possible gaps within the knowledgebase were identified during this effort. First, there are uncertainties with regard to radionuclide release from metal fuel in the molten state. Another knowledge gap appears in the available thermodynamic data regarding the behavior of lanthanides and actinides in liquid sodium. While not necessarily a phenomenological knowledge gap, a determination of the data requirements for MST development should be formally made prior to the expenditure of significant future research efforts. That is, if additional experimentation is performed in support of MST development, it is important to identify the proper quality assurance requirements for licensing. ii! . ' RTDP%–%Sodium%Fast%Reactor% Table.of.Contents. 1% INTRODUCTION......................................................................................................................................1% 2% BACKGROUND........................................................................................................................................3% 2.1% OBJECTIVES%AND%SCOPE%OF%CURRENT%WORK%..........................................................................................................%3% 2.2% SOURCE%TERM%HISTORY%...............................................................................................................................................%7% 2.2.1% CURRENT%REGULATORY%REQUIREMENTS%......................................................................................................................................................%7% 2.2.2% REGULATORY%HISTORY%OF%MST%DEVELOPMENT%......................................................................................................................................%12% 2.3% SODIUM%FAST%REACTOR%SOURCE%TERM%EXPERIENCE%.........................................................................................%17% 2.3.1% CLINCH%RIVER%BREEDER%REACTOR%............................................................................................................................................................%18% 2.3.2% THE%ALMR%PROGRAM:%PRISM%AND%SAFR%.............................................................................................................................................%20% 2.4% CURRENT%SFR%VENDOR%DESIGNS%AND%REFERENCE%DESIGN%.............................................................................%22% 3% SFR.SOURCES.OF.RADIONUCLIDES.....................................................................................................26% 3.1% FUEL%..............................................................................................................................................................................%26% 3.1.1% TYPES%OF%FUEL%..............................................................................................................................................................................................%26% 3.1.2% CORE%DESIGN%.................................................................................................................................................................................................%29% 3.1.3% SPENT%FUEL%HANDLING%...............................................................................................................................................................................%32%
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