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(Non-LWR) Reactor Mechanistic Source Term Analysis

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(Non-LWR) Reactor Mechanistic Source Term Analysis ANL/NSE-20/39 SAND2021-3250 R Survey and Assessment of Computational Capabilities for Advanced (Non-LWR) Reactor Mechanistic Source Term Analysis Nuclear Science and Engineering Division, Argonne National Laboratory Nuclear Energy Safety Technologies Group, Sandia National Laboratories 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. About Sandia National Laboratories Sandia National Laboratories is a multimission laboratory managed and operated by National Technology and Engineering Solutions of Sandia, LLC, a wholly owned subsidiary of Honeywell International Inc., for the U.S. Department of Energy’s National Nuclear Security Administration under contract DE-NA0003525. 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 at OSTI.GOV (http://www.osti.gov/), a service of the US Dept. of Energy’s Office of Scientific and Technical Information. 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. 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/NSE-20/39 SAND2021-3250 R Survey and Assessment of Computational Capabilities for Advanced (Non-LWR) Reactor Mechanistic Source Term Analysis Prepared by Shayan Shahbazi, David Grabaskas Nuclear Science and Engineering Division, Argonne National Laboratory Sara Thomas Chemical & Fuel Cycle Technologies Division, Argonne National Laboratory Nathan Andrews, Andrew Clark, David Luxat, Michael Higgins, Mariah Smith Nuclear Energy Safety Technologies Group, Sandia National Laboratories September 2020 Acknowledgements This report is part of a larger collaboration between Argonne National Laboratory and Sandia National Laboratories to assist in the development of modern mechanistic source term modeling and simulation tools, as part of the U.S. Department of Energy (DOE) Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. The authors would like to thank Nathan Bixler, Brad Beeny, Daniel Clayton, Jennifer Leute, Raymond Fasano, Andrew Hahn, and Tina Nenoff from Sandia and Bo Feng, Tanju Sofu, Tom Fanning, Jim Jerden, Nathan Hoyt, Adam Nelson, Micheal Smith, Alisha Kasam-Griffith, and Scott Richards from Argonne for their support in this effort. Executive Summary A vital part of the licensing process for advanced (non-LWR) nuclear reactor developers in the United States is the assessment of the reactor’s source term, i.e., the potential release of radionuclides from the reactor system to the environment during normal operations and accident sequences. In comparison to source term assessments which follow a bounding approach with conservative assumptions, a mechanistic approach to modeling radionuclide transport, which realistically accounts for transport and retention phenomena, is expected to be used for advanced reactor systems. As the designs of advanced reactors increase in maturity and progress towards licensing, there is a need to advance modeling and simulation capabilities in analyzing the mechanistic source term (MST) of a prospective reactor concept. In the present work, a survey is provided of existing computational capabilities for the modeling of advanced reactors MSTs. The following reactors are considered: high temperature gas reactors (HTGR); molten salt reactors (MSR) which include salt-fueled reactors and fluoride salt-cooled high temperature reactors (FHR); and sodium- and lead-cooled fast reactors (SFR, LFR). A review of relevant codes which may be useful in providing information to MST analyses is also completed, including codes that have been used for source term analyses of LWRs, as well as those being developed for other aspects of advanced reactor system modeling such as reactor physics, thermal hydraulics, and chemistry. A discussion of MST modeling capabilities for each reactor type is provided with additional focus on important phenomena and functional requirements. Additionally, a comprehensive survey is provided of tools for consequence modeling such as atmospheric transport and dispersion (ATD). Based on the survey of available modeling and simulation capabilities, key gaps and uncertainties were identified to aid in future code development. In general, there has been extensive research in modeling the reactor physics and thermal hydraulics of advanced reactors. Tools for fuel depletion and fuel performance are relatively robust but gaps remain for proper validation of certain fuel types and geometries. In addition, high-fidelity fuel performance codes have generally not been applied for the assessment of radionuclide behavior in the context of source term analyses. Determination of radionuclide chemical form is important to multiple areas of MST analysis, but modeling tool and database development is ongoing. Significant knowledge concerning aerosol behavior is available from LWR research, but there are areas of uncertainty regarding specific radionuclide aerosols and their chemical forms for different advanced reactor types. In the area of ATD, there are remaining development items concerning near-field dispersion and the behavior of reactor-specific radionuclides in the environment. Lastly, one overarching advanced reactor MST area of consideration is the potential increased importance of radionuclide sources located outside of the traditional core volume. This can be the result of the utilization of a liquid fuel, or the use of auxiliary processing or purification systems. The findings of the study presented here will be utilized in FY21 to construct an advanced reactor MST modeling and simulation development pathway, which will outline a course to expedite gap resolution and capability expansion. Table of Contents 1 INTRODUCTION ........................................................................................................................................................ 1 1.1 PROJECT OBJECTIVES ................................................................................................................................................ 1 1.2 PROJECT METHODOLOGY ......................................................................................................................................... 2 1.3 REGULATORY CONSIDERATIONS ............................................................................................................................. 3 1.4 ASME/ANS NON-LWR PROBABILISTIC RISK ASSESSMENT STANDARD ..................................................... 6 2 MODELING AND SIMULATION OVERVIEW ............................................................................................................. 7 2.1 NEAMS CODES .......................................................................................................................................................... 8 2.2 SCALE SUITE .......................................................................................................................................................... 12 2.3 THERMOCHEMICAL MODELING ............................................................................................................................ 14 2.4 AEROSOL MODELING.............................................................................................................................................. 16 2.5 MELCOR ................................................................................................................................................................. 18 2.5.1 REPRESENTATION OF LWR IN-VESSEL PHENOMENA IN MELCOR ......................................................................................................... 18 2.5.2 REPRESENTATION OF LWR LOWER PLENUM IN-VESSEL PHENOMENA IN MELCOR ........................................................................
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