REVISION 1 JANUARY 31, 2020 NRC Non-Light Water Reactor (Non- LWR) Vision and Strategy, Volume 3 – Computer Code Development Plans for Severe Accident Progression, Source Term, and Consequence Analysis TABLE OF CONTENTS 1. INTRODUCTION ...................................................................................................... 10 1.1. Regulatory Need for Source Term Analysis ...................................................... 12 1.2. Description of Computer Codes ........................................................................ 15 2. MELCOR Development Plans for non-LWRs ........................................................ 24 2.1. HTGR ............................................................................................................... 32 2.2. SFR .................................................................................................................. 41 2.3. MSR ................................................................................................................. 49 3. SCALE Development Plans for non-LWRs ........................................................... 55 3.1. HTGR/FHR ....................................................................................................... 64 3.2. SFR/HPR .......................................................................................................... 73 3.3. MSR ................................................................................................................. 78 4. MACCS Development Plans for non-LWRs .......................................................... 83 4.1. MACCS Development for Non-LWR Site- and Location-Related Issues ........... 88 4.2. MACCS Development for Non-LWR Design-Specific Issues ............................ 91 5. Concluding Remarks ............................................................................................. 95 REFERENCES ................................................................................................................ 96 Appendix A. MELCOR Modeling of HTGRs ........................................................... 103 Appendix B. MELCOR Modeling of SFRs.............................................................. 131 Appendix C. MELCOR Modeling of MSRs ............................................................. 155 Appendix D. SAS4A Computer Code .................................................................... 164 Appendix E. U.S. Sodium Experiments Reviewed by SNL .................................. 167 Appendix F. Example of MELCOR Application for HTGR by International Community 174 2 FIGURES Figure 1-1. NRC Evaluation Model for NGNP (from April 2011) ....................................... 11 Figure 1-2. Radionuclide transport paths for LWR designs. ............................................. 14 Figure 1-3. Role of Accident Progression, Source Term, and Consequence Analysis Computer Codes and Applicable Regulatory Requirements .......................... 15 Figure 2-1. Timeline of MELCOR Advanced Reactor Model Development. ...................... 24 Figure 2-2. MELCOR 2.2 Validation Cases. ..................................................................... 26 Figure 2-3. Radionuclide transport paths in HTGR designs. ............................................. 33 Figure 2-4. Proposed NRC Evaluation Model for HTGRs ................................................. 35 Figure 2-5. Radionuclide release paths for pool-type SFR designs. ................................. 42 Figure 2-6. Proposed NRC Evaluation Model for Sodium Fast Reactors .......................... 42 Figure 2-7. Radionuclide release paths for salt-cooled designs. ....................................... 49 Figure 2-8. Radionuclide release paths for salt-fueled reactor designs. ........................... 50 Figure 2-9. Proposed NRC Evaluation Model for Salt Cooled Reactor ............................. 51 Figure 2-10. Proposed NRC Evaluation Model for Salt Fueled Reactor ........................... 52 Figure 3-1. Integrated capabilities of modernized SCALE 6.2. ......................................... 56 Figure 3-2. SCALE Monte Carlo Model of HTR-10 Benchmark. ....................................... 65 Figure 3-3. SCALE/Shift model of EBR-II SFR (top left: radial view of core, top right: axial view of core, lower left: radial detail of fuel assembly, lower right: radial view of fuel pin). ........................................................................................................ 75 Figure 3-4. SCALE MSR Delayed Neutron Precursor Drift Modeling. ............................... 79 3 TABLES Table 1-1. Generic Listing of Non-LWR Designs. ............................................................. 23 Table 2-1. MELCOR Non-LWR Development Plan Start Dates. ....................................... 28 Table 2-2. Yearly Deliverables – MELCOR Development Plan ........................................ 32 Table 2-3. Input/Output for MELCOR in the HTGR EM .................................................... 36 Table 2-4. Key Accident Progression Phenomena for HTGRs. ........................................ 38 Table 2-5. Proposed MELCOR Assessment Matrix for HTGRs ........................................ 39 Table 2-6. MELCOR Maturity for HTGR Analysis ............................................................. 40 Table 2-7. Proposed Input/Output for MELCOR in the SFR EM ....................................... 43 Table 2-8. Key Accident Progression Phenomena for SFRs. ........................................... 45 Table 2-9. Proposed MELCOR Assessment Matrix for SFRs ........................................... 47 Table 2-10. MELCOR Maturity for SFR Analysis .............................................................. 48 Table 2-11. Proposed input/output table for MELCOR in the MSR EM ............................. 52 Table 2-12. Key Accident Progression Development Issues for MSRs ............................. 53 Table 2-13. Proposed MELCOR Assessment Matrix for MSRs ........................................ 54 Table 2-14. MELCOR Maturity for MSR Analysis ............................................................. 54 Table 3-1. Summary of major SCALE capabilities ............................................................ 56 Table 3-2. SCALE Non-LWR Development Plan Start Dates. .......................................... 60 Table 3-3. SCALE Maturity for HTGR Analysis ................................................................ 70 Table 3-4. SCALE Maturity for FHR Analysis ................................................................... 73 Table 3-5. SCALE Maturity for SFR/HPR Analysis ........................................................... 78 Table 3-6. SCALE Maturity for MSR Analysis .................................................................. 82 Table 4-1. MACCS Non-LWR Development Plan Start Dates. ......................................... 84 Table 4-2. MACCS Maturity for Non-LWR Analysis .......................................................... 87 4 EXECUTIVE SUMMARY This report summarizes proposed code development efforts to extend NRC’s modeling and simulation capabilities for accident progression, source term, and consequence analysis for non-LWR technologies. This report describes the different types of non-LWRs as well as the modeling gaps for NRC’s computer codes including MELCOR for accident progression and source term analysis, MACCS for consequence analysis, and SCALE for radionuclide inventories. Severe accident progression, source term, and consequence analysis are deeply embedded in the NRC’s regulatory policy and practices. The licensing process is based on the concept of defense-in-depth, in which power plant design, operation, siting, and emergency planning comprise independent layers of nuclear safety. This approach encourages nuclear plant designers to incorporate several lines of defense in order to maintain the effectiveness of physical barriers between radiation hazards and workers, members of the public, and the environment – for both normal operation and accident conditions. The various regulatory source terms, used in conjunction with the design basis accidents, establish and confirm the design basis of the nuclear facility, including items important to safety, ensuring that the plant design meets the safety and numerical radiological criteria set forth in the U.S. Code of Federal Regulations (CFR) (e.g., 10 CFR 100.11, “Determination of Exclusion Area, Low Population Zone, and Population Center Distance”; 10 CFR 50.67, “Accident Source Term”; 10 CFR 50.34(a)(1)(iv); General Design Criterion 19, “Control Room,” of Appendix A, “General Design Criteria for Nuclear Power Plants,” to 10 CFR Part 50, “Domestic Licensing of Production and Utilization Facilities”) and in subsequent staff guidance. General Design Criteria (GDCs) are applicable to light- water reactors (LWRs). Non-LWRs will have principal design criteria (PDCs) which may have a similar requirement. MELCOR is the state-of-the-art computer code developed by Sandia National Laboratories for NRC to perform nuclear reactor severe accident progression and source term analyses. MELCOR is a flexible, integrated computer code designed to characterize and track the evolution of severe accidents, and the transport of associated radionuclides within a confinement such as a containment or building. It is a knowledge repository comprised of hundreds of millions of dollars’ worth of experiments and model development, with particular focus on LWR phenomenology as well as extended