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Methodology for Source Term Analysis of a Molten Salt Reactor

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Methodology for Source Term Analysis of a Molten Salt Reactor University of Tennessee, Knoxville TRACE: Tennessee Research and Creative Exchange Masters Theses Graduate School 5-2020 Methodology for Source Term Analysis of a Molten Salt Reactor Sarah Creasman University of Tennessee, [email protected] Follow this and additional works at: https://trace.tennessee.edu/utk_gradthes Recommended Citation Creasman, Sarah, "Methodology for Source Term Analysis of a Molten Salt Reactor. " Master's Thesis, University of Tennessee, 2020. https://trace.tennessee.edu/utk_gradthes/5588 This Thesis is brought to you for free and open access by the Graduate School at TRACE: Tennessee Research and Creative Exchange. It has been accepted for inclusion in Masters Theses by an authorized administrator of TRACE: Tennessee Research and Creative Exchange. For more information, please contact [email protected]. To the Graduate Council: I am submitting herewith a thesis written by Sarah Creasman entitled "Methodology for Source Term Analysis of a Molten Salt Reactor." I have examined the final electronic copy of this thesis for form and content and recommend that it be accepted in partial fulfillment of the requirements for the degree of Master of Science, with a major in Nuclear Engineering. Thomas Harrison, Major Professor We have read this thesis and recommend its acceptance: Lawrence Heilbronn, Richard Wood, Laurance Miller Accepted for the Council: Dixie L. Thompson Vice Provost and Dean of the Graduate School (Original signatures are on file with official studentecor r ds.) Methodology for Source Term Analysis of a Molten Salt Reactor A Thesis Presented for the Master of Science Degree The University of Tennessee, Knoxville Sarah Elizabeth Creasman May 2020 © by Sarah Elizabeth Creasman, 2020 All Rights Reserved. ii Acknowledgments First, I would like to thank the Department of Energy Office of Nuclear Energy for funding this research and express my deepest appreciation to my committee. I cannot begin to express my thanks to my thesis advisor, Dr. T. Jay Harrison, for the encouragement, patience, valuable advice and unparalleled support. The completion of this thesis would not have been possible without Dr. Ben Betzler and his unparalleled knowledge of ChemTriton. I would also like to thank him for his patience and very helpful contributions. I would like to extend my sincere thanks to Dr. Lawrence Heilbronn who was instrumental in my decision to attend graduate school and helped me find a project that suits my interests. I very much appreciate Dr. Richard Wood who gave invaluable insight into advanced reactors. My sincerest thanks to Dr. Laurence Miller for his valuable advice and guidance. Thanks also to Dr. Scott Greenwood and Dr. Richard Mayes for their assistance and extensive knowledge. I would also like to thank Dr. Michael Muhlheim for his encouragement and support, someone who was instrumental in the timely completion of this thesis. I would also like to thank the members of the RNSD team at ORNL for their help and support. I cannot begin to express my thanks to Ben Creasman who offered invaluable advice and pushed me to attend graduate school; without him this thesis would not have been possible. My success would not have been possible without the support of my parents and grandparents: Cindy Rogers, Scotty Creasman, Ruth and Herman Creasman, and Jim and Kathryn Everette. Thank you for raising me to become a successful adult. I would also like to extend my deepest gratitude to my husband, Duncan Brocklehurst, for his unwavering support, encouragement, and patience throughout the duration of this project. I would like to thank Emily Stirrup and Michelle Holmes for their patience, support, and encouragement. Special thanks to all of my family, friends, and colleagues who supported me during this time. iii Abstract Interest in Molten Salt Reactors (MSRs) has increased in recent years due to the changing needs and goals of nuclear energy. The Aircraft Reactor Experiment (ARE) and Molten Salt Reactor Experiment (MSRE) set the stage early on for MSR research. There are many different options for MSRs including different salt compositions, neutron spectrum, fuel choice, and purposes. Thermophysical and thermochemical properties, such as melting point, density, and viscosity must be known for the salt composition chosen for the reactor. Some neutronics differences that set MSRs apart from Light Water Reactors (LWRs) are changes in treatment of delayed neutron precursors and point reactor kinetics. MSRs allow the ability to process the fuel during operation to remove or add desired nuclides. The Nuclear Regulatory Commission (NRC)'s experience is mainly focused on LWRs, but they have started exploring some advanced reactor source terms from a regulatory perspective. A Molten Salt Demonstration Reactor (MSDR) and MSRE were modeled in SCALE/TRITON and when compared to a Westinghouse Pressurized Water Reactor (PWR), the results showed that MSR source terms were less than 20% compared to PWR dependent of radionuclides. iv Table of Contents 1 Introduction1 1.1 Characteristics..................................5 1.2 Comparison with LWRs.............................7 1.3 Source Term....................................8 2 History 10 2.1 Aircraft Reactor Experiment........................... 10 2.2 Molten Salt Reactor Experiment......................... 11 2.3 Source Term History............................... 13 2.3.1 TID-14844................................. 13 2.3.2 WASH-1400................................ 14 2.3.3 NUREG-0956............................... 14 2.3.4 NUREG-1465............................... 14 2.3.5 Regulatory Guide 1.183......................... 15 2.3.6 NRC's Perspective on Mechanistic Source Term............ 15 3 Salt Composition 19 3.1 Fluorides vs. Chlorides vs. Nitrates....................... 20 3.2 Stoichiometry................................... 20 3.3 Neutron Spectrum................................ 23 3.4 Fuel Choice.................................... 24 3.5 Purpose (Breeder vs. Burner).......................... 25 3.6 Salt Composition Requirements......................... 25 v 4 Thermochemical and Thermophysical Properties 29 4.1 Use of Properties in Reactor Design....................... 30 4.2 Melting Point and Heat of Fusion at Boiling Point............... 30 4.3 Vapor Pressure and Vapor Species........................ 32 4.4 Density...................................... 34 4.5 Viscosity...................................... 35 4.6 Thermal Conductivity.............................. 36 4.7 Heat Capacity................................... 37 4.8 Material Solubility................................ 38 4.8.1 Soluble fission products.......................... 38 4.8.2 Noble and Semi Noble Fission Products................. 39 5 System Analysis 40 5.1 Delayed Neutron Precursors........................... 40 5.2 Point Reactor Kinetics.............................. 41 5.3 Reactivity Controls and Effects......................... 43 5.4 Neutron Flux................................... 44 5.5 Frequency Response and Reactor Stability................... 45 5.6 Thermochemistry and Corrosion......................... 46 5.7 Radiation Stability................................ 48 5.8 Modeling and Simulation............................. 49 5.8.1 Gaps in Modeling and Simulation.................... 50 6 Operational Fuel Management 52 6.1 Fuel Qualification................................. 53 6.2 Front End Processing............................... 54 6.3 During Operations................................ 55 6.3.1 Important Radioisotopes......................... 55 6.3.2 Reprocessing............................... 59 6.4 End of Life.................................... 62 6.5 Tritium...................................... 64 vi 7 Regulatory Considerations 65 7.1 NUREG-1537................................... 65 7.1.1 Chapter 4: "Reactor Description".................... 66 7.1.2 Chapter 5: "Reactor Coolant Systems"................. 66 7.1.3 Chapter 6: "Engineered Safety Features"................ 66 7.1.4 Chapter 9: "Auxiliary Systems"..................... 66 7.1.5 Chapter 11: "Radiation Protection Program and Waste Management" 67 7.1.6 Chapter 13 "Accident Analyses"..................... 67 7.2 NUREG-0800................................... 67 7.2.1 Chapter 4: "Reactor"........................... 68 7.2.2 Chapter 5: "Reactor Coolant System and Connected Systems".... 69 7.2.3 Chapter 6: "Engineered Safety Features"................ 69 7.2.4 Chapter 9: "Auxiliary Systems"..................... 70 7.2.5 Chapter 11: "Radioactive Waste Management"............ 70 8 Source Term Analysis 72 8.1 Intro to Analysis................................. 72 8.2 Modeling and Simulation............................. 73 8.3 Time Effect on Source Term........................... 74 8.3.1 Initial Loading.............................. 74 8.3.2 During Operations............................ 75 8.3.3 End of Life................................ 75 8.4 Other Considerations for Source Term Analysis................ 76 8.5 Possible Accidents................................ 77 8.6 Results....................................... 78 9 Conclusions & Future Work 86 Bibliography 88 Appendix 104 vii A Phase Diagrams 105 B Modeling Inputs 115 A MSDR Input................................... 115 B MSRE Input.................................... 119 Vita 124 viii List of Tables 1.1 Deployed Reactor Classes and Their Properties. Information from [4]....4 1.2 Generation IV Concepts and Their Properties [5], [6].............5 2.1 Main Properties of ARE[17]..........................
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