Steady-State and Transient Benchmarks of GOTHIC to the Molten Salt Reactor Experiment

References

• R. C. ROBERTSON, “MSRE Design and Operations Report, Part I, Description of Reactor Design,” ORNL-TM-728, Oak Ridge National Laboratory (1965).
   Google Scholar
• GOTHIC Thermal Hydraulic Analysis Package, Version 8.3 (QA), Electric Power Research Institute (2018).
   Google Scholar
• J. J. CARBAJO, D. DE WET, and N. R. BROWN, “Modeling the Molten Salt Reactor Experiment with the RELAP5-3D Code,” Trans. Am. Nucl. Soc., 116, 1503 (2017).
   Google Scholar
• C. SHI, M. CHENG, and G. LIU, “Development and Application of a System Analysis Code for Liquid Fueled Molten Salt Reactors Based on RELAP5 Code,” Nucl. Eng. Des., 305, 378 (Aug. 2016); https://doi.org/https://doi.org/10.1016/jnucengdes201605034.
   Web of Science ®Google Scholar
• X. HE, “Validation of the TRACE Code for the System Dynamic Simulations of the Molten Salt Reactor Experiment and the Preliminary Study on the Dual Fluid Molten Salt Reactor,” PhD Thesis, Technical University of Munich (2016).
   Google Scholar
• A. M. LEANDRO et al., “Thermal Hydraulic Model of the Molten Salt Reactor Experiment with the NEAMS System Analysis Module,” Ann. Nucl. Energy, 126, 59 (2019); https://doi.org/https://doi.org/10.1016/j.anucene.2018.10.060.
   Web of Science ®Google Scholar
• M. SCOTT GREENWOOD et al., “Demonstration of the Advanced Dynamic System Modeling Tool TRANSFORM in a Molten Salt Reactor Application via a Model of the Molten Salt Demonstration Reactor,” Nucl. Technol., 206, 3, 478504 (2020); https://doi.org/https://doi.org/10.1080/00295450.2019.1627124.
   Web of Science ®Google Scholar
• K. PODILA, Q. CHEN, and Y. RAO, “CFD Simulations of Molten Salt Reactor Experiment Core,” Nucl. Sci. Eng., 193, 12, 1379 (2019); https://doi.org/https://doi.org/10.1080/00295639.2019.1627177.
   Web of Science ®Google Scholar
• P. N. HAUBENREICH and J. R. ENGEL, “Experience with the Molten Salt Reactor Experiment,” Nucl. Appl. Technol., 8, 118 (1970).
   Web of Science ®Google Scholar
• C. W. HIRT and J. M. SICILIAN, “A Porosity Technique for the Definition of Obstacles in Rectangular Cell Meshes,” Proc. 4th Int. Conf. Numerical Ship Hydrodynamics, Washington, D.C., September 24–27, 1985, Naval Ship Research and Development Center (1985).
   Google Scholar
• E. W. LEMMON, M. L. HUBER, and M. O. McLINDEN, “NIST Standard Reference Database 23: Fluid Thermodynamic and Transport Properties— REFPROP,” Version 9.0, Users Guide, National Institute of Standards and Technology.
   Google Scholar
• D. F. WILLIAMS, L. M. TOTH, and K. T. CLARNO, “Assessment of Candidate Molten Salt Coolants for the Advanced High-Temperature Reactor (AHTR),” ORNL/TM-2006/12, Oak Ridge National Laboratory (2006).
   Google Scholar
• D. F. WILLIAMS, “Assessment of Candidate Molten Salt Coolants for the NGNP/NHI Heat-Transfer Loop,” ORNL/TM-2006/69. Oak Ridge National Laboratory (2006).
   Google Scholar
• A. PADILLA JR., “High-Temperature Thermodynamic Properties of Sodium,” HEDL-TME 77–27, Hanford Engineering Development Laboratory (1978).
   Google Scholar
• J. K. FINK, “Computer Codes Used in the Calculation of High-Temperature Thermodynamic Properties of Sodium,” ANL-CEN-RSD-79-2, Argonne National Laboratory (1979).
   Google Scholar
• J. K. FINK and L. LEIBOWITZ, “Thermodynamic and Transport Properties of Sodium Liquid and Vapor,” ANL/RE-95/2, Argonne National Laboratory (1995).
   Google Scholar
• J. W. LANE et al., “Benchmark of GOTHIC to EBR-II SHRT-17 and SHRT-45R Tests,” Nucl. Technol., 206, 7, 1019 (2020); https://doi.org/https://doi.org/10.1080/00295450.2019.1698896.
   Web of Science ®Google Scholar
• J. W. LANE et al., “Applicability of GOTHIC 8.3(QA) for Non-LWR Simulation, Aerosol Modeling & Hydrogen Management,” Proc. 18th Int. Topl. Mtg. Nuclear Reactor Thermal Hydraulics (NURETH-18), Portland, Oregon, August 18–22, 2019, American Nuclear Society (2019).
   Google Scholar
• J. W. LANE and T. L. GEORGE, “Development of a Multiple Liquid Component Capability in GOTHIC to Better Support Boric Acid, Radiological, and GSI-191 Analyses,” Proc. 16th Int. Topl. Mtg. Nuclear Reactor Thermal Hydraulics (NURETH-16), Chicago, Illinois, August 30–September 4, 2015, American Nuclear Society (2015).
   Google Scholar
• J. R. ENGEL and P. N. HAUBENREICH, “Temperatures in the MSRE Core During Steady-State Power Operation,” ORNL-TM-378, Oak Ridge National Laboratory (1962).
   Google Scholar
• R. J. KEDLE, “Fluid Dynamic Studies of the Molten-Salt Reactor Experiment (MSRE) Core,” ORNL-TM-3229, Oak Ridge National Laboratory (1970).
   Google Scholar
• R. HARVILL et al., “Benchmarking GOTHIC to the Molten Salt Reactor Experiment,” Trans. Am. Nucl. Soc., 121, 1745 (2019).
   Google Scholar
• R. C. ROBERTSON et al., “MSRE Design and Operations Report, Part III, Nuclear Analysis,” ORNL-TM-730, Oak Ridge National Laboratory (1964).
   Google Scholar
• P. N. HAUBENREICH et al., “MSRE Design and Operations Report, Part V-A, Safety Analysis of Operation with 233U,” ORNL-TM-2111, Oak Ridge National Laboratory (1969).
   Google Scholar
• “Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition,” NUREG-0800, Rev. 2, Sec. 9.2.5, “Ultimate Heat Sink,” Branch Technical Position ASB 9-2, “Residual Decay Energy for Light-Water Reactors for Long-Term Cooling,” U.S. Nuclear Regulatory Commission.
   Google Scholar
• “MSRE Systems and Components Performance,” ORNL-TM-3039, R. H. GUYMON, Ed., Oak Ridge National Laboratory (1973).
   Google Scholar
• B. E. PRINCE et al., “Zero-Power Physics Experiments on the Molten-Salt Reactor Experiment,” ORNL-4233, Oak Ridge National Laboratory (1968).
   Google Scholar
• P. N. HAUBENREICH and J. R. ENGEL, “Safety Calculations for MSRE,” ORNL-TM-251, Oak Ridge National Laboratory (1962).
   Google Scholar
• R. C. STEFFEY JR. and P. J. WOOD, “Theoretical Dynamic Analysis of the MSRE with 233U Fuel,” ORNL-TM-2571, Oak Ridge National Laboratory (1969).
   Google Scholar
• R. B. BRIGGS, “Molten-Salt Reactor Program Semiannual Progress Report for Period Ending August 31, 1965,” ORNL-3872, Oak Ridge National Laboratory (1965).
   Google Scholar
• S. GLASSTONE and A. SESONSKE, Nuclear Reactor Engineering, 3rd ed. (revised), Van Nostrand Reinhold Company, New York (1981).
   Google Scholar
• B. E. PRINCE, “Period Measurement on the Molten Salt Reactor Experiment During Fuel Circulation: Theory and Experiment,” ORNL-TM-1626, Oak Ridge National Laboratory (1966).
   Google Scholar
• J. R. TALLACKSON, “MSRE Design and Operations Report, Part IIA, Nuclear and Process Instrumentation,” ORNL-TM-729, Oak Ridge National Laboratory (1968).
   Google Scholar
• M. W. ROSENTHAL, R. B. BRIGGS, and P. R. KASTEN, “Molten-Salt Reactor Program Semiannual Progress Report for Period Ending February 28, 1969,” ORNL-4396, Oak Ridge National Laboratory (1969).
   Google Scholar
• N. R. BROWN et al., “Thermal-Hydraulic and Neutronic Phenomena Important in Modeling and Simulation of Liquid-Fuel Molten Salt Reactors,” Nucl. Technol., 206, 2, 322 (2020); https://doi.org/https://doi.org/10.1080/00295450.2019.1590077.
   Web of Science ®Google Scholar