Development, Verification, and Validation of an Advanced Systems Code KP-SAM for Kairos Power Fluoride Salt–Cooled High-Temperature Reactor (KP-FHR)

References

• E. BLANDFORD et al., “Kairos Power Thermal Hydraulics Research and Development,” Nucl. Eng. Des., 364, 110636 (Aug. 2020); https://doi.org/10.1016/j.nucengdes.2020.110636.
   Web of Science ®Google Scholar
• B. HAUGH et al., “Kairos Power Thermal Hydraulics R&D: Advances, Challenges, and Lessons Learned Applied to Validation,” Proc. 19th Int. Topl. Mtg. Nuclear Reactor Thermal Hydraulics (NURETH-19), Virtual Meeting, March 6–11, 2022, American Nuclear Society (2022).
   Google Scholar
• “Uranium Oxycarbide (UCO) Tristructural Isotropic (TRISO) Coated Particle Fuel Performance,” EPRI‐AR(NP)‐A, Electric Power Research Institute (Nov. 2020).
   Google Scholar
• R. HU, “SAM Theory Manual,” ANL/NE-17/4, Argonne National Laboratory (Mar. 2017).
   Google Scholar
• H. ZHAO et al., “Overview of Kairos Power Systems Code KP-SAM Development,” Proc. 18th Int. Topl. Mtg. Nuclear Reactor Thermal Hydraulics, Portland, Oregon, August 18–22, 2019, American Nuclear Society (2019).
   Google Scholar
• “MOOSE Framework,” Idaho National Laboratory; https://mooseframework.org ( current as of 2022).
   Google Scholar
• LibMesh website; http://libmesh.github.io/index.html ( current as of 2022).
   Google Scholar
• PETSc website; https://www.mcs.anl.gov/petsc/ ( current as of 2022).
   Google Scholar
• B. E. BOYACK et al., “Quantifying Reactor Safety Margins Part 1: An Overview of the Code Scaling, Applicability, and Uncertainty Evaluation Methodology,” Nucl. Eng. Des., 119, 1 (1990). G. E. WILSON et al., “Quantifying Reactor Safety Margins Part 2: Characterization of Important Contributors to Uncertainty,” Nucl. Eng. Des., 119, 17 (1990). W. WULFF et al., “Quantifying Reactor Safety Margins Part 3: Assessment and Ranging of Parameters,” Nucl. Eng. Des., 119, 33 (1990). G. S. LELLOUCHE et al., “Quantifying Reactor Safety Margins Part 4: Uncertainty Evaluation of LBLOCA Analysis Based on TRAC-PF1/MOD 1,” Nucl. Eng. Des., 119, 67 (1990). N. ZUBER et al., “Quantifying Reactor Safety Margins Part 5: Evaluation of Scale-Up Capabilities of Best Estimate Codes,” Nucl. Eng. Des., 119, 97 (1990). I. CATTON et al., “Quantifying Reactor Safety Margins Part 6: A Physically Based Method of Estimating PWR Large Break Loss of Coolant Accident PCT,” Nucl. Eng. Des., 119, 109 (1990).
   Web of Science ®Google Scholar
• “Transient and Accident Analysis Methods,” Regulatory Guide 1.203, U.S. Nuclear Regulatory Commission (Dec. 2005).
   Google Scholar
• “KP-FHR Mechanistic Source Term Methodology Topical Report,” KP-TR-012, Rev. 1, Kairos Power (Aug. 2021).
   Google Scholar
• R. HU, “A Fully-Implicit High-Order System Thermal-Hydraulics Model for Advanced Non-LWR Safety Analyses,” Ann. Nucl. Energy, 101, 174 (2017); https://doi.org/10.1016/j.anucene.2016.11.004.
   Web of Science ®Google Scholar
• D. A. KNOLL and D. E. KEYES, “Jacobian-Free Newton-Krylov Methods: A Survey of Approaches and Applications,” J. Comput. Phys., 193, 357 (2004); https://doi.org/10.1016/j.jcp.2003.08.010.
   Web of Science ®Google Scholar
• V. A. MOUSSEAU, “Accurate Solution of the Nonlinear Partial Differential Equations from Thermal Hydraulics: Thermal Hydraulics,” Nucl. Technol., 158, 1, 26 (2007); https://doi.org/10.13182/NT07-A3822.
   Web of Science ®Google Scholar
• V. A. MOUSSEAU, H. ZHANG, and H. ZHAO, “High Fidelity Coupled Thermal Hydraulics and Neutronics for Reactor Safety Simulations,” Proc. Int. Conf. Reactor Physics, Nuclear Power: A Sustainable Resource (PHYSOR’08), Interlaken, Switzerland, September 14–19, 2008.
   Google Scholar
• H. ZHAO, “A Pilot Framework for Modern Reactor System Analysis Codes,” presented at 2011 International RELAP5 User’s Group Seminar, Salt Lake City, Utah, July 25–28, 2011.
   Google Scholar
• D. ANDRS et al., “RELAP-7 Level 2 Milestone Report: Demonstration of a Steady State Single Phase PWR Simulation with RELAP-7,” INL/EXT-12-25924, Idaho National Laboratory (May 2012).
   Google Scholar
• H. ZHAO et al., “A Strongly Coupled Reactor Core Isolation Cooling System Model for Extended Station Black-Out 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
• L. ZOU, H. ZHAO, and H. ZHANG, “Application of Jacobian-Free Newton-Krylov Method in Implicitly Solving Two-Fluid Six-Equation Two-Phase Flow Problems: Implementation, Validation and Benchmark,” Nucl. Eng. Des., 300, 268 (2016); https://doi.org/10.1016/j.nucengdes.2016.01.033.
   Web of Science ®Google Scholar
• W. L. OBERKAMPF and C. J. ROY, Verification and Validation in Scientific Computing, Cambridge University Press (2010).
   Google Scholar
• L. FICK and H. ZHAO, “Numerical Verification of the Kairos Power Systems Code KP-SAM,” Proc. 18th Int. Topl. Mtg. Nuclear Reactor Thermal Hydraulics, Portland, Oregon, August 18–22, 2019, American Nuclear Society (2019).
   Google Scholar
• L. FICK et al., “On the Progress of Verification and Validation of KP-SAM for FHR Systems Analysis,” Proc. 19th Int. Topl. Mtg. Nuclear Reactor Thermal Hydraulics (NURETH-19), Virtual Meeting, March 6–11, 2022, American Nuclear Society (2022).
   Google Scholar
• N. ZWEIBAUM, “Scaling Methodology for the Kairos Power Testing Program,” KP-TR-006-NP, Kairos Power (2019).
   Google Scholar
• K. BRUMBACK et al., “Empirical Derivation of Heat Transfer Correlations for Fluoride Salt-Cooled Pebble Bed Reactor Cores,” Proc. 19th Int. Topl. Mtg. Nuclear Reactor Thermal Hydraulics (NURETH-19), Virtual Meeting, March 6–11, 2022, American Nuclear Society (2022).
   Google Scholar
• B. METAIS and E. R. G. ECKERT, “Forced, Mixed, and Free Convection Regimes,” J. Heat Transfer, 86, 2, 295 (1964); https://doi.org/10.1115/1.3687128.
   Google Scholar
• F. RUBIO et al., “Scaled Validation Test for High Prandtl Number Fluid Mixed Convection Between Parallel Plates,” Proc. 19th Int. Topl. Mtg. Nuclear Reactor Thermal Hydraulics (NURETH-19), Virtual Meeting, March 6–11, 2022, American Nuclear Society (2022).
   Google Scholar
• N. ZWEIBAUM, “Experimental Validation of Passive Safety System Models: Application to Design and Optimization of Fluoride-Salt-Cooled, High-Temperature Reactors,” PhD Thesis, University of California, Berkeley (2015).
   Google Scholar