Development and Application of an Integrated Fuel Performance and Subchannel Model for Analysis of Sodium Fast Reactors

References

• J. M. DEUTCH et al., “Update to the MIT 2003 Future of Nuclear Power,” Massachusetts Institute of Technology (2009).
   Google Scholar
• “A Technology Roadmap for Generation IV Nuclear Energy System: Ten Nations Preparing Today for Tomorrow’s Energy Needs,” U.S. Department of Energy, Nuclear Energy Research Advisory Committee and the Generation IV International Forum (2002).
   Google Scholar
• A. KARAHAN and J. BUONGIORNO, “A New Code for Predicting the Thermo-Mechanical and Irradiation Behavior of Metallic Fuels in Sodium Fast Reactors,” J. Nucl. Mater., 396, 2, 283 (2010).
   Web of Science ®Google Scholar
• A. KARAHAN, “Modeling of Thermo-Mechanical and Irradiation Behavior of Metallic and Oxide Fuels for Sodium Fast Reactors,” PhD Thesis, Massachusetts Institute of Technology (2009).
   Google Scholar
• H. OHSIMA, H. NARITA, and H. NINOKATA, “Analysis of Thermal-Hydraulic Behavior in a Fast Reactor Fuel Subassembly with Porous Blockages,” Proc. Int. Topl. Mtg. Advanced Reactor Safet., pp. 1157–1164, American Nuclear Society (1997).
   Google Scholar
• C. L. WHEELER, C. W. STEWART, R. J. CENA, D. S. ROWE, and A. M. SUTEY, “COBRA-IV-I: An Interim Version of COBRA for Thermal-Hydraulic Analysis of Rod Bundle Nuclear Fuel Elements and Cores,” BNWL-1962, Battelle Pacific Northwest Laboratories (1976).
   Google Scholar
• C. W. STEWART, C. L. WHEELER, R. J. CENA, C. A. McMONAGLE, J.M. CUTA, and D.S. TRENT, “COBRA-IV: The Model and the Method,” BNWL-2214, Battelle Pacific Northwest Laboratories (1977).
   Google Scholar
• C. W. STEWART et al., “Core Thermal Model: COBRA-IV Development and Applications,” BNWL-2212, Battelle Pacific Northwest Laboratories (1977).
   Google Scholar
• W. KIM, Y.-G. KIM, and Y.-J. KIM, “A Subchannel Analysis Code MATRA-LMR for Wire Wrapped LMR Subassem-bly,” Ann. Nucl. Energ., 29, 303 (2001).
   Web of Science ®Google Scholar
• M. J. MEMMOTT, “Thermal-Hydraulic Analysis of Innovative Fuel Configurations for the Sodium Fast Reactor,” PhD Thesis, Massachusetts Institute of Technology (2005).
   Google Scholar
• J. MacDOUGALL and J. LILLINGTON, “The SABRE Code for Fuel Rod Cluster Thermohydraulics,” Nucl. Eng. Des., 82, 171 (1984).
   Web of Science ®Google Scholar
• S. YANG, “An LMR Core Thermal-Hydraulics Code Based on the ENERGY Model,” J. Korean Nucl. Soc., 29, 4. (1977).
   Google Scholar
• E. U. KHAN, W. M. ROHSENOW, A. A. SONIN, and N. E. TODREAS, “A Porous Body Model for Predicting Temperature Distribution in Wire-Wrapped Fuel Rod Assemblies,” Nucl. Eng. Des., 35, 1, (1975).
   Web of Science ®Google Scholar
• C. W. STEWART, “VIPRE01: Thermal Hydraulic Code for Reactor,” EPRINP-2511-CCM-A (1983).
   Google Scholar
• J. W. FRICANO and J. BUONGIORNO, “COBRA4i-MIT: An Updated Sub-Channel Analysis Code for Sodium Fast Reactor Design,” Proc. ICAP., Nice, France, American Nuclear Society (2011).
   Google Scholar
• S.-K. CHENG and N. E. TODREAS, “Hydrodynamic Models and Correlations for Bare and Wire-Wrapped Hexagonal Rod Bundles–Bundle Friction Factors, Subchannel Friction Factors and Mixing Parameters,” Nucl. Eng. Desig., 92, 227 (1986).
   Web of Science ®Google Scholar
• E. H. NOVENDSTERN, “Turbulent Flow Pressure Drop Model for Fuel Rod Assemblies Utilizing a Helical Wire-Wrap Spacer System,” Nucl. Eng. Des., 22, 1, 19 (1972).
   Web of Science ®Google Scholar
• T. S. RO and N.E. TODREAS. “Energy Transfer Mechanisms in LMR Rod Bundles Under Mixed Convection Conditions,” Nucl. Eng. Des., 108, 343 (1988).
   Web of Science ®Google Scholar
• K. RHEME, “The Structure of Turbulence in Rod Bundles and the Implications on Natural Mixing Between the Subchannels,” Int. J. Heat Mass Transfe., 35, 567 (1992).
   Web of Science ®Google Scholar
• K. MIKITYUK, “Heat Transfer to Liquid Metal: Review of Data and Correlations for Tube Bundles,” Nucl. Eng. Des., 239, 680 (2009).
   Web of Science ®Google Scholar
• J. W. FRICANO, “Integrated Fuel Performance and Thermal-Hydraulic Sub-channel Models for Analysis of Sodium Fast Reactors,” PhD Thesis, Massachusetts Institute of Technology, Nuclear Science and Engineering (2012).
   Google Scholar
• M. H. FONTANA, R. E. MacPHERSON, P. A. GNADT, and L. F. PARSLY, “Temperature Distribution in the Duct Wall and at the Exit of a 19-Rod Simulated LMFBR Fuel Assembly ( F FM Bundle 2A),” Nucl. Technol., 24, 176 (1974).
   Web of Science ®Google Scholar
• J. L. WANTLAND, P. P. GNADT, R. E. MacPHERSON, M.H. FONTANA, H. HANUS, and C.M. SMITH, “The Effects of Duct Configuration on Flow and Temperature Structure in Sodium-Cooled 19-Rod Simulated LMFBR Fuel Bundles with Helical Wire-Wrap Spacers.” Proc. 16th National Heat Transfer Conf., St. Louis, Missouri (1976).
   Google Scholar
• E. C. ENGEL, B. MINUSHKIN, R. J. ATKINS, and R. A. MARKLEY, “Characterization of Heat Transfer and Temperature Distributions in an Electrically Heated Model of an LMFBR Blanket Assembly,” Nucl. Eng. Des., 62, 335 (1980).
   Web of Science ®Google Scholar
• F. NAMEKAWA, A. ITO, and K. MAWATARI, “Buoyancy Effects on Wire-Wrapped Rod Bundle Heat Transfer in an LMFBR Fuel Assembly,” AIChE Symp. Ser., Niagara Falls, New York, 1984, pp. 128–133.
   Google Scholar
• “ADINA Theory and Modeling Guide Vol. I: ADINA,” ARD 09-2, ADINA R&D, Inc., Watertown, Massachusetts (2009).
   Google Scholar
• “ADINA Theory andModeling Guide Vol. II: ADINA Heat Transfer,” ARD 09-3, ADINA R&D, Inc., Watertown, Massachusetts (2009).
   Google Scholar
• M. C. CHUANG, R. E. ROTHMANN, M. J. PECHERSKY, and R. A. MARKLEY, “Cladding Circumferential Hot Spot Factors for Fuel and Blanket Rods,” Nucl. Eng. Des., 35, 21 (1975).
   Web of Science ®Google Scholar
• R. G. PAHL, D. L. PORTER, D. C. CRAWFORD, and L. C. WALTERS, “Irradiation Behavior of Metallic Fast Reactor Fuels,” J. Nucl. Mater., 188, 3 (1992).
   Web of Science ®Google Scholar
• C. E. LAHM, J. F. KOENIG, R. G. PAHL, D. L. PORTER, and D. C. CRAWFORD, “Experience with Advanced Driver Fuels in EBR-II,” J. Nucl. Mater., 204, 119 (1993).
   Web of Science ®Google Scholar
• D.C. CRAWFORD, S.L. HAYES, and R.G. PAHL, “Large-Diameter, High Plutonium Metallic Fuel Testing in EBR-II,” Trans. Am. Nucl. Soc., 71, 178 (1994).
   Google Scholar
• R. G. PAHL, R. S. WISNER, M. C. BILLONE, and G. L. HOFMAN, “Steady-State Irradiation Testing of U-Pu-Zr Fuel to >18 at.% Burnup,” Proc. 1990 Int. Fast Reactor Safety Mtg., Snowbird, Utah, August 12-16, 1990, p. 129, American Nuclear Society (1990).
   Google Scholar
• B. O. LEE, J. S. CHEON, and C. B. LEE, “Performance Limit Analysis of Metallic Fuel for Kalimer,” Proc. Global 200., Boise, Idaho, pp. 397–403.
   Google Scholar
• R. G. PAHL, C. E. LAHM, and S. L. HAYES, “Performance of HT9 Clad Metallic Fuel at High Temperature,” J. Nucl. Mater., 204, 141 (1993).
   Web of Science ®Google Scholar
• A. B. COHEN, H. TSAI, and L. A. NEIMARK, “Fuel/Cladding Compatibility in U-19Pu-10Zr/HT9-Clad Fuel at Elevated Temperatures,” J. Nucl. Mater., 204, 244 (1993).
   Web of Science ®Google Scholar