A Review of Void Drift Models in Subchannel Analysis

References

• D. S. ROWE, “Cross-Flow Mixing Between Parallel Flow Channels During Boiling. Part I. COBRA: Computer Program for Coolant Boiling in Rod Arrays,” Pacific Northwest Laboratory, Battelle-Northwest (1967).
   Google Scholar
• R. T. LAHEY and F. J. MOODY, The Thermal Hydraulics of a Boiling Water Nuclear Reactor, American Nuclear Society, La Grange Park, Illinois (1993).
   Google Scholar
• B. PANG, Numerical Study of Void Drift in Rod Bundle with Subchannel and CFD Codes, KIT Scientific Publishing, Karlsruhe, Germany (2014).
   Google Scholar
• C. L. WHEELER et al., “COBRA-IV-I: An Interim Version of COBRA for Thermal-Hydraulic Analysis of Rod Bundle Nuclear Fuel Elements and Cores,” Battelle Pacific Northwest Laboratory (1976).
   Google Scholar
• J. M. GONZALEZ-SANTALO and P. GRIFFITH, “Two-Phase Flow Mixing in Rod Bundle Subchannels,” Proc. ASME Winter Annual Mtg., New York, 1972, American Society of Mechanical Engineers (1972).
   Google Scholar
• F. A. SCHRAUB, R. L. SIMPSON, and E. JANSSEN, “Two-Phase Flow and Heat Transfer in Multirod Geometries: Air-Water Flow Structure Data for a Round Tube, Concentric and Eccentric Annulus, and Nine-Rod Bundle,” GEAP-5739, General Electric Company (1969).
   Google Scholar
• R. T. LAHEY, B. S. SHIRALKAR, and D. W. RADCLIFFE, “Mass Flux and Enthalpy Distribution in a Rod Bundle for Single- and Two-Phase Flow Conditions,” J. Heat Transfer, 93, 2, 197 (1971); https://doi.org/10.1115/1.3449786.
   Google Scholar
• M. SADATOMI, A. KAWAHARA, and Y. SATO, “Flow Redistribution due to Void Drift in Two-Phase Flow in a Multiple Channel Consisting of Two Subchannels,” Nucl. Eng. Des., 148, 2–3, 463 (1994); https://doi.org/10.1016/0029-5493(94)90126-0.
   Web of Science ®Google Scholar
• M. SADATOMI et al., “Void Fraction and Pressure Drop in Two-Phase Equilibrium Flows in a Vertical 2 × 3 Rod Bundle Channel—Assessment of Correlations Against the Present Subchannel Data,” JSME, 49, 2, 279 (2006); https://doi.org/10.1299/jsmeb.49.279.
   Web of Science ®Google Scholar
• K. KANO et al., “Flow Distribution and Slug Flow Characteristics for Hydraulically Equilibrium Air-Water Two-Phase Flows in a Vertical 2 × 3 Rod Channel,” Proc. ASME/JSME 2003 4th Joint Fluids Summer Engineering Conf., Honolulu, Hawaii, July 6–10, 2003, p. 1407, American Society of Mechanical Engineers (2003).
   Google Scholar
• H. REICHARDT, “Complete Representation of the Turbulent Velocity Distribution in Smooth Pipes,” Z. Angew. Math. Mech., 31, 7, 208 (1951); https://doi.org/10.1002/zamm.19510310704.
   Google Scholar
• D. S. ROWE and C. W. ANGLE, “Crossflow Mixing Between Parallel Flow Channels During Boiling. Part III. Effect of Spacers on Mixing Between Two Channels,” Pacific Northwest Laboratory, Battelle-Northwest (1969).
   Google Scholar
• J. T. ROGERS and R. G. ROSEHART, “Mixing by Turbulent Interchange in Fuel Bundles, Correlations and Inferences,” ASME 72-HT-53, American Society of Mechanical Engineers (1972).
   Google Scholar
• S. LEVY, “Prediction of Two-Phase Pressure Drop and Density Distribution from Mixing Length Theory,” J. Heat Transfer, 85, 2, 137 (1963); https://doi.org/10.1115/1.3686033.
   Google Scholar
• R. T. LAHEY et al., “Out-of-Pile Subchannel Measurements in a Nine-Rod Bundle for Water at 1000 psia,” Prog. Heat Mass Transfer, 6, 345 (1972); https://doi.org/10.1016/B978-0-08-017035-0.50023-9.
   Google Scholar
• R. T. LAHEY, “Subchannel Measurements of the Equilibrium Quality and Mass Flux Distribution in a Rod Bundle,” Proc. 8th Int. Heat Transfer Conf., San Francisco, California, August 17–22, 1986.
   Google Scholar
• L. WOLF et al., “Comprehensive Assessment of the ISPRA BWR and PWR Subchannel Experiments and Code Analysis with Different Two-Phase Models and Solution Schemes,” Nucl. Eng. Des., 99, 329 (1987); https://doi.org/10.1016/0029-5493(87)90131-2.
   Web of Science ®Google Scholar
• D. S. ROWE, “COBRA-IIIC: A Digital Computer Program for Steady-State and Transient Thermal-Hydraulic Analysis of Rod Bundle Nuclear Fuel Elements,” BNWL-1695, Battelle Pacific Northwest Laboratory (1973).
   Google Scholar
• J. LOOMIS et al., “THERMIT—A Computer Program for Three-Dimensional Thermal-Hydraulic Analysis of Light Water Reactor Cores,” EPRI NP-2032, Electric Power Research Institute (1981).
   Google Scholar
• H. HERKENRATH et al., “Experimental Investigation of the Enthalpy and Mass Flow Distribution in 16-Rod Clusters with BWR-PWR Geometries and Conditions,” Joint Research Centre of the European Communities (1981).
   Google Scholar
• M. SADATOMI et al., “Flow Characteristics in Hydraulically Equilibrium Two-Phase Flows in a Vertical 2 × 3 Rod Bundle Channel,” Int. J. Multiphase Flow, 30, 9, 1093 (2004); ISSN 0301-9322; https://doi.org/10.1016/j.ijmultiphaseflow.2004.05.012.
   Web of Science ®Google Scholar
• J. E. KELLY, S. P. KAO, and M. S. KAZIMI, “THERMIT-2: A Two-Fluid Model for Light Water Reactor Subchannel Transient Analysis,” MIT-EL-81-014, Massachusetts Institute of Technology Energy Laboratory Electric Utility Program (1981).
   Google Scholar
• D. S. ROWE, R. B. MACDUFF, and R. E. COLLINGHAM, “Thermal Hydraulic Subchannel Model Based on Void Drift,” Proc. 9th Int. Heat Transfer Conf. (Heat Transfer 1990), Jerusalem, Israel, August 19–24, 1990, Vol. 3, p. 401 (1990).
   Google Scholar
• S. SUGAWARA and Y. MIYAMOTO, “FIDAS: Detailed Subchannel Analysis Code Based on the Three-Fluid and Three-Field Model,” Nucl. Eng. Des., 120, 147 (1990); https://doi.org/10.1016/0029-5493(90)90369-9.
   Web of Science ®Google Scholar
• S. SUGAWARA et al., “Subchannel Analysis by the FIDAS Code Based on the Three-Fluid Model,” Nucl. Eng. Des., 132, 253 (1991); https://doi.org/10.1016/0029-5493(91)90269-N.
   Web of Science ®Google Scholar
• D. H. HWANG et al., “Assessment of the Interchannel Mixing Model with a Subchannel Analysis Code for BWR and PWR Conditions,” Nucl. Eng. Des., 199, 3, 257 (2000); https://doi.org/10.1016/S0029-5493(00)00234-X.
   Web of Science ®Google Scholar
• Y. J. YOO, D. H. HWANG, and D. S. SOHN, “Evaluation of Interchannel Exchange Models with a Subchannel Analysis Code,” Proc. 11th Int. Heat Transfer Conf. (Heat Transfer 1998), Kyongju, Korea, August 23–28, 1998, Vol. 6, p. 23 (1998).
   Google Scholar
• D. H. HWANG, J. J. JEONG, and B. D. CHUNG, “NUPEC BFBT Subchannel Void Distribution Analysis Using the MATRA and MARS Code,” Nucl. Eng. Technol., 41, 3, 295 (2009); https://doi.org/10.5516/NET.2009.41.3.295.
   Web of Science ®Google Scholar
• M. GLÜCK, “Validation of the Sub-Channel Code F-COBRA-TF: Part II. Recalculation of Void Measurements,” Nucl. Eng. Des., 238, 9, 2317 (2008); https://doi.org/10.1016/j.nucengdes.2008.02.006.
   Web of Science ®Google Scholar
• G. BEUS, “Two-Phase Turbulent Mixing Model for Flow in Rod Bundles,” Proc. Annual Mtg. American Society of Mechanical Engineers, Washington, District of Columbia, 1971, American Society of Mechanical Engineers (1971).
   Google Scholar
• A. J. FAYA, “Development of a Method for BWR Subchannel Analysis,” PhD Thesis, Massachusetts Institute of Technology Department of Nuclear Engineering (1979).
   Google Scholar
• G. B. WALLIS, One Dimensional Two-Phase Flow, McGraw-Hill Book Company, New York (1969).
   Google Scholar
• D. W. RADCLIFFE, R. T. LAHEY, and B. S. SHIRLAKAR, “Two-Phase Flow and Heat Transfer in Multirod Geometries: Subchannel and Pressure Drop Measurements in a Nine-Rod Bundle for Diabatic and Adiabatic Conditions,” General Electric Company (1970).
   Google Scholar
• Y. J. YOO, D. H. HWANG, and D. S. SOHN, “Evaluation of Interchannel Exchange Models with a Subchannel Analysis Code,” Heat Transfer, 6, 33 (1998).
   Google Scholar
• Y. J. YOO, D. H. HWANG, and D. S. SOHN, “Development of a Subchannal Analysis Code MATRA Applicable to PWRs and ALWRs,” J. Korean Nucl. Soc., 31, 314 (1999).
   Google Scholar
• D. S. ROWE and C. W. ANGLE, “Crossflow Mixing Between Parallel Flow Channels During Boiling. Part II. Measurement of Flow and Enthalpy in Two Parallel Channels,” Pacific Northwest Laboratory, Battelle-Northwest (1967).
   Google Scholar
• B. PANG and X. CHENG, “Assessment of Two-Phase Interchannel Mixing Models for BWR and PWR Conditions,” Ann. Nucl. Energy, 99, 283 (2017); https://doi.org/10.1016/j.anucene.2016.09.009.
   Web of Science ®Google Scholar
• B. NEYKOV et al., “NUPEC BWR Full-Size Finemesh Bundle Test (BFBT) Benchmark, Volume I: Specifications. Technical Report,” NEA/NSC/DOC(2006)5, Version 02, Organisation for Economic Co-operation and Development/Nuclear Energy Agency (2006).
   Google Scholar
• R. W. STERNER and R. T. LAHEY, “Air/Water Subchannel Measurements of the Equilibrium Quality and Mass-Flux Distribution in a Rod Bundle,” Rensselaer Polytechnic Institute, Department of Nuclear Engineering (1983).
   Google Scholar
• P. TYE, A. TEYSSEDOU, and A. TAPUCU, “An Investigation of the Constitutive Relations for Inter-Subchannel Transfer Mechanisms in Horizontal Flows as Applied in the ASSERT-4 Subchannel Code,” Nucl. Eng. Des., 149, 1–3, 207 (1994); https://doi.org/10.1016/0029-5493(94)90287-9.
   Web of Science ®Google Scholar
• A. TAPUCU et al., “The Effect of Turbulent Mixing Models on the Predictions of Subchannel Codes,” Nucl. Eng. Des., 149, 1–3, 221 (1994); https://doi.org/10.1016/0029-5493(94)90288-7.
   Web of Science ®Google Scholar
• A. KAWAHARA et al., “Void Diffusion Coefficient in Two-Phase Void Drift for Several Channels of Two- and Multi-Subchannel Systems,” Multiphase Sci. Technol., 18, 1, 2006 (2006); https://doi.org/10.1615/MultScienTechn.v18.i1.20.
   Google Scholar
• A. TAPUCU et al., “Experimental Investigation of Mass Exchanges Between Two Laterally Interconnected Two-Phase Flows,” Nucl. Eng. Des., 105, 3, 295 (1988); https://doi.org/10.1016/0029-5493(88)90251-8.
   Web of Science ®Google Scholar
• A. KAWAHARA and A. SADATOMI, “Modeling of Void Diffusion Coefficient in a Two-Phase Subchannel Flow,” Proc. 4th JSME-KSME Thermal Engineering Conf., Kobe, Japan, October 1–6, 2000, Vol. 2, p. 611, Japan Society of Mechanical Engineers (2000).
   Google Scholar