Thermal hydraulic analysis of nuclear reactor core and its associated systems can be performed using analysis system, subchannel or computational fluid dynamics (CFD) codes to estimate the different thermal hydraulic safety margins. The safety margins and operating power limits under different conditions of the primary as well as secondary cooling system such as the system pressure, coolant inlet temperature, coolant flow rate, and thermal power and its distributions are considered as key parameters for thermal hydraulic analysis. Considering the complexity of rod bundle geometry, boiling heat transfer and different turbulent scales bring about the many challenges in performing the thermal hydraulic analysis to ensure the safe design and operation of nuclear reactor systems under normal and abnormal conditions. A comprehensive review is presented of past, present and future challenges in state-of-the-art thermal hydraulic analysis c overing various aspects of experimental, analytical and computational approaches.
Abderrahim, H., Baeten, P., Fernandez, R., De Bruyn, D. 2010. MYRRHA: An innovative and unique irradiation research facility. In: Proceedings of the 11IEMPT.
Abro, E., Johansen, G. 1999. Improved void fraction determination by means of multibeam gamma-ray attenuation measurements. Flow Meas Instrum, 10: 99–108.
Agee, L. J., Duffey, R. B., Hughes, E. D., et al. 1978. Some aspects of two-fluid models for two-phase flow and their numerical solution. In: Proceedings of the 2nd CSNI Specialists Meeting on Transient Two-Phase Flow.
Albrecht, R. W., Crowe, R. D., Dailey, D., Damborg, M. J. 1982. Measurement of two-phase flow properties using the nuclear reactor instrument. Prog Nucl Energy, 9: 37–50.
Anglart, H., Caraghiaur, D. 2011. CFD modeling of boiling annularmist flow for dryout investigations. Multiphase Sci Technol, 23: 22–251.
Anglart, H., Nylund, O. 1996. CFD application to prediction of void distribution in bubbly flows in rod bundles. Nucl Eng Des, 163: 81–98.
Anh, K.-I., Kin, D.-H. 2003. A state-pf-the-art review of the reactor lower head models employed in three representative U.S. severe accident codes. Prog Nucl Energy, 3: 361–382.
Ardron, K. H., Banerjee, S. 1986. Flooding in an elbow between a vertical and a horizontal or near horizontal pipe, Part I. Theory. Int J Multiphase Flow, 12: 543–558.
Asmolov, V. G., Khabenski, V. B., Bechta, S. V., et al. 2003. MA-3 and MA-4 tests: Zirconium and uranium partitioning between oxidic and metallic phases of molten corium. OECD MASCA Project MP-TR-9, Russian Research Center, Kurchatov Institute.
Baglietto, E., Ninokata, H. 2004. CFD modeling of secondary flows in fuel rod bundles. NUTHOS-6.
Baglietto, E., Ninokata, H. 2005. A turbulence model study for simulating flow inside tight lattice rod bundles. Nucl Eng Des, 235: 773–784.
Baglietto, E., Ninokata, H., Misawa, T. 2006. CFD and DNS methodologies development for fuel bundle simulations. Nucl Eng Des, 236: 1503–1510.
Banerjee, S., Chan, A. M. C. 1980. Separated flow model I. Analysis of the averaged and local instantaneous formulations. Int J Multiphase Flow, 6: 1–24.
Bankoff, S. G., Lee, S. C. 1985. A brief review of countercurrent flooding models applicable to PWR geometries. Nucl Safety, 26: 139–152.
Banowski, M., Beyer, M., Szalinski, L., Lucas, D., Hampel, U. 2016. Comparative study of ultrafast X-ray tomography and wire-mesh sensors for vertical gas-liquid pipe flows. Flow Meas Instrum, 53: 95–106.
Belfroid, S. P. C., Cargnelutti, M. F., Schiferli, W., van Osch, M. 2010. Forces on bends and T-joints due to multiphase flow. In: Proceedings of the ASME 2010 3rd Joint US–European Fluids Engineering Summer Meeting.
Bennett, A. W., Hewitt, G. F., Kearsey, H. A., Keeys, R. K. F., Lacey, P. M. C. 1965. Flow visualization studies of boiling at high pressure. Proc Inst Mech Engrs, 180: 260–270.
Bergles, A. E., Roos, J. P., Bourne, J. G. 1968. Investigation of boiling flow regimes and critical heat flux. NYO-3304-13, Dynatech Corp.
Besnard, D. C., Harlow, F. H. 1988. Turbulence in multiphase flow. Int J Multiphase Flow, 14: 679–699.
Bestion, D. 2014. The difficult challenge of a two-phase CFD modelling for all flow regimes. Nucl Eng Des, 279: 116–125.
Biemüller, M., Meyer, L., Rehme, K. 1996. Large eddy simulation and measurement of the structure of turbulence in two rectangular channels connected by a gap. In: Proceedings of the 3rd International Symposium on Engineering Turbulence Modelling and Experiments, 249–258.
Bourne, J. A., Bergles, A. E., Tong, L. S. 1973. Review of two-phase flow instabilities. Nucl Eng Des, 25: 165–192.
Boyer, C., Duquenne, A. M., Wild, G. 2002. Measuring techniques in gas–liquid and gas–liquid–solid reactors. Chem Eng Sci, 57: 3185–3215.
Buongiorno, J., Hu, L. W., Kim, S. J., Hannink, R., Truong, B., Forrest, E. 2008. Nanofluids for enhanced economics and safety of nuclear reactors: An evaluation of the potential features, issues, and research gap. Nucl Technol, 162: 80–91.
Burger, M., Cho, S. H., Berg, E. V., Schatz, A. 1995. Breakup of melt jets as pre-condition for premixing: Modeling and experimental verification. Nucl Eng Des, 155: 15–25.
Caraghiaur, D. 2012. On drops and turbulence in nuclear fuel assemblies of boiling water reactors. Ph.D. Thesis. KTH.
Cargenlutti, M. F., Belfroid, S. P. C., Schiferli, W. 2010. Two-phase flow-induced forces on bends in small scale tubes. J Press Vess Technol, 132: 1–7.
Carver, M. B., Tahir, A., Rowe, D. S., Tapucu, A., Ahmad, S. Y. 1984. Computational analysis of two-phase flow in horizontal bundles. Nucl Eng Des, 82: 215–226.
Chabot, J., Farag, H., de Lasa, H. 1998. Fluid dynamics of bubble columns at elevated temperature modelling and investigation with refractive fiber optic sensors. Chem Eng J, 70: 105–113.
Chandra, L., Roelofs, F., Houkema, M., Jonker, B. 2009. A stepwise development and validation of a RANS based CFD modelling approach for the hydraulic and thermal-hydraulic analyses of liquid metal flow in a fuel assembly. Nucl Eng Des, 239: 1988–2003.
Chang, D., Tavoularis, S. 2008. Simulations of turbulence, heat transfer and mixing across narrow gaps between rod-bundle subchannels. Nucl Eng Des, 238: 109–123.
Chelemer, H., Hochreiter, L. E., Boman, L. H., Chu, P. T. 1977. An improved thermal-hydraulic analysis method for rod bundle cores. Nucl Eng Des, 41: 219–229.
Cheng, X., Muller, U. 2003. Review on critical heat flux in water cooled reactors. FZKA-6825, Forschngszentrum Karlsruhe GmbH, Karlsruhe.
Cheung, S. C. P., Vahaji, S., Yeoh, G. H., Tu, J. Y. 2014. Modeling subcooled flow boiling in vertical channels at low pressures— Part 1: Assessment of empirical correlations. Int J Heat Mass Transfer, 75: 736–753.
Cheung, S. C. P., Yeoh, G. H., Tu, J. Y. 2007a. On the modeling of population balance in isothermal vertical bubbly flows—Average bubble number density approach. Chem Eng Proc, 46: 742–756.
Cheung, S. C. P., Yeoh, G. H., Tu, J. Y. 2007b. On the numerical study of isothermal bubbly flow using two population balance approaches. Chem Eng Sci, 31: 164–1072.
Cheung, S. C. P., Yeoh, G. H., Tu, J. Y. 2008. Population balance modelling of bubbly flows considering the hydrodynamics and thermomechanical processes. AIChE J, 54: 1689–1710.
Choi, K. H., Lee, W. K. 1990. Comparison of probe methods for measurement of bubble properties. Chem Eng Commun, 91: 35–47.
Chu, I. C., Chung, H. J., Lee, S. 2011. Flow-induced vibration of nuclear steam generator u-tubes in two-phase flow. Nucl Eng Des, 241: 1508–1515.
Colombo, M., Fairweather, M. 2016. Accuracy of Eulerian–Eulerian, two-fluid CFD boiling models of subcooled boiling flows. Int J Heat Mass Transfer, 103: 28–44.
Cooper, K. D., Hewitt, G. F., Pinchin, B. 1963. Photography of twophase flow. AERE-R4301.
Corradini, M. L. 1991. Vapor explosion: A review of experiments for accident analysis. Nucl Safety, 32: 337–362.
Corradini, M. L., Kim, B. J., Oh, M. D. 1988. Vapor explosions in light water reactor: A review of theory and modeing. Prog Nucl Energy, 22: 1–117.
Crowe, R. D., Eisenhawer, S. W., McAfee, F. D., Albrecht, R. W. 1977. A study of two-phase flow characteristics using reactor noise techniques. Prog Nucl Energy, 1: 85–97.
De Bertodano, M. L. 1994. Countercurrent gas–liquid flow in a pressurized water reactor hot leg. Nucl Sci Eng, 117: 126–133.
Deendarlianto Höhne, T., Lucas, D., Vallée, C. 2010. Numerical simulation of air–water counter-current two-phase flow in a model of the hot-leg of a pressurized water reactor (PWR). In: Proceeding of the 7th International Conference of the Multiphase Flow.
Deendarlianto Höhne, T., Lucas, D., Vierow, K. 2012. Gas–liquid countercurrent two-phase flow in a PWR led: A comprehensive research review. Nucl Eng Des, 243: 214–233.
Deendarlianto Vallée, C., Lucas, D., Beyer, M., Pietruske, H., Carl, H. 2008. Experimental study on the air/water counter-current flow limitation in a model of the hot leg of a pressurized water reactor. Nucl Eng Des, 238: 3389–3402.
Delhaye, J. M., Achard J. L. 1976. On the averaging operators introduced in two-phase flow modelling. In: Proceedings of the CSNI Specialists Meeting on Transient Two-Phase Flow.
Dominguez-Ontiveros, E. E., Hassan, Y. A., Conner, M. E., Karoutas, Z. 2012. Experimental benchmark data for PWR rod bundle with spacer-grids. Nucl Eng Des, 253: 396–405.
Dominguez-Ontiveros, E., Fortenberry, S., Hassan, Y. A. 2010. Experimental observations of flow modifications in nanofluid boiling utilizing particle image velocimetry. Nucl Eng Des, 240: 299–304.
Doroshchuk, V. E., Levitan, I. L., Lantzman, F. P. 1975. Investigation into Burnout in uniformly heated tubes. ASME Paper 75-WA/HT-22.
Drew, D. A. 1983. Mathematical modeling of two-phase flow. Ann Rev Fluid Mech, 15: 261–291.
Drew, D. A., Passman, S. L. 1999. Theory of Multicomponent Fluids. Springer-Verlag, Berlin.
Feenstra, P., Weaver, D. S., Nakamura, T. 2009. Two-phase flow induced vibration of parallel triangular tube arrays with asymmetric support stiffness. J Press Vess Technol, 131: 1–9.
Fletcher, D. F. 1995. Steam explosion triggering: A review of theoretical and experimental investigation. Nucl Eng Des, 155: 27–36.
Frank, T., Shi, J., Burns, F. A. D. 2004. Validation of eulerian multiphase flow models for nuclear safety application. In: Proceedings of the 3rd Symposium on Two-Phase Modeling and Experimentation.
Geweke, M., Beckmann, H., Mewes, D. 1992. Experimental studies of two-phase flow. In: Proceeding of the European Two-Phase Flow Group Meeting.
Ginox, J. J. 1978. Two-Phase Flows and Heat Transfer with Application to Nuclear Reactor Design Problems. Hemisphere Publishing Corporation.
Glaeser, H. 1992. Downcomer and tie plate countercurrent flow in the upper plenum test facility (UPTF). Nucl Eng Des, 133: 259–283.
Gluck, M. 2007. Sub-channel analysis with F-COBRA-TF-Code validation and approaches to CHF prediction. Nucl Eng Des, 237: 655–667.
Groeneveld, D. C., Leung, L. K. H., Kirillov, P. L., Bobkov, V. P., Smogalov, I. P., Vinogradov, V. N., Huang, X. C., Royer, E. 1996. The 1995 look-up table for critical heat flux in tubes. Nucl Eng Des, 163: 1–23.
Groeneveld, D. C., Shan, J. Q., Vasic, A. Z., Leung, L. K. H., Durmayaz, A., Yang, J., Cheng, S. C., Tanase, A. 2007. The 2006 CHF look-up table. Nucl Eng Des, 237: 1909–1922.
Hewitt, G. F., Roberts, D. N. 1969. Studies of two-phase flow patterns by simultaneous X-ray and flash photography. AERE-M 2159.
Hibiki, T., Ishii, M., Xiao, Z. 2001. Axial interfacial area transport of vertical bubbly flows. Int J Heat Mass Transfer, 44: 1869–1888.
Höhne, T. 2009. Experiments and numerical simulations of horizontal two-phase flow regimes. In: Proceeding of the 7th International Conference on CFD in the Minerals and Process Industries.
Hubbard, M. G., Dukler, A. E. 1966. The characterization of flow regimes for horizontal two-phase flow. In: Proceedings of the 1966 Heat Transfer and Fluid Mechanics Institute, 100–121.
Hibiki, T., Ishii, M. 1999. Experimental study on interfacial area transport in bubbly two phase flows. Int J Heat Mass Transfer, 42: 3019–3035.
Hibiki, T., Ishii, M. 2000. Experimental study on hot-leg U-bend two-phase natural circulation in a loop with a large diameter pipe. Nucl Eng Des, 195: 69–84.
Hibiki, T., Ishii, M. 2002. Development of one-group interfacial area transport equation in bubbly flow systems. Int J Heat Mass Transfer, 45: 2351–2372.
Hibiki, T., Ishii, M. 2003a. One-dimensional drift-flux model and constitutive equations for relative motion between phases in various two-phase flow regimes. Int J Heat Mass Transfer, 46: 4935–4948.
Hibiki, T., Ishii, M. 2003b. One-dimensional drift-flux model for twophase flow in a large diameter pipe. Int J Heat Mass Transfer, 46: 1773–1790.
IAEA-TECDOC-1451. 2005. Innovative small and medium sized reactors: Design features, safety approaches and R&D trends.
Ishida, I., Kusunoki, T., Murata, H., Yokomura, T., Kobayashi, M., Nariai, H. 1990. Thermal hydraulic behavior of a marine reactor under oscillations. Nucl Eng Des, 120: 213–225.
Ishii, M., Hibiki, T. 2011. Thermo-fluid Dynamics of Two-phase Flow, 2nd edn. Springer-Verlag, Berlin.
Jones, O. C., Zuber, N. 1974. Statistical methods for measurement and analysis in two-phase flow. In: Proceedings of the 5th International Heat Transfer Conference, 200–204.
Jones, O. C., Zuber, N. 1975. The interrelation between void fraction fluctuations and flow patterns in two-phase flow. Int J Multiphase Flow, 2: 273–306.
Joseph, D. D., Lundgren, T. S., Jackson, R., Saville, D. A. 1990. Ensemble averaged and mixture theory equations for incompressible fluid-particle suspensions. Int J Multiphase Flow, 16: 35–42.
JSME. 2003. Flow Induced Vibrations—Classification and Lessons from Practical Experiences. Gihondo-Publishing Co. Ltd.
Kanizawa, F. T., Oliveira, L. P. R., Ribatski, G. 2012. State of the art review on flow patterns, superficial void fraction and flow induced vibration during two-phase flows across tube bundles. In: Proceedings of the ASME 2012 Fluids Engineering Division.
Kinoshita, I., Utanohara, Y., Murase, M., Minami, N., Tomiyama, A. 2009. Numerical calculations on countercurrent gas–liquid flow in a PWR hot leg (2) steam–water flow under PWR plant conditions. In: Proceedings of the 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics.
Kolev, N. I. 2005. Multiphase Flow Dynamics 1: Fundamentals, 2nd edn. Springer-Verlag, Berlin.
Kondic, N. D., Lassahn, G. D. 1978. Nonintrusive density distribution measurement in dynamic high temperature systems. In: Proceedings of the 24th International Instrumentation Symposium.
Kosaly, G., Albrecht, R. W., Crowe, R. D., Dailey, D. 1982. Neutronic response to two-phase flow in a nuclear reactor. Prog Nucl Energy, 9: 23–36.
Konno, H., Saito, K. 1985. Identification of non-linear random vibration of the structural components in nuclear reactors. Prog Nucl Energy, 15: 331–339.
Krepper, E., Frank, T., Lucas, D., Prasser, H.-M., Zwart, P. J. 2007a. Inhomogeneous MUSIG model—A population balance approach for polydispersed bubbly flows. In: Proceedings of the 6th International Conference on Multiphase Flow.
Krepper, E., Končar, B., Egorov, Y. 2007b. CFD modelling of subcooled boiling-concept, validation and application to fuel assembly design. Nucl Eng Des, 237: 716–731.
Krepper, E., Rzehak, R., Lifante, C., Frank, T. 2013. CFD for subcooled flow boiling: Coupling wall boiling and population balance models. Nucl Eng Des, 255: 330–346.
Kataoka, I., Ishii, M. 1987. Drift-flux model for large diameter pipe and new correlation for pool void fraction. Int J Heat Mass Transfer, 30: 1927–1939.
Kawaji, M., Thomson, L. A., Krishnan, V. S. 1991. Countercurrent flooding in vertical to inclined pipes. Exp Heat Trans, 4: 95–110.
Kawaji, M., Lotocki, P. A., Krishnan, V. S. 1993. Countercurrent flooding in pipes containing multiple elbows and an orifice. JSME Int Ser B, 36: 397–403.
Kocamustafaogullari, G., Ishii, M. 1995. Foundation of the interfacial area transport equation and its closure relations. Int J Heat Mass Transfer, 38: 481–493.
Krepper, E., Lucas, D., Prasser, H.-M. 2005. On the modelling of bubbly flow in vertical pipes. Nucl Eng Des, 235: 597–611.
Kurul, N., Podowski, M. Z. 1990. Multi-dimensional effects in forced convection sub-cooled boiling. In: Proceedings of the 9th Heat Transfer Conference, 21–26.
Lahey, R. T. Jr., Drew, D. A. 1988. The three-dimensional time and volume averaged conservative equations of two-phase flow. In: Advances in Nuclear Science and Technology. Lewins, J., Becker, M. Eds. Springer, 1–69.
Lassahn, G. D. 1977. LOFT three-beam densitometer data interpretation. EG&G Idaho, Inc., TREE NUREG-1111.
Lay, J. H., Dhir, V. K. 1995. Shape of a vapor steam during nucleate boiling of saturated liquids. Trans ASME J Heat Transfer, 117: 394–401.
Lee, K., Lee, K. H., Lee, J. I., Jeong, Y. H., Lee, P. S. 2013. A new design concept for offshore nuclear power plants with enhanced safety features. Nucl Eng Des, 254: 129–141.
Lee, Y. G., Won, W. Y., Lee, B. A., Kim, S. 2017. A dual conductance sensor for simultaneous measurement of void fraction and structure velocity of downward two-phase flow in a slightly inclined pipe. Sensors, 17: 1063.
Lo, S. M. 1996. Application of population balance to CFD modeling of bubbly flow via the MUSIG model. AEAT-1096, AEA Technology.
Lo, S., Zhang, D. 2009. Modelling of break-up and coalescence in bubbly two-phase flows. J Comput Multiphase Flow, 1: 23–38.
Mandhane, J. M., Gregory, G. A., Aziz, K. 1974. A flow pattern map for gas–liquid flow in horizontal pipes. Int J Multiphase Flow, 1: 537–553.
Merzari, E., Ninokata, H., Baglietto, E. 2007. Unsteady Reynolds averaged Navier–Stokes simulation for an accurate prediction of the flow insight tight rod bundles. In: Proceedings of the 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics.
Merzari, E., Ninokata, H., Baglietto, E. 2008. Numerical simulation of flows in tight lattice fuel bundles. Nucl Eng Des, 238: 1703–1719.
Minami, N., Murase, M., Nishiwaki, D., Tomiyama, A. 2008. Countercurrent gas–liquid flow in a rectangular channel simulating a PWR hot leg (2): Analytical evaluation of counter-current flow limitation. Jpn J Multiphase Flow, 22: 413–422.
Minami, N., Murase, M., Tomiyama, A. 2010. Countercurrent gas–liquid flow in a PWR hot leg under reflux cooling (II) numerical simulation of 1/15-scale air–water tests. J Nucl Sci Tech, 47: 149–155.
Minami, N., Utanohara, Y., Kinoshita, I., Murase, M., Tomiyama, A. 2009. Numerical calculations on countercurrent gas–liquid flow in a PWR hot leg (1) air–water flow in a 1/15-scale model. In: Proceeding of the 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics.
Mishima, K., Ishii, M. 1984. Flow regime transition criteria for upward two-phase flow in vertical tubes. Int J Heat Mass Transfer, 27: 723–737.
Mitra, D., Dhir, V. K., Catton, I. 2009. Fluid elastic instability in tube arrays subjected to air-water and steam-water cross-flow. J Fluids Struct, 25: 1213–1235.
Miwa, S., Liu, Y., Hibiki, T., Ishii, M., Kondo, Y., Morita, H., Tanimoto, K. 2014. Two-phase flow induced vibration. In: Proceedings of the 22nd International Conference on Nuclear Engineering.
Moorthi, A., Sharma, A. K., Velusamy, K. 2018. A review of subchannel thermal hydraulic codes for nuclear reactor core and future directions. Nucl Eng Des, 332: 329–344.
Morel, C., Laviéville, J. M. 2009. Modeling of multisize bubbly flow and application to the simulation of boiling flows with the Neptune CFD code. Sci Tech Nucl Installation, 2009: 953527.
Murase, M., Utanohara, Y., Kinoshita, I., Minami, N., Tomiyama, A. 2009. Numerical calculations on countercurrent air–water flow in small-scale models of a PWR hot leg using a VOF model. In: Proceeding of the 17th International Conference on Nuclear Engineering.
Nariai, T., Tomiyama, A., Vallee, C., Lucas, D., Murase, M. 2010. Countercurrent flow limitation in a scale-down model of a PWR hot leg. In: Proceeding of the 8th International Topical meeting on Nuclear Thermal-Hydraulics, Operation and Safety.
Ninokata, H., Atake, N., Baglietto, E., Misawa, T., Kano, T. 2004. Direct numerical simulation of turbulence flows in a subchannel of tight lattice fuel pin bundles of nuclear reactors. Available at http://www.Jamstec.go.jp/esc/publication/annual/annual2004/.
Ninokata, H., Merzari, E. 2007. Computational fluid dynamics and simulation based-design approach for tight lattice nuclear fuel pin subassemblies. NURETH-12.
Nusret, A. 2007. Selected examples of natural circulation for small break LOCA and some severe accidents. IAEA course on natural circulation in water-cooled nuclear power plants, International Centre for Theoretical Physics, Trieste, Italy.
Ohnuki, A., Adachi, H., Murao, Y. 1988. Scale effects on counter current gas–liquid flow in a horizontal tube connected to an inclined riser. Nucl Eng Des, 107: 283–294.
Ohnuki, A., Akimoto, H. 2000. Experimental study on transition of flow pattern and phase distribution in upward air-water twophase flow along a large vertical pipe. Int J Multiphase Flow, 26: 367–386.
Okano, Y., Koshizuka, S., Oka, Y. 1997. Safety analysis of a supercritical pressure, light water cooled and moderated reactor with double tube water rods. Ann Nucl Energy, 24: 1447–1456.
Olmos, E., Gentric, C., Vial, Ch., Wild, G., Midoux, N. 2001. Numerical simulation of multiphase flow in bubble column reactors. Influence of bubble coalescence and break-up. Chem Eng Sci, 56: 6359–6365.
Panton, R. J. 1968. Flow properties for the continuum viewpoint of a non-equilibrium gas particle mixture. J Fluid Mech, 31: 273–304.
Patel, P., Theofanous, T. G. 1976. Universal relations for bubble growth. Int J Heat Mass Transfer, 19: 425–429.
Pettigrew, M. J., Taylor, C. E. 1994. Two-phase flow-induced vibration: An overview. J Press Vess Technol, 116: 233–253.
Pettigrew, M. J., Taylor, C. E., Fisher, N. J., Yetisir, M., Smith, B. A. W. 1998. Flow-induced vibration: Recent findings and open questions. Nucl Eng Des, 185: 249–276.
Pham, Q. T., Kim, T. I., Lee, S. S., Chang, S. H. 2012. Enhancement of critical heat flux using nanofluids for invessel retention-external vessel cooling. Appl Thermal Eng, 35: 157–165.
Piper, T. C. 1974. Dynamic gamma attenuation density measurements. Aerojet Nuclear Co., ANCR-1160.
Pochorecki, R., Moniuk, W., Bielski, P., Zdrojkwoski, A. 2001. Modeling of the coalescence/redispersion processes in bubble columns. Chem Eng Sci, 56: 6157–616.
Pontaza, J. P., Menon, R. G. 2011. Flow-induced vibrations of subsea jumpers due to internal multi-phase flow. In: Proceedings of the 30th International Conference on Ocean, Offshore and Arctic Engineering.
Prasser, H. M., Böttger, A., Zschau, J. 1998. A new electrode-mesh tomograph for gas liquid flows. Flow Meas Instrum, 9: 111–119.
Prassinos, P. G., Liao C. K. 1979. An investigation of two-phase flow regimes in LOFT piping during loss-of-coolant experiments. EG&G Idaho, Inc., NUREG/CR-0606, TREE-1244.
Pasamehmetoglu, K. O., Gunnerson, F. S. 1985. Theoretical considerations of transient critical heat flux. In: Proceedings of the 3rd International Topical Meeting on Reactor-Thermal Hydraulics, 2: Paper 18-F.
Pasamehmetoglu, K. O., Nelson, R. A., Gunnerson, F. S. 1987. A theoretical prediction of critical heat flux in saturated pool boiling during power transients. Nonequilibrium Transport Phenomena, ASME, New York, HTD-77: 57–64.
Prasad, G. V. D., Pandey, M., Kalra, M. S. 2007. Review of research on flow instabilities in natural circulation boiling systems. Prog Nucl Energy, 49: 429–451.
Rahim, R. A., Rahiman, M. F., Chan, K., Nawawi, S. 2007. Non-invasive imaging of liquid/gas flow using ultrasonic transmission-mode tomograph. Sens Actuators A, 135: 337–345.
Roelofs, F., Gopala, V. R., Jayaraju, S., Shams, A., Konen, E. 2013. Review of fuel assembly and pool thermal hydraulics for fast reactors. Nucl Eng Des, 265: 1205–1222.
Rouhani, S. Z., Sohal, M. S. 1983. Two-phase flow patterns: A review of research letters. Prog Nucl Energy, 11: 219–259.
Rowinski, M. K., Zhao, J., White, T. J., Soh, Y. C. 2018. Safety analysis of super-critical water reactors—A review. Prog Nucl Energy, 106: 87–101.
Sasakawa, T., Serizawa, A., Kawara, Z. 2005. Fluid-elastic vibration in two-phase cross flow. Exp Thermal Fluid Sci, 29: 403–413.
Schlegel, J. P., Hibiki, T., Shen, X., Appathurai, S., Subramani, H. 2017. Prediction of interfacial area transport in a coupled two fluid model computation. J Nucl Sci Technol, 54: 58–73.
Schlegel, J. P., Macke, C., Hibiki, T., Ishii, M. 2013. Modified distribution parameter for churn-turbulent flows in large diameter channels. Nucl Eng Des, 263: 138–150.
Schlegel, J. P., Miwa, S., Chen, S., Hibiki, T., Ishiii, M. 2012. Experimental study of two phase flow structure in large diameter pipes. Exp Therm Fluid Sci, 41: 12–22.
Schlegel, J. P., Sawant, P., Paranjape, S., Ozar, B., Hibiki, T., Ishii, M. 2009. Void fraction and flow regime in adiabatic upward twophase flow in large diameter vertical pipes. Nucl Eng Des, 239: 2864–2874.
Seidel, T., Vallée, C., Lucas, D., Beyer, M., Deendarlianto. 2010. Two-phase flow experiments in a model of the hot leg of a pressurised water reactor. Wissenschaftlich-Technische Berichte/Forschungszentrum Dresden-Rossendorf, FZD-531.
Sha, W. T. 1980. An overview of rod-bundle thermal-hydraulic analysis. Nucl Eng Des, 62: 1–24.
Shen, X., HIbiki, T. 2013. One-group interfacial area transport equation and its sink and source terms in narrow rectangular channel. Int J Heat Fluid Flow, 44: 312–326.
Shen, X., Mishima, K., Nakamura, H. 2010a. Measurement of twophase flow in a vertical large diameter pipe using hot-film anemometer. Jpn J Multi-phase Flow, 23: 605–613. (in Japanese)
Shen, X., Mishima, K., Nakamura, H. 2010b. Flow-induced void fraction transition phenomenon in two-phase flow. In: Proceedings of the 18th International Conference on Nuclear Engineering.
Shen, X., Saito, Y., Mishima, K., Nakamura, H. 2006. A study on the characteristics of upward air-water two-phase flow in a large pipe. Exp Therm Fluid Sci, 31: 21–36.
Shen, X., Schlegel, J. P., Hibiki, T., Nakamura, H. 2018. Some characteristics of gas–liquid two-phase flow in vertical largediameter channels. Nucl Eng Des, 333: 87–98.
Shi, J. M., Zwart, P. J., Frank, T., Rohde, U., Prasser, H. M. 2004. Development of a multiple velocity multiple size group model for poly-dispersed multiphase flows. Annual Report of Institute of Safety Research, Forschungszentrum Rossendorf, Germany.
Siddiqui, H., Banerjee, S., Ardron, K. H. 1986. Flooding in an elbow between a vertical and a horizontal or near horizontal pipe, Part I. Experiments. Int J Multiphase Flow, 12: 531–541.
Smith, T. R. 2002. Two-group interfacial area transport equation in large diameter pipes. Ph.D. Thesis. Purdue University.
Smith, T. R., Schlegel, J. P., Hibiki, T., Ishii, M. 2012. Mechanistic modelling of interfacial area transport in large diameter pipes. Int J Multiphase Flow, 47: 1–16.
Son, G., Dhir, V. K. 2008. Numerical simulation of nucleate boiling on a horizontal surface at high heat fluxes. Int J Heat Mass Transfer, 51: 2566–2582.
Son, G., Dhir, V. K., Ramanujapu, N. 1999. Dynamics and heat transfer associated with a single bubble during nucleate boiling on a horizontal surface. Trans ASME J Heat Transfer, 121: 623–631.
Sorokin, A. P., Efanov, A. D., Ivanov, E. F., Martsinyuk, D. E., Bogoslovskaya, G. P. 1999. Heat transfer during boiling of a liquid metal during emergency cool down of a fast neutron reactor. At Energy, 87: 801–807.
Staub, F. W., Zuber, N. 1964. A program of two-phase flow investigation. General Electric Co., Report, EURAEC 1171, GEAP 4631.
Subbotin, V., Sorokin, D., Kudryavtsev, A. 1970. Generalized relationship for calculating heat transfer in the developed boiling of alkali metals. At Energy, 29: 730–731.
Sun, X., Smith, T. R., Kim, S., Ishii, M., Uhle, J. 2002. Interfacial area of bubbly flow in a relatively large diameter pipe. Exp Therm Fluid Sci, 27: 97–109.
Taitel, Y., Barnea, D., Dukler, A. E. 1980. Modeling flow pattern transitions for steady upward gas–liquid flow in vertical tubes. AIChE J, 26: 345–354.
Taitel, Y., Dukler, A. E. 1976. A model for predicting flow regime transition iii horizontal and near horizontal gas-liquid flow. AIChE J, 22: 47–55.
Takada, N., Misawa, M., Tomiyama, A., Fujiwara, S. 2000. Numerical simulation of two- and three-dimensional two-phase fluid motion by lattice Boltzmann method. Comp Phys Comm, 129: 233–246.
Taylor, C. E., Pettigrew, M. J. 2001. Effect of flow regime and void fraction on tube bundle vibration. J Press Vess Technol, 123: 407–413.
Thakrar, R., Murallidharan, J., Walker, S. P. 2015. Simulations of high-pressure subcooled boiling flows in rectangular channels. In: Proceedings of the 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics.
Theofanous, T. G. 1980. The boiling crisis in nuclear reactor safety and performance. Int J Multiphase Flow, 6: 69–95.
Theofanous, T. G., Boher, T. H., Chang, M. C., Patel, P. 1978. Experiments and universal growth relations for vapor bubbles with micro-layers. J Heat Transfer, 100: 41–48.
Theofanous, T. G., Liu, C., Additon, S., Angelini, S., Kynalainen, O., Salmassi, T. 1994. In-vessel coolability and retention of a core melt. DOE/ID-l0460, Vol. 1: Peer Review Version.
Theofanous, T. G., Liu, C., Additon, S., Angelini, S., Kynalainen, O., Salmassi, T. 1995. In-vessel coolability and retention of a core melt. DOE/ID-10460, Vol. 2: Peer Re-Review Version.
Thomas, S., Narayanan, C., Lakehal, D. 2013. Progress in modelling convective boiling flows using the n-phase approach in TransAT. In: Proceedings of the 15th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics.
Todreas, N. E., Kazimi, M. S. 2001. Nuclear Systems II-Elements of Thermal Hydraulic Design. Taylor and Francis.
Tomiyama, A. 1998. Struggle with computational bubble dynamics. Multiphase Sci Tech, 10: 369–405.
Tong, L. S., Hewitt, G. F. 1972. Overall view point of flow boiling CHF mechanisms. ASME Paper 72-HT-54.
Tu, J. Y., Yeoh, G. H. 2002. On numerical modeling of low-pressure subcooled boiling flows. Int J Heat Mass Transfer, 45: 1197–1209.
Tu, J. Y., Yeoh, G. H., Park, G.-C., Kim, M.-O. 2005. On population balance approach for subcooled boiling flow prediction. ASME J Heat Transfer, 127: 253–264.
Utanohara, Y., Kinoshita, I., Murase, M., Minami, N., Tomiyama, A. 2009. Effects of interfacial friction correlations on numerical calculations for countercurrent gas–liquid flow in a PWR hot leg. In: Proceeding of the 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics.
Vernier, P., Delhaye, J. M. 1968. General two-phase flow equation applied to the thermohydrodynamics of boiling nuclear reactors. Acta Tech Belg Energie Primaire, 4: 3–43.
Vince, M. A., Lahey, R. T. Jr. 1980. Flow regime identification and void fraction measurement techniques in two-phase flow. Rensselaer Polytechnic Institute, NUREG/CR-1692.
Wang, M. J., Mayinger, F. 1995. Simulation and analysis of thermalhydraulic phenomena in a PWR hot leg related SBLOCA. Nucl Eng Des, 155: 643–652.
Weisman, J., Duncan, D., Gibson, J., Crawford, T. 1979. Effects of fluid properties and pipe diameter on two-phase flow patterns in horizontal lines. Int J Multiphase Flow, 5: 437–462.
Weisman, J., Kang, S. Y. 1981. Flow pattern transitions in vertical and upwardly inclined tubes. Int J Multiphase Flow, 7: 271–291.
Wu, J. M., Zhao, J. 2013. A review of nanofluid heat transfer and critical heat flux enhancement—Research gap to engineering application. Prog Nucl Energy, 66: 13–24.
Wu, Q., Kim, S., Ishii, M., Beus, S. G. 1998. One-group interfacial area transport in vertical bubbly flow. Int J Heat Mass Transfer, 41: 1103–1112.
Wu, Y. W., Luo, S., Wang, L., Hou, Y., Su, G. H., Tuan, W., Qiu, S. 2018. Review on heat transfer and flow characteristics of liquid sodium (2): Two-phase. Prog Nucl Energy, 103: 151–164.
Wulff, W. 2011. Critical review of conservation equations for twophase flow in the US NRC TRACE code. Nucl Eng Des, 241: 4237–4260.
Yadigaroglu, G. 2014. CMFD and the critical-heat-flux grand challenge in nuclear thermal-hydraulics—A letter to the editor of this special issue. Int J Multiphase Flow, 67: 3–12.
Yadigaroglu, G., Lahey, R. T. Jr. 1976. On the various forms of the conservation equations in two-phase flow. Int J Multiphase Flow, 2: 477–494.
Yamaji, A., Oka, Y., Ishiwatari, Y., Liu, J., Suzuki, M. 2006. Principle of rationalizing the criteria for abnormal transients of the Super LWR with fuel rod analyses. Ann Nucl Energy, 33: 984–993.
Yamano, H., Tanaka, M., Kimura, N., Ohshima, H., Kamide, H., Watanabe, O. 2011. Development of flow induced vibration evaluation methodology for large diameter piping with elbow in Japan sodium-cooled fast reactor. Nucl Eng Des, 241: 4464–4475.
Yan, B. H. 2017. Review of the nuclear reactor thermal hydraulic research in ocean motions. Nucl Eng Des, 313: 370–385.
Yao, W., Morel, C. 2004. Volumetric interfacial area prediction in upwards bubbly two-phase flow. Int J Heat Mass Transfer, 47: 307–328.
Yeoh, G. H., Cheung, S. C. P., Tu, J. Y., Ho, M. K. M. 2008. Fundamental consideration of wall heat partition of vertical subcooled boiling flows. Int J Heat Mass Transfer, 51: 3840–3853.
Yeoh, G. H., Tu, J. Y. 2004. Population balance modelling for bubbly flows with heat and mass transfer. Chem Eng Sci, 59: 3125–3139.
Yeoh, G. H., Tu, J. Y. 2005. Thermal-hydrodynamic modelling of bubbly flows with heat and mass transfer. AIChE J, 51: 8–27.
Yeoh, G. H., Tu, J. Y. 2006a. Numerical modelling of bubbly flows with and without heat and mass transfer. Appl Math Model, 30: 1067–1095.
Yeoh, G. H., Tu, J. Y. 2006b. Two-fluid and population balance models for subcooled boiling flow. Appl Math Model, 30: 1370–1391.
Yeoh, G. H., Tu, J. Y. 2010. Computational Techniques for Multiphase Flows. Elsevier Science and Technology.
Yeoh, G. H., Tu, J. Y. 2017. Basic theory and conceptual framework of multiphase flows. In: Handbook of Multiphase Flow Science and Technology. Yeoh, G. H. Ed. Springer Science: 1–47.
Yeoh, G. H., Vahaji, S., Cheung, S. C. P., Tu, J. Y. 2014. Modeling subcooled flow boiling in vertical channels at low pressures—Part 2: Evaluation of mechanistic approach. Int J Heat Mass Transfer, 75: 754–768.
Zhang, C., Pettigrew, M. J., Mureithi, N. W. 2007. Vibration excitation force measurements in a rotated triangular tube bundle subjected to two-phase cross flow. J Press Vess Technol, 129: 21–27.
Zhou, L., Ge, C., Zan, Y. F., Yan, X., Chen, B. D. 2015. Study on generation expression for liquid force per unit mass under noninertial reference frame. Nucl Power Eng, 2: 37–41.
The original version of this article was revised due to a retrospective Open Access order.
About this article
Cite this article
Yeoh, G.H. Thermal hydraulic considerations of nuclear reactor systems: Past, present and future challenges. Exp. Comput. Multiph. Flow 1, 3–27 (2019). https://doi.org/10.1007/s42757-019-0002-5
- nuclear reactor systems thermal hydraulic analysis critical heat flux