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Review Article | Open Access

Thermal hydraulic considerations of nuclear reactor systems: Past, present and future challenges

School of Mechanical and Manufacturing Engineering, University of New South Wales, Sydney, NSW 2052, Australia
Australian Nuclear Science and Technology Organisation (ANSTO), Locked Bag 2001, Kirrawee DC, NSW 2232, Australia
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An erratum to this article is available online at:
https://doi.org/10.1007/s42757-020-0093-z

Abstract

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 covering various aspects of experimental, analytical and computational approaches.

Keywords

nuclear reactor systemsthermal hydraulic analysiscritical heat flux

References

 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
Anglart, H., Caraghiaur, D. 2011. CFD modeling of boiling annular- mist flow for dryout investigations. Multiphase Sci Technol, 23: 22–251.
Crossref Google Scholar
 
Anglart, H., Nylund, O. 1996. CFD application to prediction of void distribution in bubbly flows in rod bundles. Nucl Eng Des, 163: 81–98.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
Baglietto, E., Ninokata, H., Misawa, T. 2006. CFD and DNS methodologies development for fuel bundle simulations. Nucl Eng Des, 236: 1503–1510.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
Bankoff, S. G., Lee, S. C. 1985. A brief review of countercurrent flooding models applicable to PWR geometries. Nucl Safety, 26: 139–152.
Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
Bestion, D. 2014. The difficult challenge of a two-phase CFD modelling for all flow regimes. Nucl Eng Des, 279: 116–125.
Crossref Google Scholar
 
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.
Crossref
 
Bourne, J. A., Bergles, A. E., Tong, L. S. 1973. Review of two-phase flow instabilities. Nucl Eng Des, 25: 165–192.
Crossref Google Scholar
 
Boyer, C., Duquenne, A. M., Wild, G. 2002. Measuring techniques in gas–liquid and gas–liquid–solid reactors. Chem Eng Sci, 57: 3185–3215.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Google Scholar
 
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..
Crossref Google Scholar
 
Choi, K. H., Lee, W. K. 1990. Comparison of probe methods for measurement of bubble properties. Chem Eng Commun, 91: 35–47.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
Cooper, K. D., Hewitt, G. F., Pinchin, B. 1963. Photography of two- phase flow. AERE-R4301.
Crossref
 
Corradini, M. L. 1991. Vapor explosion: A review of experiments for accident analysis. Nucl Safety, 32: 337–362.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
De Bertodano, M. L. 1994. Countercurrent gas–liquid flow in a pressurized water reactor hot leg. Nucl Sci Eng, 117: 126–133.
Crossref Google Scholar
 
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.
Crossref
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
Doroshchuk, V. E., Levitan, I. L., Lantzman, F. P. 1975. Investigation into Burnout in uniformly heated tubes. ASME Paper 75-WA/ HT-22.
Crossref
 
Drew, D. A. 1983. Mathematical modeling of two-phase flow. Ann Rev Fluid Mech, 15: 261–291.
Crossref Google Scholar
 
Drew, D. A., Passman, S. L. 1999. Theory of Multicomponent Fluids. Springer-Verlag, Berlin.
Crossref
 
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.
Crossref Google Scholar
 
Fletcher, D. F. 1995. Steam explosion triggering: A review of theoretical and experimental investigation. Nucl Eng Des, 155: 27–36.
Crossref Google Scholar
 
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.
Crossref
 
Geweke, M., Beckmann, H., Mewes, D. 1992. Experimental studies of two-phase flow. In: Proceeding of the European Two-Phase Flow Group Meeting.
Crossref
 
Ginox, J. J. 1978. Two-Phase Flows and Heat Transfer with Application to Nuclear Reactor Design Problems. Hemisphere Publishing Corporation.
Crossref
 
Glaeser, H. 1992. Downcomer and tie plate countercurrent flow in the upper plenum test facility (UPTF). Nucl Eng Des, 133: 259–283.
Crossref Google Scholar
 
Gluck, M. 2007. Sub-channel analysis with F-COBRA-TF -Code validation and approaches to CHF prediction. Nucl Eng Des, 237: 655–667.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
Hewitt, G. F., Roberts, D. N. 1969. Studies of two-phase flow patterns by simultaneous X-ray and flash photography. AERE-M 2159.
Crossref
 
Hibiki, T., Ishii, M., Xiao, Z. 2001. Axial interfacial area transport of vertical bubbly flows. Int J Heat Mass Transfer, 44: 1869–1888.
Crossref Google Scholar
 
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.
Crossref
 
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.
Crossref
 
Hibiki, T., Ishii, M. 1999. Experimental study on interfacial area transport in bubbly two phase flows. Int J Heat Mass Transfer, 42: 3019–3035.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Google Scholar
 
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.
Crossref Google Scholar
 
Hibiki, T., Ishii, M. 2003b. One-dimensional drift-flux model for two- phase flow in a large diameter pipe. Int J Heat Mass Transfer, 46: 1773–1790.
Crossref Google Scholar
 
IAEA-TECDOC-1451. 2005. Innovative small and medium sized reactors: Design features, safety approaches and R&D trends.
Crossref
 
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.
Crossref Google Scholar
 
Ishii, M., Hibiki, T. 2011. Thermo-fluid Dynamics of Two-phase Flow, 2nd edn. Springer-Verlag, Berlin.
Crossref
 
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.
Crossref
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
JSME. 2003. Flow Induced Vibrations—Classification and Lessons from Practical Experiences. Gihondo-Publishing Co. Ltd.
Crossref
 
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.
Crossref
 
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.
Crossref
 
Kolev, N. I. 2005. Multiphase Flow Dynamics 1: Fundamentals, 2nd edn. Springer-Verlag, Berlin.
Crossref
 
Kondic, N. D., Lassahn, G. D. 1978. Nonintrusive density distribution measurement in dynamic high temperature systems. In: Proceedings of the 24th International Instrumentation Symposium.
Crossref
 
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.
Crossref Google Scholar
 
Konno, H., Saito, K. 1985. Identification of non-linear random vibration of the structural components in nuclear reactors. Prog Nucl Energy, 15: 331–339.
Crossref Google Scholar
 
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.
Crossref
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
Kawaji, M., Thomson, L. A., Krishnan, V. S. 1991. Countercurrent flooding in vertical to inclined pipes. Exp Heat Trans, 4: 95–110.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
Kocamustafaogullari, G., Ishii, M. 1995. Foundation of the interfacial area transport equation and its closure relations. Int J Heat Mass Transfer, 38: 481–493.
Crossref Google Scholar
 
Krepper, E., Lucas, D., Prasser, H.-M. 2005. On the modelling of bubbly flow in vertical pipes. Nucl Eng Des, 235: 597–611.
Crossref Google Scholar
 
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.
Crossref
 
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.
Crossref
 
Lassahn, G. D. 1977. LOFT three-beam densitometer data interpretation. EG&G Idaho, Inc., TREE NUREG-1111.
Crossref
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
Lo, S. M. 1996. Application of population balance to CFD modeling of bubbly flow via the MUSIG model. AEAT-1096, AEA Technology.
Crossref
 
Lo, S., Zhang, D. 2009. Modelling of break-up and coalescence in bubbly two-phase flows. J Comput Multiphase Flow, 1: 23–38.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref
 
Merzari, E., Ninokata, H., Baglietto, E. 2008. Numerical simulation of flows in tight lattice fuel bundles. Nucl Eng Des, 238: 1703–1719.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref
 
Moorthi, A., Sharma, A. K., Velusamy, K. 2018. A review of sub- channel thermal hydraulic codes for nuclear reactor core and future directions. Nucl Eng Des, 332: 329–344.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref
 
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.
Crossref
 
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/.
Crossref
 
Ninokata, H., Merzari, E. 2007. Computational fluid dynamics and simulation based-design approach for tight lattice nuclear fuel pin subassemblies. NURETH-12.
Crossref
 
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.
Crossref
 
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.
Crossref Google Scholar
 
Ohnuki, A., Akimoto, H. 2000. Experimental study on transition of flow pattern and phase distribution in upward air-water two- phase flow along a large vertical pipe. Int J Multiphase Flow, 26: 367–386.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
Panton, R. J. 1968. Flow properties for the continuum viewpoint of a non-equilibrium gas particle mixture. J Fluid Mech, 31: 273–304.
Crossref Google Scholar
 
Patel, P., Theofanous, T. G. 1976. Universal relations for bubble growth. Int J Heat Mass Transfer, 19: 425–429.
Crossref Google Scholar
 
Pettigrew, M. J., Taylor, C. E. 1994. Two-phase flow-induced vibration: An overview. J Press Vess Technol, 116: 233–253.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Google Scholar
 
Piper, T. C. 1974. Dynamic gamma attenuation density measurements. Aerojet Nuclear Co., ANCR-1160.
Crossref
 
Pochorecki, R., Moniuk, W., Bielski, P., Zdrojkwoski, A. 2001. Modeling of the coalescence/redispersion processes in bubble columns. Chem Eng Sci, 56: 6157–616.
Crossref Google Scholar
 
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.
Crossref
 
Prasser, H. M., Böttger, A., Zschau, J. 1998. A new electrode-mesh tomograph for gas liquid flows. Flow Meas Instrum, 9: 111–119.
Crossref Google Scholar
 
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.
Crossref
 
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.
Crossref
 
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.
Crossref
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
Rouhani, S. Z., Sohal, M. S. 1983. Two-phase flow patterns: A review of research letters. Prog Nucl Energy, 11: 219–259.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
Sasakawa, T., Serizawa, A., Kawara, Z. 2005. Fluid-elastic vibration in two-phase cross flow. Exp Thermal Fluid Sci, 29: 403–413.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
Schlegel, J. P., Sawant, P., Paranjape, S., Ozar, B., Hibiki, T., Ishii, M. 2009. Void fraction and flow regime in adiabatic upward two- phase flow in large diameter vertical pipes. Nucl Eng Des, 239: 2864–2874.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
Sha, W. T. 1980. An overview of rod-bundle thermal-hydraulic analysis. Nucl Eng Des, 62: 1–24.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
Shen, X., Mishima, K., Nakamura, H. 2010a. Measurement of two- phase flow in a vertical large diameter pipe using hot-film anemometer. Jpn J Multi-phase Flow, 23: 605–613. (in Japanese)
Crossref Google Scholar
 
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.
Crossref
 
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.
Crossref Google Scholar
 
Shen, X., Schlegel, J. P., Hibiki, T., Nakamura, H. 2018. Some characteristics of gas–liquid two-phase flow in vertical large- diameter channels. Nucl Eng Des, 333: 87–98.
Crossref Google Scholar
 
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.
Crossref
 
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.
Crossref Google Scholar
 
Smith, T. R. 2002. Two-group interfacial area transport equation in large diameter pipes. Ph.D. Thesis. Purdue University.
Crossref
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
Staub, F. W., Zuber, N. 1964. A program of two-phase flow investigation. General Electric Co., Report, EURAEC 1171, GEAP 4631.
Crossref
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref
 
Theofanous, T. G. 1980. The boiling crisis in nuclear reactor safety and performance. Int J Multiphase Flow, 6: 69–95.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref
 
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.
Crossref
 
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.
Crossref
 
Todreas, N. E., Kazimi, M. S. 2001. Nuclear Systems II-Elements of Thermal Hydraulic Design. Taylor and Francis.
Crossref
 
Tomiyama, A. 1998. Struggle with computational bubble dynamics. Multiphase Sci Tech, 10: 369–405.
Crossref Google Scholar
 
Tong, L. S., Hewitt, G. F. 1972. Overall view point of flow boiling CHF mechanisms. ASME Paper 72-HT-54.
Crossref
 
Tu, J. Y., Yeoh, G. H. 2002. On numerical modeling of low-pressure subcooled boiling flows. Int J Heat Mass Transfer, 45: 1197–1209.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref
 
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.
Crossref Google Scholar
 
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.
Crossref
 
Wang, M. J., Mayinger, F. 1995. Simulation and analysis of thermal- hydraulic phenomena in a PWR hot leg related SBLOCA. Nucl Eng Des, 155: 643–652.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
Weisman, J., Kang, S. Y. 1981. Flow pattern transitions in vertical and upwardly inclined tubes. Int J Multiphase Flow, 7: 271–291.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
Wulff, W. 2011. Critical review of conservation equations for two- phase flow in the US NRC TRACE code. Nucl Eng Des, 241: 4237–4260.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
Yan, B. H. 2017. Review of the nuclear reactor thermal hydraulic research in ocean motions. Nucl Eng Des, 313: 370–385.
Crossref Google Scholar
 
Yao, W., Morel, C. 2004. Volumetric interfacial area prediction in upwards bubbly two-phase flow. Int J Heat Mass Transfer, 47: 307–328.
Crossref Google Scholar
 
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.
Crossref Google Scholar
 
Yeoh, G. H., Tu, J. Y. 2004. Population balance modelling for bubbly flows with heat and mass transfer. Chem Eng Sci, 59: 3125–3139.
Google Scholar
 
Yeoh, G. H., Tu, J. Y. 2005. Thermal-hydrodynamic modelling of bubbly flows with heat and mass transfer. AIChE J, 51: 8–27.
Google Scholar
 
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.
Google Scholar
 
Yeoh, G. H., Tu, J. Y. 2006b. Two-fluid and population balance models for subcooled boiling flow. Appl Math Model, 30: 1370–1391.
Google Scholar
 
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.
Google Scholar
 
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.
Google Scholar
 
Zhou, L., Ge, C., Zan, Y. F., Yan, X., Chen, B. D. 2015. Study on generation expression for liquid force per unit mass under non- inertial reference frame. Nucl Power Eng, 2: 37–41.
Google Scholar
Experimental and Computational Multiphase Flow
Volume 1 Issue 1,
March 2019
Pages 3-27
DOI: 10.1007/s42757-019-0002-5
Cite this article:
Heng Yeoh G. Thermal hydraulic considerations of nuclear reactor systems: Past, present and future challenges. Experimental and Computational Multiphase Flow, 2019, 1(1): 3-27. https://doi.org/10.1007/s42757-019-0002-5

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Received: 23 December 2018
Revised: 31 January 2019
Accepted: 31 January 2019
Published: 05 March 2019
© Tsinghua University Press 2019
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