Development of one-dimensional two-fluid model with consideration of void fraction covariance effect

References

• Boyack B, Duffey R, Griffith P, et al. Quantifying reactor safety margins: application of code scaling, applicability, and uncertainty evaluation methodology to a large-break, loss-of-coolant accident. Washington (DC): US NRC; 1989. (NUREG/CR-5249/EGG-2552).
   Google Scholar
• American Society of Mechanical Engineers. Standard for verification and validation in computational fluid dynamics and heat transfer. New York (NY): ASME; 2009. ( V&V 20-2009).
   Google Scholar
• Atomic Energy Society of Japan. Guideline for credibility assessment of nuclear simulations: 2015. Tokyo: AESJ; 2016. ( AESJ-SC-A008:2015). Japanese.
   Google Scholar
• US Nuclear Regulation Committee. Transient and accident analysis methods. Washington (DC): US NRC; 2005. ( US NRC Regulatory Guide 1.203).
   Google Scholar
• Atomic Energy Society of Japan. Standard method for safety evaluation using best estimate code based on uncertainty and scaling analyses with statistical approach: 2008. Tokyo: AESJ; 2009. ( AESJ-SC-S001:2008). Japanese.
   Google Scholar
• US Nuclear Regulation Committee. TRACE/V5.0 theory manual; Field equations, solutions methods, and physical models. Washington (DC): US NRC; 2008.
   Google Scholar
• Information System Laboratories. RELAP5/MOD3.3 code manual volume IV: models and correlations. Washington (DC): US NRC; 2001. (NUREG/CR-5535/Rev 1-Vol.IV).
   Google Scholar
• Borkowski JA., Wade NL. TRAC-BF1/MOD1 models and correlations. Washington (DC): US NRC; 1992. (NUREG/CR-4391/EGG-2680R4).
   Google Scholar
• Andersen JGM, Chu KH. BWR refill-reflood program constitutive correlations for shear and heat transfer for the BWR version of TRAC. Washington (DC): US NRC; 1983. (NURG/CR-2134/GEAP-24940).
   Google Scholar
• Ishii M. One-dimensional drift-flux model and constitutive relations for relative motion between phases in various two-phase flow regimes. Argonne (IL): ANL; 1977. (ANL-77-47).
   Google Scholar
• Ozaki T, Suzuki R, Mashiko H, et al. Development of drift-flux model based on 8×8 BWR rod bundle geometry experiments under prototypic temperature and pressure conditions. J Nucl Sci Technol. 2013;50:563–580.
   Web of Science ®Google Scholar
• Ozaki T, Hibiki T. Drift-flux model for rod bundle geometry. Prog Nucl Energy. 2015;83:229–247.
   Web of Science ®Google Scholar
• Brooks CS, Ozar B, Hibiki T, et al. Two-group drift-flux model in boiling flow. Int J Heat Mass Transfer. 2012;55:6121–6129.
   Web of Science ®Google Scholar
• Brooks CS, Liu Y, Hibiki T, et al. Effect of void fraction covariance on relative velocity in gas-dispersed two-phase flow. Prog Nucl Energy. 2014;70:209–220.
   Web of Science ®Google Scholar
• Ishii M, Mishima K. Two-fluid model and hydrodynamic constitutive relations. Nucl Eng Des. 1984;82:107–126.
   Web of Science ®Google Scholar
• Hibiki T, Ozaki T. Modeling of distribution parameter, void fraction covariance and relative velocity covariance for upward steam-water boiling flow in vertical pipe. Int J Heat Mass Transfer. 2017;112:620–629.
   Web of Science ®Google Scholar
• Ozaki T, Hibiki T. Modeling of distribution parameter, void fraction covariance and relative velocity covariance for upward steam-water boiling flow in vertical rod bundle. J Nucl Sci Technol. 2018; 55:386–399.
   Web of Science ®Google Scholar
• Ishii M, Hibiki T. Thermo-fluid dynamics of two-phase flow. 2nd ed. New York (NY): Springer; 2011.
   Google Scholar
• Tomiyama A, Kataoka, I, Fukuda, T, et al. Drag Coefficients of Bubbles: 2nd Report, Drag Coefficient for a Swarm of Bubbles and its Applicability to Transient Flow. Trans Japan Soc Mech Eng Series B. 2008;61:2810–2817. Japanese.
   Google Scholar
• Garnier J, Manon E, Cubizolles G. Local measurements on flow boiling of refrigerant 12 in a vertical tube. Multiphase Sci Technol. 2001;13:1–111.
   Google Scholar
• Roy R, Kang S, Zarate J, et al. Turbulent subcooled boiling flow in – experiments and simulations. J Heat Mass Transfer. 2002;124:73–93.
   Google Scholar
• Situ R, Hibiki T, Sun X, et al. Flow structure of subcooled boiling flow in an internally heated annulus. Int J Heat Mass Transfer. 2004;47:5351–5364.
   Web of Science ®Google Scholar
• Lee T, Situ R, Hibiki T, et al. Axial developments of interfacial area and void concentration profiles in subcooled boiling flow of water. Int J Heat Mass Transfer. 2009;52:473–487.
   Web of Science ®Google Scholar
• Yun B, Bae B, Euh D, et al. Characteristics of the local bubble parameters of a subcooled boiling flow in an annulus. Nucl Eng Des. 2010;240:2295–2303.
   Web of Science ®Google Scholar
• Morooka S, Inoue A, Oishi M, et al. In-bundle void measurement of BWR fuel assembly by X-ray CT scanner. Proceedings of ICONE-1; 1991 Nov 4–7. Tokyo: JASME ( Paper No. 38).
   Google Scholar
• Hibiki T, Ishii M. One dimensional drift-flux model for two-phase flow in a large diameter pipe. Int J Heat Mass Transfer. 2003;46:1773–1790.
   Web of Science ®Google Scholar
• Kataoka I, Ishii M. Drift flux model for large diameter pipe and new correlation for pool void fraction. Int J Heat Mass Transfer. 1987;30:1927–1939.
   Web of Science ®Google Scholar
• Nuclear Power Engineering Test Center. Report on proving test on reliability for BWR fuel bundles (void fraction tests in BWR fuel bundles – support document). Tokyo: NUPEC; 1992. Japanese.
   Google Scholar
• Nuclear Power Engineering Corporation. Report on proving test on reliability for BWR fuel bundles (supplement) (void fraction tests for high burnup fuel bundle – overall evaluation). Tokyo: NUPEC; 1993. Japanese.
   Google Scholar
• Moody LF. An approximate formula for pipe friction factors. Trans ASME. 1947;69:1005–1011.
   Google Scholar
• Martinelli RC, Nelson DB. Prediction of pressure drop during forced circulation boiling of water. Trans ASME. 1948;70:695–702.
   Google Scholar
• Zuber N, Findlay JA. Average volumetric concentration in two-phase flow systems. J Heat Transfer. 1965;87:453–468.
   Web of Science ®Google Scholar
• Ruspini LC, Marcel CP, Clausse A. Two-phase flow instabilities: a review. Int J Heat Mass Transfer. 2014;71:521–548.
   Web of Science ®Google Scholar