Improved drift-flux correlation to enhance the prediction of void fraction in nuclear reactor fuel bundles at low flow and elevated pressure conditions

References

• Yeoh GH. Thermal hydraulic considerations of nuclear reactor systems: past, present and future challenges. Exp Comput Multiphase Flow. 2019;1(1):3–27.
   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
• Chen SW, Liu Y, Hibiki T, et al. Experimental study of air-water two-phase flow in 8 × 8 rod bundle under pool condition for one-dimensional drift-flux analysis. Int J Heat Fluid Flow. 2012;33:168–181.
   Web of Science ®Google Scholar
• Chen SW, Liu Y, Hibiki T, et al. One-dimensional drift-flux model for two-phase flow in pool rod bundle systems. Int J Multiphase Flow. 2012;40:166–177.
   Web of Science ®Google Scholar
• Clark C, Griffiths M, Chen SW, et al. Experimental study of void fraction in an 8 × 8 rod bundle at low pressure and low liquid flow conditions. Int J Multiphase Flow. 2014;62:87–100.
   Web of Science ®Google Scholar
• Clark C, Griffiths M, Chen SW, et al. Drift-flux correlation for rod bundle geometries. Int J Heat Fluid Flow. 2014;48:1–14.
   Web of Science ®Google Scholar
• Abbs T, Hibiki T. One-dimensional drift-flux correlation for vertical upward two-phase flow in a large size rectangular channel. Prog Nucl Energy. 2019;110:311–324.
   Web of Science ®Google Scholar
• Zhao Q, Hibiki T. One-dimensional drift-flux correlation for vertical upward two-phase flow in large size concentric and eccentric annuli. Int J Multiphase Flow. 2019;113:33–44.
   Web of Science ®Google Scholar
• Ohnuki A, Akimoto H. An experimental study on developing air-water two-phase flow along a large vertical pipe: effect of air injection method. Int J Multiphase Flow. 1996;22:1143–1154.
   Web of Science ®Google Scholar
• Ohnuki A, Akimoto H. Developed flow pattern and phase distribution under gas-liquid two-phase flow in a large vertical pipe and prediction of phase distribution by multidimensional two-fluid model. Proceedings of the 8th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-8); 1997 September 30-October 4; Kyoto, Japan.
   Google Scholar
• Hibiki T, Ozaki T, Shen X, et al. Constitutive equations for vertical upward two-phase flow in rod bundle. Int J Heat Mass Transfer. 2018;127:1252–1266.
   Web of Science ®Google Scholar
• Hibiki T, Ozaki T. Modeling of void fraction covariance and relative velocity covariance for upward 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
• Ozaki T, Hibiki T, Miwa S, et al. Development of one-dimensional two-fluid model with consideration of void fraction covariance effect. J Nucl Sci Technol. 2018;55:720–732.
   Web of Science ®Google Scholar
• Ozaki T, Hibiki T, Miwa S, et al. Effect of void fraction covariance on two-fluid model based code calculation in pipe flow. Prog Nucl Energy. 2018;108:319–333.
   Web of Science ®Google Scholar
• Chexal B, Lellouche G, Horowitz J, et al. The Chexal-Lellouche void fraction correlation for generalized applications. Nuclear Safety Analysis Center of the Electric Power Research Institute, NSAC-139; 1991.
   Google Scholar
• Ozaki T, Suzuki R, Mashika 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
• RELAP5/MOD3.3, Code Manual Volume IV: Models and Correlations, NUREG/CR-5535/Rev1, Prepared for the Division of Systems Research, Office of Nuclear Regulatory Research. USNRC by Information Systems Laboratories, Inc., Rockville, Maryland, 2001.
   Google Scholar
• Anklam TM, Miller RJ. Void fraction under high pressure, low flow conditions in rod bundle geometry. Nucl Eng Des. 1982;75:99–108.
   Web of Science ®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
• Griffiths MJ, Schlegel JP, Clark C, et al. Uncertainty evaluation of the Chexal-Lellouche correlation for void fraction in rod bundles. Prog Nucl Energy. 2014;74:143–153.
   Web of Science ®Google Scholar
• Hibiki T. One-dimensional drift-flux correlations for two-phase flow in medium-size channels. Exp Comput Multiphase Flow. 2019;1(2):85–100.
   Google Scholar
• Ishii M. One-dimensional drift-flux model and constitutive equations for relative motion between phases in various two-phase flow regimes. Argonne National Laboratory, ANL-77-47; 1977.
   Google Scholar
• Katoaka 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
• Shi S, Schlegel JP, Brooks CS, et al. Experimental investigation of natural circulation instability in a BWR-type small modular reactor. Prog Nucl Energy. 2015;85:96–107.
   Web of Science ®Google Scholar