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Research Article

Effect of two-group void fraction covariance correlations on interfacial drag predictions for two-fluid model calculations in large diameter pipes

Alexander Swearingen1Sean Drewry1Joshua P. Schlegel1( )Takashi Hibiki2
Department of Nuclear Engineering and Radiation Sciences, Missouri University of Science and Technology, 301 W 14th St, Rolla, MO 65401, USA
School of Nuclear Engineering, Purdue University, 516 Northwestern Ave, West Lafayette, IN 47907, USA
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Abstract

Void fraction covariance has been introduced into the interfacial drag calculation used to close the one-dimensional two-fluid model. A model for void fraction covariance has been developed for large diameter pipes. The newly developed model has been compared with two previously developed models in terms of void fraction prediction accuracy. The effects of these additions on the void fraction prediction uncertainty have been evaluated utilizing a computational tool developed in MATLAB. The results indicate that there are small differences in the void fraction prediction between the models evaluated and the two-fluid model without void fraction covariance. Higher void fractions above 0.7 show the most significant changes. However, the differences in the uncertainty are not significant when compared to the uncertainty in the data used for the comparison. The results highlight a need for additional data for higher void fractions, collected with steam–water systems in large diameter pipes.

References

 
Andersen, J. G. M., Chu, K. H. 1982. BWR refill-reflood program Task 4.7: Constitutive correlations for shear and heat transfer for the BWR version of TRAC. Technical Report. NUREG/CR-2134. Office of Scientific and Technical Information, U.S. Department of Energy.
 
Baily, R. V., Zmola, P. C., Taylor, F. M., Planchet, R. J. 1956. Transport of gases through liquid–gas mixture. In: Proceedings of the AIChE New Orleans Meeting.
 
Brooks, C. S., Liu, Y., Hibiki, T., Ishii, M. 2014. Effect of void fraction covariance on relative velocity in gas-dispersed two-phase flow. Progress in Nuclear Energy, 70: 209–220.
 
Carrier, F. 1963. Steam separation technology under the Euroatom Program. Technical Report. ACNP-63021. Office of Scientific and Technical Information, U.S. Department of Energy.
 
Dandekar, A. V., Brooks, C. S. 2016. Modeling of void fraction covariance in two-phase flows with phase change. International Journal of Heat and Mass Transfer, 100: 231–242.
 
Garnier, J., Manon, E., Cubizolles, G. 2001. Local measurements on flow boiling of refrigerant 12 in a vertical tube. Multiphase Science and Technology, 13: 1–111.
 
Hall, W. H., Prueter, W. P., Thorne, T. L., Wall, J. R. 1988. High-pressure steam/water void fraction profiles in a large-diameter, vertical pipe with non-developed entrance flow. In: Proceedings of the Thermal Hydraulics of Nuclear Steam Generators/Heat Exchangers.
 
Hashemi, A., Kim, J. H., Sursock, J. P. 1986. Effect of diameter and geometry on two-phase flow regimes and carry-over in a model PWR hot leg. In: Proceeding of the 8th International Heat Transfer Conference, 2443–2451.
 
Hibiki, T., Ishii, M. 2000. Experimental study on hot-leg U-bend two-phase natural circulation in a loop with a large diameter pipe. Nuclear Engineering and Design, 195: 69–84.
 
Hibiki, T., Ishii, M. 2001. Effect of inlet geometry on hot-leg U-bend two-phase natural circulation in a loop with a large diameter pipe. Nuclear Engineering and Design, 203: 209–228.
 
Hibiki, T., Ozaki, T. 2017. Modeling of void fraction covariance and relative velocity covariance for upward boiling flow in vertical pipe. International Journal of Heat and Mass Transfer, 112: 620–629.
 
Hills, J. H. 1976. The operation of a bubble column at high throughputs: I. Gas holdup measurements. The Chemical Engineering Journal, 12: 89–99.
 
Inoue, Y. 2001. Measurement of interfacial area concentration of gas–liquid two-phase flows in a large diameter pipe. Ph.D. Thesis. Kyoto University, Japan.
 
Ishii, M. 1977. One-dimensional drift–flux model and constitutive equations for relative motion between phases in various two-phase flow regimes. Technical Report. ANL-77-47. Office of Scientific and Technical Information, U.S. Department of Energy.
 
Ishii, M., Mishima, K. 1984. Two-fluid model and hydrodynamic constitutive relations. Nuclear Engineering and Design, 82: 107–126.
 
Kim, S., Fu, X. Y., Wang, X., Ishii, M. 2000. Development of the miniaturized four-sensor conductivity probe and the signal processing scheme. International Journal of Heat and Mass Transfer, 43: 4101–4118.
 
Manera, A., Ozar, B., Paranjape, S., Ishii, M., Prasser, H. M. 2009. Comparison between wire-mesh sensors and conductive needle-probes for measurements of two-phase flow parameters. Nuclear Engineering and Design, 239: 1718–1724.
 
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. International Journal of Multiphase Flow, 26: 367–386.
 
Ohnuki, A., Akimoto, H., Sudo, Y. 1995. Flow pattern and its transition in gas–liquid two-phase flows along a large vertical pipe. In: Proceedings of the 2nd International Conference on Multiphase Flow.
 
Omebere-Iyari, N. K., Azzopardi, B. J., Lucas, D., Beyer, M., Prasser, H. M. 2008. The characteristics of gas/liquid flow in large risers at high pressures. International Journal of Multiphase Flow, 34: 461–476.
 
Ozaki, T., Hibiki, T., Miwa, S., Mori, M. 2018. Development of one-dimensional two-fluid model with consideration of void fraction covariance effect. Journal of Nuclear Science and Technology, 55: 720–732.
 
Prasser, H. M., Beyer, M., Böttger, A., Carl, H., Lucas, D., Schaffrath, A., Schütz, P., Weiss, F. P., Zschau, J. 2005. Influence of the pipe diameter on the structure of the gas–liquid interface in a vertical two-phase pipe flow. Nuclear Technology, 152: 3–22.
 
Prasser, H. M., Beyer, M., Carl, H., Gregor, S., Lucas, D., Pietruske, H., Schütz, P., Weiss, F. P. 2007. Evolution of the structure of a gas–liquid two-phase flow in a large vertical pipe. Nuclear Engineering and Design, 237: 1848–1861.
 
Saito, M. 1998. Dispersion characteristics of gas–liquid two-phase pools. In: Proceedings of the 6th International Conference on Nuclear Engineering.
 
Sawant, P., Schelegel, J., Paranjape, S., Ozar, B., Hibiki, T., Ishii, M. 2009. Flow regime identification in large diameter pipe. In: Proceedings of the 16th International Conference on Nuclear Engineering, 341–351.
 
Schlegel, J. P., Hibiki, T., Ishii, M. 2015. Two-group modeling of interfacial area transport in large diameter channels. Nuclear Engineering and Design, 293: 75–86.
 
Schlegel, J. P., Hibiki, T., Shen, X., Appathurai, S., Subramani, H. 2017. Prediction of interfacial area transport in a coupled two-fluid model computation. Journal of Nuclear Science and Technology, 54: 58–73.
 
Schlegel, J. P., Macke, C. J., Hibiki, T., Ishii, M. 2013. Modified distribution parameter for churn-turbulent flows in large diameter channels. Nuclear Engineering and Design, 263: 138–150.
 
Schlegel, J. P., Miwa, S., Chen, S., Hibiki, T., Ishii, M. 2012. Experimental study of two-phase flow structure in large diameter pipes. Experimental Thermal and Fluid Science, 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 two-phase flow in large diameter vertical pipes. Nuclear Engineering and Design, 239: 2864–2874.
 
Schlegel, J. P., Sharma, S., Cuenca, R. M., Hibiki, T., Ishii, M. 2014. Local flow structure beyond bubbly flow in large diameter channels. International Journal of Heat and Fluid Flow, 47: 42–56.
 
Shawkat, M. E., Ching, C. Y., Shoukri, M. 2008. Bubble and liquid turbulence characteristics of bubbly flow in a large diameter vertical pipe. International Journal of Multiphase Flow, 34: 767–785.
 
Shen, X. Z., Mishima, K., Nakamura, H. 2011. Flow-induced void fraction transition phenomenon in two-phase flow. In: Proceedings of the 18th International Conference on Nuclear Engineering, 597–604.
 
Shen, X., Hibiki, T., Nakamura, H. 2015. Bubbly-to-cap bubbly flow transition in a long-26 m vertical large diameter pipe at low liquid flow rate. International Journal of Heat and Fluid Flow, 52: 140–155.
 
Smith, T. R., Schlegel, J. P., Hibiki, T., Ishii, M. 2012. Two-phase flow structure in large diameter pipes. International Journal of Heat and Fluid Flow, 33: 156–167.
 
Spore, J. W., Jolly-Woodruff, S. J., Knight, T. K., Lin, J. C., Nelson, R. A., Pasamehmetoglu, R. A., Steinke, R. G., Unal, C. 1993. TRAC-PF1/MOD2, vol. 1: Theory Manual. LA-12031-M, vol. I, NUREG/CR-5673, USA.
 
Styrikovich, M. A., Kutateladze, S. S. 1976. Hydrodynamics of liquid–gas system. Energy, 196.
 
Sun, X., Ishii, M., Kelly, J. M. 2003. Modified two-fluid model for the two-group interfacial area transport equation. Annals of Nuclear Energy, 30: 1601–1622.
 
Swearingen, A., Schlegel, J. P., Hibiki, T. 2022. Sensitivity of two-fluid model calculations to two-group drift-flux correlations used in the prediction of interfacial drag. Experimental and Computational Multiphase Flow, 4: 318–335.
 
The RELAP5 Development Team. 1995. RELAP5/MOD3 code manual: Code structure, system models, and solution methods. Volume 1. Technical Report. NUREG/CR-5535-Vol.1. Office of Scientific and Technical Information, U.S. Department of Energy.
 
Wang, J., Li, X., Allison, C., Hohorst, J. 2020. Nuclear Power Plant Design and Analysis Codes, 2nd edn. Woodhead Publishing.
 
Wilson, J. F., Grenda, R. J., Patterson, J. F. 1961. Steam volume fraction in a bubbling two-phase mixture. Transactions of the American Nuclear Society, 4: 356–357.
 
Worosz, T. 2015. Interfacial area transport equation for bubbly to cap-bubbly transition flows. Ph.D. Thesis. Penn State University.
 
Yondeda, K., Akira, Y., Fumio, I., Tomio, O., Hirokai, T., Akira, O. 2009. Improvement of void fraction prediction method for two-phase flow in a large diameter pipe. Denryoku Chuo Kenkyujo Komae Kenkyujo Hokoku T99059.
Experimental and Computational Multiphase Flow
Pages 221-231
Cite this article:
Swearingen A, Drewry S, Schlegel JP, et al. Effect of two-group void fraction covariance correlations on interfacial drag predictions for two-fluid model calculations in large diameter pipes. Experimental and Computational Multiphase Flow, 2023, 5(2): 221-231. https://doi.org/10.1007/s42757-022-0138-6

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Received: 22 December 2021
Revised: 25 March 2022
Accepted: 24 May 2022
Published: 08 August 2022
© Tsinghua University Press 2022
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