Document Type : Research Article
Authors
School of Mechanical Engineering, Shiraz University, Shiraz, Iran
Abstract
Highlights
[1] B. S. Shin, S. H. Chang, Experimental study on the effect of angles and positions of mixing vanes on CHF in a 2×2 rod bundle with working fluid R-134a, nuclear engineering and design, (235) (2005) 1749-1759.
[2] C. M. Lee, J. Soo,Y. D. Choi, Thermo-Hydraulic Characteristics of Hybrid Mixing Vanes in a 17x17 Nuclear Rod Bundle, Journal of mechanical science and technology, (21) (2007) 1263-1270.
[3] E. Krepper, B. Koncar, Y. Egorov, CFD modelling of subcooled boiling—Concept, validation and application to fuel assembly design, Nuclear engineering and design, (237) (2007) 716-731.
[4] M. A. Navarro, A. C. Santos, 2009. Evaluation of a numeric procedure for flow simulation of a 5X5 PWR rod bundle with a mixing vane spacer, International Nuclear Atlantic Conference, Rio de Janeiro,RJ, Brazil, (2009) September27 to October 2.
[5] B. S. Shin, S. H. Chang, CHF experiment and CFD analysis in a 2×3 rod bundle with mixing vane, nuclear engineering and design, (239) (2009) 899-912.
[6] M. Damsohn, H. M. Prasser, Experimental studies of the effect of functional spacers to annular flow in subchannels of a BWR fuel element, Nuclear engineering and desing, (240) (2010) 3126-3144.
[7] I. S. Jun, K. H. Bae, Y. J. Chung, Validation of the TASS/ SMR-S Code for the Core Heat Transfer Model on the Steady Experimental Conditions. Journal of energy and power engineering, (2012) 338-345.
[8] A. W. Bennett, G. F. Hewitt, H. A. Kearsey, et al. Heat transfer to steam-water mixtures flowing in uniformly heated tubes in which the critical heat flux has been exceeded, Proceedings of the Institution of Mechanical Engineers, Sept. (1976) 258-267.
[9] S. Jayanti, K. R. Reddy, Effect of spacer grids on CHF in nuclear rod bundles, Nuclear engineering and desing, (261) (2013) 66-75.
[10] M. Nazififard, P. SOROUSH, M.R. Nematollahi, Heat Transfer and safety enhancement analysis of fuel assembly an advanced pressurized water reactors: A CFD approach, Indian J. Sci. Res, 1(2) (2014) 487-495.
[11] X. Zhu, S. Morooka, Y. Oka, Numerical investigation of grid spacer effect on heat transfer of supercritical water flows in a tight rod bundle, International journal of thermal sciences, (76) (2014) 245-257.
[12] H. Seo, S. D. Park, S. B. Seo, H. Heo, I. C. Bang, Swirling performance of flow-driven rotating mixing vane toward critical heat flux enhancement, International journal of Heat and Mass transfer, (89) (2015) 1216-1229.
[13] S. Mimouni, C. Baudry, M. Guingo, J. Lavieville, N. Merigoux, N. Mechitoua, Computational multi-fluid dynamics predictions of critical heat flux in boiling flow, Nuclear engineering and design, (299) (2015) 59-80.
[14] D. Chen, Y. Xiao, S. Xie, D. Yuan, X. Lang, Z. Yang, Y. Zhong, Q. Lu, Thermal–hydraulic performance of a 5×5 rod bundle with spacer grid in a nuclear reactor, Applied thermal engineering, (103) (2016) 1416-1426.
[15] M. Zhao, H. Y. Gu, H. B. Li, X. Cheng, Heat transfer of water flowing upward in vertical annuli with spacers at high pressure conditions, Annals of nuclear energy, (87) (2016) 209-216.
[16] H. Li, H. Punekar, S. A. Vasquez, R. Muralikrishnan, Prediction of Boiling and Critical Heat Flux using an Eulerian Multiphase Boiling Model, Proceedings of the ASME, International Mechanical Engineering Congress & Exposition, Colorado,USA (2011).
[17] N. Kurul, M. Z. Podowski, On the modeling of multidimensional effects in boiling channels. In: Proceedings of the 27th National Heat Transfer Conference, Minneapolis, Minnesota, USA, July (1991).
[18] V. H. D. Vall, D. B. R. Kenning, Subcooled flow boiling at high heat flux, Int. J. Heat Mass Transfer, (28) (195) 1907-1920.
[19] R. Cole, A photographic study of pool boiling in the region of the critical heat flux. AICHE J. (6) (1960) 533- 542.
[20] M. Lemmert, J. M. Chawla, Influence of flow velocity on surface boiling heat transfer coefficient. Heat Transfer in Boiling, (1977) 237-247.
[21] V. I. Tolubinski, D. M. Kostanchuk, Vapor bubbles growth rate and heat transfer intensity at subcooled water boiling. In: 4th International Heat Transfer Conference, Paris, France (1970).
[22] H. Li, S. A. Vasquez, H. Punekar, Prediction of Boiling and Critical Heat Flux Using an Eulerian Multiphase Boiling Model. Proceedings of the ASME 2010, International Mechanical Engineering Congress & Exposition, canada (2010).
[23] A. Ioilev, M. Samigulin, V. Ustinenko, Advances in the modeling of cladding heat transfer and critical heat flux in boiling water reactor fuel assembly, NURETH-12, Pittsburgh, Pennsylvania, USA (2007).
[24] A. A. Troshko, Y. A. Hassan, A two-equation turbulence model of turbulent bubbly flow, Int. J. Multiphase Flow, 22(11) (1965) 2000-2001.
[25] Z. Karoutas, C. Gu, B. Sholin, 3-D flow analyses for design of nuclear fuel spacer. In: Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-hydraulics NURETH-7, New York, USA, 3153e3174 (1995).
[26] W. K. In, D. S. Oh, T. H. Chun, CFD Analysis of Turbulent Flow in Nuclear Fuel Bundle with Flow Mixing Device, KAERI report, TR-1296/99, (1999) 54.
[27] C. B. Mullins, D. K. Felde, A. G. Sutton, S. S. Gould, D. G. Morris, J. J. Robinson, ORNL Rod-Bundle Heat- Transfer Test Data, Vol. 3, Thermal-hydraulic test facility experimental data report for test 3.06.6B-transient film boiling in upflow, technical report (1982).
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