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Numerical Analysis of Critical Heat Flux Phenomenon in a Nuclear Power Plant Core Channel in the Presence of Mixing Vanes | ||
| AUT Journal of Mechanical Engineering | ||
| مقاله 1، دوره 1، شماره 2، اسفند 2017، صفحه 119-130 اصل مقاله (1.58 M) | ||
| نوع مقاله: Research Article | ||
| شناسه دیجیتال (DOI): 10.22060/mej.2017.12928.5472 | ||
| نویسندگان | ||
| A. Rabiee* ؛ L. Moradi؛ A. Atf | ||
| School of Mechanical Engineering, Shiraz University, Shiraz, Iran | ||
| چکیده | ||
| The necessity and importance of a high heat removal potential in various areas particularly in nuclear applications are in a direct relationship with the excessively applied heat flux level. One way to increase the heat transfer performance and subsequently enhance the threshold of the critical heat flux is to employ spacer grids accompanied by mixing vanes. In this study, the effect of the spacers with mixing vanes on the critical heat flux characteristics in the dryout condition has been numerically investigated employing the benefits of the Eulerian-Eulerian framework. In the current research, several vane angles, including vane with 0, 15 and 25 degrees in comparison with the effect of the bare spacer without any mixing vanes on the flow characteristics were examined. It was shown that the existence of the spacer alone, delays the temperature jump under critical heat flux conditions. It was also concluded that increasing the angle of the mixing vanes, further improves the heat transfer performance of the system by postponing the sudden temperature jump occurring in the channel; however, the presence of the spacers and vanes in the flow field imposes an increase of the pressure drop due to the constriction on the coolant flow area. | ||
تازه های تحقیق | ||
[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). | ||
| کلیدواژه‌ها | ||
| Boiling؛ Critical heat flux؛ Mixing vane؛ Spacer grid | ||
| مراجع | ||
|
[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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