Heat transfer study in one eighth of the high performance supercritical water-cooled reactor fuel assembly
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Abstract
The high-performance light-water nuclear reactor is the European version of the supercritical water-cooled reactor, proposed as one of the candidates for Generation IV of nuclear reactors. This paper assesses the ability to predict the heat transfer behavior to supercritical water in the sub-channels of the 1/8 HPLWR fuel assembly by codes of Computational Fluid Dynamics using two models of turbulence (the Reynolds stress model developed by Speziale and the k-? shear stress transport model). A mesh sensitivity study was performed to guarantee the independence of the numerical results regardless the size and distribution of the mesh elements. Temperature distributions were calculated in the fuel rods, in the clad, and in water both in the cooling zone and moderator zone. The results of the two turbulence models were compared. No appreciable difference was obtained in the values of the supercritical water average temperature calculated with the turbulence models used. However, the numerical results using the SST turbulence model show higher values regarding the temperature of both fuel rods and clad surface compared to those calculated with the SSG model.
Article Details
How to Cite
Castro González, L. Y., Alfonso Barrera, R., García Hernández, C., & Rosales García, J. (1). Heat transfer study in one eighth of the high performance supercritical water-cooled reactor fuel assembly. Nucleus, (61), 32-38. Retrieved from http://nucleus.cubaenergia.cu/index.php/nucleus/article/view/16
Section
Ciencias Nucleares
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[2] SCHULENBERG T, STARFLINGER J, MARSAULT P, et. al. European supercritical water cooled reactor. Nucl. Eng. Des. 2011; 241(9): 3505-3513.
[3] NAIDIN M, MOKRY S, BAIG F, et. al. Thermal-design options for pressure-channel SCWRs with co-generation of hydrogen. J. Eng. Gas Turbines Power. 2009; 131(1).
[4] FARAH A. Assessment of FLUENT CFD code as an analysis tool for SCW applications [master of Applied Science in Nuclear Engineering]. Canada: University of Ontario Institute of Technology; 2012.
[5] AMMIRABILE L. Studies on supercritical water reactor fuel assemblies using the sub-channel code COBRA-EN. Nucl. Eng. Des. 2010; 240(10): 3087-3094.
[6] CHENG X, SCHULENBERG T. Heat transfer at supercritical pressures - literature review and application to an HPLWR. FZKA 6609. Karlsruhe: Forschungszentrum Karlsruhe, 2001. http://bibliothek.fzk.de/zb/berichte/FZKA6609.pdf.
[7] ROELOFS F. CFD Analyses of heat transfer to supercritical water flowing vertically upward in a tube. 2004.
[8] CHENG X, KUANG B, YANG YH. Numerical analysis of heat transfer in supercritical water cooled flow channels. Nucl. Eng. Des. 2007; 237(3): 240-252.
[9] GU HY, CHENG X, YANG YH. CFD analysis of thermal-hydraulic behavior of supercritical water in sub-channels. Nucl. Eng. Des. 2010; 240(2): 364-374.
[10] WEN QL, GU HY. Numerical simulation of heat transfer deterioration phenomenon in supercritical water through vertical tube. Ann. Nucl. Energy. 2010; 37(10): 1272-1280.
[11] JAROMIN M, ANGLART H. A numerical study of heat transfer to supercritical water flowing upward in vertical tubes under normal and deteriorated conditions. Nucl. Eng. Des. 2013; 264: 61-70.
[12] PODILA K, RAO YF. Assessment of CFD for the Canadian SCWR bundle with wire wraps. Prog. Nucl. Energy. 2014; 77: 373-380. doi:10.1016/j.pnucene.2014.02.009.
[13] XIONG J, CHENG X, YANG Y. Numerical analysis on supercritical water heat transfer in a 2×2 rod bundle. Ann. Nucl. Energy. 2015; 80: 123-134. doi:10.1016/j.anucene.2015.02.005.
[14] PODILA K, RAO Y. CFD modelling of supercritical water flow and heat transfer in a 2×2 fuel rod bundle. Nucl. Eng. Des. 2016; 301: 279-289. doi:10.1016/j.nucengdes.2016.03.019.
[15] ROWINSKI MK, ZHAO J, WHITE TJ, et. al. Numerical investigation of supercritical water flow in a vertical pipe under axially non-uniform heat flux. Prog. Nucl. Energy. 2017; 97: 11-25. doi:10.1016/j.pnucene.2016.12.009.
[16] ZHAO CR, ZHANG Z., JIANG PX, BO HL. Influence of various aspects of low Reynolds number k-? turbulence models on predicting in-tube buoyancy affected heat transfer to supercritical pressure fluids. Nucl. Eng. Des. 2017; 313: 401-413. doi:10.1016/j.nucengdes.2016.12.033.
[17] ANSYS CFX Solver Theory Guide. Canonsburg, PA: ANSYS Inc, 2010.
[18] HOFMEISTER J, WAATA C, STARFLINGER J, et. al. Fuel assembly design study for a reactor with supercritical water. Nucl. Eng. Des. 2007; 237(14): 1513-1521.
[19] WAATA CL. Coupled Neutronics thermal hydraulics analysis of a high-performance light-water Reactor Fuel Assembly [doctoral thesis]. Karlsruhe: Forschungszentrum Karlsruhe GmbH, 2006.
[20] REISS T, FEHÉR S, CZIFRUS S. Coupled neutronics and thermohydraulics calculations with burn-up for HPLWRs. Prog. Nucl. Energy. 2008; 50(1): 52-61.
[21] XI X, XIAO Z, YAN X, et. al. The axial power distribution validation of the SCWR fuel assembly with coupled neutronics-thermal hydraulics method. Nucl. Eng. Des. 2013; 258: 157-163.
[22] CHAUDRI KS, SU Y, CHEN R, et. al. Development of sub-channel code SACoS and its application in coupled neutronics/thermal hydraulics system for SCWR. Ann. Nucl. Energy. 2012; 45: 37-45.