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湍流模型对5×5格架棒束通道流动传热数值模拟影响分析 被引量:9

CFD Evaluation of Turbulence Model on Heat Transfer in 5×5 Rod Bundles
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摘要 同一软件工具采用不同湍流模型进行燃料组件格架棒束通道CFD分析时会得到不同的数值结果,本文采用ANSYS CFX软件,建立了包含典型5×5格架的棒束通道CFD模型,研究了涡粘和雷诺应力两大类6种典型湍流模型对燃料组件压降与换热特性数值结果的影响,计算了压降和Nu分布结果与相似的实验结果进行对比,通过分析3个典型搅混效果评价因子,探讨了搅混翼流动与换热的内在影响关系,同时对比了不同湍流模型对结果的影响。通过与相似实验数据对比分析,认为雷诺应力模型较适宜计算本文所研究的定位格架及棒束通道内流动传热特性。 Different turbulence models may lead to different results w hen analyzing fuel assemblies using computational fluid dynamics (CFD) method .In this paper ,a 5 × 5 rod bundle model was built to analyze the relationship between flow and heat transfer .The pressure drop and Nu were calculated using ANSYS CFX . Three factors evaluating swirling flow and cross-flow were used to analyze the inner relationship between flow field and heat transfer .T he performances of various turbulence models ,including eddy viscosity model and Reynold stress model ,were evaluated .The comparison between numerical and similar experimental results indicates that Reynold stress model is more appropriate for modeling flow features and heat transfer in spacer grids discussed in this paper .
出处 《原子能科学技术》 EI CAS CSCD 北大核心 2014年第10期1782-1789,共8页 Atomic Energy Science and Technology
关键词 格架 热工水力 CFD 湍流模型 spacer grid thermal-hydraulic CFD turbulence model
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参考文献10

  • 1LEE K B, JANG H G. A numerical prediction on the turbulent flow in closely spaced bare rod arrays by a nonlinear be model[J]. Nuclear En- gineering and Design, 1997, 172: 351-357.
  • 2SMITH L D, CONNER M E, LIU B, et al. Benchmarking computational fluid dynamics for application to PWR fuel[C] // Proceedings of the 10th International Conference on Nuclear Engi- neering. USA: ASME, 2002: 823-830.
  • 3YADIGAROGLU G, ANDERANI M, DREIER J, et aI. Trends and needs in experimentation and numerical simulation for LWR safety[J].Nucle- ar Engineering and Design, 2003, 221 : 205-223.
  • 4LIUCC, FERNGYM, SHIHCK. CFDeval uation of turbulence models for flow the fuel Applied 396. rod bundle with a spacer a simulation of ssembly[J]. Thermal Engineering, 2012 (40) : 389.
  • 5HOLLOWAY M V, McCLUSKY H L, BEAS- LEY D E, et al. The effect of support grid lea tures on local, single phase heat urements in rod bundles[J]. AS meas- rnal of Heat Transfer, 2004, 126: 43 -53.
  • 6HOLLOWAY M V, CONOVER T A, Mc- CLUSKY H L, et al. The effect of support grid design on azimuthal variation in heat transfer co- efficient for rod bundles[J]. ASME Journal of Heat Transfer, 2005, 127: 598-605.
  • 7CUI X Z, KIM K Y. Three-dimensional analysis of turbulent heat transfer and flow through mix- ing vane in a subehannel of nuclear reactor[J].Journal of Nuclear Science and Technology, 2003, 40(10): 719-724.
  • 8LEE C M, CHOI Y D. Comparison of thermo hydraulic performances of large scale vortex flow (LSVF) and small scale vortex flow (SSVF) mixing vanes in 17 X 17 nuclear rod bundle[J]. Nuclear Engineering and Design, 2007, 237 (24): 2 322-2 331.
  • 9IN WK, CHUN T H, OH DS, et al. CFDa nalysis of turbulent flows in rod Bundles for nu- clear fuel spacer design[C]//Transactions of the 15th International Conference on Structural Me- chanics in Reactor Technology. Seoul: SMiRT, 1999:365- 372.
  • 10IN W K, CHUN T H, SHIN C H, et al. Nu- merical computation of heat transfer enhancement of a PWR rod bundle with mixing vane spacers [J]. Nuclear Technology, 2008, 161(1): 69-79.

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