压水堆燃料棒热力计算与(火用)分析
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摘要
为了研究燃料棒核能转化为热能过程中的(火用)损,采用压水堆燃料棒稳态传热偏微分方程和热力学第一、第二定律,创新性地将(火用)分析方法与燃料棒温度场数值计算相结合,编制数值计算程序对燃料棒及传热通道进行模拟计算,并分析了核能转换为热能以及与冷却剂换热过程中温度分布、(火用)损的分布和能量的利用效率.结果表明:燃料棒(火用)损沿轴向先增大后减小,沿径向不断变大,在燃料芯块边缘处达到最大,(火用)损系数约为0.207;而对流换热过程中(火用)损主要与传热温差有关,(火用)损沿热通道先增大后减小,该过程累积(火用)损系数约为0.304.
To investigate the exergy loss of fuel rod during converting nuclear energy into thermal energy,the partial differential equations of steady heat transfer of pressurized water reactor fuel rods and the first and second laws of thermodynamics were used. The exergy analysis method was innovatively combined with the numerical calculation of temperature field. The numerical calculation program was compiled to simulate the fuel rods and heat transfer channels and analyze the temperature distribution,the exergy loss distribution and the energy utilization efficiency during converting nuclear energy into heat energy and during coolant heat transfer. The results show that the fuel rod exergy loss is increased with latter decreasing in the axial direction and increased in the radial direction with the maximum exergy loss coefficient of 0. 207 at the edge of fuel core. The exergy loss in the convective heat transfer process is mainly related to the heat transfer temperature difference and increased with latter decreasing along the thermal channel with the total exergy loss coefficient of 0. 304.
To investigate the exergy loss of fuel rod during converting nuclear energy into thermal energy,the partial differential equations of steady heat transfer of pressurized water reactor fuel rods and the first and second laws of thermodynamics were used. The exergy analysis method was innovatively combined with the numerical calculation of temperature field. The numerical calculation program was compiled to simulate the fuel rods and heat transfer channels and analyze the temperature distribution,the exergy loss distribution and the energy utilization efficiency during converting nuclear energy into heat energy and during coolant heat transfer. The results show that the fuel rod exergy loss is increased with latter decreasing in the axial direction and increased in the radial direction with the maximum exergy loss coefficient of 0. 207 at the edge of fuel core. The exergy loss in the convective heat transfer process is mainly related to the heat transfer temperature difference and increased with latter decreasing along the thermal channel with the total exergy loss coefficient of 0. 304.
引文
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[3]卢川,严明宇,毕树茂,等.基于CFD方法的行波堆19燃料棒束流固耦合传热特性研究[J].原子能科学技术,2015,49(12):2170-2175.LU C,YAN M Y,BI S M,et al.Study on fluid-solid coupling heat transfer characteristics of TWR assembly with 19 fuel pins based on CFD method[J].Atomic Energy Science and Technology,2015,49(12):2170-2175.(in Chinese)
[4]罗磊,陈文振,陈志云,等.单个燃料元件热工水力三维数值模拟[J].海军工程大学学报,2011,23(1):63-66.LUO L,CHEN W Z,CHEN Z Y,et al.Numerical simulation of thermal hydrodynamic of single reactor fuel rod[J].Journal of Naval University of Engineering,2011,23(1):63-66.(in Chinese)
[5]SALAMA A,EL-DIN EL-MORSHEDY S.CFD simulation of flow blockage through a coolant channel of atypical material testing reactor core[J].Annals of Nuclear Energy,2012,41:26-39.
[6]LI X C,GAO Y.Methods of simulating large scale rod bundle and application to a 17×17 fuel assembly with mixing vane spacer grid[J].Nuclear Engineering and Design,2014,267:10-22.
[7]FRICANO J W,BAGLIETTO E.A quantitative CFD benchmark for solidium fast reactor fuel assembly modeling[J].Annals of Nuclear Energy,2014,64:32-42.
[8]PIRO M H A,LEITCH B W.Conjugate heat transfer simulations of advanced research reactor fuel[J].Nuclear Engineering and Design,2014,274:30-43.
[9]RASU N G,VELUSAMY K,SUNDARARAJAN T,et al.Simultaneous development of flow and temperature fields in wire-wrapped fuel pin bundles of sodium cooled fast reactor[J].Nuclear Engineering and Design,2014,267:44-60.
[10]LIU C C,FERNG Y M,SHIH C K.CFD evaluation of turbulence models for flow simulation of the fuel rod bundle with a spacer assembly[J].Applied Thermal Engineering,2012,40:389-396.
[11]ZHANG M.Modeling of radiative heat transfer and diffusion processes using unstructured grid[D].Cookeville:Tennessee Technological University,2000.
[12]闵元佑,黄云.秦山核电二期工程反应堆及反应堆冷却剂系统设计[J].核动力工程,2003,24(2):1-7.MIN Y Y,HUANG Y.Design of the reactor and reactor coolant system for qingshan phase II NPP project[J].Nuclear Power Engineering,2003,24(2):1-7.(in Chinese)
[13]陈文振,于雷,郝建立.核动力装置热工水力[M].北京:中国原子能出版社,2013.
[14]彭敏俊,田兆斐.核动力装置热力分析[M].哈尔滨:哈尔滨工程大学出版社,2012.