VVER反应堆燃料组件流动传热特性CFD分析
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摘要
采用计算流体力学(CFD)方法对俄罗斯水-水高能反应堆(VVER)先进燃料组件(AFA)的流动传热特性进行模拟,获得了额定工况下燃料组件冷却剂流场、流动压降和温度分布等。结果表明:与内部含交混翼的格架相比,AFA燃料组件定位格架的压力损失较小;定位格架围板导向翼附近存在滞流现象,导致燃料组件外围区域冷却剂温度偏高;不同的测量管周向棒功率比K_c对燃料组件出口冷却剂温度的测量值有较大影响。该分析结果可为核电厂堆芯温升预警值ΔT_t的设定提供参考。
The flow and heat transfer characteristics of AFA fuel assembly for VVER reactors have been investigated using computational fluid dynamics(CFD) simulation. The flow field, pressure drop and temperature distribution of the coolant in AFA under normal regime have been calculated. The results show that the pressure drop of the spacer grid of AFA is lower than that of the grid having mixing vane. The stagnation zone of coolant appears around the rim of the spacer grid and causes higher temperature in the periphery region of AFA. The power ratio of the circumferential pin around instrumental tube(K_c) with different values has a great effect on the measured temperature of the coolant at FA outlet. The results can be referred in the setting of temperature warning value(ΔT_t) for the reactor core during the operation of nuclear power plants.
The flow and heat transfer characteristics of AFA fuel assembly for VVER reactors have been investigated using computational fluid dynamics(CFD) simulation. The flow field, pressure drop and temperature distribution of the coolant in AFA under normal regime have been calculated. The results show that the pressure drop of the spacer grid of AFA is lower than that of the grid having mixing vane. The stagnation zone of coolant appears around the rim of the spacer grid and causes higher temperature in the periphery region of AFA. The power ratio of the circumferential pin around instrumental tube(K_c) with different values has a great effect on the measured temperature of the coolant at FA outlet. The results can be referred in the setting of temperature warning value(ΔT_t) for the reactor core during the operation of nuclear power plants.
引文
[1]朱继洲,单建强,张斌.压水堆核电厂的运行[M].北京:原子能出版社,2008:141-143.
[2]Hofmann F,Archambeau F,Chaize C.Computational fluid dynamic analysis of a guide tube in a PWR[J].Nuclear Engineering&Design,2000,200(1-2):117-126.
[3]HJ Ji,BS Han.Coolant flow field in a real geometry of PWR downcomer and lower plenum[J].Annals of Nuclear Energy,2008,35(4):610-619.
[2]Hofmann F,Archambeau F,Chaize C.Computational fluid dynamic analysis of a guide tube in a PWR[J].Nuclear Engineering&Design,2000,200(1-2):117-126.
[3]HJ Ji,BS Han.Coolant flow field in a real geometry of PWR downcomer and lower plenum[J].Annals of Nuclear Energy,2008,35(4):610-619.