AP1000燃料组件的热工水力研究
发布时间：2018-10-16 09:44
【摘要】：在核电厂的正常运行及事故工况下,都会出现非常复杂的热工水力现象,而反应堆又是一种结构紧密、单位体积释热率极高的热源,因此需要提高其设计的可靠性来保证堆芯不会损坏;并确保反应堆即使在严重事故工况下也不会导致放射性物质的泄漏。为了获得堆芯热工参数在稳定运行及事故中的变化过程,需要进行可靠的反应堆热工水力分析。本文通过计算流体力学程序Fluent和一维子通道程序COBRA-Ⅳ的计算,研究组件内部温度场和速度场的分布情况,并将Fluent与COBRA-Ⅳ的计算结果进行对比,以验证Fluent程序在计算三维组件时的准确性。首先,通过这两种软件计算3×3组件在发生失流事故时出口温度的变化过程,再对AP1000八分之一组件稳态运行时的内部温度场和速度场的分布情况进行模拟计算,并初步研究格架对流动的影响及计算在不同模型下格架的阻力系数。计算表明,Fluent程序在不同工况下计算得到的子通道内的温度分布与一维子通道程序COBRA-Ⅳ的计算所得结果相比,两者趋势一致,且出口温度的误差在1%以内,从而证明改进后的Fluent程序适用于AP1000的堆芯计算;在分析流场时,通过CFD程序与子通道程序相结合的分析方法,一方面可以更直观的表示通道内轴向流速沿堆芯高度的分布情况,并能观察格架对冷却工质横向流动的影响,另一方面还可以准确的计算出子通道间的横向和轴向的流速大小;由于模型的简化及相关尺寸参数的缺乏,通过Fluent程序计算得到的格架阻力系数与COBRA-Ⅳ文件内的参考值有较大偏差,但仍有一定的参考性。通过Fluent程序与COBRA-Ⅳ程序的结合使用,既能得到全面直观的三维结果和局部热工流体特征,又能快速有效的得出DNBR以及燃料棒内部温度的分布情况。因此在热工分析中,通过两者的配合使用,既能提高计算效率,也可满足不同的计算要求。
[Abstract]:Under the normal operation and accident conditions of the nuclear power plant, there will be very complicated thermohydraulic phenomena, and the reactor is a kind of heat source with a compact structure and extremely high heat release rate per unit volume. Therefore, it is necessary to improve the reliability of its design to ensure that the core will not be damaged and that the reactor will not cause leakage of radioactive material even under serious accident conditions. In order to obtain the variation process of reactor core thermal parameters in stable operation and accident, reliable reactor thermohydraulic analysis is needed. In this paper, the distribution of temperature field and velocity field in the module is studied by the calculation of the computational fluid dynamics program Fluent and the one-dimensional subchannel program COBRA- 鈪，
本文编号：2273961
[Abstract]:Under the normal operation and accident conditions of the nuclear power plant, there will be very complicated thermohydraulic phenomena, and the reactor is a kind of heat source with a compact structure and extremely high heat release rate per unit volume. Therefore, it is necessary to improve the reliability of its design to ensure that the core will not be damaged and that the reactor will not cause leakage of radioactive material even under serious accident conditions. In order to obtain the variation process of reactor core thermal parameters in stable operation and accident, reliable reactor thermohydraulic analysis is needed. In this paper, the distribution of temperature field and velocity field in the module is studied by the calculation of the computational fluid dynamics program Fluent and the one-dimensional subchannel program COBRA- 鈪，
本文编号：2273961
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