超临界水堆子通道分析  

Subchannel Analysis of Supercritical Water Cooled Reactor

作　　者：赵冬建[1] 路璐[1] 史国宝[1] 

机构地区：[1]上海核工程研究设计院,上海200233

出　　处：《原子能科学技术》2009年第6期543-547,共5页Atomic Energy Science and Technology

基　　金：国家重点基础研究发展计划资助项目(2007CB029800)

摘　　要：超临界水堆作为6种第4代未来堆型中唯一的水冷堆,具有一些独特的特点,受到了广泛重视。本工作以上海核工程研究设计院的常规压水堆子通道程序为基础,开发编制了适用于超临界水堆的子通道程序,并对典型带有慢化剂水棒的超临界水堆燃料组件进行了模拟计算,得到了堆芯子通道内的温度、燃料棒包壳温度、表面传热系数等参数的分布规律。此外,研究了不同超临界流体换热关系式对计算结果的影响,结果显示,各传热关系式的计算结果存在一定差异。As the only water cooled reactor among the six generation-Ⅳ reactor, supercritical water cooled reactor （SCWR） has its special characteristics, and takes up attentions extensively. Based on the PWR subchannel analysis code of Shanghai Nuclear Engineering Research and Design Institute, a subchannel code for SCWR was developed. Furthermore, the thermohydralic characteristics of the typical SCWR fuel assembly with the moderator water rod, such as subchannel temperature, fuel rod cladding temperature, heat transfer coefficients and so on, were investigated using the developed subchannel code. Finally, a sensitivity study of different heat transfer correlations was performed on the thermohydralic characteristics. The results show that obvious difference occurs when using different heat transfer correlations.

关 键 词：超临界水堆 第4代反应堆 子通道程序 

分 类 号：TL364.4[核科学技术—核技术及应用]