钠冷快堆组件冷却剂沸腾子通道分析方法研究  

Study on the Sub-channel Analysis Method of Coolant Boiling in the Sodium Cooled Fast Reactor Assembly

作　　者：方闻韬 佟立丽[1] 曹学武[1] FANG Wentao;TONG Lili;CAO Xuewu(School of Mechanical Engineering,Shanghai Jiao Tong University,Shanghai 200240,China)

机构地区：[1]上海交通大学机械与动力工程学院,上海200240

出　　处：《核科学与工程》2023年第3期544-552,共9页Nuclear Science and Engineering

基　　金：国家自然科学基金资助项目(U1967202)。

摘　　要：钠冷快堆发生超设计基准事故时,组件内冷却剂可能沸腾甚至干涸,准确预测其温度分布对钠冷快堆的安全评估具有重要意义。基于均相流模型构建守恒方程,采用Mikityuk对流传热模型以及Cheng-Todreas阻力模型等关系式,开发了适用于钠冷快堆两相流动模拟的子通道分析方法,与FFM-2A单相流动实验数据和KNS-37钠沸腾实验结果进行了对比验证,并与同类子通道分析程序的计算结果作比较,验证了方法的合理性。When the beyond-design-basis accident occurs for the sodium cooled fast reactor,the coolant in the assembly may boil or even dry up.Therefore,accurate prediction of coolant temperature distribution is important to the safety assessment of the sodium fast reactor.In this paper,the conservation equation is constructed based on the homogeneous flow model,and adopting the Mikityuk convective heat transfer model and the Cheng-Todreas resistance model,a sub-channel analysis method suitable for the two-phase flow simulation of sodium-cooled fast reactors is developed.The results are compared and verified with the data of FFM-2A steady state experiment and the KNS-37 loss-flow sodium boiling experiment and also compared with the calculation results of similar sub-channel analysis codes,which shows the rationality of the method.

关 键 词：钠冷快堆 钠沸腾 子通道分析 

分 类 号：TL333[核科学技术—核技术及应用]