CFR600堆芯热工水力设计程序初步研发  

Primary Development of Thermal-hydraulics Design Code for CFR600 Core

作　　者：周志伟[1] 杨红义[1] 李淞[1] 林超[1] ZHOU Zhiwei;YANG Hongyi;LI Song;LIN Chao(China Institute of Atomic Energy,P.O.Box 275-95,Beijing 102413,China)

机构地区：[1]中国原子能科学研究院反应堆工程技术研究部,北京102413

出　　处：《原子能科学技术》2018年第1期56-63,共8页Atomic Energy Science and Technology

摘　　要：针对快堆结构,自主开发了适用于CFR600快堆热工水力设计及优化的程序,并进行了相关测试验证。结果表明,该程序具备快堆全堆图形建模、精细化子通道自动划分、考虑组件间换热的热工水力分析以及流量自动分区优化等功能,可为后续自主知识产权的商业快堆技术研发提供支持。A code was developed for CFR600 fast reactor core thermal-hydraulics design and optimization,and was tested and validated.It supplies the abilities such as quick modelling of full core by GUI(graphical user interface),automatic fine sub-channel meshing,thermal-hydraulic analysis considering heat transfer among neighbor assemblies,and automatically optimizing flow zoning.Furthermore,it will be used to support the independent intellectual technology innovation of commercial fast reactor power plant.

关 键 词：CFR600 快堆 堆芯设计 子通道分析 流量分区 

分 类 号：TL333[核科学技术—核技术及应用]