原子能科学技术2018，Vol.52Issue(1)：56-63,8.DOI:10.7538/yzk.2018.52.01.0056
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
周志伟 1杨红义 1李淞 1林超1
作者信息
- 1. 中国原子能科学研究院反应堆工程技术研究部,北京 102413
- 折叠
摘要
Abstract
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 assem-blies ,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Key words
CFR600/fast reactor/core design/sub-channel analysis/flow zoning分类
能源科技引用本文复制引用
周志伟,杨红义,李淞,林超..CFR600堆芯热工水力设计程序初步研发[J].原子能科学技术,2018,52(1):56-63,8.
