Primary Development of Thermal-hydraulics Design Code for CFR600 Core

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.