Code development for transient thermal-hydraulics and safety analysis of sodium-cooled fast reactor

Sodium-cooled fast reactor is one of the Generation-IV nuclear reactors. Research on its transient thermal-hydraulic and safety characteristics is the significant work for its design and licensing, which requires specific analysis tools. Based on the modular modeling idea, the thermal-hydraulic models and auxiliary models were established for the key components of the sodium-cooled fast reactor system. The Gear algorithm with high stability and automatic step-changing capacity was adopted to develop the transient thermal-hydraulic analysis code for sodium-cooled fast reactors (THACS), which was validated by the protected loss-of-flow test SHRT-17 of the international benchmark EBR-II. The results show that THACS can simulate the test transient process well, which indicates that THACS code has possessed the capability to perform the transient thermal-hydraulics and safety analysis for sodium-cooled fast reactor, and support the development of China sodium-cooled fast reactor. (authors)