Development of reactor vessel thermal-hydraulic analysis method in natural circulation conditions; Applicability investigation for transient analysis

Hamase, Erina ; Miyake, Yasuhiro*; Imai, Yasutomo*; Doda, Norihiro  ; Ono, Ayako ; Tanaka, Masaaki  

In a design study of sodium-cooled fast reactors, we have developed the practical reactor vessel thermal-hydraulic analysis method (RV-CFD) that had a low computational cost about the thermal-hydraulics in the core to evaluate the core-plenum interactions occurred in the natural circulation decay heat removal during the dipped-type direct heat exchanger operation. In this study, the non-equilibrium thermal model which considered the thermal inertia of fuel pins was developed and incorporated into the core of RV-CFD. Through the transient analysis simulating the power reduction due to reactor scram, the applicability of RV-CFD to the transient analysis was confirmed.