Development of reactor vessel thermal-hydraulic analysis method in natural circulation conditions; Investigation of interwrapper Gap model

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

We have been developed a thermal-hydraulic analysis model in the reactor vessel using the computational fluid dynamics code with a low computational cost to evaluate core-plenum interactions during a natural circulation decay heat removal using a dipped-type direct heat exchanger in a design of sodium-cooled fast reactors. In this study, we investigate the coarse mesh modeling of interwrapper gap (IWG) using correlations for the purpose of the development of a practical model which can reduce the computational cost maintaining the prediction accuracy. An influence of combinations of the coarse mesh and the correlation for pressure loss in the IWG on the thermal-hydraulics and the core temperature distribution is revealed through the numerical analysis of a sodium experiment.