Numerical study on the thermohydraulics of near-critical water in rod bundle with spacer grids

Abstract

The supercritical water cooled reactor (SCWR) is stands out among the Generation IV reactors with high safety and economy. Understanding the fluid flow and heat transfer characteristics within rod bundles under near-critical conditions is crucial for ensuring the safety of the SCWR core. In this study, computational models utilizing the SST k-ω turbulence model and boundary-resolved grids were validated against heat transfer data for near-critical water within a rod bundle. Subsequently, the validated model was utilized to examine the impacts of system pressure, fluid temperature at inlet, mass flux, and heat flux on the heat transfer characteristics of rod bundle with spacer grids under subcritical and supercritical conditions. The study revealed that the heat-transfer coefficient under supercritical conditions is significantly larger than that under subcritical conditions. Additionally, the heat transfer coefficient increases with mass flux, while independent from pressure and heat flux.