High-fidelity thermal-hydraulic investigation on the whole RPV with integrated CFD multiscale modeling method

Accurate prediction of coolant flow and heat transfer characteristics within the reactor pressure vessel is fundamental to ensuring the safety margins of nuclear reactors. Conventional thermal-hydraulic analyses often decouple the plenum flow from the core heat transfer processes, making it difficult to capture the complex cross-scale coupling effects within the RPV. In this study, an integrated multiscale simulation framework for the entire PWR RPV was constructed by combining the core sub-channel-scale precision of the CorTAF code with high-fidelity CFD models of the upper and lower plenums. Based on the ROCOM benchmark, the reliability of the proposed coupling scheme in predicting flow fields was validated. On this basis, the flow distribution patterns in the lower plenum, the thermal-hydraulic distributions induced by lateral turbulent mixing within the core, and the coolant mixing characteristics in the upper plenum of the typical advanced passive PWR were analyzed in depth, systematically elucidating the flow field and heat transfer evolution mechanism throughout the entire vessel. The research elucidates the physical feedback mechanisms among key components and provides an in-depth analysis of the thermal-hydraulic phenomena under the coupled framework. The calculation results provide refined thermal-hydraulic parameter distributions within the core that are difficult to obtain using traditional porous media models. This work provides an important reference for improving the accuracy of thermal-hydraulic safety analysis for large-scale pressurized water reactors.