Review on flow and heat transfer processes in rod bundle channels of water-cooled nuclear reactor

The thermal–hydraulic analysis of coolant flow in the reactor core plays a significant role in the optimized design of nuclear fuel assemblies and nuclear safety. This article reviews the experimental and Computational Fluid Dynamics (CFD) methods for the thermal–hydraulic characteristics of subchannels with water as the coolant in different rod bundle geometries and flow conditions over the past few decades. It summarizes the effects of design parameters such as rod spacing and spacer grid type, as well as flow parameters like pressure, mass flow rate, and heat flux on fluid flow and heat transfer. Additionally, various models used for numerical simulations of rod bundle channels are introduced, providing valuable references for research and practice in related fields.