A review of sub-channel thermal hydraulic codes for nuclear reactor core and future directions

Thermal hydraulic analysis of nuclear reactor core is mainly performed using the sub-channel analysis codes to estimate different thermal hydraulic safety margins. The safety margins and the operating power limits of nuclear reactor core under different conditions of primary system i.e. system pressure, coolant inlet temperature, coolant flow rate and thermal power and its distributions are considered as the key parameters for sub-channel analysis. Considering the complexity of rod bundle geometry, different turbulent scales and due to their limitations of computational resources, performing the full scale computational fluid dynamic (CFD) analysis of nuclear reactor core is a cumbersome and time consuming task. Hence, the thermal hydraulic safety margins of most of the reactors operating in the world are carried out using the sub-channel analysis codes. In these codes, the governing equations of mass, momentum and energy are solved in control volumes which are connected in both radial and axial directions. The flow distributions in the rod bundle geometry are estimated by considering lateral momentum balance and the inter channel mixing models to account for the cross flow between the adjacent sub-channels. The accurate estimations of the local conditions of the sub-channels are required to predict fuel temperature, critical heat flux ratio (CHFR) and critical power ratio (CPR). In this paper, various experiments conducted on the different geometries of sub-channels related inter sub-channel mixing in rod bundles are identified. A comprehensive review of sub-channel thermal hydraulic codes used for the analysis of nuclear reactor core is presented. This review covers various aspects of experimental, analytical and computational works related to rod bundles carried out in the past and brings out future directions derived from earlier research works.