A thermal neutronics coupling analysis method for lead based reactor core

Abstract In this research paper, a sub-channel thermal hydraulic analysis code is coupled with the point reactor neutron kinetics model with six group delayed neutron. The coupling code is mainly used to perform the transient calculation of ADS/lead based alloy cooled fast reactor. The thermal hydraulic model is used for calculating temperature distribution profile and the feedback temperature information, providing input parameters for point kinetic model. This sub-channel analysis model can provide a new approach to solve the problem of one-dimension thermal hydraulic model and simulate the temperature distribution accurately. Furthermore the accuracy and reliability of calculated results are tested by another coupled code named FLUENT/PK and good agreements are achieved. To improve computational speed, one equivalent assembly is used to replace the whole core and the study shows that using of equivalent assembly which has the same average outlet temperature with the core obtained more reasonable results. The effects of fuel rods pitch diameter P/D ratio on simulation results are discussed. The code is capable to the quick calculations and safety analysis for reactivity accidents.