Thermal-hydraulic feedback model for the neutronic analysis of PWR cores

A computational system BGU-CORE for the analysis of the PWR cores was developed in the nuclear Engineering Department of the Ben-Gurion University in recent years. This thesis presents an effort to expand and to improve the 'BGU-CORE' system by addition of the thermal - hydraulic feedback module. The main purpose of the module is to improve tile quality of the neutronic analysis by taking into account the influence of the total core power output and especially power distribution between fuel assemblies. Basic approach was to account for different power density levels in assemblies by evaluating average moderator density in each fuel assembly considered as a single coolant channel. Simplified but adequately accurate numerical models were developed for the calculation of the moderator density variations with power and adjustment of the assembly macroscopic cross - sections to the moderator density variations. All models were discussed, tested and finally implemented into overall calculational system. Calculational sequence of the reactor core and detailed discussion of the Calculational approach to thermal, hydraulic feedback is devoted to the evaluation of the moderator density in the coolant channel as a function of the power generated in fuel rod. It should be noted that 'BGU-CORE' is based on a 2-D description of the reactor core. The result of the analysis is a simple and consistent numerical model for the calculation of the channel water (moderator) density. Sensitivity calculations of the macroscopic lattice cross - sections to the variations of the water density are presented. Further investigations of the dependence of the infinite multiplication factor on tile variations in tile macroscopic cross - sections caused by the changes in the water density was carried out, and as a result a set of a few cross - sections entries was selected an 'important one' for the feedback representation. (author). 37 tabs., 26 figs., 17 refs