Using software based on the Monte-Carlo method for receiving the few-group homogenized macroscopic interaction cross-section

The constant preparation issues are important for performing the few-group analysis for different states of the reactor core. Besides, the constant preparation method influences the further calculations accuracy and quality. The transport software products (deterministic codes) are usually used for the few-group characteristics preparation. On the basis of the neutron transport theory these codes calculate neutron fluxes depending on the energy and on the location in the cell. In present paper the description of fuel assembly calculation scheme for preparing the few-group characteristics is given for the Serpent code. This code uses the Monte-Carlo method and energy continuous microscopic data library. Serpent code was developed for calculating the fuel assembly characteristics including burnup calculations and preparation of the few-group homogenized macroscopic interaction cross-sections for the core calculating. The calculation scheme for the Serpent code for FAA and the results of the basic neutron-physical characteristics comparative calculations with PHOENIX-H and WIMSD5B codes are presented