Mcu-fcp Program For First-collisions Probability Calculations Of Neutron Transport

摘要

A general scheme, methods, and algorithms, implemented in the MCU-FCP program, making it possible to calculate the neutron-physical characteristics of two- and three-dimensional RBMK cells and polycells taking account of fuel burnup during reactor operation by the first-collisions probability method, are described. Test calculations are preformed and the results are compared with the MCU-REA program.