The neutron transport equation solving coupled with the quasi-diffusion equations solving in r-z geometry

The improvement of 2D leakage estimation throughout top and bottom of reactor active zone on the basis of coupled decision of transport equation and quasi-diffusion equations for effective one group has been suggested. Coefficients for this system have been determinate throw ought averaging muligroup cross-sections with 1D calculation spectrum. It has been shown that scalar neutron flux calculated in 1D model with account of 2D leakage is correspond to integral average of neutron flux calculated in 2D model on height of active zone.