Low-cost calculation of multigroup neutron transport equations for time-to-time recalculation of averaged over spectrum cross-sections

Methodology for multigroup neutron transport equation calculation for 2D r-z geometry on the basis of quasi-diffusion method is described. The application of quasi-diffusion method for solving eigenvalues problem of neutron transport leads to essential decreasing of needed amount of source iterations with simultaneous increasing of the accuracy. Results for determining of parameters of active zone of uranium-plutonium fast reactor of BN-800 type which can operate in self-adjustable mode are posted.