Quasidiffusion method realization for fast reactor critical parameters calculation in 3D hexagonal geometry

Strategy for multigroup neutron transport equation calculation on the basis of quasi-diffusion method, aimed at finding critical parameters of fast reactors, capable to operate in self-adjustable mode, is described. The numerical method is based on Gol’din’s quasi-diffusion method for multigroup neutron transport equation solving. Approximation for all types of high and low orders equations are suggested. The method of high order transport equation solving is based on developed earlier conservative method. Application of quasi-diffusion method for solving eigenvalues problem of neutron transport leads to essential decreasing of required number of source iterations with simultaneous increasing of the accuracy. Computations of parameters of active zone of uranium-plutonium fast reactor of BN-800 type are carried out for 3D $x-y-z$ hexagonal geometry, reflecting structure of the reactor active zone.