Implementation of the Quasi-Diffusion Method for Calculating the Critical Parameters of a Fast Neutron Reactor in 3D Hexagonal Geometry

摘要

A numerical technique for solving a multigroup neutron transport equation with quasi-dif-fusion aimed at determining critical parameters of fast reactors capable of long-term operation in a self-adjustable neutron nuclear mode (SANNM) is described. The method for solving a multigroup neutron transport equation is based on the Gol'din quasi-diffusion method. The conservative-characteristic method for solving the transport equation which was proposed earlier is extended to the case of three-dimensional (3D) hexagonal geometry. Approximation of quasi-diffusion equations is suggested. The effective algorithm is constructed based on all reactor arrangement symmetries possible in the case of the reactor operation in a self-adjustable mode. The calculations are performed for a 3D model of the active zone of the BN-800 type reactor capable of operation in SANNM. The results of the study can be used for dynamic simulation of active zones in fast reactors.