Calculation of neutron flux distribution in WWER-440 core

The application is described of the difference network method for solving the multigroup diffusion equation in x-y geometry. Difference equations are derived for the triangular network with nodular points identical with the centres of the fuel assemblies. The methods are given of solving the difference equations. The HEXA programme is described designed for the calculation of the two-dimensional x-y distribution of the neutron flux in the WWER-440 reactor core and the possible applications of this programme in the calculation of the physical characteristics of the reactor in the course of the campaign. The results are given of calculations of the first fuel charge of the reactor of the V-1 nuclear power plant. (J.P.)