POLEX - a code for coarse-mesh calculation of neutron flux distribution in cores composed of hexagonal fuel assemblies

The POLEX code is applied to calculations of neutron flux and power density distributions within the core of WWER-type reactors. It allows to refine the meshes step by step both in the radial planes of the lattice and over the height of the core. The few-group diffusion equation is discretized over a (hex; z) lattice and transferred into a set of meshcentred finite-difference equations. The numerical solution of the eigenvalue problem is found by the power method using well-known numerical acceleration techniques. (author)