Abstract
In many deterministic methods, concerning the resolution of the neutron transport equation, a more global and practical representation of the angular flux is needed to provide us with useful and complete information on the neutron population in a reactor core. The purpose of this paper is to provide a reference quality calculation for the angular flux. The discrete ordinate method (S-N) is processed with a new matrix form which is used to model an isotropic and anisotropic multiplicative system in a region, multi-region for one energy group in a cartesian geometry. The obtained results are compared to those obtained by MCNP6 code.