Corner neutronic code

The neutronic code CORNER is based on the SN discrete ordinates method [1] and the PM scattering cross-section approximation. It is intended for high-precision deterministic neutronic calculations of fast-neutron reactors and can be used to solve two types of steady-state neutron transport and gamma quanta problems in a 3D hexagonal geometry: Keff problems (homogeneous) and source problems (inhomogeneous).The code is developed in the Fortran language and has a modular structure. Its key modules are a module for the preparation of neutron constants in the ANISN format; a geometric module containing a description of the core's loading map and fuel assembly types, including their axial meshing and material composition; a module for preparing angular quadrature sets; an input data module containing the parameters of the approximation used and the control parameters; a neutronic calculation module and a calculation data processing module.The Directional Theta-Weighted (DTW) difference scheme [2] has been built to approximate the spatial dependence. It has advantages over the DD (Diamond Difference) scheme broadly used in coarse-mesh problems.The energy dependence is represented by multigroup approximation. The angular variable is discretized by introduction of the angular quadrature set. Quadrature sets can be also defined by the user.An iterative solution process is used, including external iterations for the fission source and internal iterations for the scattering source. The paper presents the results of a cross-verification against the Monte Carlo MMK code [3] and on a model of the BN-800 reactor core.