A three-dimensional two-energy-group finite-difference neutron-diffusion code with discontinuity factors

Full-core neutronic calculations are usually carried out in diffusion theory, using fuel-bundle-size homogenized nodes. A simple way to evaluate the neutronic parameters for the homogenized nodes is to calculate flux-weighted fuel-bundle-averaged multigroup macroscopic cross sections. This is known as Standard Homogenization (SH). SH produces satisfactory results for mildly heterogeneous configurations but fails to do so for more heterogeneous ones. To address this issue, advanced homogenization techniques have been developed, such as Superhomogenization (SPH) and Generalized Equivalence Theory (GET). While SPH can be used with any neutron diffusion code without requiring it to be altered, GET requires that some changes be implemented into the numerical formalism employed for the solution of the multigroup diffusion equations, in particular the introduction of discontinuity factors at node interfaces. The present work describes the implementation of discontinuity factors into a simple two-energy-group three-dimensional finite-difference diffusion code. Test results for simple configurations are also presented. (author)