Numerical Solution of a Two-Dimensional Multigroup Diffusion Equation for the Analysis of The Miniature Neutron Source Reactor (Mnsr)

Abstract

The lattice code WIMS was used to generate a two-group macroscopic cross-section data base for all homogeneous zones of the prototype Miniature Neutron Source Reactor (MNSR) for radial and axial directions. The data base takes into account the effect of increments in burnup, temperature and reactor power. The error analysis has shown that the data base is accurate enough for the purpose intended. The maximum deviation from the actual value is 0.2%. The two-dimensional two-group neutron diffusion eguation was solved numerically using the finite difference technique. A computer code called KWABEN is being developed to solve numerically the diffusion equation. The numerical methods and techniques used in the development of the code are presented in this work. Preliminary calculations with the code using the data base were carried out at 2 0 kW thermal power and the computed zone average thermal fluxes in the radial direction follow the same trend of results available from experiments. The value of the flux for the annular Be reflector where the inner irradiation sites are 11 2 located was determined by KWABEN to be 7.7 8x10 n/cm -s as 11 2 compared with the experimental value of 7.69x10 n/cm -s.