Neutron flux distribution calculation for various spatial mesh of finite slab geometry using one-dimensional diffusion equation

Abstract

One of the crucial problems of the nuclear reactor core is to predict the distribution of neutron flux. Neutron diffusion equation is widely used as an approximation to solve the neutron transport problem. In this study, neutron flux distribution with variation of spatial mesh in the finite slab geometry using one-dimensional diffusion equation has been evaluated. It provides a simple example of a slab geometry reactor design for fixed diameters, but the spatial meshes are varied in all region. The computation procedures to calculate the neutron flux distribution are conducted as follows: The boundary conditions of initial neutron source and neutron flux are determined. Macroscopic cross section of Uranium is regulated in homogeneous region and taken from the reference. The diffusion coefficient is arranged as fixed value. Furthermore, evaluation of numerical neutron flux value was carried out using the Jacobi method. The results show that with increasing spatial mesh, the neutron flux will decrease, but the neutron flux distribution pattern remained the same.