Introduction to Neutron Diffusion Theory (general aspects, limitations, and applications); finite difference method to approximate the neutron flux derivative; discretization of the diffusion equation for a homogeneous one-dimensional and two-dimensional reactor; discretization of the diffusion equation using a mesh-centered integration scheme for a heterogeneous slab reactor; discretization of the diffusion equation using an interface-centered integration scheme for a heterogeneous slab reactor; power method and its application in Neutron Diffusion Theory; outer and inner iterations; solution methods for the diffusion equation (Thomas algorithm, Jacobi method, and Gauss–Seidel method); convergence criteria (eigenvalue and neutron flux); convergence acceleration techniques; introduction to the Fortran programming language and algorithms.

Credits: 3.0 (three)