ABSTRACT
The physical phenomenon of neutrons transport associated with eigenvalue problems appears in the criticality calculations of nuclear reactors and can be treated as a diffusion process. This paper presents a method to solve eigenvalue problems of neutron diffusion in slab geometry and one energy group. This formulation combines the Finite Element Method, considered an intermediate mesh method, with the Spectral Green's Function Method, which is free of truncation errors, and it is considered a coarse mesh method. The novelty of this formulation is to approach the spatial moments of the neutron flux distribution by the first-order polynomials obtained from the spectral analysis of diffusion equation. The approximations provided by the proposed formulation allow obtaining accurate results in coarse mesh calculations. To validate the proposed method, we compared its results with methods described in the literature. The accuracy and computational performance of our formulation were characterized by solving a benchmark problem with a high degree of heterogeneity.
Keywords:
Eigenvalue problems; Neutron diffusion equation; Spectral Green's function
