Numerical analysis of multigroup neutron flux in a bare fast reactor

Abstract

This study consists of the development of a computer program to numerically solve the space and energy dependent multigroup neutron diffusion equations in a bare homogeneous fast reactor core or reactor material assembly. The resulting program is unique in that it was designed for future use by Naval Postgraduate School students undertaking experimental studies in neutron diffusion with limited time to determine numerical solutions for verification of their results . The equations are solved iteratively in cylindrical geometry using a point successive overrelaxation technique. Convergence between -6 successive iterations was less than 10 after fifty iterations. The program was tested using ANL three group data. Flux shapes and energy spectra were determined for a typical fast reactor core and for a solid iron cylinder with a source at its center. The program was also used to determine criticality. Computation times were from one to ten minutes with less than 15 OK words of core storage using the IBM 360/67.