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.
Description
Approved for public release; distribution unlimited.
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