Accession Number:

AD0728657

Title:

Numerical Analysis of Multigroup Neutron in a Bare Fast Reactor.

Descriptive Note:

Master's thesis,

Corporate Author:

NAVAL POSTGRADUATE SCHOOL MONTEREY CALIF

Personal Author(s):

Report Date:

1971-06-01

Pagination or Media Count:

131.0

Abstract:

The 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 successive iterations was less than 0.000001 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 composition with and without a centrally located Pu-Be point source 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 150K words of core storage using the IBM 36067. Author

Subject Categories:

  • Computer Programming and Software
  • Fission Reactor Physics

Distribution Statement:

APPROVED FOR PUBLIC RELEASE