Accession Number:

AD0621121

Title:

A TRANSPORT CALCULATION OF THE FLUX IN THE NUCLEAR ENGINEERING TEST REACTOR TEST-CELLS

Descriptive Note:

Master's thesis

Corporate Author:

AIR FORCE INST OF TECH WRIGHT-PATTERSON AFB OH SCHOOL OF ENGINEERING

Personal Author(s):

Report Date:

1965-06-01

Pagination or Media Count:

77.0

Abstract:

This report describes a calculation of the neutron flux in the test- cells using the discrete S sub n approximation to the transport equation in xy- geometry. The calculation was done by writing a computer program, S4C40, for the IBM 7094 using three energy groups and 1600 mesh points. Fast cross- sections were generated with the General Atomics code GAM-I, and thermal cross- sections were calculated by hand assuming a Maxwell-Boltzmann distribution. The results, which were compared with the multigroup, two-dimensional diffusion theory code PDQ, show a significantly higher thermal flux over the entire reactor and a more slowly decreasing flux for all groups in the test-cell.

Subject Categories:

  • Nuclear Power Plants and Fission Reactor Engineering

Distribution Statement:

APPROVED FOR PUBLIC RELEASE