A TRANSPORT CALCULATION OF THE FLUX IN THE NUCLEAR ENGINEERING TEST REACTOR TEST-CELLS
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.
Document Details
- Document Type
- Technical Report
- Publication Date
- Jun 01, 1965
- Accession Number
- AD0621121
Entities
People
- Lee H. Livingston
Organizations
- Air Force Institute of Technology