A TRANSPORT CALCULATION OF THE FLUX IN THE NUCLEAR ENGINEERING TEST REACTOR TEST-CELLS
AIR FORCE INST OF TECH WRIGHT-PATTERSON AFB OH SCHOOL OF ENGINEERING
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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.
- Nuclear Power Plants and Fission Reactor Engineering