NEUTRON TRANSPORT CALCULATIONS OF THE AIR FORCE NUCLEAR ENGINEERING TEST FACILITY CORE.
AIR FORCE INST OF TECH WRIGHT-PATTERSON AFB OHIO SCHOOL OF ENGINEERING
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The neutron flux in segments of the core of the Air Force Nuclear Engineering Test Reactor was investigated using transport theory. Two portions of the core were treated the fuel plates and the control rods. These regions were represented as a one-dimensional array of thin slabs. A matrix collapsing technique for the solution of many-region, slab-geometry problems using the P-3 approximation was developed and a computer code to make the calculations was written. It was determined that no flux weighting was necessary in calculating homogeneous cross sections for the fuel regions. A technique was developed to compute a simplified control rod model, but computing difficulty prevented its being tested. Author