SUBJECT: M.S. Thesis Presentation
BY: Christopher Edgar
TIME: Friday, June 21, 2013, 1:00 p.m.
PLACE: Boggs, 3-39
TITLE: Improvements to the Pool Critical Assembly Benchmark using 3-D Discrete Ordinate Transport with Adaptive Difference
COMMITTEE: Dr. Glenn Sjoden, Chair (NRE)
Dr. Bojan Petrovic (NRE)
Dr. Ce Yi (NRE)
Dr. Chaitanya Deo (NRE)


The Pool Critical Assembly Pressure Vessel Benchmark (PCA Benchmark), developed and performed by Oak Ridge National Laboratory is an industry standard benchmark which can be performed for partial establishment of the qualification of a radiation transport methodology for pressure vessel neutron fluence calculations. The geometry presented in the PCA Benchmark is well representative of the shielding methods used to reduce neutron fluence in the pressure vessel in a modern pressurized water reactor. The geometry, source definition, and material specifications are generally well characterized in the PCA Benchmark documentation, as are the supplied measured results for various foil sample interactions at specific locations varying in distance from the face of the reactor core.

As originally implemented, the PCA benchmark was performed with the DORT 2-D Sn computer code and the multigroup cross sections libraries BUGLE-93, SAILOR-95, and BUGLE-96. The 2-D flux synthesis method was utilized to combine one and two-dimensional transport calculations to obtain an approximate three-dimensional result using the DOTSYN code. The benchmark was performed in this research using the 3-D Parallel Environment Neutral-Particle TRANsport Sn code (PENTRAN). The PENTRAN code system can be used for fully decomposed (angle, energy, spatial) parallel 3-D Cartesian multigroup forward or adjoint discrete ordinates (Sn) simulations. The SN method is a deterministic approach that discretizes the angle, energy, and physical spatial variables into a finite number of discrete angular ordinates, energy groups, and spatial grids over the entire phase space system. Thus, the use of flux synthesis was not required. Furthermore, this research expanded on the methodology provided in the benchmark to provided a more representative source term in the reactor core, as well as flux and volume weighted neutron cross sections for the reactor core region.