The internationally circulated Pool Critical Assembly (PCA) Pressure Vessel Benchmark was analyzed using the PENTRAN Parallel SN code system for the geometry, material, and source specifications as described in the PCA Benchmark documentation. Improvements to the benchmark are proposed through the application of more representative flux and volume weighted homogenized cross sections for the PCA reactor core, which were obtained from a rigorous heterogeneous modeling of all fuel assembly types in the core. A new source term definition is also proposed based on calculated relative power in each core fuel assembly with a spectrum based on the Uranium-235 fission spectra. This research focused on utilizing the BUGLE-96 cross section library and accompanying reaction rates, while examining both adaptive differencing on a coarse mesh basis, as well as the sole use of Directional Theta-Weighted (DTW) SN differencing scheme in order to compare the calculated PENTRAN results to measured data. The results show good comparison with the measured data, which suggests PENTRAN is a viable and reliable code system for calculation of light water reactor neutron shielding and dosimetry calculations. Furthermore, the improvements to the benchmark methodology resulting from this work provide a 6 percent increase in accuracy of the calculation (based on the average of all calculation points), when compared with experimentally measured results at the same spatial location in the PCA pressure vessel simulator.