The coarse mesh radiation transport (COMET) code uses response functions to solve the neutron transport equation. Most nuclear codes used today have a very steep learning curve; COMET is no exception. To ease the user's onus of learning how to create correctly formatted COMET input-files, a graphical user interface (GUI) was created. The GUI allows the user to select values for all the relevant variables while simultaneously minimizing the errors a typical new user would make. To this end, the GUI creates all of the input files required to run COMET. The GUI also provides a visualization tool that the user may use to check the problem geometry before running COMET. The GUI is also responsible for post-processing the COMET output for visualization with TecPlot. In addition to the GUI, multi-group cross section libraries were generated as part of the MHTGR-350 (Modular High Temperature Gas Reactor) benchmark problem under development at Georgia Tech. This project aims to couple COMET with a thermal hydraulics code to best model the true physics of the reactor design. In order for this goal to be actualized, six-group cross sections were generated over the operational temperature range of the MHTGR using the current coupling and collision probability code HELIOS.

MANE (MCNP ACE from NJOY & ENDF), a code for generating continuous energy cross sections at arbitrary temperatures, was created. Cross sections were evaluated using NJOY99 such that they would agree with the cross sections provided by MCNP5. The MANE cross sections were found to be in very good agreement with those provided by MCNP5 with some minor exceptions caused by round-off errors and some differences in the unresolved resonance region. Differences in the resonance region are caused by differences in the random number generator used to start the cross section calculations. The MANE cross sections were verified against the MCNP5 cross sections in five unique MCNP configurations: an 8.7% enriched MOX fuel pin cell, a UO₂ assembly (controlled and uncontrolled), a MOX assembly, and a whole core configuration containing the 3 assemblies. In each of these cases, eigenvalue and tally density results were found to be in very good agreement with one another.

This thesis presents a whole-core benchmark of the European Pressurized Reactor (EPR) using multiple transport methods. The core specifications were taken directly from the Final Safety Analysis Report (FSAR) submitted to the Nuclear Regulatory Commission (NRC) and the reactor was modeled in a stylized manner while maintaining full heterogeneity at the pin and assembly level. The geometry and material specifications are given as well as problem-specific cross sections for 2, 4, and 8 energy group calculations. Cross sections were generated using HELIOS, a lattice depletion code based on the Collision Probability Method (CPM). The multi-group cross sections were utilized in the reference calculation, COMET calculation, and response function generation. The reference solution was obtained via an MCNP model identical to the one implemented in COMET. Specific steps towards constructing and running a COMET calculation are outlined. Detailed results including assembly eigenvalues, core eigenvalues, and pin fission densities are presented.