Uncertainty calculations with openMOC for lattice reactor physics

OpenMOC is a deterministic neutron transport code using the method of characteristics (MOC) for full-core 3D reactor physics analysis developed at Massachusetts Institute of Technology. OpenMOC relies on user input for macroscopic cross-sections data and currently does not perform self-shielding of the multi-group macroscopic cross-sections and any burnup calculations. This work aims to add perturbation methods to OpenMOC concerning cross-section uncertainties and lattice perturbations. Typical measurement uncertainties of the thermal neutron flux in commercial LWRs are about 3 to 5% at the positions of measurement (fuel assembly guide tubes). In this work, a total 100 different input files were prepared for OpenMOC based on covariance information from the ENDF/B-VII. The final objective is to estimate the impact that covariance data have in the multi-group flux obtained by OpenMOC at a PWR lattice configuration domain. In the end, this project has successfully demonstrated how the total Monte Carlo method can in principle be used in conjunction with OpenMOC.