Investigation of the Fuel Shape Impact on the MTR Reactor Parameters Using the OpenMC Code

Abstract

The goal of this study was to evaluate the impact of simulating different fuel shapes for the material testing reactor (MTR). Two OpenMC codes were built, and the first OpenMC model was simulated using a curved shape fuel element to mimic the real dimensions and shape of the MTR. The code core parameters were validated with the collected parameters from the experimental work and two well-known Monte Carlo simulation codes (MCNP and SCALE). The validation process included the axial flux profile and criticality. After the OpenMC curve fuel model was validated, the MTR fuel was simulated as flat fuel elements with the exact amount of fuel as in the curve fuel model. By comparing the two OpenMC models’ calculations, it was observed that the radial flux distribution has only a slight difference due to fuel mass similarity. In conclusion, simulating the MTR fuel as flat elements provided a good agreement calculation compared to the real shape, but it was also observed that this might carry some discrepancies for in-depth simulation studies.