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

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