Hello, I recently completed my hexagonal RBWR. There are a few ways to accomplish your objective using tallies with Filter like MeshFilter and CellFilter : Tally Arithmetic — OpenMC Documentation
Alternatively specify a material and geometry zone separately and utilize results.get_reaction_rate : Pincell Depletion — OpenMC Documentation
Additionally, you may view my post here; my code there may correspond to what you’re attempting to accomplish : Difference in Fission rate Tally value from MeshFilter and CellFilter