I’m new in OpenMC. I tried to run a depletion calculation on a hexagonal assembly (similar to the example) i used
diff_burnable_mats = True
it produced so many materials with many id’s. How can i know which mat files belongs to which cell??(position inside the assembly) also,
I can plot & see the changes of concentration with burnup for those materials with different id’s individually. Can i find the Keff value for each individual cells inside that assembly as well?