Hello Forums,
I’m trying to run a fixed-source calculation with the fixed source defined based on a depleted materials file. My method is similar to the rigorous 2-step method described in the documentation: 11. Decay Sources — OpenMC Documentation
I’m getting all the way to Step 6 with no issues but get a segmentation fault as soon as transport calculations start. Has anyone seen this before?
I’m running v0.13.3, and should note that the source definition is very complex (~400 source objects), as each fuel assembly material is defined individually to capture variations in material depletion due to variations in flux.