Hi OpenMC community,
I am figuring out how I can tally reaction rates, say (n,p), for activation foils/wires without explicitly modelling them in the geometry. In this scenario I am assuming self-shielding effects can be neglected. In MCNP, this can be done using just material numbers and a tally multiplier.
I tried tally.nuclide = [‘Al27’] in conjunction with (n,p) score and a mesh filter over the moderator region. This region do not contain actual cells filled with Al-27, so material is not added in openmc.Materials. Results I got is mostly zero except for 1 mesh bin, which is not what I expected, either I got non-zero values or all zero values.
Does OpenMC require that the material is present in the model for it to perform the reaction rate scoring?
If the above is true, how can I proceed with tallying reaction rates without the target materials added to the model? I suppose there might be a way using EnergyFunctionFilter. Can you provide guidance on how I can use the xsection data this manner?
I greatly appreciate your attention to this matter.