Hello,
I am trying to perform a depletion followed by a photon transport of a material by using the depleted material as a photon source for the transport. I am using MicroXS for the depletion and extracting the results object from it, and manually creating a depleted material object using a method similar to this post Create materials.xml from depletion results? - #2 by paulromano. The issue I am running into is that many of the nuclides I need for the photon transport are not included in the standard XS library, such as Al28, and ~20 other nuclides.
This is not ideal since according to this documentation openmc.deplete.Results — OpenMC Documentation, if the nuclide does not have XS data, it cannot be used for transport calculations. So when openmc.run() is called on my photon transport, I get this error
RuntimeError: Could not find nuclide Al28 in the nuclear data library. Abort(-1) on node 0 (rank 0 in comm 0): application called MPI_Abort(MPI_COMM_WORLD, -1) - process 0
Is there a way around this issue to not require the neutron data for these nuclides in order to do a photon transport? I have the source.particle = ‘photon’ and settings.photon_transport = True.
Thanks,
Alex