I’m interested in using the IRDFF-II cross section library ([International Reactor Dosimetry and Fusion File (IRDFF-II)]) for simulating certain dosimetry reactions, but I’m having a tough time getting the IRDFF data into the OpenMC HDF5 cross section format. On the website linked above there are download links for the library in ENDF-6 and ACE format, which I’ve tried to convert to the OpenMC format using the openmc.data module with no success.
When I try using the ENDF file I get the following NotImplementedError, which I assume means that this feature simply isn’t available yet:
NotImplementedError: Cannot export incident neutron data that originated from an ENDF file.
When I try using the ACE files I get the following TypeError:
TypeError: <ACE Table: 14028.34y> is not a continuous-energy neutron ACE table.
Am I going about this correctly, or is it possible at all in the current version of OpenMC?
Thanks,
John Ball