Using IRDFF-II Cross Section Data in OpenMC

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 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?

John Ball

@jlball welcome to the community and thanks for your question. You are going about this correct – unfortunately we don’t have a way at present to easily utilize dosimetry data. However, with a few lines of Python, it’s possible to extract data out of the ACE files and use them in an EnergyFunctionFilter for an OpenMC tally. I’ve put together a quick notebook demonstrating this. If you can do me a favor, give this a try and let me know if it works out for you. If so, I’ll add a function in OpenMC with this logic to make it easier in the future to get data out of those dosimetry ACE files.