Hello,
I’ve been playing around with nuclear data lately and was trying to use the GitHub - openmc-dev/data: Collection of scripts for managing data for OpenMC script collection to build the CENDL-3.2 cross-sections for OpenMC.
The dedicated script runs fine and the resulting files are valid, but they seem to be incomplete. For instance, the Pu239.h5 does not contain reaction MT=18 (fission), even though the CENDL-3.2 obviously has fission data for Pu239.
Has anyone tried using this script? I’m mostly interested in retrieving MT=301 for nuclides that are faulty in JEFF/ENDFB (see this) so I am not too interested in the fission cross-section by itself, but I am in still intrigued.
Cheers,
Nicolas