Hello everyone,
I am running simulations and have set the tally as ‘heating-local’ using the FENDL-3.2 nuclear data library. However, I am encountering an issue where I am not getting meaningful values from the simulations, only zeros.
I have also tried using ‘heating’ as the tally with the same library, and it works correctly. This makes me think that there might be a problem specifically with the ‘heating-local’ tally in conjunction with the FENDL-3.2 library (using different libraries it works).
Has anyone else experienced similar issues or have any suggestions on how to resolve this problem? It’s possible that there might be some limitations or factors I am overlooking in the FENDL-3.2 library that affect the ‘heating-local’ tally.
Any help or advice you can provide would be greatly appreciated.
Thank you in advance.
Davide Pettinari
Hi Davide. The “heating-local” cross section is only present when a data library is generated starting from the original ENDF files (using openmc.data.IncidentNeutron.from_njoy). For the FENDL 3.2 data library, that was generated from ACE files in which case we don’t have enough information to calculate “heating-local” since it requires some custom NJOY runs to get the right data. If you need heating-local for simulations with FENDL 3.2, I would recommend trying to generate your own library starting with the original FENDL 3.2 files. Our data repo has many examples of data library generation.