Using MGXS from NJOY

Development
1

Hi all,

I have a desire to use MGXS generated with NJOY in OpenMC. I am leaning towards a new method analogous to “from_endf” under the openmc.data.IncidentNeutron class, namely “from_gendf”.

I am envisioning that, with this workflow, one would still run in continuous mode, but with obviously far less energy resolution.

Without spending much time under the hood, I thought I would make a post here for buy-in from the developers. Im thinking that because OpenMC does not unionize the energy grids across all isotopes that this method would work fine, but I don’t know for certain.

Any and all help is greatly appreciated!

Thanks,

Matt

2

Hi @mfnash and welcome to the community! What would be the benefit of still using continuous energy mode if you have multigroup cross sections? Why not just have the GENDF data converted to OpenMC’s multigroup format cross sections and then run in MG mode?

3

Thanks for getting back so quickly. Please correct me if I’m wrong, but running in continuous you can tally over specific reaction channels as opposed to only the lumped “fission”, or “absorption”, for example in MG.

4

Hi Matt,

I have a full build that reads in GENDF XS.
My application of the multigroup XS is in OpenMC Deplete.

https://github.com/yrrepy/openmc/blob/isomeric-gendf_Sync2/openmc/deplete/gendf.py