Hi, I am trying to model the depletion of a core and would like to model some sort of off-gas system that removed gaseous poisons like Xenon-135. Is there a functionality in openmc to remove or change concentrations of nuclides during or in between depletion runs? Currently I am using the restart integrator and editing the depletion_results.h5 file in between each run to set Xenon concentrations to 0, although I am not sure if this is the best way to do it and have hit a segmentation fault after 15 or so iterations. Any help or suggestions would be appreciated.
Hi llgunnel7 and welcome to the forum!
We don’t have this functionality at this time. There have been some good discussions in
and
relating to how you could use some existing and new features (exporting depletion data to xml) to help achieve this.
Cheers,
Andrew
Thank you! I’ll look into those.
Hello @llgunnell7. Welcome to the forum. Can I know how you are using the restart integrator to edit depletion_results.h5 file? Because I happen to look the python APIs, I found no python API with the name restart integrator in the openmc.depletion section. Thank you!
Hi fsabab, the restart integrator just allows me to stop depletion runs between iterations. I manually edit the h5 file in between each run with the h5py library to change the count of some nuclides, but I’m sure there are other ways to do it.
Thank you for the answers 