Hi,
I’m getting this error when I try to run the depletion example after integrator.integrate(). I’m using the ENDF/B-VII.1 chain, so I’ve also downloaded the ENDF/B-VII.1 cross_sections.xml . I believe I’m getting this error because I haven’t properly configured my code to read this cross sections file. Sorry if this is on the cross sections configuration page somewhere and I’m not getting it. I’ve tried to use openmc.Materials.cross_sections but I end up with the same error. I have cross_sections.xml open in the jupyter notebook. How could I fix this?
Thank you
If settings openmc.Materials.cross_sections doesn’t work, the alternative would be to set the OPENMC_CROSS_SECTIONS environment variable as you’ve probably seen in the documentation. Note that you’ll have to configure it such that the Jupyter notebook inherits the environment in which that environment variable is set (a good check is to call os.environ['OPENMC_CROSS_SECTIONS'] from your notebook and see if it’s set).
I also see that you’re running from Windows – none of our developers routinely run on Windows, so it’s possible there may be some issues to resolve. Perhaps others on the forum have experience getting it to work on Windows and can comment?
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I’ve directed the path to the cross sections file in settings.xml and materials.xml and renamed cross_sections.xml to Across_sections.xml but still get the same missing nuclide error. Is there another way to possibly fill in this data somehow? I know this is Windows but I appreciate any help or ideas.
Thank you
@cparrott For reference, did you compile OpenMC through WSL?
I did that and then tried to run on Ubuntu, but I’m not sure how to access the Ubuntu files, so I don’t know how to put the cross_sections.xml and chain_casl_pwr.xml/chain_endfb71_pwr.xml in the bin file with openmc executable. This was what my initial plan was. However, I was able to get a Sm149 build up model in a TRIGA on Jupyter, as the TRIGA example doesn’t use any material with missing nuclide nuclear data. My purpose for using OpenMC is to model SM149 build up in TRIGA reactors, so I believe this will suffice, although I would like to fix the missing nuclide errors if I could for the future.
Thank you
@cparrott You might want to have a look at the Microsoft documentation on working across filesystems.
