Hello Everyone,
I’m trying to run a simple simulation of a TRISO fuel cylinder with a neutron flux tally but get the following error:
RuntimeError: Could not find nuclide N7 in the nuclear data library.
Initially, I thought this was because I had an “air” material specified in my code and there was an issue with the nitrogen element I was adding to that material. However, since then I’ve tried running the code without any instance of nitrogen and I still get the error.
I’ve tried using both endfb-vii.1-hdf5 and endfb-viii.1-hdf5 cross-section XML files with no success.
Here is how I am calling the cross-sections into my project:
openmc.config['cross_sections'] = 'Absolute Path'