When I intend to tally macrosection of Xe135 in different depletion steps , the progress repoort " ERROR: Could not find the nuclide Xe135 specified in tally 1 in any material".
The filter set in fuel cell which concludes U235 and U238. In addtion, tally file is added before integrator
@Lucky_Kevin welcome to the community! This is a limitation in the current version of OpenMC but has actually already been fixed in our developmental version and will be part of the next release.
Okay ,thank you for replying me!