Hallo Everyone…
I want to look for the distribution of fission reactions but when calling the program the tally file cannot be found even though I have included the tally program when running burn up. Please help me to fix this error🙏
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Hi @Rahma
I’m not sure, but the problem may be occur if tally data was not proprely loaded from the obtained files : simulation_nx.h5, also you need to check that your tally was taken into consideration by your model when running openmc, this could be verified by the creation of not-empty file tallies.xml and tallies.out at the end of depletion calculation.
In any case, a shot of the corresponding part of the used python code will be usefull to find the error source efficiently.
Regards
Sorry sir, can you show me how to fix it so the tally file can be found?