I’m unfortunately not familiar with the neutron tally files from FLUKA, so it’s difficult for me to give specific advice on this issue. Are you suggesting that you want a source of different intensity in each of the cylindrical sub-volumes? If you could be more specific with what you’re trying to do, perhaps I can advise further.
Beyond the rudimentary source options (source uniformly distributed in a box, point source), there is also an option to use a source file which contains the positions, directions, and energies of source sites. However, normally this file is generated directly from an OpenMC run. It would be possible to have a script create a source file though.
Best,
Paul
Hi NiuOBA,
I’ve attached a Python script that shows an example of creating a source file. The documentation for the upcoming release of OpenMC has a detailed file specification for the source file, but you can see from the script what goes into it. The source sites are stored in a numpy array with a compound datatype. Note that if you are using OpenMC 0.7.0 (as opposed to the current developmental branch), you should remove the ‘delayed_group’ field (just get rid of the last tuple in the source_dtype list). Hopefully the example is self-explanatory, but let me know if you have any questions.
Best,
Paul
make_source.py (927 Bytes)
You can find the source for the documentation on output file formats here:
https://github.com/mit-crpg/openmc/tree/develop/docs/source/usersguide/output
Best,
Paul