Is there a way to tally the neutron flux from a specific reaction? For example, the neutron flux from (n,2n) reactions only. I have been checking openmc.Tally and openmc.Filter documentation, but I have not figured out if this is possible. Thanks
There is no direct way to get the flux specifically produced from a reaction like that. The closest way I could think of would be to tally the (n,2n) reaction rate in your problem and then use that to define a starting source for a subsequent calculation.
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That could work well. Thank you very much @paulromano