Hello everyone!
I have a question regarding tally normalization in OpenMC.
I understand that the neutron flux tally should be normalized using the reactor power (via a global heating or heating-local tally) to obtain units of n/cm²/s.
My question is: **Should the same normalization factor also be applied to reaction rate tallies such as absorption, fission, total, and scatter?
If so, should these tallies simply be multiplied by the normalization factor (to obtain reactions/s), while only the flux tally is additionally divided by the tally cell volume?
Yes, this is one way to obtain a normalisation factor. Yes, you should use the same normalization factor for all tallies. Your normalisation factor in this case has units <source particles> / <second> and your other tallies have units <reactions> / <source>. Always check your units.
As for the division by cell volume, I think that depends on whether you want to report the total reaction rate per cell or the reaction rate density in that cell. In the latter case you’d need to divide the tally results by the cell volumes.