Hi Enej, welcome to the community.
Regarding the tally normalization, from what I understand in your case, you use a source strength that came from fuel assembly power/energy per fission tally (either from kappa fission or fission q recoverable) source/sec. That’s similar to what I usually did, but then you use fuel volume to get the flux value (n/cm2-sec) when flux-tallying the instrumentation tube on the center of your fuel assembly.
I think for getting flux (n/cm2-sec), you need to use the volume of cell that you have tallied (the instrumentation tube) so the flux tally which has a unit of (#-cm/source) can be multiplied by source (source/sec) and divided by tallied cell volume (cm3) and you get the average neutron flux on your tallied cell n/cm2-sec.
Then regarding the other score, i.e. reaction rate: fission, absorption, capture, etc. the tally unit itself has a unit as it already described on the URL you sent before, i.e. fission score has a unit of fission/source. So I think you only need to multiply these tally values with the source and you get your reaction rate in reaction/sec.
So all tally has been integrated to its volume, same as their equation Sig.rx(macroscopic cross sections) × flux × corresponding cell volume, besides that it has been normalized to the source strength which came from the fuel assembly power in your case.
You also can check it from your power tally i.e. if you have tallied kappa fission on all your fuel pins individually, or as you want it to be, then the total fuel assembly power will be the same as the sum of individual pin power, or #fuel pin × 1 fuel pin power if you think you have almost similar pin power fission rate and kappa.
So, if you want to know the reaction rate on the instrumentation tube cell, you can try to tally and then multiply it by source, and you get your rx. rate in rx./sec.
I hope it could help you
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