Flux distribution and unit of reaction rate for the research reactor fuel assembly

[wahidluthfi]

I think you can read other discussions regarding tally normalization, Paul has been describing tally normalization properly in this discussion

NOrmalizing Tally to get Flux value

The tally is not normalized by volume, so the units are something like [neutrons-cm/source]. This can be understood quite easily: in a collision estimator tally, OpenMC accumulates 1/Sigma_t which has units of cm, and similarly for a tracklength tally, OpenMC accumulates the tracklength in cm. When you divide by volume, you get [neutrons/cm^2-source]. The normalization factor Pnu/(Qk) gives you [J/secneutrons/fission/(J/fissionneutrons/source)] = [source/sec] so that when you multiply the flux b…