Keep in mind that a reactor’s fission rate is limited by heat transfer, not neutronics. An operating power reactor can freely increase its fission rate until it hits heat transfer constraints. This means that the OpenMC values for fission rate are not useful for comparing different reactor designs.

However, you also mentioned criticality which is a useful performance metric that OpenMC computes. Another interesting quantity to study with OpenMC is the ratio of absorption in the fuel to absorption elsewhere in the system. If very few neutrons are absorbed outside of the fuel, the reactor design is neutronically efficient and is better for purposes like breeding fuel. You can also use OpenMC to compute values for each factor of Fermi’s four-factor formula. These factors are useful for comparing different reactor designs.

That said, if you do still want to normalize those tally values, here is one way to do it:

That pincell model comes from the BEAVRS benchmark which is a model of a Westinghouse PWR. That PWR produces 3400 MW of thermal power. It has 193 fuel assemblies. Each assembly has 264 fuel rods. That means the average fuel pin produces 3400 [MW] / (193 * 264) = 67 [kW] of power. Fission of U-235 produces about 200 MeV of energy so there are about 67 [kW] / 1.6e-16 [kJ / MeV] / 200 [MeV] = 2.1e15 [fissions / s] in the average fuel pin. Notice that we have computed the total fission rate without the use of a neutronics solver. It also assumes we are talking about an “average fuel pin”, but the pins in a real reactor can of course have higher or lower fission rates.

If we want our tally results to match that 2.1e15 [fissions / s] then we must multiply all the tally results by a normalization factor of 2.1e15 / 0.547 = 3.8e15. We are essentially saying that there must be 3.8e15 source particles per second in this system.

Note that you also have to divide by volume for a flux value in units of [1 / cm^2 / s] . The radius of the fuel is 0.39 [cm] and the height is 366 [cm] for a total volume of 175 [cm^3] in cell 1. If you add a flux tally to that pincell model with the same cell filter, you’ll get a value of 11.02. We can normalize that with 11.02 * 3.8e15 / 175 = 2.4e14 [1 / cm^2 / s].