How to get the fast neutron fluence in depletion calculation?

Hi all,
I wonder if there is a way to get the fast neutron fluence in the depletion calculation? I learned from forums how to normalize neutron fluence. Although I have some guesses on how to calculate the fast neutron flux, I am not sure of its correctness. So I hope someone can teach me how to get the fluence of fast neutrons knowing the power. I would be very grateful for any advice


You can create a tally for the fast neutron flux (this example doesn’t have a spatial filter but you may want to include one):

tally = openmc.Tally()
tally.filters = [
    openmc.EnergyFilter([E_low, E_high])
tally.scores = ['flux']

Then, when you run a depletion calculation, each statepoint file (openmc_simulation_n#.h5) will have the result of that tally. You can multiply the tally result (flux) by the depletion timestep length to get a fluence. Also, it is up to you to define what “fast” is, that is, you would need to select E_low and E_high.