How to get the fast neutron fluence in depletion calculation?

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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']
...
model.tallies.append(tally)

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.

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