How to get neutron flux for depletion per depletion time

[Pranto]

Please take a look at the following link where @andrewjohnson gave answer which is similar to your question.

Flux variation with respect to time

Hi Farshid, After each transport simulation, the depletion module will write tally data to statepoint files, just like you would obtain from a single transport solution. These are named “openmc_simulation_n.h5”, where is the current time step. You can open each file using the StatePoint class. With one StatePoint instance per file (and each file at a unique point in time) you can view and plot tally data over time manually. I hope this is helpful, Andrew