To find neutron flux, atomic density, and burn-up level through depletion calculation

Dear all,

I wanna ask to you all about how to find neutron flux, atomic density and burn-up level through depletion calculation?
In example for pincell depletion only exist k effective and fission reaction.


Hi Aulia,

By writing flux score in tallies.xml file, you can find out neutron flux in each burnup step. Depletion module will write tally data to statepoint files for each burnup steps in openmc_simulation_n.h5, current time step.

By opening each statepoint file, you will get neutron flux for different burnup step.

If you use openmc develop branch, you can calculate atomic density in " atom/b-cm, atom/cm3" directly from depletion statepoint file. Andrew johnson addded this Use atoms/b-cm, atoms/cm3 in ResultsList.get_atoms in develop branch. If you use openmc 0.11, th you have to convert unit by yourself [ Divide by volume( mat ) to get density ].

  • Pranto