Deplet without consider fuel

Hi, I want to use openmc to solve the transmutation problem of long-lived fission products. I set the fuel to be non-degradable to maintain the neutron flux during the irradiation cycle. Set the time step to 2 years, total cycle time to 20 years, number of particles to 10000, number of batches to 11000, and skip 100. but the depletion results seem to be wrong because the nuclide atoms in the material are incredible.
I would like to know how I can get long irradiation results at constant neutron flux if I cannot set the fuel to depletable?
Many thanks!

Hi @xiangyang and welcome to the forum. I believe the problem here is that if you set all your fuel to be non-depletable, when it tries to normalizes fluxes based on the power it relies on the fission reaction rates from depletable materials (which are in this case now zero). What I would recommend is to:

  • Use the “energy-depostion” normalization mode: Operator(..., normalization_mode="energy-deposition") which will perform the normalization based on the overall heating (this doesn’t rely on which materials are tagged as depletable).
  • Since this mode depends on the heating-local tally score, make sure you use one of our official data libraries which have the appropriate cross section for accumulating this score.
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Thank you very much for your advice and help, I have followed your suggestion to set up the program, and the result is partly as expected but the other part has a large error, I am trying to improve the simulation accuracy by increasing the number of particles.