Some question about fixed source depletion

Hi, everyone

I used MCNP for my research before but it still lacks some functions for me. After reading some others’ questions on forum, It seems that openMC would probably give a big help to me for the fixed source calculation.
I have two questions:

  1. I’m doing the research on ADS, and I saw the (User’s Guide » 10.1. Fixed-Source Transmutation), it means I can do fixed source depltion with high energy(about 0.6 to 1 GeV)? and the source will be generated from openMC or I can use the source file from MCNP output?
  2. I found a lot of simulation program, and there are no program could change neutron spectrum. e.g., when I do the depletion on MCNP, the neutron source should be fission source and the energy spectrum is from prompt neutron. but if I wanna change the prompt neutron energy spectrum to my specified neutron spectrum instead of fission spectrum. Is that possible for openMC?

Thanks for watching my poor English, I will try my best to express what I am thinking.
also thanks for your time and effort.

Hi @Rong and welcome to the community! To answer your questions:

OpenMC will transport particles up to whatever energy is available in the nuclear data library. Generally, the libraries that we use (ENDF/B, JEFF) only go up to about 20 MeV, so it would not be possible to simulate particles at ~GeV energies. However, if you’re able to generate a neutron source that is below 20 MeV, you could feed that into OpenMC.

I would recommend reading the user’s guide section on fixed source transmutation, which explains how to handle this.

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Hi paulromano,

Thanks for your response.
OpenMC could transport particles up to 20 MeV, it means that if I want to generate an external source(openmc.SourceParticle), the highest energy of incident particles would be set as 20 MeV(E=20000000)? or I can just feed a source file generated from other simulation program with energy below 20 MeV into OpenMC?

For the second question, it’s my fault for the bad expression. I want to know is it possible to use a specific neutron spectrum in depletion calculation? I have a idea for this question: first, set openmc.deplete.Operator( reaction_rate_mode=‘flux’), then call openmc.deplete.helpers.FluxCollapseHelper(energies=,,_). Will it work?

Hope these questions would not too stupid to waste your time.
Thanks a lot for watching my problems.

Yes, the energy of any neutrons would have to be 20 MeV or below no matter how the source is produced.

Right now, our Operator class with the “flux” reaction_rate_mode will set up tallies for a multigroup flux and then use that flux to collapse with the cross sections in order to obtain 1 group reaction rates. However, if you already have a multigroup flux spectrum, at present we can’t take that directly and use it to deplete. That may be something that’s added in the future (e.g., there is an open pull request that moves in that direction).