About constant power normalization during eigenvalue calculation and depletion and photo-fission contribution to the eigenvalue

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After a week of poking around i have successfully gotten agreeable results between OpenMC and Serpent:

  1. Easiest way was to change fission energy release in Serpent: set U235H 193.4054000, for which i read the value from OpenMC depletion chain file. NOTE: Actual power output is higher as we are disregarding any delayed power from fission product decays and any power produced by absorption reactions other than fission.
    For detectors i used Serpent predefined energy grid: ene egridd 4 scale238, for which the bins are defined in the Serpent manual. I converted them into an array multiplied by 1e6 (MeV to eV). In 2-D simulations Serpent gives some results normalised per unit length, so to get correct results you may need to multiply by 100. For OpenMC normalization use the tally “kappa-fission” and follow the guide. Watch out for volumes if any cell or material filters are applied.

  2. For the solver I used the OpenMC default cram48 method, which is called in Serpent by: set bumode 2 -48 .
    The problem was the integrator, only CE, CE/LI and LE/QI are supported by both programs. CE is quite unprecise and for the other two I am not sure in the same implementation. Nonetheless LE/QI integrator in OpenMC and for Serpent: set pcc 5 1 2 2 2 yielded results agreeing within 60 pcm at worst. NOTE: calculation time is quite a bit longer in OpenMC than in Serpent (maybe parallelization).

  3. Make sure that the libraries used were compiled from the same ENDF file and if run on a cluster that slurm files have the correct PATH for the library.
    Neutronics simulation is the default value in settings, photons are desregarded.

Note to any developers:
Can we get a local energy deposition addition for Q_capture like in Serpent: set edepmode 1 [some extra heating]?

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