openmc.deplete.get_microxs_and_flux(model: Model, domains, nuclides: Optional[Iterable[str]] = None, reactions: Optional[Iterable[str]] = None, energies: Optional[Union[Iterable[float], str]] = None, chain_file: Optional[Union[str, PathLike]] = None, run_kwargs=None) Tuple[List[ndarray], List[MicroXS]][source]

Generate a microscopic cross sections and flux from a Model

New in version 0.14.0.

  • model (openmc.Model) – OpenMC model object. Must contain geometry, materials, and settings.

  • domains (list of openmc.Material or openmc.Cell or openmc.Universe, or openmc.MeshBase) – Domains in which to tally reaction rates.

  • nuclides (list of str) – Nuclides to get cross sections for. If not specified, all burnable nuclides from the depletion chain file are used.

  • reactions (list of str) – Reactions to get cross sections for. If not specified, all neutron reactions listed in the depletion chain file are used.

  • energies (iterable of float or str) – Energy group boundaries in [eV] or the name of the group structure

  • chain_file (str, optional) – Path to the depletion chain XML file that will be used in depletion simulation. Used to determine cross sections for materials not present in the inital composition. Defaults to openmc.config['chain_file'].

  • run_kwargs (dict, optional) – Keyword arguments passed to


  • list of numpy.ndarray – Flux in each group in [n-cm/src] for each domain

  • list of MicroXS – Cross section data in [b] for each domain