openmc.deplete.helpers.TalliedFissionYieldHelper¶
- class openmc.deplete.helpers.TalliedFissionYieldHelper(chain_nuclides)[source]¶
Abstract class for computing fission yields with tallies
Generates a basic fission rate tally in all burnable materials with
generate_tallies()
, and set nuclides to be tallied withupdate_tally_nuclides()
. Subclasses will need to implementunpack()
andweighted_yields()
.- Parameters
chain_nuclides (iterable of openmc.deplete.Nuclide) – Nuclides tracked in the depletion chain. Not necessary that all have yield data.
- Variables
constant_yields (dict of str to
openmc.deplete.FissionYield
) – Fission yields for all nuclides that only have one set of fission yield data. Can be accessed as{parent: {product: yield}}
results (None or numpy.ndarray) – Tally results shaped in a manner useful to this helper.
- generate_tallies(materials, mat_indexes)[source]¶
Construct the fission rate tally
- Parameters
materials (iterable of
openmc.lib.Material
) – Materials to be used inopenmc.lib.MaterialFilter
mat_indexes (iterable of int) – Indices of tallied materials that will have their fission yields computed by this helper. Necessary as the
openmc.deplete.CoupledOperator
that uses this helper may only burn a subset of all materials when running in parallel mode.
- abstract unpack()[source]¶
Unpack tallies after a transport run.
Abstract because each subclass will need to arrange its tally data.
- update_tally_nuclides(nuclides)[source]¶
Tally nuclides with non-zero density and multiple yields
Must be run after
generate_tallies()
.- Parameters
nuclides (iterable of str) – Potential nuclides to be tallied, such as those with non-zero density at this stage.
- Returns
nuclides – Union of input nuclides and those that have multiple sets of yield data. Sorted by nuclide name
- Return type
list of str
- Raises
AttributeError – If tallies not generated