openmc.deplete.abc.TalliedFissionYieldHelper¶
-
class
openmc.deplete.abc.
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.Operator
that uses this helper may only burn a subset of all materials when running in parallel mode.
- materials (iterable of
-
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
- constant_yields (dict of str to