openmc.deplete.abc.FissionYieldHelper¶
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class
openmc.deplete.abc.
FissionYieldHelper
(chain_nuclides)[source]¶ Abstract class for processing energy dependent fission yields
Parameters: chain_nuclides (iterable of openmc.deplete.Nuclide) – Nuclides tracked in the depletion chain. All nuclides are not required to have fission yield data. Variables: constant_yields (collections.defaultdict) – Fission yields for all nuclides that only have one set of fission yield data. Dictionary of form {str: {str: float}}
representing yields for{parent: {product: yield}}
. Default return object is an empty dictionary-
classmethod
from_operator
(operator, **kwargs)[source]¶ Create a new instance by pulling data from the operator
All keyword arguments should be identical to their counterpart in the main
__init__
methodParameters: - operator (openmc.deplete.TransportOperator) – Operator with a depletion chain
- kwargs (optional) – Additional keyword arguments to be used in constuction
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static
generate_tallies
(materials, mat_indexes)[source]¶ Construct tallies necessary for computing fission yields
Called during the operator set up phase prior to depleting. Not necessary for subclasses to implement
Parameters: - materials (iterable of C-API materials) – Materials to be used in
openmc.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 C-API materials) – Materials to be used in
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static
unpack
()[source]¶ Unpack tally data prior to compute fission yields.
Called after a
openmc.deplete.Operator.__call__()
routine during the normalization of reaction rates.Not necessary for all subclasses to implement, unless tallies are used.
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update_tally_nuclides
(nuclides)[source]¶ Return nuclides with non-zero densities and yield data
Parameters: nuclides (iterable of str) – Nuclides with non-zero densities from the openmc.deplete.Operator
Returns: nuclides – Union of nuclides that the openmc.deplete.Operator
says have non-zero densities at this stage and those that have yield data. Sorted by nuclide nameReturn type: list of str
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weighted_yields
(local_mat_index)[source]¶ Return fission yields for a specific material
Parameters: local_mat_index (int) – Index for the material with requested fission yields. Should correspond to the material represented in mat_indexes[local_mat_index]
duringgenerate_tallies()
.Returns: library – Dictionary-like object mapping {str: {str: float}
. This reflects fission yields for{parent: {product: fyield}}
.Return type: collections.abc.Mapping
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classmethod