openmc.deplete.abc.FissionYieldHelper¶
- 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__
method- Parameters
operator (openmc.deplete.abc.TransportOperator) – Operator with a depletion chain
kwargs (optional) – Additional keyword arguments to be used in constuction
- 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.CoupledOperator
that uses this helper may only burn a subset of all materials when running in parallel mode.
- static unpack()[source]¶
Unpack tally data prior to compute fission yields.
Called after a
openmc.deplete.abc.TransportOperator.__call__()
routine during the normalization of reaction rates.Not necessary for all subclasses to implement, unless tallies are used.
- update_tally_nuclides(nuclides: Sequence[str]) list [source]¶
Return nuclides with non-zero densities and yield data
- Parameters
nuclides (iterable of str) – Nuclides with non-zero densities from the
openmc.deplete.abc.TransportOperator
- Returns
nuclides – Union of nuclides that the
openmc.deplete.abc.TransportOperator
says have non-zero densities at this stage and those that have yield data. Sorted by nuclide name- Return type
list of str
- abstract 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