openmc.deplete.abc.FissionYieldHelper

class openmc.deplete.abc.FissionYieldHelper(chain_nuclides)

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)

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)

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()

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)

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)

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] during generate_tallies().

Returns

library – Dictionary-like object mapping {str: {str: float}. This reflects fission yields for {parent: {product: fyield}}.

Return type

collections.abc.Mapping