# 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 with update_tally_nuclides(). Subclasses will need to implement unpack() and weighted_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
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