- class openmc.deplete.abc.TalliedFissionYieldHelper(chain_nuclides)¶
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
chain_nuclides (iterable of openmc.deplete.Nuclide) – Nuclides tracked in the depletion chain. Not necessary that all have yield data.
- generate_tallies(materials, mat_indexes)¶
Construct the fission rate tally
mat_indexes (iterable of int) – Indices of tallied materials that will have their fission yields computed by this helper. Necessary as the
openmc.deplete.Operatorthat uses this helper may only burn a subset of all materials when running in parallel mode.
- abstract unpack()¶
Unpack tallies after a transport run.
Abstract because each subclass will need to arrange its tally data.
Tally nuclides with non-zero density and multiple yields
Must be run after
nuclides (iterable of str) – Potential nuclides to be tallied, such as those with non-zero density at this stage.
nuclides – Union of input nuclides and those that have multiple sets of yield data. Sorted by nuclide name
- Return type
list of str
AttributeError – If tallies not generated