openmc.deplete.helpers.AveragedFissionYieldHelper¶

class
openmc.deplete.helpers.
AveragedFissionYieldHelper
(chain_nuclides)[source]¶ Class that computes fission yields based on average fission energy
Computes average energy at which fission events occured with
\[\bar{E} = \frac{ \int_0^\infty E\sigma_f(E)\phi(E)dE }{ \int_0^\infty\sigma_f(E)\phi(E)dE }\]If the average energy for a nuclide is below the lowest energy with yield data, that set of fission yields is taken. Conversely, if the average energy is above the highest energy with yield data, that set of fission yields is used. For the case where the average energy is between two sets of yields, the effective fission yield computed by linearly interpolating between yields provided at the nearest energies above and below the average.
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  results (None or numpy.ndarray) – If tallies have been generated and unpacked, then the array will
have shape
(n_mats, n_tnucs)
, wheren_mats
is the number of materials where fission reactions were tallied andn_tnucs
is the number of nuclides with multiple sets of fission yields. Data in the array are the average energy of fission events for tallied nuclides across burnable materials.

classmethod
from_operator
(operator, **kwargs)[source]¶ Return a new helper with data from an 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 – Additional keyword arguments to be used in construction
Returns: Return type:

generate_tallies
(materials, mat_indexes)[source]¶ Construct tallies to determine average energy of fissions
Parameters:  materials (iterable of
openmc.lib.Material
) – Materials to be used inopenmc.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

update_tally_nuclides
(nuclides)[source]¶ Tally nuclides with nonzero density and multiple yields
Must be run after
generate_tallies()
.Parameters: nuclides (iterable of str) – Potential nuclides to be tallied, such as those with nonzero 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: tuple of str Raises: AttributeError
– If tallies not generated

weighted_yields
(local_mat_index)[source]¶ Return fission yields for a specific material
Use the computed average energy of fission events to determine fission yields. If average energy is between two sets of yields, linearly interpolate bewteen the two. Otherwise take the closet set of yields.
Parameters: local_mat_index (int) – Index for specific burnable material. Effective yields will be produced using self.results[local_mat_index]
Returns: library – Dictionary of {parent: {product: fyield}}
. Default return value is an empty dictionaryReturn type: collections.defaultdict
 constant_yields (collections.defaultdict) – Fission yields for all nuclides that only have one set of
fission yield data. Dictionary of form