openmc.deplete.helpers.FissionYieldCutoffHelper¶
- class openmc.deplete.helpers.FissionYieldCutoffHelper(chain_nuclides, n_bmats, cutoff=112.0, thermal_energy=0.0253, fast_energy=500000.0)[source]¶
Helper that computes fission yields based on a cutoff energy
Tally fission rates above and below the cutoff energy. Assume that all fissions below cutoff energy have use thermal fission product yield distributions, while all fissions above use a faster set of yield distributions.
Uses a limit of 20 MeV for tallying fission.
- Parameters
chain_nuclides (iterable of openmc.deplete.Nuclide) – Nuclides tracked in the depletion chain. All nuclides are not required to have fission yield data.
n_bmats (int, optional) – Number of burnable materials tracked in the problem
cutoff (float, optional) – Cutoff energy in [eV] below which all fissions will be use thermal yields. All other fissions will use a faster set of yields. Default: 112 [eV]
thermal_energy (float, optional) – Energy of yield data corresponding to thermal yields. Default: 0.0253 [eV]
fast_energy (float, optional) – Energy of yield data corresponding to fast yields.
- Variables
n_bmats (int) – Number of burnable materials tracked in the problem. Must be set prior to generating tallies
thermal_yields (dict) – Dictionary of the form
{parent: {product: yield}}
with thermal yieldsfast_yields (dict) – Dictionary of the form
{parent: {product: yield}}
with fast yieldsconstant_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 dictionaryresults (numpy.ndarray) – Array of fission rate fractions with shape
(n_mats, 2, n_nucs)
.results[:, 0]
corresponds to the fraction of all fissions that occurred belowcutoff
. The number of materials in the first axis corresponds to the number of materials burned by theopenmc.deplete.CoupledOperator
- classmethod from_operator(operator, **kwargs)[source]¶
Construct a helper from an operator
All keyword arguments should be identical to their counterpart in the main
__init__
method- Parameters
operator (openmc.deplete.CoupledOperator) – Operator with a chain and burnable materials
kwargs – Additional keyword arguments to be used in construction
- Return type
- generate_tallies(materials, mat_indexes)[source]¶
Use C API to produce a fission rate tally in burnable materials
Include a
openmc.lib.EnergyFilter
to tally fission rates above and below cutoff energy.- 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.CoupledOperator
that uses this helper may only burn a subset of all materials when running in parallel mode.
- weighted_yields(local_mat_index)[source]¶
Return fission yields for a specific material
For nuclides with both yield data above and below the cutoff energy, the effective yield for nuclide
A
will be a weighted sum of fast and thermal yields. The weights will be the fraction ofA
fission events in the above and below the cutoff energy.If
A
has fission product distributionF
for fast fissions andT
for thermal fissions, and 70% ofA
fissions are considered thermal, then the effective fission product yield distributions forA
is0.7 * T + 0.3 * F
- 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}}
- Return type