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Add linear and log interpolation for energy dependent fission yields

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ymchan 2026-03-20 11:17:46 +01:00
parent 6581af8487
commit 0b94726944
1 changed files with 28 additions and 24 deletions
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52
openmc/deplete/helpers.py
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@ -1025,6 +1025,33 @@ class AveragedFissionYieldHelper(TalliedFissionYieldHelper):
class LogLinInterpolateFissionYieldHelper(TalliedFissionYieldHelper):
r"""Helper that computes fission yields from log-lin weighted groups.
Parameters
----------
chain_nuclides : iterable of openmc.deplete.Nuclide
Nuclides tracked in the depletion chain. All nuclides are
not required to have fission yield data.
energy_bins : iterable of float, optional
Energy points [eV] that define ``len(energy_bins)`` log-lin
groups. Must be strictly increasing and positive.
Defaults to ``(0.025, 500e3, 14e6)``.
Attributes
----------
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_groups, n_tnucs)``, where ``n_mats`` is the
number of materials where fission reactions were tallied,
``n_groups = len(energy_bins)``, and ``n_tnucs`` is the number
of nuclides with multiple sets of fission yields.
Data are normalized group fractions corresponding to ``y1..yN``.
Notes
-----
Fission events are partitioned into ``len(energy_bins)`` group
fractions using the provided energy points. Adjacent groups share each
interval and are blended using log-linear interpolation in incident energy.
@ -1104,30 +1131,7 @@ class LogLinInterpolateFissionYieldHelper(TalliedFissionYieldHelper):
``\vec{f} \approx 0.3697\,\vec{f}_1 + 0.6303\,\vec{f}_2`` for this
interval.
Parameters
----------
chain_nuclides : iterable of openmc.deplete.Nuclide
Nuclides tracked in the depletion chain. All nuclides are
not required to have fission yield data.
energy_bins : iterable of float, optional
Energy points [eV] that define ``len(energy_bins)`` log-lin
groups. Must be strictly increasing and positive.
Defaults to ``(0.025, 500e3, 14e6)``.
Attributes
----------
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_groups, n_tnucs)``, where ``n_mats`` is the
number of materials where fission reactions were tallied,
``n_groups = len(energy_bins)``, and ``n_tnucs`` is the number
of nuclides with multiple sets of fission yields.
Data are normalized group fractions corresponding to ``y1..yN``.
"""
def __init__(self, chain_nuclides, energy_bins=(0.025, 500.0e3, 14.0e6)):