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Merge pull request #2083 from yardasol/move-tallied-fission-yield-helpers-to-helpers-py
Move TalliedFissionYieldHelper from abc.py to helpers.py
This commit is contained in:
Paul Romano
GitHub
commit
edf913b80f
2 changed files with 98 additions and 98 deletions
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@ -19,7 +19,6 @@ from warnings import warn
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from numpy import nonzero, empty, asarray
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from uncertainties import ufloat
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from openmc.lib import MaterialFilter, Tally
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from openmc.checkvalue import check_type, check_greater_than
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from openmc.mpi import comm
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from .stepresult import StepResult
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@ -30,7 +29,7 @@ from .pool import deplete
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__all__ = [
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"OperatorResult", "TransportOperator", "ReactionRateHelper",
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"NormalizationHelper", "FissionYieldHelper", "TalliedFissionYieldHelper",
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"NormalizationHelper", "FissionYieldHelper",
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"Integrator", "SIIntegrator", "DepSystemSolver", "add_params"]
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@ -480,99 +479,6 @@ class FissionYieldHelper(ABC):
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return cls(operator.chain.nuclides, **kwargs)
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class TalliedFissionYieldHelper(FissionYieldHelper):
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"""Abstract class for computing fission yields with tallies
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Generates a basic fission rate tally in all burnable materials with
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:meth:`generate_tallies`, and set nuclides to be tallied with
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:meth:`update_tally_nuclides`. Subclasses will need to implement
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:meth:`unpack` and :meth:`weighted_yields`.
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Parameters
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----------
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chain_nuclides : iterable of openmc.deplete.Nuclide
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Nuclides tracked in the depletion chain. Not necessary
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that all have yield data.
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Attributes
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----------
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constant_yields : dict of str to :class:`openmc.deplete.FissionYield`
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Fission yields for all nuclides that only have one set of
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fission yield data. Can be accessed as ``{parent: {product: yield}}``
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results : None or numpy.ndarray
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Tally results shaped in a manner useful to this helper.
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"""
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_upper_energy = 20.0e6 # upper energy for tallies
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def __init__(self, chain_nuclides):
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super().__init__(chain_nuclides)
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self._local_indexes = None
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self._fission_rate_tally = None
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self._tally_nucs = []
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self.results = None
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def generate_tallies(self, materials, mat_indexes):
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"""Construct the fission rate tally
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Parameters
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----------
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materials : iterable of :class:`openmc.lib.Material`
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Materials to be used in :class:`openmc.lib.MaterialFilter`
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mat_indexes : iterable of int
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Indices of tallied materials that will have their fission
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yields computed by this helper. Necessary as the
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:class:`openmc.deplete.Operator` that uses this helper
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may only burn a subset of all materials when running
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in parallel mode.
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"""
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self._local_indexes = asarray(mat_indexes)
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# Tally group-wise fission reaction rates
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self._fission_rate_tally = Tally()
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self._fission_rate_tally.writable = False
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self._fission_rate_tally.scores = ['fission']
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self._fission_rate_tally.filters = [MaterialFilter(materials)]
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def update_tally_nuclides(self, nuclides):
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"""Tally nuclides with non-zero density and multiple yields
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Must be run after :meth:`generate_tallies`.
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Parameters
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----------
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nuclides : iterable of str
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Potential nuclides to be tallied, such as those with
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non-zero density at this stage.
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Returns
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-------
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nuclides : list of str
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Union of input nuclides and those that have multiple sets
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of yield data. Sorted by nuclide name
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Raises
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------
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AttributeError
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If tallies not generated
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"""
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assert self._fission_rate_tally is not None, (
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"Run generate_tallies first")
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overlap = set(self._chain_nuclides).intersection(set(nuclides))
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nuclides = sorted(overlap)
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self._tally_nucs = [self._chain_nuclides[n] for n in nuclides]
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self._fission_rate_tally.nuclides = nuclides
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return nuclides
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@abstractmethod
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def unpack(self):
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"""Unpack tallies after a transport run.
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Abstract because each subclass will need to arrange its
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tally data.
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"""
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def add_params(cls):
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cls.__doc__ += cls._params
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return cls
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@ -2,6 +2,7 @@
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Class for normalizing fission energy deposition
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"""
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import bisect
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from abc import abstractmethod
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from collections import defaultdict
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from copy import deepcopy
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from itertools import product
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@ -17,14 +18,107 @@ from openmc.lib import (
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Tally, MaterialFilter, EnergyFilter, EnergyFunctionFilter)
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import openmc.lib
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from .abc import (
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ReactionRateHelper, NormalizationHelper, FissionYieldHelper,
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TalliedFissionYieldHelper)
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ReactionRateHelper, NormalizationHelper, FissionYieldHelper)
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__all__ = (
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"DirectReactionRateHelper", "ChainFissionHelper", "EnergyScoreHelper"
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"SourceRateHelper", "ConstantFissionYieldHelper", "FissionYieldCutoffHelper",
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"SourceRateHelper", "TalliedFissionYieldHelper",
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"ConstantFissionYieldHelper", "FissionYieldCutoffHelper",
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"AveragedFissionYieldHelper", "FluxCollapseHelper")
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class TalliedFissionYieldHelper(FissionYieldHelper):
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"""Abstract class for computing fission yields with tallies
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Generates a basic fission rate tally in all burnable materials with
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:meth:`generate_tallies`, and set nuclides to be tallied with
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:meth:`update_tally_nuclides`. Subclasses will need to implement
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:meth:`unpack` and :meth:`weighted_yields`.
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Parameters
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----------
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chain_nuclides : iterable of openmc.deplete.Nuclide
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Nuclides tracked in the depletion chain. Not necessary
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that all have yield data.
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Attributes
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----------
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constant_yields : dict of str to :class:`openmc.deplete.FissionYield`
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Fission yields for all nuclides that only have one set of
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fission yield data. Can be accessed as ``{parent: {product: yield}}``
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results : None or numpy.ndarray
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Tally results shaped in a manner useful to this helper.
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"""
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_upper_energy = 20.0e6 # upper energy for tallies
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def __init__(self, chain_nuclides):
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super().__init__(chain_nuclides)
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self._local_indexes = None
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self._fission_rate_tally = None
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self._tally_nucs = []
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self.results = None
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def generate_tallies(self, materials, mat_indexes):
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"""Construct the fission rate tally
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Parameters
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----------
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materials : iterable of :class:`openmc.lib.Material`
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Materials to be used in :class:`openmc.lib.MaterialFilter`
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mat_indexes : iterable of int
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Indices of tallied materials that will have their fission
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yields computed by this helper. Necessary as the
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:class:`openmc.deplete.Operator` that uses this helper
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may only burn a subset of all materials when running
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in parallel mode.
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"""
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self._local_indexes = asarray(mat_indexes)
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# Tally group-wise fission reaction rates
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self._fission_rate_tally = Tally()
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self._fission_rate_tally.writable = False
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self._fission_rate_tally.scores = ['fission']
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self._fission_rate_tally.filters = [MaterialFilter(materials)]
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def update_tally_nuclides(self, nuclides):
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"""Tally nuclides with non-zero density and multiple yields
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Must be run after :meth:`generate_tallies`.
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Parameters
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----------
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nuclides : iterable of str
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Potential nuclides to be tallied, such as those with
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non-zero density at this stage.
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Returns
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-------
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nuclides : list of str
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Union of input nuclides and those that have multiple sets
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of yield data. Sorted by nuclide name
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Raises
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------
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AttributeError
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If tallies not generated
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"""
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assert self._fission_rate_tally is not None, (
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"Run generate_tallies first")
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overlap = set(self._chain_nuclides).intersection(set(nuclides))
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nuclides = sorted(overlap)
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self._tally_nucs = [self._chain_nuclides[n] for n in nuclides]
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self._fission_rate_tally.nuclides = nuclides
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return nuclides
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@abstractmethod
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def unpack(self):
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"""Unpack tallies after a transport run.
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Abstract because each subclass will need to arrange its
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tally data.
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"""
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# -------------------------------------
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# Helpers for generating reaction rates
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# -------------------------------------
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