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Implement FluxCollapseHelper that is used by Operator
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Paul Romano
parent
b7df2d95e1
commit
76fc6fb3e7
3 changed files with 150 additions and 15 deletions
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@ -225,11 +225,11 @@ class TransportOperator(ABC):
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class ReactionRateHelper(ABC):
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"""Abstract class for generating reaction rates for operators
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Responsible for generating reaction rate tallies for burnable
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materials, given nuclides and scores from the operator.
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Responsible for generating reaction rate tallies for burnable materials,
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given nuclides and scores from the operator.
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Reaction rates are passed back to the operator for be used in
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an :class:`openmc.deplete.OperatorResult` instance
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Reaction rates are passed back to the operator for be used in an
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:class:`openmc.deplete.OperatorResult` instance.
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Parameters
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----------
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@ -246,7 +246,6 @@ class ReactionRateHelper(ABC):
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def __init__(self, n_nucs, n_react):
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self._nuclides = None
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self._rate_tally = None
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self._results_cache = empty((n_nucs, n_react))
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@abstractmethod
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@ -262,7 +261,6 @@ class ReactionRateHelper(ABC):
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def nuclides(self, nuclides):
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check_type("nuclides", nuclides, list, str)
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self._nuclides = nuclides
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self._rate_tally.nuclides = nuclides
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@abstractmethod
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def get_material_rates(self, mat_id, nuc_index, react_index):
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@ -281,8 +279,7 @@ class ReactionRateHelper(ABC):
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def divide_by_adens(self, number):
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"""Normalize reaction rates by number of nuclides
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Acts on the current material examined by
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:meth:`get_material_rates`
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Acts on the current material examined by :meth:`get_material_rates`
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Parameters
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----------
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@ -559,7 +556,6 @@ class TalliedFissionYieldHelper(FissionYieldHelper):
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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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@ -7,13 +7,14 @@ from numbers import Real
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import bisect
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from collections import defaultdict
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from numpy import dot, zeros, newaxis
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from numpy import dot, zeros, newaxis, asarray
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from . import comm
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from openmc.checkvalue import check_type, check_greater_than
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from openmc.data import JOULE_PER_EV
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from openmc.data import JOULE_PER_EV, REACTION_NAME
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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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@ -21,7 +22,7 @@ from .abc import (
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__all__ = (
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"DirectReactionRateHelper", "ChainFissionHelper", "EnergyScoreHelper"
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"SourceRateHelper", "ConstantFissionYieldHelper", "FissionYieldCutoffHelper",
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"AveragedFissionYieldHelper")
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"AveragedFissionYieldHelper", "FluxCollapseHelper")
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# -------------------------------------
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# Helpers for generating reaction rates
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@ -43,6 +44,14 @@ class DirectReactionRateHelper(ReactionRateHelper):
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nuclides : list of str
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All nuclides with desired reaction rates.
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"""
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def __init__(self, n_nuc, n_react):
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super().__init__(n_nuc, n_react)
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self._rate_tally = None
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@ReactionRateHelper.nuclides.setter
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def nuclides(self, nuclides):
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ReactionRateHelper.nuclides.fset(self, nuclides)
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self._rate_tally.nuclides = nuclides
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def generate_tallies(self, materials, scores):
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"""Produce one-group reaction rate tally
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@ -92,6 +101,107 @@ class DirectReactionRateHelper(ReactionRateHelper):
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return self._results_cache
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class FluxCollapseHelper(ReactionRateHelper):
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"""Class that generates tallies for one-group rates
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.. versionadded:: 0.12.1
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Parameters
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----------
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n_nucs : int
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Number of burnable nuclides tracked by :class:`openmc.deplete.Operator`
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n_react : int
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Number of reactions tracked by :class:`openmc.deplete.Operator`
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energies : iterable of float
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Energy group boundaries for flux spectrum in [eV]
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Attributes
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----------
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nuclides : list of str
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All nuclides with desired reaction rates.
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"""
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def __init__(self, n_nucs, n_reacts, energies):
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super().__init__(n_nucs, n_reacts)
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self._energies = asarray(energies)
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def generate_tallies(self, materials, scores):
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"""Produce multigroup flux spectrum tally
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Uses the :mod:`openmc.lib` module to generate a multigroup flux tally
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for each burnable material.
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Parameters
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----------
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materials : iterable of :class:`openmc.Material`
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Burnable materials in the problem. Used to construct a
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:class:`openmc.MaterialFilter`
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scores : iterable of str
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Reaction identifiers, e.g. ``"(n, fission)"``, ``"(n, gamma)"``,
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needed for the reaction rate tally.
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"""
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self._materials = materials
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# Convert reactions to MT values (needed when collapsing)
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mt_values = {v: k for k, v in REACTION_NAME.items()}
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mt_values['fission'] = 18
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self._mts = [mt_values[x] for x in scores]
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# Create flux tally with material and energy filters
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self._flux_tally = Tally()
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self._flux_tally.writable = False
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self._flux_tally.filters = [
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MaterialFilter(materials),
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EnergyFilter(self._energies)
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]
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self._flux_tally.scores = ['flux']
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def get_material_rates(self, mat_index, nuc_index, react_index):
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"""Return an array of reaction rates for a material
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Parameters
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----------
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mat_index : int
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Index for material
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nuc_index : iterable of int
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Index for each nuclide in :attr:`nuclides` in the
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desired reaction rate matrix
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react_index : iterable of int
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Index for each reaction scored in the tally
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Returns
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-------
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rates : numpy.ndarray
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Array with shape ``(n_nuclides, n_rxns)`` with the reaction rates in
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this material
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"""
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self._results_cache.fill(0.0)
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# Get flux for specified material
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shape = (len(self._materials), len(self._energies) - 1)
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mean_value = self._flux_tally.mean.reshape(shape)
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flux = mean_value[mat_index]
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mat = self._materials[mat_index]
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for name, i_nuc in zip(self.nuclides, nuc_index):
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# Determine density of nuclide
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# TODO: O(N) search for nuclide is not ideal
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for nucname, density in zip(mat.nuclides, mat.densities):
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if nucname == name:
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break
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else:
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raise ValueError("Couldn't find {} in material {}".format(name, mat_id))
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for mt, i_react in zip(self._mts, react_index):
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# Use flux to collapse reaction rate (per N)
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nuc = openmc.lib.nuclides[name]
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rate_per_nuc = nuc.collapse_rate(mt, self._energies, flux)
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# Multiply by density to get absolute reaction rate
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self._results_cache[i_nuc, i_react] = rate_per_nuc * density
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return self._results_cache
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# ------------------------------------------
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# Helpers for obtaining normalization factor
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# ------------------------------------------
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@ -27,7 +27,7 @@ from .results_list import ResultsList
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from .helpers import (
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DirectReactionRateHelper, ChainFissionHelper, ConstantFissionYieldHelper,
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FissionYieldCutoffHelper, AveragedFissionYieldHelper, EnergyScoreHelper,
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SourceRateHelper)
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SourceRateHelper, FluxCollapseHelper)
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__all__ = ["Operator", "OperatorResult"]
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@ -113,6 +113,18 @@ class Operator(TransportOperator):
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``fission_yield_mode``. Will be passed directly on to the
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helper. Passing a value of None will use the defaults for
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the associated helper.
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reaction_rate_mode : {"direct", "flux"}
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Indicate how one-group reaction rates should be calculated. The "direct"
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method tallies transmutation reaction rates directly. The "flux" method
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tallies a multigroup flux spectrum and then collapses one-group reaction
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rates after a transport solve.
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.. versionadded:: 0.12.1
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reaction_rate_energies : iterable of float
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Energy group boundaries that are to be used for calculating a multigroup
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flux spectrum when the "flux" based ``reaction_rate_mode`` is being used.
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.. versionadded:: 0.12.1
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reduce_chain : bool, optional
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If True, use :meth:`openmc.deplete.Chain.reduce` to reduce the
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depletion chain up to ``reduce_chain_level``. Default is False.
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@ -170,6 +182,7 @@ class Operator(TransportOperator):
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diff_burnable_mats=False, normalization_mode="fission-q",
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fission_q=None, dilute_initial=1.0e3,
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fission_yield_mode="constant", fission_yield_opts=None,
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reaction_rate_mode="direct", reaction_rate_energies=None,
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reduce_chain=False, reduce_chain_level=None):
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check_value('fission yield mode', fission_yield_mode,
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self._fission_helpers.keys())
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@ -239,8 +252,24 @@ class Operator(TransportOperator):
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self.local_mats, self._burnable_nucs, self.chain.reactions)
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# Get classes to assist working with tallies
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self._rate_helper = DirectReactionRateHelper(
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self.reaction_rates.n_nuc, self.reaction_rates.n_react)
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if reaction_rate_mode == "direct":
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self._rate_helper = DirectReactionRateHelper(
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self.reaction_rates.n_nuc, self.reaction_rates.n_react)
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elif reaction_rate_mode == "flux":
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# Ensure energy group boundaries were specified
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if reaction_rate_energies is None:
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raise ValueError(
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"Energy group boundaries must be specified in the "
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"reaction_rate_energies argument when reaction_rate_mode is"
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"set to 'flux'.")
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self._rate_helper = FluxCollapseHelper(
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self.reaction_rates.n_nuc,
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self.reaction_rates.n_react,
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reaction_rate_energies
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)
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else:
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raise ValueError("Invalid reaction rate mode.")
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if normalization_mode == "fission-q":
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self._normalization_helper = ChainFissionHelper()
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