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Change EnergyHelper -> NormalizationHelper
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Paul Romano
parent
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commit
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4 changed files with 38 additions and 40 deletions
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@ -207,7 +207,7 @@ OpenMC simulations back on to the :class:`abc.TransportOperator`
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:nosignatures:
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:template: myclass.rst
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abc.EnergyHelper
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abc.NormalizationHelper
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abc.FissionYieldHelper
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abc.ReactionRateHelper
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abc.TalliedFissionYieldHelper
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@ -29,7 +29,7 @@ from .pool import deplete
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__all__ = [
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"OperatorResult", "TransportOperator", "ReactionRateHelper",
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"EnergyHelper", "FissionYieldHelper", "TalliedFissionYieldHelper",
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"NormalizationHelper", "FissionYieldHelper", "TalliedFissionYieldHelper",
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"Integrator", "SIIntegrator", "DepSystemSolver"]
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@ -302,14 +302,14 @@ class ReactionRateHelper(ABC):
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return results
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class EnergyHelper(ABC):
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"""Abstract class for obtaining energy produced
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class NormalizationHelper(ABC):
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"""Abstract class for obtaining normalization factor on tallies
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The ultimate goal of this helper is to provide instances of
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:class:`openmc.deplete.Operator` with the total energy produced
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in a transport simulation. This information, provided with the
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power requested by the user and reaction rates from a
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:class:`ReactionRateHelper` will scale reaction rates to the
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This helper class determines how reaction rates calculated by an instance of
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:class:`openmc.deplete.Operator` should be normalized for the purpose of
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constructing a burnup matrix. Based on the method chosen, the power or
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source rate provided by the user, and reaction rates from a
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:class:`ReactionRateHelper`, this class will scale reaction rates to the
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correct values.
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Attributes
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@ -14,7 +14,7 @@ from openmc.checkvalue import check_type, check_greater_than
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from openmc.lib import (
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Tally, MaterialFilter, EnergyFilter, EnergyFunctionFilter)
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from .abc import (
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ReactionRateHelper, EnergyHelper, FissionYieldHelper,
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ReactionRateHelper, NormalizationHelper, FissionYieldHelper,
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TalliedFissionYieldHelper)
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__all__ = (
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@ -91,13 +91,13 @@ class DirectReactionRateHelper(ReactionRateHelper):
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return self._results_cache
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# ----------------------------
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# Helpers for obtaining energy
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# ----------------------------
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# ------------------------------------------
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# Helpers for obtaining normalization factor
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# ------------------------------------------
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class ChainFissionHelper(EnergyHelper):
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"""Computes energy using fission Q values from depletion chain
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class ChainFissionHelper(NormalizationHelper):
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"""Computes normalization using fission Q values from depletion chain
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Attributes
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----------
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@ -159,7 +159,7 @@ class ChainFissionHelper(EnergyHelper):
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self._energy += dot(fission_rates, self._fission_q_vector)
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class EnergyScoreHelper(EnergyHelper):
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class EnergyScoreHelper(NormalizationHelper):
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"""Class responsible for obtaining system energy via a tally score
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Parameters
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@ -172,7 +172,7 @@ class EnergyScoreHelper(EnergyHelper):
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----------
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nuclides : list of str
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List of nuclides with reaction rates. Not needed, but provided
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for a consistent API across other :class:`EnergyHelper`
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for a consistent API across other :class:`NormalizationHelper`
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energy : float
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System energy [eV] computed from the tally. Will be zero for
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all MPI processes that are not the "master" process to avoid
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@ -18,6 +18,7 @@ import numpy as np
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from uncertainties import ufloat
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import openmc
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from openmc.checkvalue import check_value
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import openmc.lib
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from . import comm
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from .abc import TransportOperator, OperatorResult
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@ -81,7 +82,7 @@ class Operator(TransportOperator):
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Whether to differentiate burnable materials with multiple instances.
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Volumes are divided equally from the original material volume.
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Default: False.
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energy_mode : {"energy-deposition", "fission-q"}
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normalization_mode : {"energy-deposition", "fission-q", "source-rate"}
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Indicator for computing system energy. ``"energy-deposition"`` will
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compute with a single energy deposition tally, taking fission energy
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release data and heating into consideration. ``"fission-q"`` will
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@ -89,7 +90,7 @@ class Operator(TransportOperator):
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fission_q : dict, optional
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Dictionary of nuclides and their fission Q values [eV]. If not given,
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values will be pulled from the ``chain_file``. Only applicable
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if ``"energy_mode" == "fission-q"``
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if ``"normalization_mode" == "fission-q"``
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dilute_initial : float, optional
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Initial atom density [atoms/cm^3] to add for nuclides that are zero
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in initial condition to ensure they exist in the decay chain.
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@ -164,23 +165,18 @@ class Operator(TransportOperator):
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}
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def __init__(self, geometry, settings, chain_file=None, prev_results=None,
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diff_burnable_mats=False, energy_mode="fission-q",
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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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reduce_chain=False, reduce_chain_level=None):
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if fission_yield_mode not in self._fission_helpers:
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raise KeyError(
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"fission_yield_mode must be one of {}, not {}".format(
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", ".join(self._fission_helpers), fission_yield_mode))
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if energy_mode == "energy-deposition":
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check_value('fission yield mode', fission_yield_mode,
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self._fission_helpers.keys())
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check_value('normalization mode', normalization_mode,
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('energy-deposition', 'fission-q', 'source-rate'))
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if normalization_mode != "fission-q":
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if fission_q is not None:
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warn("Fission Q dictionary not used if energy deposition "
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"is used")
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warn("Fission Q dictionary will not be used")
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fission_q = None
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elif energy_mode != "fission-q":
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raise ValueError(
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"energy_mode {} not supported. Must be energy-deposition "
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"or fission-q".format(energy_mode))
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super().__init__(chain_file, fission_q, dilute_initial, prev_results)
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self.round_number = False
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self.settings = settings
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@ -243,11 +239,13 @@ class Operator(TransportOperator):
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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 energy_mode == "fission-q":
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self._energy_helper = ChainFissionHelper()
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else:
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if normalization_mode == "fission-q":
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self._normalization_helper = ChainFissionHelper()
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elif normalization_mode == "energy-deposition":
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score = "heating" if settings.photon_transport else "heating-local"
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self._energy_helper = EnergyScoreHelper(score)
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self._normalization_helper = EnergyScoreHelper(score)
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else:
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self._normalization_helper = ...
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# Select and create fission yield helper
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fission_helper = self._fission_helpers[fission_yield_mode]
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@ -286,7 +284,7 @@ class Operator(TransportOperator):
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self._update_materials()
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nuclides = self._get_tally_nuclides()
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self._rate_helper.nuclides = nuclides
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self._energy_helper.nuclides = nuclides
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self._normalization_helper.nuclides = nuclides
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self._yield_helper.update_tally_nuclides(nuclides)
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# Run OpenMC
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@ -493,7 +491,7 @@ class Operator(TransportOperator):
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materials = [openmc.lib.materials[int(i)]
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for i in self.burnable_mats]
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self._rate_helper.generate_tallies(materials, self.chain.reactions)
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self._energy_helper.prepare(
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self._normalization_helper.prepare(
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self.chain.nuclides, self.reaction_rates.index_nuc, materials)
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# Tell fission yield helper what materials this process is
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# responsible for
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@ -643,7 +641,7 @@ class Operator(TransportOperator):
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# Keep track of energy produced from all reactions in eV per source
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# particle
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self._energy_helper.reset()
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self._normalization_helper.reset()
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self._yield_helper.unpack()
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# Store fission yield dictionaries
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@ -674,14 +672,14 @@ class Operator(TransportOperator):
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fission_yields.append(self._yield_helper.weighted_yields(i))
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# Accumulate energy from fission
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self._energy_helper.update(tally_rates[:, fission_ind], mat_index)
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self._normalization_helper.update(tally_rates[:, fission_ind], mat_index)
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# Divide by total number and store
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rates[i] = self._rate_helper.divide_by_adens(number)
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# Reduce energy produced from all processes
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# J / s / source neutron
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energy = comm.allreduce(self._energy_helper.energy)
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energy = comm.allreduce(self._normalization_helper.energy)
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# Guard against divide by zero
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if energy == 0:
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