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Return energy in J/s/source neutron for EnergyHelper
Addressing comments in review for #1278 - Documentation cleanup - Better naming convention regarding ReactionRateHelper results cache - The power in Operator tally unpacking and normalizing is no longer converted to eV/s, since the EnergyHelper.energy property is now returned in J/s/source neutron
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Andrew Johnson
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4 changed files with 55 additions and 41 deletions
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@ -75,8 +75,8 @@ data, such as number densities and reaction rates for each material.
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:template: myclass.rst
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AtomNumber
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ChainFissHelper
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DirectRxnRateHelper
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ChainFissionHelper
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DirectReactionRateHelper
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OperatorResult
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ReactionRates
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Results
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@ -92,7 +92,7 @@ The following classes are abstract classes that can be used to extend the
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:template: myclass.rst
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ReactionRateHelper
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FissionEnergyHelper
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EnergyHelper
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TransportOperator
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Each of the integrator functions also relies on a number of "helper" functions
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@ -11,9 +11,9 @@ from abc import ABC, abstractmethod
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from xml.etree import ElementTree as ET
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from warnings import warn
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from numpy import zeros, nonzero
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from numpy import nonzero, empty
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from openmc.data import DataLibrary
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from openmc.data import DataLibrary, JOULE_PER_EV
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from openmc.checkvalue import check_type
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from .chain import Chain
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@ -188,7 +188,7 @@ class ReactionRateHelper(ABC):
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@abstractmethod
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def generate_tallies(self, materials, scores):
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"""Use the capi to build tallies needed for reaction rates"""
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"""Use the C API to build tallies needed for reaction rates"""
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@property
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def nuclides(self):
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@ -201,13 +201,15 @@ class ReactionRateHelper(ABC):
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self._nuclides = nuclides
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self._rate_tally.nuclides = nuclides
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def _reset_results_cache(self, nnucs, nreact):
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"""Cache for results for a given material"""
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def _get_results_cache(self, nnucs, nreact):
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"""Cache for results for a given material
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Creates an empty array of shape ``(nnucs, nreact)``
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if the shape does not match the current cache.
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"""
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if (self._results_cache is None
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or self._results_cache.shape != (nnucs, nreact)):
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self._results_cache = zeros((nnucs, nreact))
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else:
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self._results_cache.fill(0.0)
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self._results_cache = empty((nnucs, nreact))
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return self._results_cache
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@abstractmethod
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@ -216,6 +218,8 @@ class ReactionRateHelper(ABC):
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Parameters
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----------
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mat_id : int
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Unique ID for the requested material
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nuc_index : list of str
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Ordering of desired nuclides
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react_index : list of str
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@ -230,17 +234,13 @@ class ReactionRateHelper(ABC):
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Parameters
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----------
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energy : float
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Energy produced in this region [W]
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number : iterable of float
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Number density [atoms/b/cm] of each nuclide tracked in the calculation.
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Ordered identically to :attr:`nuclides`
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Returns
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-------
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results : `numpy.ndarray`
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2D array ``[n_nuclides, n_rxns]`` of reaction rates normalized by
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the number of nuclides
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results : numpy.ndarray
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Array of reactions rates of shape ``(n_nuclides, n_rxns)``
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normalized by the number of nuclides
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"""
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mask = nonzero(number)
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@ -266,7 +266,7 @@ class EnergyHelper(ABC):
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All nuclides with desired reaction rates. Ordered to be
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consistent with :class:`openmc.deplete.Operator`
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energy : float
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Total energy [eV/s] produced in a transport simulation.
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Total energy [J/s/source neutron] produced in a transport simulation.
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Updated in the material iteration with :meth:`update`.
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"""
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@ -276,7 +276,7 @@ class EnergyHelper(ABC):
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@property
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def energy(self):
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return self._energy
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return self._energy * JOULE_PER_EV
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def reset(self):
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"""Reset energy produced prior to unpacking tallies"""
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@ -14,7 +14,14 @@ from .abc import ReactionRateHelper, EnergyHelper
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class DirectReactionRateHelper(ReactionRateHelper):
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"""Class that generates tallies for one-group rates"""
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"""Class that generates tallies for one-group rates
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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. Ordered to be
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consistent with :class:`openmc.deplete.Operator`
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"""
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def generate_tallies(self, materials, scores):
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"""Produce one-group reaction rate tally
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@ -41,7 +48,7 @@ class DirectReactionRateHelper(ReactionRateHelper):
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Parameters
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----------
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mat_id : int
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Unique id for the requested material
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Unique ID for the requested 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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@ -50,11 +57,12 @@ class DirectReactionRateHelper(ReactionRateHelper):
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Returns
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-------
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rates : :class:`numpy.ndarray`
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2D matrix ``(len(nuc_index), len(react_index))`` with the
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rates : numpy.ndarray
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Array with shape ``(n_nuclides, n_rxns)`` with the
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reaction rates in this material
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"""
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results = self._reset_results_cache(len(nuc_index), len(react_index))
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results = self._get_results_cache(len(nuc_index), len(react_index))
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results.fill(0.0)
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full_tally_res = self._rate_tally.results[mat_id, :, 1]
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for i_tally, (i_nuc, i_react) in enumerate(
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product(nuc_index, react_index)):
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@ -63,13 +71,23 @@ class DirectReactionRateHelper(ReactionRateHelper):
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return results
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# ------------------------------------
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# Helpers for obtaining fission energy
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# ------------------------------------
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# ----------------------------
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# Helpers for obtaining energy
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# ----------------------------
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class ChainFissionHelper(EnergyHelper):
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"""Fission Q-values are pulled from chain"""
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"""Computes energy using fission Q values from depletion chain
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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. Ordered to be
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consistent with :class:`openmc.deplete.Operator`
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energy : float
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Total energy [J/s/source neutron] produced in a transport simulation.
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Updated in the material iteration with :meth:`update`.
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"""
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def __init__(self):
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super().__init__()
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@ -78,8 +96,8 @@ class ChainFissionHelper(EnergyHelper):
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def prepare(self, chain_nucs, rate_index, _materials):
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"""Populate the fission Q value vector from a chain.
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Paramters
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---------
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Parameters
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----------
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chain_nucs : iterable of :class:`openmc.deplete.Nuclide`
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Nuclides used in this depletion chain. Do not need
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to be ordered
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@ -19,7 +19,6 @@ import numpy as np
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import openmc
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import openmc.capi
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from openmc.data import JOULE_PER_EV
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from . import comm
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from .abc import TransportOperator, OperatorResult
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from .atom_number import AtomNumber
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@ -138,7 +137,11 @@ class Operator(TransportOperator):
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openmc.reset_auto_ids()
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self.burnable_mats, volume, nuclides = self._get_burnable_mats()
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self.local_mats = _distribute(self.burnable_mats)
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self._mat_index_map = {}
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# Generate map from local materials => material index
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self._mat_index_map = {
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lm: self.burnable_mats.index(lm) for lm in self.local_mats}
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# Determine which nuclides have incident neutron data
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self.nuclides_with_data = self._get_nuclides_with_data()
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@ -158,7 +161,6 @@ class Operator(TransportOperator):
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self._rate_helper = DirectReactionRateHelper()
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self._energy_helper = ChainFissionHelper()
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def __call__(self, vec, power, print_out=True):
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"""Runs a simulation.
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@ -387,10 +389,6 @@ class Operator(TransportOperator):
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self._energy_helper.prepare(
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self.chain.nuclides, self.reaction_rates.index_nuc, materials)
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# Generate map from local materials => material index
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self._mat_index_map = {
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lm: self.burnable_mats.index(lm) for lm in self.local_mats}
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# Return number density vector
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return list(self.number.get_mat_slice(np.s_[:]))
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@ -564,11 +562,9 @@ class Operator(TransportOperator):
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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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# Determine power in eV/s
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power /= JOULE_PER_EV
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# Scale reaction rates to obtain units of reactions/sec
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rates *= power / energy
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