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Partial refactor of MGXS subclasses
This commit is contained in:
Will Boyd
parent
17f4927e8b
commit
9e843de1bd
1 changed files with 95 additions and 350 deletions
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@ -215,10 +215,50 @@ class MGXS(object):
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@property
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def tallies(self):
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"""Construct the OpenMC tallies needed to compute the cross section."""
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# Instantiate tallies if they do not exist
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if self._tallies is None:
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# Initialize a collection of Tallies
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self._tallies = OrderedDict()
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# Create a domain Filter object
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domain_filter = openmc.Filter(self.domain_type, self.domain.id)
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# Create each Tally needed to compute the multi group cross section
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for score, key, filters in zip(self.scores, self.keys, self.filters):
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self.tallies[key] = openmc.Tally(name=self.name)
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self.tallies[key].scores = [score]
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self.tallies[key].estimator = self.estimator
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self.tallies[key].filters = [domain_filter]
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# If a tally trigger was specified, add it to each tally
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if self.tally_trigger:
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trigger_clone = copy.deepcopy(self.tally_trigger)
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trigger_clone.scores = [score]
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self.tallies[key].triggers.append(trigger_clone)
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# Add non-domain specific Filters (e.g., 'energy') to the Tally
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for add_filter in filters:
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self.tallies[key].filters.append(add_filter)
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# If this is a by-nuclide cross-section, add nuclides to Tally
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if self.by_nuclide and score != 'flux':
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all_nuclides = self.get_all_nuclides()
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for nuclide in all_nuclides:
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self.tallies[key].nuclides.append(nuclide)
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else:
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self.tallies[key].nuclides.append('total')
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return self._tallies
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@property
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def rxn_rate_tally(self):
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if self._rxn_rate_tally is None:
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self._rxn_rate_tally = self.tallies[self.rxn_type]
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self._rxn_rate_tally.sparse = self.sparse
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return self._rxn_rate_tally
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@property
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@ -263,6 +303,24 @@ class MGXS(object):
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def derived(self):
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return self._derived
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@property
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def scores(self):
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return ['flux', self.rxn_type]
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@property
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def filters(self):
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group_edges = self.energy_groups.group_edges
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energy_filter = openmc.Filter('energy', group_edges)
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return [[energy_filter] * len(self.scores)]
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@property
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def tally_keys(self):
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return self.scores
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@property
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def estimator(self):
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return 'tracklength'
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@name.setter
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def name(self, name):
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cv.check_type('name', name, basestring)
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@ -504,63 +562,6 @@ class MGXS(object):
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return densities
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def _create_tallies(self, scores, all_filters, keys, estimator):
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"""Instantiates tallies needed to compute the multi-group cross section.
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This is a helper method for MGXS subclasses to create tallies
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for input file generation. The tallies are stored in the tallies dict.
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This method is called by each subclass' tallies property getter
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which define the parameters given to this parent class method.
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Parameters
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----------
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scores : Iterable of str
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Scores for each tally
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all_filters : Iterable of tuple of openmc.Filter
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Tuples of non-spatial domain filters for each tally
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keys : Iterable of str
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Key string used to store each tally in the tallies dictionary
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estimator : {'analog', 'tracklength'}
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Type of estimator to use for each tally
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"""
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cv.check_iterable_type('scores', scores, basestring)
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cv.check_length('scores', scores, len(keys))
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cv.check_iterable_type('filters', all_filters, openmc.Filter, 1, 2)
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cv.check_type('keys', keys, Iterable, basestring)
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cv.check_value('estimator', estimator, ['analog', 'tracklength'])
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self._tallies = OrderedDict()
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# Create a domain Filter object
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domain_filter = openmc.Filter(self.domain_type, self.domain.id)
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# Create each Tally needed to compute the multi group cross section
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for score, key, filters in zip(scores, keys, all_filters):
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self.tallies[key] = openmc.Tally(name=self.name)
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self.tallies[key].scores = [score]
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self.tallies[key].estimator = estimator
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self.tallies[key].filters = [domain_filter]
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# If a tally trigger was specified, add it to each tally
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if self.tally_trigger:
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trigger_clone = copy.deepcopy(self.tally_trigger)
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trigger_clone.scores = [score]
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self.tallies[key].triggers.append(trigger_clone)
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# Add all non-domain specific Filters (e.g., 'energy') to the Tally
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for add_filter in filters:
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self.tallies[key].filters.append(add_filter)
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# If this is a by-nuclide cross-section, add all nuclides to Tally
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if self.by_nuclide and score != 'flux':
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all_nuclides = self.get_all_nuclides()
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for nuclide in all_nuclides:
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self.tallies[key].nuclides.append(nuclide)
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else:
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self.tallies[key].nuclides.append('total')
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def _compute_xs(self):
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"""Performs generic cleanup after a subclass' uses tally arithmetic to
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compute a multi-group cross section as a derived tally.
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@ -1492,100 +1493,30 @@ class TotalXS(MGXS):
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groups, by_nuclide, name)
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self._rxn_type = 'total'
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@property
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def tallies(self):
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"""Construct the OpenMC tallies needed to compute this cross section.
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This method constructs two tracklength tallies to compute the 'flux'
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and 'total' reaction rates in the spatial domain and energy groups
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of interest.
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"""
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# Instantiate tallies if they do not exist
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if self._tallies is None:
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# Create a list of scores for each Tally to be created
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scores = ['flux', 'total']
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estimator = 'tracklength'
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keys = scores
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# Create the non-domain specific Filters for the Tallies
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group_edges = self.energy_groups.group_edges
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energy_filter = openmc.Filter('energy', group_edges)
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filters = [[energy_filter], [energy_filter]]
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# Initialize the Tallies
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self._create_tallies(scores, filters, keys, estimator)
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return self._tallies
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@property
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def rxn_rate_tally(self):
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if self._rxn_rate_tally is None :
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self._rxn_rate_tally = self.tallies['total']
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self._rxn_rate_tally.sparse = self.sparse
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return self._rxn_rate_tally
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class TransportXS(MGXS):
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"""A transport-corrected total multi-group cross section.
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Attributes
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----------
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use_nu : bool
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Whether or not to account for scattering multiplicity in the
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correction. If False, a "scatter-1" score is used (default);
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if True, a "nu-scatter-1" score is used. This should be
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set to False if using a ScatterMatrixXS and True if using
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a NuScatterMatrixXS to preserve neutron balance.
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"""
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"""A transport-corrected total multi-group cross section."""
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def __init__(self, domain=None, domain_type=None,
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groups=None, by_nuclide=False, name=''):
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super(TransportXS, self).__init__(domain, domain_type,
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groups, by_nuclide, name)
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self._rxn_type = 'transport'
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self._use_nu = False
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@property
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def use_nu(self):
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return self._use_nu
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def scores(self):
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return ['flux', 'total', 'scatter-1']
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@property
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def tallies(self):
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"""Construct the OpenMC tallies needed to compute this cross section.
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def filters(self):
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group_edges = self.energy_groups.group_edges
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energy_filter = openmc.Filter('energy', group_edges)
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energyout_filter = openmc.Filter('energyout', group_edges)
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return [[energy_filter], [energy_filter], [energyout_filter]]
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This method constructs three analog tallies to compute the 'flux',
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'total' and 'scatter-1' reaction rates in the spatial domain and
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energy groups of interest.
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"""
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# Instantiate tallies if they do not exist
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if self._tallies is None:
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# Create a list of scores for each Tally to be created
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scores = ['flux', 'total']
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if self.use_nu:
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scores.append('nu-scatter-1')
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else:
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scores.append('scatter-1')
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estimator = 'analog'
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keys = ['flux', 'total', 'scatter-1']
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# Create the non-domain specific Filters for the Tallies
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group_edges = self.energy_groups.group_edges
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energy_filter = openmc.Filter('energy', group_edges)
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energyout_filter = openmc.Filter('energyout', group_edges)
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filters = [[energy_filter], [energy_filter], [energyout_filter]]
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# Initialize the Tallies
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self._create_tallies(scores, filters, keys, estimator)
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return self._tallies
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@property
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def estimator(self):
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return 'analog'
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@property
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def rxn_rate_tally(self):
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@ -1609,33 +1540,12 @@ class NuTransportXS(TransportXS):
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self._rxn_type = 'nu-transport'
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@property
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def tallies(self):
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"""Construct the OpenMC tallies needed to compute this cross section.
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def scores(self):
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return ['flux', 'total', 'nu-scatter-1']
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This method constructs three analog tallies to compute the 'flux',
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'total' and 'nu-scatter-1' reaction rates in the spatial domain and
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energy groups of interest.
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"""
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# Instantiate tallies if they do not exist
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if self._tallies is None:
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# Create a list of scores for each Tally to be created
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scores = ['flux', 'total', 'nu-scatter-1']
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keys = ['flux', 'total', 'scatter-1']
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estimator = 'analog'
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# Create the non-domain specific Filters for the Tallies
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group_edges = self.energy_groups.group_edges
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energy_filter = openmc.Filter('energy', group_edges)
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energyout_filter = openmc.Filter('energyout', group_edges)
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filters = [[energy_filter], [energy_filter], [energyout_filter]]
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# Initialize the Tallies
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self._create_tallies(scores, filters, keys, estimator)
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return self._tallies
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@property
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def keys(self):
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return ['flux', 'total', 'scatter-1']
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class AbsorptionXS(MGXS):
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@ -1647,41 +1557,6 @@ class AbsorptionXS(MGXS):
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groups, by_nuclide, name)
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self._rxn_type = 'absorption'
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@property
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def tallies(self):
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"""Construct the OpenMC tallies needed to compute this cross section.
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This method constructs two tracklength tallies to compute the 'flux'
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and 'absorption' reaction rates in the spatial domain and energy
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groups of interest.
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"""
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# Instantiate tallies if they do not exist
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if self._tallies is None:
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# Create a list of scores for each Tally to be created
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scores = ['flux', 'absorption']
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estimator = 'tracklength'
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keys = scores
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# Create the non-domain specific Filters for the Tallies
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group_edges = self.energy_groups.group_edges
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energy_filter = openmc.Filter('energy', group_edges)
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filters = [[energy_filter], [energy_filter]]
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# Initialize the Tallies
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self._create_tallies(scores, filters, keys, estimator)
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return self._tallies
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@property
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def rxn_rate_tally(self):
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if self._rxn_rate_tally is None:
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self._rxn_rate_tally = self.tallies['absorption']
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self._rxn_rate_tally.sparse = self.sparse
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return self._rxn_rate_tally
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class CaptureXS(MGXS):
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"""A capture multi-group cross section.
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@ -1700,32 +1575,8 @@ class CaptureXS(MGXS):
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self._rxn_type = 'capture'
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@property
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def tallies(self):
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"""Construct the OpenMC tallies needed to compute this cross section.
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This method constructs two tracklength tallies to compute the 'flux'
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and 'capture' reaction rates in the spatial domain and energy
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groups of interest.
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"""
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# Instantiate tallies if they do not exist
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if self._tallies is None:
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# Create a list of scores for each Tally to be created
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scores = ['flux', 'absorption', 'fission']
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estimator = 'tracklength'
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keys = scores
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# Create the non-domain specific Filters for the Tallies
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group_edges = self.energy_groups.group_edges
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energy_filter = openmc.Filter('energy', group_edges)
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filters = [[energy_filter], [energy_filter], [energy_filter]]
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# Initialize the Tallies
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self._create_tallies(scores, filters, keys, estimator)
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return self._tallies
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def scores(self):
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return ['flux', 'absorption', 'fission']
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@property
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def rxn_rate_tally(self):
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@ -1735,80 +1586,36 @@ class CaptureXS(MGXS):
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self._rxn_rate_tally.sparse = self.sparse
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return self._rxn_rate_tally
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class FissionXSBase(MGXS):
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"""A fission production multi-group cross section base class
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for NuFission and KappaFission
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"""
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# This is an abstract class which cannot be instantiated
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__metaclass__ = abc.ABCMeta
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def __init__(self, rxn_type, domain=None, domain_type=None,
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groups=None, by_nuclide=False, name=''):
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super(FissionXSBase, self).__init__(domain, domain_type,
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groups, by_nuclide, name)
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self._rxn_type = rxn_type
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@property
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def tallies(self):
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"""Construct the OpenMC tallies needed to compute this cross section.
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This method constructs two tracklength tallies to compute the 'flux'
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and 'rxn_type' reaction rates in the spatial domain and energy
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groups of interest.
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"""
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# Instantiate tallies if they do not exist
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if self._tallies is None:
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# Create a list of scores for each Tally to be created
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scores = ['flux', self._rxn_type]
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estimator = 'tracklength'
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keys = scores
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# Create the non-domain specific Filters for the Tallies
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group_edges = self.energy_groups.group_edges
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energy_filter = openmc.Filter('energy', group_edges)
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filters = [[energy_filter], [energy_filter]]
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# Initialize the Tallies
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self._create_tallies(scores, filters, keys, estimator)
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return self._tallies
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@property
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def rxn_rate_tally(self):
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if self._rxn_rate_tally is None:
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self._rxn_rate_tally = self.tallies[self._rxn_type]
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self._rxn_rate_tally.sparse = self.sparse
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return self._rxn_rate_tally
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class FissionXS(FissionXSBase):
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class FissionXS(MGXS):
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"""A fission multi-group cross section."""
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def __init__(self, domain=None, domain_type=None,
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groups=None, by_nuclide=False, name=''):
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super(FissionXS, self).__init__('fission', domain, domain_type,
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super(FissionXS, self).__init__(domain, domain_type,
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groups, by_nuclide, name)
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self._rxn_type = 'fission'
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class NuFissionXS(FissionXSBase):
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class NuFissionXS(MGXS):
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"""A fission production multi-group cross section."""
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def __init__(self, domain=None, domain_type=None,
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groups=None, by_nuclide=False, name=''):
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super(NuFissionXS, self).__init__('nu-fission', domain, domain_type,
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super(NuFissionXS, self).__init__(domain, domain_type,
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groups, by_nuclide, name)
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self._rxn_type = 'nu-fission'
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class KappaFissionXS(FissionXSBase):
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class KappaFissionXS(MGXS):
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"""A recoverable fission energy production rate multi-group cross section."""
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def __init__(self, domain=None, domain_type=None,
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groups=None, by_nuclide=False, name=''):
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super(KappaFissionXS, self).__init__('kappa-fission', domain, domain_type,
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super(KappaFissionXS, self).__init__(domain, domain_type,
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groups, by_nuclide, name)
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self._rxn_type = 'kappa-fission'
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class ScatterXS(MGXS):
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"""A scatter multi-group cross section."""
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@ -1819,41 +1626,6 @@ class ScatterXS(MGXS):
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groups, by_nuclide, name)
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self._rxn_type = 'scatter'
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@property
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def tallies(self):
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"""Construct the OpenMC tallies needed to compute this cross section.
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This method constructs two tracklength tallies to compute the 'flux'
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and 'scatter' reaction rates in the spatial domain and energy
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groups of interest.
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"""
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# Instantiate tallies if they do not exist
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if self._tallies is None:
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# Create a list of scores for each Tally to be created
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||||
scores = ['flux', 'scatter']
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estimator = 'tracklength'
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keys = scores
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||||
# Create the non-domain specific Filters for the Tallies
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group_edges = self.energy_groups.group_edges
|
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energy_filter = openmc.Filter('energy', group_edges)
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filters = [[energy_filter], [energy_filter]]
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|
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# Intialize the Tallies
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||||
self._create_tallies(scores, filters, keys, estimator)
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||||
|
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return self._tallies
|
||||
|
||||
@property
|
||||
def rxn_rate_tally(self):
|
||||
if self._rxn_rate_tally is None:
|
||||
self._rxn_rate_tally = self.tallies['scatter']
|
||||
self._rxn_rate_tally.sparse = self.sparse
|
||||
return self._rxn_rate_tally
|
||||
|
||||
|
||||
class NuScatterXS(MGXS):
|
||||
"""A nu-scatter multi-group cross section."""
|
||||
|
|
@ -1864,41 +1636,6 @@ class NuScatterXS(MGXS):
|
|||
groups, by_nuclide, name)
|
||||
self._rxn_type = 'nu-scatter'
|
||||
|
||||
@property
|
||||
def tallies(self):
|
||||
"""Construct the OpenMC tallies needed to compute this cross section.
|
||||
|
||||
This method constructs two analog tallies to compute the 'flux'
|
||||
and 'nu-scatter' reaction rates in the spatial domain and energy
|
||||
groups of interest.
|
||||
|
||||
"""
|
||||
|
||||
# Instantiate tallies if they do not exist
|
||||
if self._tallies is None:
|
||||
|
||||
# Create a list of scores for each Tally to be created
|
||||
scores = ['flux', 'nu-scatter']
|
||||
estimator = 'analog'
|
||||
keys = scores
|
||||
|
||||
# Create the non-domain specific Filters for the Tallies
|
||||
group_edges = self.energy_groups.group_edges
|
||||
energy_filter = openmc.Filter('energy', group_edges)
|
||||
filters = [[energy_filter], [energy_filter]]
|
||||
|
||||
# Initialize the Tallies
|
||||
self._create_tallies(scores, filters, keys, estimator)
|
||||
|
||||
return self._tallies
|
||||
|
||||
@property
|
||||
def rxn_rate_tally(self):
|
||||
if self._rxn_rate_tally is None:
|
||||
self._rxn_rate_tally = self.tallies['nu-scatter']
|
||||
self._rxn_rate_tally.sparse = self.sparse
|
||||
return self._rxn_rate_tally
|
||||
|
||||
|
||||
class ScatterMatrixXS(MGXS):
|
||||
"""A scattering matrix multi-group cross section for one or more Legendre
|
||||
|
|
@ -1935,6 +1672,14 @@ class ScatterMatrixXS(MGXS):
|
|||
def legendre_order(self):
|
||||
return self._legendre_order
|
||||
|
||||
@property
|
||||
def scores(self):
|
||||
return ['flux', 'total', 'nu-scatter-1']
|
||||
|
||||
@property
|
||||
def keys(self):
|
||||
return ['flux', 'total', 'scatter-1']
|
||||
|
||||
@property
|
||||
def tallies(self):
|
||||
"""Construct the OpenMC tallies needed to compute this cross section.
|
||||
|
|
|
|||
Loading…
Add table
Add a link
Reference in a new issue