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Removed MGXS subclass create_tallies routines in place of tallies property getters
This commit is contained in:
Will Boyd
parent
e57ca85a7c
commit
370c942977
3 changed files with 310 additions and 260 deletions
File diff suppressed because one or more lines are too long
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@ -248,7 +248,6 @@ class Library(object):
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if isinstance(mgxs, openmc.mgxs.ScatterMatrixXS):
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mgxs.correction = self.correction
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mgxs.create_tallies()
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self.all_mgxs[domain.id][mgxs_type] = mgxs
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def add_to_tallies_file(self, tallies_file, merge=True):
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@ -41,7 +41,7 @@ DOMAIN_TYPES = ['cell',
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# Supported domain classes
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# TODO: Implement Mesh domains
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_DOMAINS = [openmc.Cell,
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DOMAINS = [openmc.Cell,
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openmc.Universe,
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openmc.Material]
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@ -106,7 +106,7 @@ class MGXS(object):
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self._domain_type = None
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self._energy_groups = None
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self._tally_trigger = None
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self._tallies = OrderedDict()
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self._tallies = None
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self._xs_tally = None
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self.name = name
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@ -409,7 +409,6 @@ class MGXS(object):
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return densities
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@abc.abstractmethod
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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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@ -437,6 +436,8 @@ class MGXS(object):
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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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@ -535,9 +536,6 @@ class MGXS(object):
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'which is not yet supported'.format(self.domain_type)
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raise ValueError(msg)
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# Create Tallies to search for in StatePoint
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self.create_tallies()
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# Use tally "slicing" to ensure that tallies correspond to our domain
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# NOTE: This is important if tally merging was used
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if self.domain_type != 'distribcell':
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@ -1263,7 +1261,8 @@ class TotalXS(MGXS):
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groups, by_nuclide, name)
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self._rxn_type = 'total'
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def create_tallies(self):
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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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@ -1272,18 +1271,24 @@ class TotalXS(MGXS):
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"""
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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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# Instantiate tallies if they do not exist
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if self._tallies is None:
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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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# 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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# Initialize the Tallies
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super(TotalXS, self).create_tallies(scores, filters, keys, estimator)
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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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super(TotalXS, self).create_tallies(scores, filters,
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keys, estimator)
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return super(TotalXS, self).tallies
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def compute_xs(self):
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"""Computes the multi-group total cross sections using OpenMC
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@ -1303,7 +1308,8 @@ class TransportXS(MGXS):
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groups, by_nuclide, name)
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self._rxn_type = 'transport'
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def create_tallies(self):
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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 three analog tallies to compute the 'flux',
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@ -1312,20 +1318,25 @@ class TransportXS(MGXS):
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"""
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# Create a list of scores for each Tally to be created
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scores = ['flux', 'total', 'scatter-P1']
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estimator = 'analog'
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keys = scores
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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 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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# Create a list of scores for each Tally to be created
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scores = ['flux', 'total', 'scatter-P1']
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estimator = 'analog'
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keys = scores
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# Initialize the Tallies
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super(TransportXS, self).create_tallies(scores, filters,
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keys, estimator)
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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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super(TransportXS, self).create_tallies(scores, filters,
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keys, estimator)
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return super(TransportXS, self).tallies
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def load_from_statepoint(self, statepoint):
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"""Extracts tallies in an OpenMC StatePoint with the data needed to
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@ -1375,7 +1386,8 @@ class AbsorptionXS(MGXS):
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groups, by_nuclide, name)
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self._rxn_type = 'absorption'
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def create_tallies(self):
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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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@ -1384,19 +1396,24 @@ class AbsorptionXS(MGXS):
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"""
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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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# Instantiate tallies if they do not exist
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if self._tallies is None:
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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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# 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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# Initialize the Tallies
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super(AbsorptionXS, self).create_tallies(scores, filters,
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keys, estimator)
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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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super(AbsorptionXS, self).create_tallies(scores, filters,
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keys, estimator)
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return super(AbsorptionXS, self).tallies
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def compute_xs(self):
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"""Computes the multi-group absorption cross sections using OpenMC
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@ -1415,7 +1432,8 @@ class CaptureXS(MGXS):
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groups, by_nuclide, name)
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self._rxn_type = 'capture'
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def create_tallies(self):
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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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@ -1424,18 +1442,24 @@ class CaptureXS(MGXS):
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"""
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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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# Instantiate tallies if they do not exist
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if self._tallies is None:
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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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# 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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# Initialize the Tallies
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super(CaptureXS, self).create_tallies(scores, filters, keys, estimator)
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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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super(CaptureXS, self).create_tallies(scores, filters,
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keys, estimator)
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return super(CaptureXS, self).tallies
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def compute_xs(self):
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"""Computes the multi-group capture cross sections using OpenMC
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@ -1455,7 +1479,8 @@ class FissionXS(MGXS):
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groups, by_nuclide, name)
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self._rxn_type = 'fission'
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def create_tallies(self):
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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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@ -1464,18 +1489,24 @@ class FissionXS(MGXS):
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"""
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# Create a list of scores for each Tally to be created
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scores = ['flux', 'fission']
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estimator = 'tracklength'
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keys = scores
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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 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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# Create a list of scores for each Tally to be created
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scores = ['flux', 'fission']
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estimator = 'tracklength'
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keys = scores
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# Initialize the Tallies
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super(FissionXS, self).create_tallies(scores, filters, keys, estimator)
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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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super(FissionXS, self).create_tallies(scores, filters,
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keys, estimator)
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return super(FissionXS, self).tallies
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def compute_xs(self):
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"""Computes the multi-group fission cross sections using OpenMC
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@ -1494,7 +1525,8 @@ class NuFissionXS(MGXS):
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groups, by_nuclide, name)
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self._rxn_type = 'nu-fission'
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def create_tallies(self):
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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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@ -1503,19 +1535,24 @@ class NuFissionXS(MGXS):
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"""
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# Create a list of scores for each Tally to be created
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scores = ['flux', 'nu-fission']
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estimator = 'tracklength'
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keys = scores
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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 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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# Create a list of scores for each Tally to be created
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scores = ['flux', 'nu-fission']
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estimator = 'tracklength'
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keys = scores
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# Initialize the Tallies
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super(NuFissionXS, self).create_tallies(scores, filters,
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keys, estimator)
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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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super(NuFissionXS, self).create_tallies(scores, filters,
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keys, estimator)
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return super(NuFissionXS, self).tallies
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def compute_xs(self):
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"""Computes the multi-group nu-fission cross sections using OpenMC
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@ -1534,7 +1571,8 @@ class ScatterXS(MGXS):
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groups, by_nuclide, name)
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self._rxn_type = 'scatter'
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def create_tallies(self):
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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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@ -1543,18 +1581,24 @@ class ScatterXS(MGXS):
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"""
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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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# Instantiate tallies if they do not exist
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if self._tallies is None:
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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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# 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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# Intialize the Tallies
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super(ScatterXS, self).create_tallies(scores, filters, keys, estimator)
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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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# Intialize the Tallies
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super(ScatterXS, self).create_tallies(scores, filters,
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keys, estimator)
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return super(ScatterXS, self).tallies
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def compute_xs(self):
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"""Computes the scattering multi-group cross sections using
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@ -1573,7 +1617,8 @@ class NuScatterXS(MGXS):
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groups, by_nuclide, name)
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self._rxn_type = 'nu-scatter'
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def create_tallies(self):
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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 analog tallies to compute the 'flux'
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@ -1582,19 +1627,24 @@ class NuScatterXS(MGXS):
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"""
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# Create a list of scores for each Tally to be created
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scores = ['flux', 'nu-scatter']
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estimator = 'analog'
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keys = scores
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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 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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# Create a list of scores for each Tally to be created
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scores = ['flux', 'nu-scatter']
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estimator = 'analog'
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keys = scores
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# Initialize the Tallies
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super(NuScatterXS, self).create_tallies(scores, filters,
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keys, estimator)
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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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super(NuScatterXS, self).create_tallies(scores, filters,
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keys, estimator)
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return super(NuScatterXS, self).tallies
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def compute_xs(self):
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"""Computes the nu-scattering multi-group cross section using OpenMC
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@ -1630,12 +1680,8 @@ class ScatterMatrixXS(MGXS):
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def correction(self):
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return self._correction
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@correction.setter
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def correction(self, correction):
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cv.check_value('correction', correction, ('P0', None))
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self._correction = correction
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def create_tallies(self):
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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 three analog tallies to compute the 'flux',
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@ -1644,24 +1690,34 @@ class ScatterMatrixXS(MGXS):
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"""
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group_edges = self.energy_groups.group_edges
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energy = openmc.Filter('energy', group_edges)
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energyout = openmc.Filter('energyout', group_edges)
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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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if self.correction == 'P0':
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scores = ['flux', 'scatter', 'scatter-P1']
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filters = [[energy], [energy, energyout], [energyout]]
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else:
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scores = ['flux', 'scatter']
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filters = [[energy], [energy, energyout]]
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group_edges = self.energy_groups.group_edges
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energy = openmc.Filter('energy', group_edges)
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energyout = openmc.Filter('energyout', group_edges)
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estimator = 'analog'
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keys = scores
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# Create a list of scores for each Tally to be created
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if self.correction == 'P0':
|
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scores = ['flux', 'scatter', 'scatter-P1']
|
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filters = [[energy], [energy, energyout], [energyout]]
|
||||
else:
|
||||
scores = ['flux', 'scatter']
|
||||
filters = [[energy], [energy, energyout]]
|
||||
|
||||
# Initialize the Tallies
|
||||
super(ScatterMatrixXS, self).create_tallies(scores, filters,
|
||||
keys, estimator)
|
||||
estimator = 'analog'
|
||||
keys = scores
|
||||
|
||||
# Initialize the Tallies
|
||||
super(ScatterMatrixXS, self).create_tallies(scores, filters,
|
||||
keys, estimator)
|
||||
|
||||
return super(ScatterMatrixXS, self).tallies
|
||||
|
||||
@correction.setter
|
||||
def correction(self, correction):
|
||||
cv.check_value('correction', correction, ('P0', None))
|
||||
self._correction = correction
|
||||
|
||||
def compute_xs(self, correction='P0'):
|
||||
"""Computes the multi-group scattering matrix using OpenMC
|
||||
|
|
@ -1943,7 +1999,8 @@ class NuScatterMatrixXS(ScatterMatrixXS):
|
|||
groups, by_nuclide, name)
|
||||
self._rxn_type = 'nu-scatter matrix'
|
||||
|
||||
def create_tallies(self):
|
||||
@property
|
||||
def tallies(self):
|
||||
"""Construct the OpenMC tallies needed to compute this cross section.
|
||||
|
||||
This method constructs three analog tallies to compute the 'flux',
|
||||
|
|
@ -1952,26 +2009,31 @@ class NuScatterMatrixXS(ScatterMatrixXS):
|
|||
|
||||
"""
|
||||
|
||||
# Create the non-domain specific Filters for the Tallies
|
||||
group_edges = self.energy_groups.group_edges
|
||||
energy = openmc.Filter('energy', group_edges)
|
||||
energyout = openmc.Filter('energyout', group_edges)
|
||||
# Instantiate tallies if they do not exist
|
||||
if self._tallies is None:
|
||||
|
||||
# Create a list of scores for each Tally to be created
|
||||
if self.correction == 'P0':
|
||||
scores = ['flux', 'nu-scatter', 'scatter-P1']
|
||||
estimator = 'analog'
|
||||
keys = ['flux', 'scatter', 'scatter-P1']
|
||||
filters = [[energy], [energy, energyout], [energyout]]
|
||||
else:
|
||||
scores = ['flux', 'nu-scatter']
|
||||
estimator = 'analog'
|
||||
keys = ['flux', 'scatter']
|
||||
filters = [[energy], [energy, energyout]]
|
||||
# Create the non-domain specific Filters for the Tallies
|
||||
group_edges = self.energy_groups.group_edges
|
||||
energy = openmc.Filter('energy', group_edges)
|
||||
energyout = openmc.Filter('energyout', group_edges)
|
||||
|
||||
# Intialize the Tallies
|
||||
super(ScatterMatrixXS, self).create_tallies(scores, filters,
|
||||
keys, estimator)
|
||||
# Create a list of scores for each Tally to be created
|
||||
if self.correction == 'P0':
|
||||
scores = ['flux', 'nu-scatter', 'scatter-P1']
|
||||
estimator = 'analog'
|
||||
keys = ['flux', 'scatter', 'scatter-P1']
|
||||
filters = [[energy], [energy, energyout], [energyout]]
|
||||
else:
|
||||
scores = ['flux', 'nu-scatter']
|
||||
estimator = 'analog'
|
||||
keys = ['flux', 'scatter']
|
||||
filters = [[energy], [energy, energyout]]
|
||||
|
||||
# Intialize the Tallies
|
||||
super(ScatterMatrixXS, self).create_tallies(scores, filters,
|
||||
keys, estimator)
|
||||
|
||||
return super(ScatterMatrixXS, self).tallies
|
||||
|
||||
class Chi(MGXS):
|
||||
"""The fission spectrum."""
|
||||
|
|
@ -1981,7 +2043,8 @@ class Chi(MGXS):
|
|||
super(Chi, self).__init__(domain, domain_type, groups, by_nuclide, name)
|
||||
self._rxn_type = 'chi'
|
||||
|
||||
def create_tallies(self):
|
||||
@property
|
||||
def tallies(self):
|
||||
"""Construct the OpenMC tallies needed to compute this cross section.
|
||||
|
||||
This method constructs two analog tallies to compute 'nu-fission'
|
||||
|
|
@ -1990,19 +2053,24 @@ class Chi(MGXS):
|
|||
|
||||
"""
|
||||
|
||||
# Create a list of scores for each Tally to be created
|
||||
scores = ['nu-fission', 'nu-fission']
|
||||
estimator = 'analog'
|
||||
keys = ['nu-fission-in', 'nu-fission-out']
|
||||
# Instantiate tallies if they do not exist
|
||||
if self._tallies is None:
|
||||
|
||||
# Create the non-domain specific Filters for the Tallies
|
||||
group_edges = self.energy_groups.group_edges
|
||||
energyout = openmc.Filter('energyout', group_edges)
|
||||
energyin = openmc.Filter('energy', [group_edges[0], group_edges[-1]])
|
||||
filters = [[energyin], [energyout]]
|
||||
# Create a list of scores for each Tally to be created
|
||||
scores = ['nu-fission', 'nu-fission']
|
||||
estimator = 'analog'
|
||||
keys = ['nu-fission-in', 'nu-fission-out']
|
||||
|
||||
# Intialize the Tallies
|
||||
super(Chi, self).create_tallies(scores, filters, keys, estimator)
|
||||
# Create the non-domain specific Filters for the Tallies
|
||||
group_edges = self.energy_groups.group_edges
|
||||
energyout = openmc.Filter('energyout', group_edges)
|
||||
energyin = openmc.Filter('energy', [group_edges[0], group_edges[-1]])
|
||||
filters = [[energyin], [energyout]]
|
||||
|
||||
# Intialize the Tallies
|
||||
super(Chi, self).create_tallies(scores, filters, keys, estimator)
|
||||
|
||||
return super(Chi, self).tallies
|
||||
|
||||
def compute_xs(self):
|
||||
"""Computes chi fission spectrum using OpenMC tally arithmetic."""
|
||||
|
|
|
|||
Loading…
Add table
Add a link
Reference in a new issue