mirror of
https://github.com/openmc-dev/openmc.git
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Clean up tally namespace
This commit is contained in:
Sterling Harper
parent
52c9898128
commit
62f8dbfc02
4 changed files with 75 additions and 73 deletions
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@ -1,3 +1,4 @@
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from openmc.arithmetic import *
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from openmc.cell import *
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from openmc.lattice import *
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from openmc.element import *
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@ -4,7 +4,7 @@ from collections import Iterable
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import numpy as np
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from openmc import Filter, Nuclide
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import openmc
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from openmc.filter import _FILTER_TYPES
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import openmc.checkvalue as cv
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@ -171,7 +171,7 @@ class CrossNuclide(object):
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string = ''
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# If the Summary was linked, the left nuclide is a Nuclide object
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if isinstance(self.left_nuclide, Nuclide):
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if isinstance(self.left_nuclide, openmc.Nuclide):
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string += '(' + self.left_nuclide.name
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# If the Summary was not linked, the left nuclide is the ZAID
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else:
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@ -180,7 +180,7 @@ class CrossNuclide(object):
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string += ' ' + self.binary_op + ' '
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# If the Summary was linked, the right nuclide is a Nuclide object
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if isinstance(self.right_nuclide, Nuclide):
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if isinstance(self.right_nuclide, openmc.Nuclide):
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string += self.right_nuclide.name + ')'
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# If the Summary was not linked, the right nuclide is the ZAID
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else:
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@ -191,13 +191,13 @@ class CrossNuclide(object):
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@left_nuclide.setter
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def left_nuclide(self, left_nuclide):
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cv.check_type('left_nuclide', left_nuclide,
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(Nuclide, CrossNuclide, AggregateNuclide))
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(openmc.Nuclide, CrossNuclide, AggregateNuclide))
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self._left_nuclide = left_nuclide
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@right_nuclide.setter
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def right_nuclide(self, right_nuclide):
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cv.check_type('right_nuclide', right_nuclide,
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(Nuclide, CrossNuclide, AggregateNuclide))
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(openmc.Nuclide, CrossNuclide, AggregateNuclide))
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self._right_nuclide = right_nuclide
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@binary_op.setter
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@ -330,14 +330,14 @@ class CrossFilter(object):
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@left_filter.setter
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def left_filter(self, left_filter):
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cv.check_type('left_filter', left_filter,
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(Filter, CrossFilter, AggregateFilter))
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(openmc.Filter, CrossFilter, AggregateFilter))
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self._left_filter = left_filter
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self._bins['left'] = left_filter.bins
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@right_filter.setter
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def right_filter(self, right_filter):
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cv.check_type('right_filter', right_filter,
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(Filter, CrossFilter, AggregateFilter))
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(openmc.Filter, CrossFilter, AggregateFilter))
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self._right_filter = right_filter
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self._bins['right'] = right_filter.bins
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@ -550,8 +550,8 @@ class AggregateNuclide(object):
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# Append each nuclide in the aggregate to the string
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string = '{0}('.format(self.aggregate_op)
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names = [nuclide.name if isinstance(nuclide, Nuclide) else str(nuclide)
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for nuclide in self.nuclides]
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names = [nuclide.name if isinstance(nuclide, openmc.Nuclide)
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else str(nuclide) for nuclide in self.nuclides]
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string += ', '.join(map(str, names)) + ')'
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return string
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@ -567,15 +567,15 @@ class AggregateNuclide(object):
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def name(self):
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# Append each nuclide in the aggregate to the string
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names = [nuclide.name if isinstance(nuclide, Nuclide) else str(nuclide)
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for nuclide in self.nuclides]
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names = [nuclide.name if isinstance(nuclide, openmc.Nuclide)
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else str(nuclide) for nuclide in self.nuclides]
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string = '(' + ', '.join(map(str, names)) + ')'
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return string
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@nuclides.setter
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def nuclides(self, nuclides):
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cv.check_iterable_type('nuclides', nuclides,
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(basestring, Nuclide, CrossNuclide))
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(basestring, openmc.Nuclide, CrossNuclide))
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self._nuclides = nuclides
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@aggregate_op.setter
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@ -700,7 +700,8 @@ class AggregateFilter(object):
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@aggregate_filter.setter
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def aggregate_filter(self, aggregate_filter):
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cv.check_type('aggregate_filter', aggregate_filter, (Filter, CrossFilter))
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cv.check_type('aggregate_filter', aggregate_filter,
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(openmc.Filter, CrossFilter))
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self._aggregate_filter = aggregate_filter
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@bins.setter
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@ -8,7 +8,7 @@ from xml.etree import ElementTree as ET
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import numpy as np
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from openmc import Mesh
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import openmc
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import openmc.checkvalue as cv
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@ -506,7 +506,7 @@ class MeshFilter(Filter):
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@mesh.setter
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def mesh(self, mesh):
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cv.check_type('filter mesh', mesh, Mesh)
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cv.check_type('filter mesh', mesh, openmc.Mesh)
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self._mesh = mesh
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self.bins = mesh.id
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@ -13,10 +13,7 @@ from xml.etree import ElementTree as ET
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import numpy as np
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from openmc import Trigger, Nuclide
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from openmc.arithmetic import CrossScore, CrossNuclide, CrossFilter, \
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AggregateScore, AggregateNuclide, AggregateFilter
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import openmc.filter
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import openmc
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import openmc.checkvalue as cv
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from openmc.clean_xml import clean_xml_indentation
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@ -36,9 +33,10 @@ _PRODUCT_TYPES = ['tensor', 'entrywise']
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# The following indicate acceptable types when setting Tally.scores,
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# Tally.nuclides, and Tally.filters
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_SCORE_CLASSES = (basestring, CrossScore, AggregateScore)
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_NUCLIDE_CLASSES = (basestring, Nuclide, CrossNuclide, AggregateNuclide)
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_FILTER_CLASSES = (openmc.filter.Filter, CrossFilter, AggregateFilter)
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_SCORE_CLASSES = (basestring, openmc.CrossScore, openmc.AggregateScore)
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_NUCLIDE_CLASSES = (basestring, openmc.Nuclide, openmc.CrossNuclide,
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openmc.AggregateNuclide)
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_FILTER_CLASSES = (openmc.Filter, openmc.CrossFilter, openmc.AggregateFilter)
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# Valid types of estimators
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ESTIMATOR_TYPES = ['tracklength', 'collision', 'analog']
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@ -118,7 +116,7 @@ class Tally(object):
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self._nuclides = cv.CheckedList(_NUCLIDE_CLASSES, 'tally nuclides')
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self._scores = cv.CheckedList(_SCORE_CLASSES, 'tally scores')
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self._estimator = None
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self._triggers = cv.CheckedList(Trigger, 'tally triggers')
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self._triggers = cv.CheckedList(openmc.Trigger, 'tally triggers')
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self._num_realizations = 0
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self._with_summary = False
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@ -187,7 +185,7 @@ class Tally(object):
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string += '{0: <16}{1}'.format('\tNuclides', '=\t')
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for nuclide in self.nuclides:
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if isinstance(nuclide, Nuclide):
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if isinstance(nuclide, openmc.Nuclide):
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string += '{0} '.format(nuclide.name)
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else:
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string += '{0} '.format(nuclide)
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@ -396,7 +394,8 @@ class Tally(object):
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@triggers.setter
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def triggers(self, triggers):
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cv.check_type('tally triggers', triggers, MutableSequence)
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self._triggers = cv.CheckedList(Trigger, 'tally triggers', triggers)
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self._triggers = cv.CheckedList(openmc.Trigger, 'tally triggers',
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triggers)
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def add_trigger(self, trigger):
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"""Add a tally trigger to the tally
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@ -700,8 +699,8 @@ class Tally(object):
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return False
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# Return False if only one tally has a delayed group filter
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tally1_dg = self.contains_filter(openmc.filter.DelayedGroupFilter)
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tally2_dg = other.contains_filter(openmc.filter.DelayedGroupFilter)
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tally1_dg = self.contains_filter(openmc.DelayedGroupFilter)
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tally2_dg = other.contains_filter(openmc.DelayedGroupFilter)
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if sum([tally1_dg, tally2_dg]) == 1:
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return False
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@ -1032,7 +1031,7 @@ class Tally(object):
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if len(self.nuclides) > 0:
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nuclides = ''
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for nuclide in self.nuclides:
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if isinstance(nuclide, Nuclide):
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if isinstance(nuclide, openmc.Nuclide):
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nuclides += '{0} '.format(nuclide.name)
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else:
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nuclides += '{0} '.format(nuclide)
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@ -1070,7 +1069,7 @@ class Tally(object):
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Parameters
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----------
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filter_type : openmc.filter.FilterMeta
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filter_type : openmc.FilterMeta
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Type of the filter, e.g. MeshFilter
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Returns
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@ -1096,7 +1095,7 @@ class Tally(object):
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Parameters
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----------
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filter_type : openmc.filter.FilterMeta
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filter_type : openmc.FilterMeta
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Type of the filter, e.g. MeshFilter
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Returns
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@ -1122,7 +1121,7 @@ class Tally(object):
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# Also check to see if the desired filter is wrapped up in an
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# aggregate
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elif isinstance(test_filter, AggregateFilter):
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elif isinstance(test_filter, openmc.AggregateFilter):
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if isinstance(test_filter.aggregate_filter, filter_type):
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filter_found = test_filter
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break
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@ -1140,7 +1139,7 @@ class Tally(object):
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Parameters
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----------
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filter_type : openmc.filter.FilterMeta
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filter_type : openmc.FilterMeta
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Type of the filter, e.g. MeshFilter
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filter_bin : int or tuple
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The bin is an integer ID for 'material', 'surface', 'cell',
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@ -1191,7 +1190,7 @@ class Tally(object):
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for i, test_nuclide in enumerate(self.nuclides):
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# If the Summary was linked, then values are Nuclide objects
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if isinstance(test_nuclide, Nuclide):
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if isinstance(test_nuclide, openmc.Nuclide):
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if test_nuclide.name == nuclide:
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nuclide_index = i
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break
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@ -1250,7 +1249,7 @@ class Tally(object):
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Parameters
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----------
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filters : Iterable of openmc.filter.FilterMeta
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filters : Iterable of openmc.FilterMeta
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An iterable of filter types
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(e.g., [MeshFilter, EnergyFilter]; default is [])
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filter_bins : Iterable of tuple
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@ -1273,7 +1272,7 @@ class Tally(object):
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"""
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cv.check_type('filters', filters, Iterable, openmc.filter.FilterMeta)
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cv.check_type('filters', filters, Iterable, openmc.FilterMeta)
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cv.check_type('filter_bins', filter_bins, Iterable, tuple)
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# Determine the score indices from any of the requested scores
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@ -1295,21 +1294,20 @@ class Tally(object):
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# If not a user-requested Filter, get all bins
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if not user_filter:
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# Create list of 2- or 3-tuples tuples for mesh cell bins
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if isinstance(self_filter, openmc.filter.MeshFilter):
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if isinstance(self_filter, openmc.MeshFilter):
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dimension = self_filter.mesh.dimension
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xyz = [range(1, x+1) for x in dimension]
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bins = list(itertools.product(*xyz))
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# Create list of 2-tuples for energy boundary bins
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elif isinstance(self_filter, (openmc.filter.EnergyFilter,
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openmc.filter.EnergyoutFilter)):
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elif isinstance(self_filter, (openmc.EnergyFilter,
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openmc.EnergyoutFilter)):
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bins = []
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for k in range(self_filter.num_bins):
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bins.append((self_filter.bins[k], self_filter.bins[k+1]))
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# Create list of cell instance IDs for distribcell Filters
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elif isinstance(self_filter,
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openmc.filter.DistribcellFilter):
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elif isinstance(self_filter, openmc.DistribcellFilter):
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bins = np.arange(self_filter.num_bins)
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# Create list of IDs for bins for all other filter types
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@ -1392,7 +1390,7 @@ class Tally(object):
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"""
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for score in scores:
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if not isinstance(score, (basestring, CrossScore)):
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if not isinstance(score, (basestring, openmc.CrossScore)):
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msg = 'Unable to get score indices for score "{0}" in Tally ' \
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'ID="{1}" since it is not a string or CrossScore'\
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.format(score, self.id)
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@ -1424,7 +1422,7 @@ class Tally(object):
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scores : list of str
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A list of one or more score strings
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(e.g., ['absorption', 'nu-fission']; default is [])
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filters : Iterable of openmc.filter.FilterMeta
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filters : Iterable of openmc.FilterMeta
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An iterable of filter types
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(e.g., [MeshFilter, EnergyFilter]; default is [])
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filter_bins : list of Iterables
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@ -1568,9 +1566,9 @@ class Tally(object):
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column_name = 'nuclide'
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for nuclide in self.nuclides:
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if isinstance(nuclide, Nuclide):
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if isinstance(nuclide, openmc.Nuclide):
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nuclides.append(nuclide.name)
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elif isinstance(nuclide, AggregateNuclide):
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elif isinstance(nuclide, openmc.AggregateNuclide):
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nuclides.append(nuclide.name)
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column_name = '{0}(nuclide)'.format(nuclide.aggregate_op)
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else:
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@ -1587,9 +1585,9 @@ class Tally(object):
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column_name = 'score'
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for score in self.scores:
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if isinstance(score, (basestring, CrossScore)):
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if isinstance(score, (basestring, openmc.CrossScore)):
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scores.append(str(score))
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elif isinstance(score, AggregateScore):
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elif isinstance(score, openmc.AggregateScore):
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scores.append(score.name)
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column_name = '{0}(score)'.format(score.aggregate_op)
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@ -1979,7 +1977,8 @@ class Tally(object):
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else:
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all_filters = [self_copy.filters, other_copy.filters]
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for self_filter, other_filter in itertools.product(*all_filters):
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new_filter = CrossFilter(self_filter, other_filter, binary_op)
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new_filter = openmc.CrossFilter(self_filter, other_filter,
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binary_op)
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new_tally.filters.append(new_filter)
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# Add nuclides to the new tally
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@ -1990,7 +1989,7 @@ class Tally(object):
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all_nuclides = [self_copy.nuclides, other_copy.nuclides]
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for self_nuclide, other_nuclide in itertools.product(*all_nuclides):
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new_nuclide = \
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CrossNuclide(self_nuclide, other_nuclide, binary_op)
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openmc.CrossNuclide(self_nuclide, other_nuclide, binary_op)
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new_tally.nuclides.append(new_nuclide)
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# Add scores to the new tally
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@ -2000,7 +1999,8 @@ class Tally(object):
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else:
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all_scores = [self_copy.scores, other_copy.scores]
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for self_score, other_score in itertools.product(*all_scores):
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new_score = CrossScore(self_score, other_score, binary_op)
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new_score = openmc.CrossScore(self_score, other_score,
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binary_op)
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new_tally.scores.append(new_score)
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# Update the new tally's filter strides
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@ -2229,12 +2229,12 @@ class Tally(object):
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# Construct lists of tuples for the bins in each of the two filters
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filters = [type(filter1), type(filter2)]
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if isinstance(filter1, openmc.filter.DistribcellFilter):
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if isinstance(filter1, openmc.DistribcellFilter):
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filter1_bins = np.arange(filter1.num_bins)
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else:
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filter1_bins = [filter1.get_bin(i) for i in range(filter1.num_bins)]
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if isinstance(filter2, openmc.filter.DistribcellFilter):
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if isinstance(filter2, openmc.DistribcellFilter):
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filter2_bins = np.arange(filter2.num_bins)
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else:
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filter2_bins = [filter2.get_bin(i) for i in range(filter2.num_bins)]
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@ -2354,11 +2354,11 @@ class Tally(object):
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raise ValueError(msg)
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# Check that the scores are valid
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if not isinstance(score1, (basestring, CrossScore)):
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if not isinstance(score1, (basestring, openmc.CrossScore)):
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msg = 'Unable to swap score1 "{0}" in Tally ID="{1}" since it is ' \
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'not a string or CrossScore'.format(score1, self.id)
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raise ValueError(msg)
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elif not isinstance(score2, (basestring, CrossScore)):
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elif not isinstance(score2, (basestring, openmc.CrossScore)):
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msg = 'Unable to swap score2 "{0}" in Tally ID="{1}" since it is ' \
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'not a string or CrossScore'.format(score2, self.id)
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raise ValueError(msg)
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@ -2876,7 +2876,7 @@ class Tally(object):
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scores : list of str
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A list of one or more score strings
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(e.g., ['absorption', 'nu-fission']; default is [])
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filters : Iterable of openmc.filter.FilterMeta
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filters : Iterable of openmc.FilterMeta
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An iterable of filter types
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(e.g., [MeshFilter, EnergyFilter]; default is [])
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filter_bins : list of Iterables
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@ -2981,13 +2981,13 @@ class Tally(object):
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for filter_bin in filter_bins[i]:
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bin_index = find_filter.get_bin_index(filter_bin)
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if filter_type in [openmc.filter.EnergyFilter,
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openmc.filter.EnergyoutFilter]:
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if filter_type in [openmc.EnergyFilter,
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openmc.EnergyoutFilter]:
|
||||
bin_indices.extend([bin_index])
|
||||
bin_indices.extend([bin_index, bin_index+1])
|
||||
num_bins += 1
|
||||
elif filter_type in [openmc.filter.DistribcellFilter,
|
||||
openmc.filter.MeshFilter]:
|
||||
elif filter_type in [openmc.DistribcellFilter,
|
||||
openmc.MeshFilter]:
|
||||
bin_indices = [0]
|
||||
num_bins = find_filter.num_bins
|
||||
else:
|
||||
|
|
@ -3019,7 +3019,7 @@ class Tally(object):
|
|||
scores : list of str
|
||||
A list of one or more score strings to sum across
|
||||
(e.g., ['absorption', 'nu-fission']; default is [])
|
||||
filter_type : openmc.filter.FilterMeta
|
||||
filter_type : openmc.FilterMeta
|
||||
Type of the filter, e.g. MeshFilter
|
||||
filter_bins : Iterable of int or tuple
|
||||
A list of the filter bins corresponding to the filter_type parameter
|
||||
|
|
@ -3058,14 +3058,14 @@ class Tally(object):
|
|||
std_dev = self.get_reshaped_data(value='std_dev')
|
||||
|
||||
# Sum across any filter bins specified by the user
|
||||
if isinstance(filter_type, openmc.filter.FilterMeta):
|
||||
if isinstance(filter_type, openmc.FilterMeta):
|
||||
find_filter = self.find_filter(filter_type)
|
||||
|
||||
# If user did not specify filter bins, sum across all bins
|
||||
if len(filter_bins) == 0:
|
||||
bin_indices = np.arange(find_filter.num_bins)
|
||||
|
||||
if isinstance(find_filter, openmc.filter.DistribcellFilter):
|
||||
if isinstance(find_filter, openmc.DistribcellFilter):
|
||||
filter_bins = np.arange(find_filter.num_bins)
|
||||
else:
|
||||
num_bins = find_filter.num_bins
|
||||
|
|
@ -3088,8 +3088,8 @@ class Tally(object):
|
|||
|
||||
# Add AggregateFilter to the tally sum
|
||||
if not remove_filter:
|
||||
filter_sum = \
|
||||
AggregateFilter(self_filter, [tuple(filter_bins)], 'sum')
|
||||
filter_sum = openmc.AggregateFilter(self_filter,
|
||||
[tuple(filter_bins)], 'sum')
|
||||
tally_sum.filters.append(filter_sum)
|
||||
|
||||
# Add a copy of each filter not summed across to the tally sum
|
||||
|
|
@ -3111,7 +3111,7 @@ class Tally(object):
|
|||
std_dev = np.sqrt(std_dev)
|
||||
|
||||
# Add AggregateNuclide to the tally sum
|
||||
nuclide_sum = AggregateNuclide(nuclides, 'sum')
|
||||
nuclide_sum = openmc.AggregateNuclide(nuclides, 'sum')
|
||||
tally_sum.nuclides.append(nuclide_sum)
|
||||
|
||||
# Add a copy of this tally's nuclides to the tally sum
|
||||
|
|
@ -3129,7 +3129,7 @@ class Tally(object):
|
|||
std_dev = np.sqrt(std_dev)
|
||||
|
||||
# Add AggregateScore to the tally sum
|
||||
score_sum = AggregateScore(scores, 'sum')
|
||||
score_sum = openmc.AggregateScore(scores, 'sum')
|
||||
tally_sum.scores.append(score_sum)
|
||||
|
||||
# Add a copy of this tally's scores to the tally sum
|
||||
|
|
@ -3166,7 +3166,7 @@ class Tally(object):
|
|||
scores : list of str
|
||||
A list of one or more score strings to average across
|
||||
(e.g., ['absorption', 'nu-fission']; default is [])
|
||||
filter_type : openmc.filter.FilterMeta
|
||||
filter_type : openmc.FilterMeta
|
||||
Type of the filter, e.g. MeshFilter
|
||||
filter_bins : Iterable of int or tuple
|
||||
A list of the filter bins corresponding to the filter_type parameter
|
||||
|
|
@ -3205,14 +3205,14 @@ class Tally(object):
|
|||
std_dev = self.get_reshaped_data(value='std_dev')
|
||||
|
||||
# Average across any filter bins specified by the user
|
||||
if isinstance(filter_type, openmc.filter.FilterMeta):
|
||||
if isinstance(filter_type, openmc.FilterMeta):
|
||||
find_filter = self.find_filter(filter_type)
|
||||
|
||||
# If user did not specify filter bins, average across all bins
|
||||
if len(filter_bins) == 0:
|
||||
bin_indices = np.arange(find_filter.num_bins)
|
||||
|
||||
if isinstance(find_filter, openmc.filter.DistribcellFilter):
|
||||
if isinstance(find_filter, openmc.DistribcellFilter):
|
||||
filter_bins = np.arange(find_filter.num_bins)
|
||||
else:
|
||||
num_bins = find_filter.num_bins
|
||||
|
|
@ -3236,8 +3236,8 @@ class Tally(object):
|
|||
|
||||
# Add AggregateFilter to the tally avg
|
||||
if not remove_filter:
|
||||
filter_sum = \
|
||||
AggregateFilter(self_filter, [tuple(filter_bins)], 'avg')
|
||||
filter_sum = openmc.AggregateFilter(self_filter,
|
||||
[tuple(filter_bins)], 'avg')
|
||||
tally_avg.filters.append(filter_sum)
|
||||
|
||||
# Add a copy of each filter not averaged across to the tally avg
|
||||
|
|
@ -3260,7 +3260,7 @@ class Tally(object):
|
|||
std_dev = np.sqrt(std_dev)
|
||||
|
||||
# Add AggregateNuclide to the tally avg
|
||||
nuclide_avg = AggregateNuclide(nuclides, 'avg')
|
||||
nuclide_avg = openmc.AggregateNuclide(nuclides, 'avg')
|
||||
tally_avg.nuclides.append(nuclide_avg)
|
||||
|
||||
# Add a copy of this tally's nuclides to the tally avg
|
||||
|
|
@ -3279,7 +3279,7 @@ class Tally(object):
|
|||
std_dev = np.sqrt(std_dev)
|
||||
|
||||
# Add AggregateScore to the tally avg
|
||||
score_sum = AggregateScore(scores, 'avg')
|
||||
score_sum = openmc.AggregateScore(scores, 'avg')
|
||||
tally_avg.scores.append(score_sum)
|
||||
|
||||
# Add a copy of this tally's scores to the tally avg
|
||||
|
|
@ -3554,7 +3554,7 @@ class Tallies(cv.CheckedList):
|
|||
already_written = set()
|
||||
for tally in self:
|
||||
for f in tally.filters:
|
||||
if isinstance(f, openmc.filter.MeshFilter):
|
||||
if isinstance(f, openmc.MeshFilter):
|
||||
if f.mesh not in already_written:
|
||||
if len(f.mesh.name) > 0:
|
||||
self._tallies_file.append(ET.Comment(f.mesh.name))
|
||||
|
|
|
|||
Loading…
Add table
Add a link
Reference in a new issue