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Merge pull request #687 from samuelshaner/mesh-domain

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Mesh domain for MGXS
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Will Boyd 2016-07-31 10:59:47 -04:00 • committed by GitHub
commit 8902860eea
9 changed files with 435 additions and 146 deletions
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2
docs/source/pythonapi/examples/mgxs-part-iii.ipynb
View file

@ -574,7 +574,7 @@
"cell_type": "markdown",
"metadata": {},
"source": [
"Now we must specify the type of domain over which we would like the `Library` to compute multi-group cross sections. The domain type corresponds to the type of tally filter to be used in the tallies created to compute multi-group cross sections. At the present time, the `Library` supports `\"material,\"` `\"cell,\"` and `\"universe\"` domain types. We will use a `\"cell\"` domain type here to compute cross sections in each of the cells in the fuel assembly geometry.\n",
"Now we must specify the type of domain over which we would like the `Library` to compute multi-group cross sections. The domain type corresponds to the type of tally filter to be used in the tallies created to compute multi-group cross sections. At the present time, the `Library` supports `\"material\"`, `\"cell\"`, `\"universe\"`, and `\"mesh\"` domain types. We will use a `\"cell\"` domain type here to compute cross sections in each of the cells in the fuel assembly geometry.\n",
"\n",
"**Note:** By default, the `Library` class will instantiate `MGXS` objects for each and every domain (material, cell or universe) in the geometry of interest. However, one may specify a subset of these domains to the `Library.domains` property. In our case, we wish to compute multi-group cross sections in each and every cell since they will be needed in our downstream OpenMOC calculation on the identical combinatorial geometry mesh."
]

14
docs/source/pythonapi/examples/mgxs-part-iv.ipynb
View file

@ -519,9 +519,9 @@
"cell_type": "markdown",
"metadata": {},
"source": [
"Now we must specify the type of domain over which we would like the `Library` to compute multi-group cross sections. The domain type corresponds to the type of tally filter to be used in the tallies created to compute multi-group cross sections. At the present time, the `Library` supports \"material,\" \"cell,\" and \"universe\" domain types. In this simple example, we wish to compute multi-group cross sections only for each material andtherefore will use a \"material\" domain type.\n",
"Now we must specify the type of domain over which we would like the `Library` to compute multi-group cross sections. The domain type corresponds to the type of tally filter to be used in the tallies created to compute multi-group cross sections. At the present time, the `Library` supports \"material\" \"cell\", \"universe\", and \"mesh\" domain types. In this simple example, we wish to compute multi-group cross sections only for each material and therefore will use a \"material\" domain type.\n",
"\n",
"**Note:** By default, the `Library` class will instantiate `MGXS` objects for each and every domain (material, cell or universe) in the geometry of interest. However, one may specify a subset of these domains to the `Library.domains` property."
"**Note:** By default, the `Library` class will instantiate `MGXS` objects for each and every domain (material, cell, universe, or mesh) in the geometry of interest. However, one may specify a subset of these domains to the `Library.domains` property."
]
},
{
@ -1437,21 +1437,21 @@
],
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"display_name": "Python 3",
"display_name": "Python 2",
"language": "python",
"name": "python3"
"name": "python2"
},
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"codemirror_mode": {
"name": "ipython",
"version": 3
"version": 2
},
"file_extension": ".py",
"mimetype": "text/x-python",
"name": "python",
"nbconvert_exporter": "python",
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"nbformat": 4,

2
openmc/__init__.py
View file

@ -9,8 +9,8 @@ from openmc.plots import *
from openmc.settings import *
from openmc.surface import *
from openmc.universe import *
from openmc.mgxs_library import *
from openmc.mesh import *
from openmc.mgxs_library import *
from openmc.filter import *
from openmc.trigger import *
from openmc.tallies import *

14
openmc/mgxs/library.py
View file

@ -54,9 +54,9 @@ class Library(object):
If true, computes cross sections for each nuclide in each domain
mgxs_types : Iterable of str
The types of cross sections in the library (e.g., ['total', 'scatter'])
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
Domain type for spatial homogenization
domains : Iterable of openmc.Material, openmc.Cell or openmc.Universe
domains : Iterable of openmc.Material, openmc.Cell, openmc.Universe or openmc.Mesh
The spatial domain(s) for which MGXS in the Library are computed
correction : {'P0', None}
Apply the P0 correction to scattering matrices if set to 'P0'
@ -183,6 +183,8 @@ class Library(object):
return self.openmc_geometry.get_all_material_cells()
elif self.domain_type == 'universe':
return self.openmc_geometry.get_all_universes()
elif self.domain_type == 'mesh':
raise ValueError('Unable to get domains for Mesh domain type')
else:
raise ValueError('Unable to get domains without a domain type')
else:
@ -273,6 +275,12 @@ class Library(object):
elif self.domain_type == 'universe':
cv.check_iterable_type('domain', domains, openmc.Universe)
all_domains = self.openmc_geometry.get_all_universes()
elif self.domain_type == 'mesh':
cv.check_iterable_type('domain', domains, openmc.Mesh)
# The mesh and geometry are independent, so set all_domains
# to the input domains
all_domains = domains
else:
raise ValueError('Unable to set domains with domain '
'type "{}"'.format(self.domain_type))
@ -474,6 +482,8 @@ class Library(object):
cv.check_type('domain', domain, (openmc.Cell, Integral))
elif self.domain_type == 'universe':
cv.check_type('domain', domain, (openmc.Universe, Integral))
elif self.domain_type == 'mesh':
cv.check_type('domain', domain, (openmc.Mesh, Integral))
# Check that requested domain is included in library
if isinstance(domain, Integral):

333
openmc/mgxs/mgxs.py
View file

@ -7,6 +7,7 @@ import os
import sys
import copy
import abc
import itertools
import numpy as np
@ -14,7 +15,6 @@ import openmc
import openmc.checkvalue as cv
from openmc.mgxs import EnergyGroups
if sys.version_info[0] >= 3:
basestring = str
@ -41,17 +41,17 @@ MGXS_TYPES = ['total',
# Supported domain types
# TODO: Implement Mesh domains
DOMAIN_TYPES = ['cell',
'distribcell',
'universe',
'material']
'material',
'mesh']
# Supported domain classes
# TODO: Implement Mesh domains
_DOMAINS = (openmc.Cell,
openmc.Universe,
openmc.Material)
openmc.Material,
openmc.Mesh)
class MGXS(object):
@ -66,9 +66,9 @@ class MGXS(object):
Parameters
----------
domain : openmc.Material or openmc.Cell or openmc.Universe
domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
The domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
The domain type for spatial homogenization
energy_groups : openmc.mgxs.EnergyGroups
The energy group structure for energy condensation
@ -86,9 +86,9 @@ class MGXS(object):
Reaction type (e.g., 'total', 'nu-fission', etc.)
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
domain : Material or Cell or Universe
domain : Material or Cell or Universe or Mesh
Domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
Domain type for spatial homogenization
energy_groups : openmc.mgxs.EnergyGroups
Energy group structure for energy condensation
@ -115,9 +115,10 @@ class MGXS(object):
is None unless the multi-group cross section has been computed.
num_subdomains : int
The number of subdomains is unity for 'material', 'cell' and 'universe'
domain types. When the This is equal to the number of cell instances
domain types. This is equal to the number of cell instances
for 'distribcell' domain types (it is equal to unity prior to loading
tally data from a statepoint file).
tally data from a statepoint file) and the number of mesh cells for
'mesh' domain types.
num_nuclides : int
The number of nuclides for which the multi-group cross section is
being tracked. This is unity if the by_nuclide attribute is False.
@ -263,6 +264,10 @@ class MGXS(object):
# Create a domain Filter object
domain_filter = openmc.Filter(self.domain_type, self.domain.id)
# If a mesh domain, give the mesh to the domain filter
if self.domain_type == 'mesh':
domain_filter.mesh = self.domain
# Create each Tally needed to compute the multi group cross section
tally_metadata = zip(self.scores, self.tally_keys, self.filters)
for score, key, filters in tally_metadata:
@ -378,6 +383,8 @@ class MGXS(object):
self._domain_type = 'cell'
elif isinstance(domain, openmc.Universe):
self._domain_type = 'universe'
elif isinstance(domain, openmc.Mesh):
self._domain_type = 'mesh'
@domain_type.setter
def domain_type(self, domain_type):
@ -432,9 +439,9 @@ class MGXS(object):
----------
mgxs_type : {'total', 'transport', 'nu-transport', 'absorption', 'capture', 'fission', 'nu-fission', 'kappa-fission', 'scatter', 'nu-scatter', 'scatter matrix', 'nu-scatter matrix', 'multiplicity matrix', 'nu-fission matrix', 'chi', 'chi-prompt', 'inverse-velocity', 'prompt-nu-fission'}
The type of multi-group cross section object to return
domain : openmc.Material or openmc.Cell or openmc.Universe
domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
The domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
The domain type for spatial homogenization
energy_groups : openmc.mgxs.EnergyGroups
The energy group structure for energy condensation
@ -675,6 +682,8 @@ class MGXS(object):
self.domain = statepoint.summary.get_universe_by_id(self.domain.id)
elif self.domain_type == 'material':
self.domain = statepoint.summary.get_material_by_id(self.domain.id)
elif self.domain_type == 'mesh':
self.domain = statepoint.meshes[self.domain.id]
else:
msg = 'Unable to load data from a statepoint for domain type {0} ' \
'which is not yet supported'.format(self.domain_type)
@ -682,7 +691,11 @@ class MGXS(object):
# Use tally "slicing" to ensure that tallies correspond to our domain
# NOTE: This is important if tally merging was used
if self.domain_type != 'distribcell':
if self.domain_type == 'mesh':
filters = [self.domain_type]
xyz = [range(1, x+1) for x in self.domain.dimension]
filter_bins = [tuple(itertools.product(*xyz))]
elif self.domain_type != 'distribcell':
filters = [self.domain_type]
filter_bins = [(self.domain.id,)]
# Distribcell filters only accept single cell - neglect it when slicing
@ -756,12 +769,19 @@ class MGXS(object):
cv.check_value('value', value, ['mean', 'std_dev', 'rel_err'])
cv.check_value('xs_type', xs_type, ['macro', 'micro'])
# FIXME: Unable to get microscopic xs for mesh domain because the mesh
# cells do not know the nuclide densities in each mesh cell.
if self.domain_type == 'mesh' and xs_type == 'micro':
msg = 'Unable to get micro xs for mesh domain since the mesh ' \
'cells do not know the nuclide densities in each mesh cell.'
raise ValueError(msg)
filters = []
filter_bins = []
# Construct a collection of the domain filter bins
if not isinstance(subdomains, basestring):
cv.check_iterable_type('subdomains', subdomains, Integral, max_depth=2)
cv.check_iterable_type('subdomains', subdomains, Integral, max_depth=3)
for subdomain in subdomains:
filters.append(self.domain_type)
filter_bins.append((subdomain,))
@ -981,16 +1001,16 @@ class MGXS(object):
cv.check_iterable_type('energy_groups', groups, Integral)
# Build lists of filters and filter bins to slice
if len(groups) == 0:
filters = []
filter_bins = []
else:
filter_bins = []
filters = []
filter_bins = []
if len(groups) != 0:
energy_bins = []
for group in groups:
group_bounds = self.energy_groups.get_group_bounds(group)
filter_bins.append(group_bounds)
filter_bins = [tuple(filter_bins)]
filters = ['energy']
energy_bins.append(group_bounds)
filter_bins.append(tuple(energy_bins))
filters.append('energy')
# Clone this MGXS to initialize the sliced version
slice_xs = copy.deepcopy(self)
@ -1135,6 +1155,9 @@ class MGXS(object):
cv.check_iterable_type('subdomains', subdomains, Integral)
elif self.domain_type == 'distribcell':
subdomains = np.arange(self.num_subdomains, dtype=np.int)
elif self.domain_type == 'mesh':
xyz = [range(1, x+1) for x in self.domain.dimension]
subdomains = list(itertools.product(*xyz))
else:
subdomains = [self.domain.id]
@ -1168,7 +1191,7 @@ class MGXS(object):
# Loop over all subdomains
for subdomain in subdomains:
if self.domain_type == 'distribcell':
if self.domain_type == 'distribcell' or self.domain_type == 'mesh':
string += '{0: <16}=\t{1}\n'.format('\tSubdomain', subdomain)
# Loop over all Nuclides
@ -1270,6 +1293,9 @@ class MGXS(object):
elif self.domain_type == 'avg(distribcell)':
domain_filter = self.xs_tally.find_filter('avg(distribcell)')
subdomains = domain_filter.bins
elif self.domain_type == 'mesh':
xyz = [range(1, x+1) for x in self.domain.dimension]
subdomains = list(itertools.product(*xyz))
else:
subdomains = [self.domain.id]
@ -1376,15 +1402,14 @@ class MGXS(object):
# Get a Pandas DataFrame for the data
df = self.get_pandas_dataframe(groups=groups, xs_type=xs_type)
# Capitalize column label strings
df.columns = df.columns.astype(str)
df.columns = map(str.title, df.columns)
# Export the data using Pandas IO API
if format == 'csv':
df.to_csv(filename + '.csv', index=False)
elif format == 'excel':
df.to_excel(filename + '.xls', index=False)
if self.domain_type == 'mesh':
df.to_excel(filename + '.xls')
else:
df.to_excel(filename + '.xls', index=False)
elif format == 'pickle':
df.to_pickle(filename + '.pkl')
elif format == 'latex':
@ -1465,7 +1490,10 @@ class MGXS(object):
distribcell_paths=distribcell_paths)
# Remove nuclide column since it is homogeneous and redundant
df.drop('nuclide', axis=1, inplace=True)
if self.domain_type == 'mesh':
df.drop('nuclide', axis=1, level=0, inplace=True)
else:
df.drop('nuclide', axis=1, inplace=True)
# If the user requested a specific set of nuclides
elif self.by_nuclide and nuclides != 'all':
@ -1479,7 +1507,10 @@ class MGXS(object):
distribcell_paths=distribcell_paths)
# Remove the score column since it is homogeneous and redundant
df = df.drop('score', axis=1)
if self.domain_type == 'mesh':
df = df.drop('score', axis=1, level=0)
else:
df = df.drop('score', axis=1)
# Override energy groups bounds with indices
all_groups = np.arange(self.num_groups, 0, -1, dtype=np.int)
@ -1535,7 +1566,12 @@ class MGXS(object):
# Sort the dataframe by domain type id (e.g., distribcell id) and
# energy groups such that data is from fast to thermal
df.sort_values(by=[self.domain_type] + columns, inplace=True)
if self.domain_type == 'mesh':
mesh_str = 'mesh {0}'.format(self.domain.id)
df.sort_values(by=[(mesh_str, 'x'), (mesh_str, 'y'), \
(mesh_str, 'z')] + columns, inplace=True)
else:
df.sort_values(by=[self.domain_type] + columns, inplace=True)
return df
def get_units(self, xs_type='macro'):
@ -1574,9 +1610,9 @@ class MatrixMGXS(MGXS):
Parameters
----------
domain : openmc.Material or openmc.Cell or openmc.Universe
domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
The domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
The domain type for spatial homogenization
energy_groups : openmc.mgxs.EnergyGroups
The energy group structure for energy condensation
@ -1594,9 +1630,9 @@ class MatrixMGXS(MGXS):
Reaction type (e.g., 'total', 'nu-fission', etc.)
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
domain : Material or Cell or Universe
domain : Material or Cell or Universe or Mesh
Domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
Domain type for spatial homogenization
energy_groups : openmc.mgxs.EnergyGroups
Energy group structure for energy condensation
@ -1623,9 +1659,10 @@ class MatrixMGXS(MGXS):
is None unless the multi-group cross section has been computed.
num_subdomains : int
The number of subdomains is unity for 'material', 'cell' and 'universe'
domain types. When the This is equal to the number of cell instances
domain types. This is equal to the number of cell instances
for 'distribcell' domain types (it is equal to unity prior to loading
tally data from a statepoint file).
tally data from a statepoint file) and the number of mesh cells for
'mesh' domain types.
num_nuclides : int
The number of nuclides for which the multi-group cross section is
being tracked. This is unity if the by_nuclide attribute is False.
@ -1716,13 +1753,20 @@ class MatrixMGXS(MGXS):
cv.check_value('value', value, ['mean', 'std_dev', 'rel_err'])
cv.check_value('xs_type', xs_type, ['macro', 'micro'])
# FIXME: Unable to get microscopic xs for mesh domain because the mesh
# cells do not know the nuclide densities in each mesh cell.
if self.domain_type == 'mesh' and xs_type == 'micro':
msg = 'Unable to get micro xs for mesh domain since the mesh ' \
'cells do not know the nuclide densities in each mesh cell.'
raise ValueError(msg)
filters = []
filter_bins = []
# Construct a collection of the domain filter bins
if not isinstance(subdomains, basestring):
cv.check_iterable_type('subdomains', subdomains, Integral,
max_depth=2)
max_depth=3)
for subdomain in subdomains:
filters.append(self.domain_type)
filter_bins.append((subdomain,))
@ -1884,6 +1928,9 @@ class MatrixMGXS(MGXS):
cv.check_iterable_type('subdomains', subdomains, Integral)
elif self.domain_type == 'distribcell':
subdomains = np.arange(self.num_subdomains, dtype=np.int)
elif self.domain_type == 'mesh':
xyz = [range(1, x+1) for x in self.domain.dimension]
subdomains = list(itertools.product(*xyz))
else:
subdomains = [self.domain.id]
@ -1992,9 +2039,9 @@ class TotalXS(MGXS):
Parameters
----------
domain : openmc.Material or openmc.Cell or openmc.Universe
domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
The domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
The domain type for spatial homogenization
groups : openmc.mgxs.EnergyGroups
The energy group structure for energy condensation
@ -2012,9 +2059,9 @@ class TotalXS(MGXS):
Reaction type (e.g., 'total', 'nu-fission', etc.)
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
domain : Material or Cell or Universe
domain : Material or Cell or Universe or Mesh
Domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
Domain type for spatial homogenization
energy_groups : openmc.mgxs.EnergyGroups
Energy group structure for energy condensation
@ -2043,7 +2090,7 @@ class TotalXS(MGXS):
is None unless the multi-group cross section has been computed.
num_subdomains : int
The number of subdomains is unity for 'material', 'cell' and 'universe'
domain types. When the This is equal to the number of cell instances
domain types. This is equal to the number of cell instances
for 'distribcell' domain types (it is equal to unity prior to loading
tally data from a statepoint file).
num_nuclides : int
@ -2110,9 +2157,9 @@ class TransportXS(MGXS):
Parameters
----------
domain : openmc.Material or openmc.Cell or openmc.Universe
domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
The domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
The domain type for spatial homogenization
groups : openmc.mgxs.EnergyGroups
The energy group structure for energy condensation
@ -2130,9 +2177,9 @@ class TransportXS(MGXS):
Reaction type (e.g., 'total', 'nu-fission', etc.)
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
domain : Material or Cell or Universe
domain : Material or Cell or Universe or Mesh
Domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
Domain type for spatial homogenization
energy_groups : openmc.mgxs.EnergyGroups
Energy group structure for energy condensation
@ -2161,7 +2208,7 @@ class TransportXS(MGXS):
is None unless the multi-group cross section has been computed.
num_subdomains : int
The number of subdomains is unity for 'material', 'cell' and 'universe'
domain types. When the This is equal to the number of cell instances
domain types. This is equal to the number of cell instances
for 'distribcell' domain types (it is equal to unity prior to loading
tally data from a statepoint file).
num_nuclides : int
@ -2240,9 +2287,9 @@ class NuTransportXS(TransportXS):
Parameters
----------
domain : openmc.Material or openmc.Cell or openmc.Universe
domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
The domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
The domain type for spatial homogenization
groups : openmc.mgxs.EnergyGroups
The energy group structure for energy condensation
@ -2260,9 +2307,9 @@ class NuTransportXS(TransportXS):
Reaction type (e.g., 'total', 'nu-fission', etc.)
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
domain : Material or Cell or Universe
domain : Material or Cell or Universe or Mesh
Domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
Domain type for spatial homogenization
energy_groups : openmc.mgxs.EnergyGroups
Energy group structure for energy condensation
@ -2291,7 +2338,7 @@ class NuTransportXS(TransportXS):
is None unless the multi-group cross section has been computed.
num_subdomains : int
The number of subdomains is unity for 'material', 'cell' and 'universe'
domain types. When the This is equal to the number of cell instances
domain types. This is equal to the number of cell instances
for 'distribcell' domain types (it is equal to unity prior to loading
tally data from a statepoint file).
num_nuclides : int
@ -2361,9 +2408,9 @@ class AbsorptionXS(MGXS):
Parameters
----------
domain : openmc.Material or openmc.Cell or openmc.Universe
domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
The domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
The domain type for spatial homogenization
groups : openmc.mgxs.EnergyGroups
The energy group structure for energy condensation
@ -2381,9 +2428,9 @@ class AbsorptionXS(MGXS):
Reaction type (e.g., 'total', 'nu-fission', etc.)
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
domain : Material or Cell or Universe
domain : Material or Cell or Universe or Mesh
Domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
Domain type for spatial homogenization
energy_groups : openmc.mgxs.EnergyGroups
Energy group structure for energy condensation
@ -2412,7 +2459,7 @@ class AbsorptionXS(MGXS):
is None unless the multi-group cross section has been computed.
num_subdomains : int
The number of subdomains is unity for 'material', 'cell' and 'universe'
domain types. When the This is equal to the number of cell instances
domain types. This is equal to the number of cell instances
for 'distribcell' domain types (it is equal to unity prior to loading
tally data from a statepoint file).
num_nuclides : int
@ -2477,9 +2524,9 @@ class CaptureXS(MGXS):
Parameters
----------
domain : openmc.Material or openmc.Cell or openmc.Universe
domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
The domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
The domain type for spatial homogenization
groups : openmc.mgxs.EnergyGroups
The energy group structure for energy condensation
@ -2497,9 +2544,9 @@ class CaptureXS(MGXS):
Reaction type (e.g., 'total', 'nu-fission', etc.)
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
domain : Material or Cell or Universe
domain : Material or Cell or Universe or Mesh
Domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
Domain type for spatial homogenization
energy_groups : openmc.mgxs.EnergyGroups
Energy group structure for energy condensation
@ -2528,7 +2575,7 @@ class CaptureXS(MGXS):
is None unless the multi-group cross section has been computed.
num_subdomains : int
The number of subdomains is unity for 'material', 'cell' and 'universe'
domain types. When the This is equal to the number of cell instances
domain types. This is equal to the number of cell instances
for 'distribcell' domain types (it is equal to unity prior to loading
tally data from a statepoint file).
num_nuclides : int
@ -2599,9 +2646,9 @@ class FissionXS(MGXS):
Parameters
----------
domain : openmc.Material or openmc.Cell or openmc.Universe
domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
The domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
The domain type for spatial homogenization
groups : openmc.mgxs.EnergyGroups
The energy group structure for energy condensation
@ -2619,9 +2666,9 @@ class FissionXS(MGXS):
Reaction type (e.g., 'total', 'nu-fission', etc.)
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
domain : Material or Cell or Universe
domain : Material or Cell or Universe or Mesh
Domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
Domain type for spatial homogenization
energy_groups : openmc.mgxs.EnergyGroups
Energy group structure for energy condensation
@ -2650,7 +2697,7 @@ class FissionXS(MGXS):
is None unless the multi-group cross section has been computed.
num_subdomains : int
The number of subdomains is unity for 'material', 'cell' and 'universe'
domain types. When the This is equal to the number of cell instances
domain types. This is equal to the number of cell instances
for 'distribcell' domain types (it is equal to unity prior to loading
tally data from a statepoint file).
num_nuclides : int
@ -2710,9 +2757,9 @@ class NuFissionXS(MGXS):
Parameters
----------
domain : openmc.Material or openmc.Cell or openmc.Universe
domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
The domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
The domain type for spatial homogenization
groups : openmc.mgxs.EnergyGroups
The energy group structure for energy condensation
@ -2730,9 +2777,9 @@ class NuFissionXS(MGXS):
Reaction type (e.g., 'total', 'nu-fission', etc.)
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
domain : Material or Cell or Universe
domain : Material or Cell or Universe or Mesh
Domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
Domain type for spatial homogenization
energy_groups : openmc.mgxs.EnergyGroups
Energy group structure for energy condensation
@ -2761,7 +2808,7 @@ class NuFissionXS(MGXS):
is None unless the multi-group cross section has been computed.
num_subdomains : int
The number of subdomains is unity for 'material', 'cell' and 'universe'
domain types. When the This is equal to the number of cell instances
domain types. This is equal to the number of cell instances
for 'distribcell' domain types (it is equal to unity prior to loading
tally data from a statepoint file).
num_nuclides : int
@ -2826,9 +2873,9 @@ class KappaFissionXS(MGXS):
Parameters
----------
domain : openmc.Material or openmc.Cell or openmc.Universe
domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
The domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
The domain type for spatial homogenization
groups : openmc.mgxs.EnergyGroups
The energy group structure for energy condensation
@ -2846,9 +2893,9 @@ class KappaFissionXS(MGXS):
Reaction type (e.g., 'total', 'nu-fission', etc.)
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
domain : Material or Cell or Universe
domain : Material or Cell or Universe or Mesh
Domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
Domain type for spatial homogenization
energy_groups : openmc.mgxs.EnergyGroups
Energy group structure for energy condensation
@ -2877,7 +2924,7 @@ class KappaFissionXS(MGXS):
is None unless the multi-group cross section has been computed.
num_subdomains : int
The number of subdomains is unity for 'material', 'cell' and 'universe'
domain types. When the This is equal to the number of cell instances
domain types. This is equal to the number of cell instances
for 'distribcell' domain types (it is equal to unity prior to loading
tally data from a statepoint file).
num_nuclides : int
@ -2939,9 +2986,9 @@ class ScatterXS(MGXS):
Parameters
----------
domain : openmc.Material or openmc.Cell or openmc.Universe
domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
The domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
The domain type for spatial homogenization
groups : openmc.mgxs.EnergyGroups
The energy group structure for energy condensation
@ -2959,9 +3006,9 @@ class ScatterXS(MGXS):
Reaction type (e.g., 'total', 'nu-fission', etc.)
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
domain : Material or Cell or Universe
domain : Material or Cell or Universe or Mesh
Domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
Domain type for spatial homogenization
energy_groups : openmc.mgxs.EnergyGroups
Energy group structure for energy condensation
@ -2990,7 +3037,7 @@ class ScatterXS(MGXS):
is None unless the multi-group cross section has been computed.
num_subdomains : int
The number of subdomains is unity for 'material', 'cell' and 'universe'
domain types. When the This is equal to the number of cell instances
domain types. This is equal to the number of cell instances
for 'distribcell' domain types (it is equal to unity prior to loading
tally data from a statepoint file).
num_nuclides : int
@ -3054,9 +3101,9 @@ class NuScatterXS(MGXS):
Parameters
----------
domain : openmc.Material or openmc.Cell or openmc.Universe
domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
The domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
The domain type for spatial homogenization
groups : openmc.mgxs.EnergyGroups
The energy group structure for energy condensation
@ -3074,9 +3121,9 @@ class NuScatterXS(MGXS):
Reaction type (e.g., 'total', 'nu-fission', etc.)
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
domain : Material or Cell or Universe
domain : Material or Cell or Universe or Mesh
Domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
Domain type for spatial homogenization
energy_groups : openmc.mgxs.EnergyGroups
Energy group structure for energy condensation
@ -3105,7 +3152,7 @@ class NuScatterXS(MGXS):
is None unless the multi-group cross section has been computed.
num_subdomains : int
The number of subdomains is unity for 'material', 'cell' and 'universe'
domain types. When the This is equal to the number of cell instances
domain types. This is equal to the number of cell instances
for 'distribcell' domain types (it is equal to unity prior to loading
tally data from a statepoint file).
num_nuclides : int
@ -3184,9 +3231,9 @@ class ScatterMatrixXS(MatrixMGXS):
Parameters
----------
domain : openmc.Material or openmc.Cell or openmc.Universe
domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
The domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
The domain type for spatial homogenization
groups : openmc.mgxs.EnergyGroups
The energy group structure for energy condensation
@ -3208,9 +3255,9 @@ class ScatterMatrixXS(MatrixMGXS):
Reaction type (e.g., 'total', 'nu-fission', etc.)
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
domain : Material or Cell or Universe
domain : Material or Cell or Universe or Mesh
Domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
Domain type for spatial homogenization
energy_groups : openmc.mgxs.EnergyGroups
Energy group structure for energy condensation
@ -3239,7 +3286,7 @@ class ScatterMatrixXS(MatrixMGXS):
is None unless the multi-group cross section has been computed.
num_subdomains : int
The number of subdomains is unity for 'material', 'cell' and 'universe'
domain types. When the This is equal to the number of cell instances
domain types. This is equal to the number of cell instances
for 'distribcell' domain types (it is equal to unity prior to loading
tally data from a statepoint file).
num_nuclides : int
@ -3531,12 +3578,19 @@ class ScatterMatrixXS(MatrixMGXS):
cv.check_value('value', value, ['mean', 'std_dev', 'rel_err'])
cv.check_value('xs_type', xs_type, ['macro', 'micro'])
# FIXME: Unable to get microscopic xs for mesh domain because the mesh
# cells do not know the nuclide densities in each mesh cell.
if self.domain_type == 'mesh' and xs_type == 'micro':
msg = 'Unable to get micro xs for mesh domain since the mesh ' \
'cells do not know the nuclide densities in each mesh cell.'
raise ValueError(msg)
filters = []
filter_bins = []
# Construct a collection of the domain filter bins
if not isinstance(subdomains, basestring):
cv.check_iterable_type('subdomains', subdomains, Integral, max_depth=2)
cv.check_iterable_type('subdomains', subdomains, Integral, max_depth=3)
for subdomain in subdomains:
filters.append(self.domain_type)
filter_bins.append((subdomain,))
@ -3681,9 +3735,12 @@ class ScatterMatrixXS(MatrixMGXS):
df['moment'] = moments
# Place the moment column before the mean column
mean_index = df.columns.get_loc('mean')
columns = df.columns.tolist()
df = df[columns[:mean_index] + ['moment'] + columns[mean_index:-1]]
mean_index = [i for i, s in enumerate(columns) if 'mean' in s][0]
if self.domain_type == 'mesh':
df = df[columns[:mean_index] + [('moment', '')] + columns[mean_index:-1]]
else:
df = df[columns[:mean_index] + ['moment'] + columns[mean_index:-1]]
# Select rows corresponding to requested scattering moment
if moment != 'all':
@ -3723,6 +3780,9 @@ class ScatterMatrixXS(MatrixMGXS):
cv.check_iterable_type('subdomains', subdomains, Integral)
elif self.domain_type == 'distribcell':
subdomains = np.arange(self.num_subdomains, dtype=np.int)
elif self.domain_type == 'mesh':
xyz = [range(1, x+1) for x in self.domain.dimension]
subdomains = list(itertools.product(*xyz))
else:
subdomains = [self.domain.id]
@ -3832,9 +3892,9 @@ class NuScatterMatrixXS(ScatterMatrixXS):
Parameters
----------
domain : openmc.Material or openmc.Cell or openmc.Universe
domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
The domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
The domain type for spatial homogenization
groups : openmc.mgxs.EnergyGroups
The energy group structure for energy condensation
@ -3856,9 +3916,9 @@ class NuScatterMatrixXS(ScatterMatrixXS):
Reaction type (e.g., 'total', 'nu-fission', etc.)
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
domain : Material or Cell or Universe
domain : Material or Cell or Universe or Mesh
Domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
Domain type for spatial homogenization
energy_groups : openmc.mgxs.EnergyGroups
Energy group structure for energy condensation
@ -3887,7 +3947,7 @@ class NuScatterMatrixXS(ScatterMatrixXS):
is None unless the multi-group cross section has been computed.
num_subdomains : int
The number of subdomains is unity for 'material', 'cell' and 'universe'
domain types. When the This is equal to the number of cell instances
domain types. This is equal to the number of cell instances
for 'distribcell' domain types (it is equal to unity prior to loading
tally data from a statepoint file).
num_nuclides : int
@ -3958,9 +4018,9 @@ class MultiplicityMatrixXS(MatrixMGXS):
Parameters
----------
domain : openmc.Material or openmc.Cell or openmc.Universe
domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
The domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
The domain type for spatial homogenization
groups : openmc.mgxs.EnergyGroups
The energy group structure for energy condensation
@ -3978,9 +4038,9 @@ class MultiplicityMatrixXS(MatrixMGXS):
Reaction type (e.g., 'total', 'nu-fission', etc.)
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
domain : Material or Cell or Universe
domain : Material or Cell or Universe or Mesh
Domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
Domain type for spatial homogenization
energy_groups : openmc.mgxs.EnergyGroups
Energy group structure for energy condensation
@ -4009,7 +4069,7 @@ class MultiplicityMatrixXS(MatrixMGXS):
is None unless the multi-group cross section has been computed.
num_subdomains : int
The number of subdomains is unity for 'material', 'cell' and 'universe'
domain types. When the This is equal to the number of cell instances
domain types. This is equal to the number of cell instances
for 'distribcell' domain types (it is equal to unity prior to loading
tally data from a statepoint file).
num_nuclides : int
@ -4105,9 +4165,9 @@ class NuFissionMatrixXS(MatrixMGXS):
Parameters
----------
domain : openmc.Material or openmc.Cell or openmc.Universe
domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
The domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
The domain type for spatial homogenization
groups : openmc.mgxs.EnergyGroups
The energy group structure for energy condensation
@ -4125,9 +4185,9 @@ class NuFissionMatrixXS(MatrixMGXS):
Reaction type (e.g., 'total', 'nu-fission', etc.)
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
domain : Material or Cell or Universe
domain : Material or Cell or Universe or Mesh
Domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
Domain type for spatial homogenization
energy_groups : openmc.mgxs.EnergyGroups
Energy group structure for energy condensation
@ -4156,7 +4216,7 @@ class NuFissionMatrixXS(MatrixMGXS):
is None unless the multi-group cross section has been computed.
num_subdomains : int
The number of subdomains is unity for 'material', 'cell' and 'universe'
domain types. When the This is equal to the number of cell instances
domain types. This is equal to the number of cell instances
for 'distribcell' domain types (it is equal to unity prior to loading
tally data from a statepoint file).
num_nuclides : int
@ -4220,9 +4280,9 @@ class Chi(MGXS):
Parameters
----------
domain : openmc.Material or openmc.Cell or openmc.Universe
domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
The domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
The domain type for spatial homogenization
groups : openmc.mgxs.EnergyGroups
The energy group structure for energy condensation
@ -4240,9 +4300,9 @@ class Chi(MGXS):
Reaction type (e.g., 'total', 'nu-fission', etc.)
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
domain : Material or Cell or Universe
domain : Material or Cell or Universe or Mesh
Domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
Domain type for spatial homogenization
energy_groups : openmc.mgxs.EnergyGroups
Energy group structure for energy condensation
@ -4271,7 +4331,7 @@ class Chi(MGXS):
is None unless the multi-group cross section has been computed.
num_subdomains : int
The number of subdomains is unity for 'material', 'cell' and 'universe'
domain types. When the This is equal to the number of cell instances
domain types. This is equal to the number of cell instances
for 'distribcell' domain types (it is equal to unity prior to loading
tally data from a statepoint file).
num_nuclides : int
@ -4499,12 +4559,19 @@ class Chi(MGXS):
cv.check_value('value', value, ['mean', 'std_dev', 'rel_err'])
cv.check_value('xs_type', xs_type, ['macro', 'micro'])
# FIXME: Unable to get microscopic xs for mesh domain because the mesh
# cells do not know the nuclide densities in each mesh cell.
if self.domain_type == 'mesh' and xs_type == 'micro':
msg = 'Unable to get micro xs for mesh domain since the mesh ' \
'cells do not know the nuclide densities in each mesh cell.'
raise ValueError(msg)
filters = []
filter_bins = []
# Construct a collection of the domain filter bins
if not isinstance(subdomains, basestring):
cv.check_iterable_type('subdomains', subdomains, Integral, max_depth=2)
cv.check_iterable_type('subdomains', subdomains, Integral, max_depth=3)
for subdomain in subdomains:
filters.append(self.domain_type)
filter_bins.append((subdomain,))
@ -4703,9 +4770,9 @@ class ChiPrompt(Chi):
Parameters
----------
domain : openmc.Material or openmc.Cell or openmc.Universe
domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
The domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
The domain type for spatial homogenization
groups : openmc.mgxs.EnergyGroups
The energy group structure for energy condensation
@ -4723,9 +4790,9 @@ class ChiPrompt(Chi):
Reaction type (e.g., 'total', 'nu-fission', etc.)
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
domain : Material or Cell or Universe
domain : Material or Cell or Universe or Mesh
Domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
Domain type for spatial homogenization
energy_groups : openmc.mgxs.EnergyGroups
Energy group structure for energy condensation
@ -4754,7 +4821,7 @@ class ChiPrompt(Chi):
is None unless the multi-group cross section has been computed.
num_subdomains : int
The number of subdomains is unity for 'material', 'cell' and 'universe'
domain types. When the This is equal to the number of cell instances
domain types. This is equal to the number of cell instances
for 'distribcell' domain types (it is equal to unity prior to loading
tally data from a statepoint file).
num_nuclides : int
@ -4818,9 +4885,9 @@ class InverseVelocity(MGXS):
Parameters
----------
domain : openmc.Material or openmc.Cell or openmc.Universe
domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
The domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
The domain type for spatial homogenization
groups : openmc.mgxs.EnergyGroups
The energy group structure for energy condensation
@ -4838,9 +4905,9 @@ class InverseVelocity(MGXS):
Reaction type (e.g., 'total', 'nu-fission', etc.)
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
domain : Material or Cell or Universe
domain : Material or Cell or Universe or Mesh
Domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
Domain type for spatial homogenization
energy_groups : openmc.mgxs.EnergyGroups
Energy group structure for energy condensation
@ -4917,8 +4984,6 @@ class InverseVelocity(MGXS):
"""
cv.check_value('xs_type', xs_type, ['macro', 'micro'])
if xs_type == 'macro':
return 'second/cm'
else:
@ -4954,9 +5019,9 @@ class PromptNuFissionXS(MGXS):
Parameters
----------
domain : openmc.Material or openmc.Cell or openmc.Universe
domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
The domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
The domain type for spatial homogenization
groups : openmc.mgxs.EnergyGroups
The energy group structure for energy condensation
@ -4974,9 +5039,9 @@ class PromptNuFissionXS(MGXS):
Reaction type (e.g., 'total', 'nu-fission', etc.)
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
domain : Material or Cell or Universe
domain : Material or Cell or Universe or Mesh
Domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe'}
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
Domain type for spatial homogenization
energy_groups : openmc.mgxs.EnergyGroups
Energy group structure for energy condensation
@ -5005,7 +5070,7 @@ class PromptNuFissionXS(MGXS):
is None unless the multi-group cross section has been computed.
num_subdomains : int
The number of subdomains is unity for 'material', 'cell' and 'universe'
domain types. When the This is equal to the number of cell instances
domain types. This is equal to the number of cell instances
for 'distribcell' domain types (it is equal to unity prior to loading
tally data from a statepoint file).
num_nuclides : int

2
openmc/tallies.py
View file

@ -1293,7 +1293,7 @@ class Tally(object):
# Create list of 2- or 3-tuples tuples for mesh cell bins
if self_filter.type == 'mesh':
dimension = self_filter.mesh.dimension
xyz = map(lambda x: np.arange(1, x+1), dimension)
xyz = [range(1, x+1) for x in dimension]
bins = list(itertools.product(*xyz))
# Create list of 2-tuples for energy boundary bins

1
tests/test_mgxs_library_mesh/inputs_true.dat Normal file
View file

@ -0,0 +1 @@
a4cd030bea212e45fdb159e75a7fb3d1947e9bf3d0384ac5d37a72298d67dcfdd1b9eb5c6af8ac6e5983bd5b47de9c17a2ea472b467b7222a4909ee070bf1ca3

132
tests/test_mgxs_library_mesh/results_true.dat Normal file
View file

@ -0,0 +1,132 @@
mesh 1 group in nuclide mean std. dev.
x y z
0 1 1 1 1 total 0.640786 0.044177
1 1 2 1 1 total 0.660597 0.128423
2 2 1 1 1 total 0.615276 0.104046
3 2 2 1 1 total 0.646999 0.186709
mesh 1 group in nuclide mean std. dev.
x y z
0 1 1 1 1 total 0.36665 0.048814
1 1 2 1 1 total 0.40784 0.096486
2 2 1 1 1 total 0.36356 0.074111
3 2 2 1 1 total 0.41456 0.160443
mesh 1 group in nuclide mean std. dev.
x y z
0 1 1 1 1 total 0.366650 0.048814
1 1 2 1 1 total 0.407840 0.096486
2 2 1 1 1 total 0.363560 0.074111
3 2 2 1 1 total 0.414593 0.160436
mesh 1 group in nuclide mean std. dev.
x y z
0 1 1 1 1 total 0.025749 0.002863
1 1 2 1 1 total 0.028400 0.005275
2 2 1 1 1 total 0.022988 0.004099
3 2 2 1 1 total 0.027589 0.010350
mesh 1 group in nuclide mean std. dev.
x y z
0 1 1 1 1 total 0.015861 0.002876
1 1 2 1 1 total 0.017280 0.004371
2 2 1 1 1 total 0.014403 0.003542
3 2 2 1 1 total 0.018061 0.010110
mesh 1 group in nuclide mean std. dev.
x y z
0 1 1 1 1 total 0.009888 0.001077
1 1 2 1 1 total 0.011121 0.002456
2 2 1 1 1 total 0.008585 0.001552
3 2 2 1 1 total 0.009527 0.003659
mesh 1 group in nuclide mean std. dev.
x y z
0 1 1 1 1 total 0.026065 0.002907
1 1 2 1 1 total 0.029084 0.006430
2 2 1 1 1 total 0.022596 0.004062
3 2 2 1 1 total 0.025066 0.009687
mesh 1 group in nuclide mean std. dev.
x y z
0 1 1 1 1 total 1.938476 0.211550
1 1 2 1 1 total 2.177360 0.480780
2 2 1 1 1 total 1.682799 0.303764
3 2 2 1 1 total 1.864890 0.715661
mesh 1 group in nuclide mean std. dev.
x y z
0 1 1 1 1 total 0.615037 0.041754
1 1 2 1 1 total 0.632196 0.123878
2 2 1 1 1 total 0.592288 0.100439
3 2 2 1 1 total 0.619410 0.177190
mesh 1 group in nuclide mean std. dev.
x y z
0 1 1 1 1 total 0.584014 0.054315
1 1 2 1 1 total 0.622514 0.111323
2 2 1 1 1 total 0.587256 0.084833
3 2 2 1 1 total 0.613792 0.168612
mesh 1 group in group out nuclide moment mean std. dev.
x y z
0 1 1 1 1 1 total P0 0.584014 0.054315
1 1 1 1 1 1 total P1 0.243427 0.025488
2 1 1 1 1 1 total P2 0.089236 0.007357
3 1 1 1 1 1 total P3 0.008994 0.005768
4 1 2 1 1 1 total P0 0.622514 0.111323
5 1 2 1 1 1 total P1 0.239376 0.042594
6 1 2 1 1 1 total P2 0.088386 0.017200
7 1 2 1 1 1 total P3 -0.001243 0.005639
8 2 1 1 1 1 total P0 0.587256 0.084833
9 2 1 1 1 1 total P1 0.245120 0.041033
10 2 1 1 1 1 total P2 0.086784 0.016255
11 2 1 1 1 1 total P3 0.008660 0.004755
12 2 2 1 1 1 total P0 0.612950 0.167940
13 2 2 1 1 1 total P1 0.226176 0.061882
14 2 2 1 1 1 total P2 0.086593 0.026126
15 2 2 1 1 1 total P3 0.009672 0.011995
mesh 1 group in group out nuclide moment mean std. dev.
x y z
0 1 1 1 1 1 total P0 0.584014 0.054315
1 1 1 1 1 1 total P1 0.243427 0.025488
2 1 1 1 1 1 total P2 0.089236 0.007357
3 1 1 1 1 1 total P3 0.008994 0.005768
4 1 2 1 1 1 total P0 0.622514 0.111323
5 1 2 1 1 1 total P1 0.239376 0.042594
6 1 2 1 1 1 total P2 0.088386 0.017200
7 1 2 1 1 1 total P3 -0.001243 0.005639
8 2 1 1 1 1 total P0 0.587256 0.084833
9 2 1 1 1 1 total P1 0.245120 0.041033
10 2 1 1 1 1 total P2 0.086784 0.016255
11 2 1 1 1 1 total P3 0.008660 0.004755
12 2 2 1 1 1 total P0 0.613792 0.168612
13 2 2 1 1 1 total P1 0.226142 0.061856
14 2 2 1 1 1 total P2 0.086174 0.025979
15 2 2 1 1 1 total P3 0.009721 0.012027
mesh 1 group in group out nuclide mean std. dev.
x y z
0 1 1 1 1 1 total 1.000000 0.088094
1 1 2 1 1 1 total 1.000000 0.160891
2 2 1 1 1 1 total 1.000000 0.126864
3 2 2 1 1 1 total 1.001374 0.305883
mesh 1 group in group out nuclide mean std. dev.
x y z
0 1 1 1 1 1 total 0.027395 0.004680
1 1 2 1 1 1 total 0.022914 0.006025
2 2 1 1 1 1 total 0.019384 0.002846
3 2 2 1 1 1 total 0.029629 0.006292
mesh 1 group out nuclide mean std. dev.
x y z
0 1 1 1 1 total 1.0 0.220956
1 1 2 1 1 total 1.0 0.316565
2 2 1 1 1 total 1.0 0.132140
3 2 2 1 1 total 1.0 0.181577
mesh 1 group out nuclide mean std. dev.
x y z
0 1 1 1 1 total 1.0 0.222246
1 1 2 1 1 total 1.0 0.316565
2 2 1 1 1 total 1.0 0.132140
3 2 2 1 1 total 1.0 0.181577
mesh 1 group in nuclide mean std. dev.
x y z
0 1 1 1 1 total 3.610522e-07 3.169931e-08
1 1 2 1 1 total 3.942353e-07 8.459167e-08
2 2 1 1 1 total 3.097784e-07 5.252025e-08
3 2 2 1 1 total 3.799163e-07 1.806470e-07
mesh 1 group in nuclide mean std. dev.
x y z
0 1 1 1 1 total 0.025735 0.002840
1 1 2 1 1 total 0.028773 0.006349
2 2 1 1 1 total 0.022306 0.004010
3 2 2 1 1 total 0.024549 0.009379

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tests/test_mgxs_library_mesh/test_mgxs_library_mesh.py Normal file
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#!/usr/bin/env python
import os
import sys
import glob
import hashlib
sys.path.insert(0, os.pardir)
from testing_harness import PyAPITestHarness
import openmc
import openmc.mgxs
class MGXSTestHarness(PyAPITestHarness):
def _build_inputs(self):
# Generate inputs using parent class routine
super(MGXSTestHarness, self)._build_inputs()
# Initialize a one-group structure
energy_groups = openmc.mgxs.EnergyGroups(group_edges=[0, 20.])
# Initialize MGXS Library for a few cross section types
# for one material-filled cell in the geometry
self.mgxs_lib = openmc.mgxs.Library(self._input_set.geometry)
self.mgxs_lib.by_nuclide = False
# Test all MGXS types
self.mgxs_lib.mgxs_types = openmc.mgxs.MGXS_TYPES
self.mgxs_lib.energy_groups = energy_groups
self.mgxs_lib.legendre_order = 3
self.mgxs_lib.domain_type = 'mesh'
# Instantiate a tally mesh
mesh = openmc.Mesh(mesh_id=1)
mesh.type = 'regular'
mesh.dimension = [2, 2]
mesh.lower_left = [-100., -100.]
mesh.width = [100., 100.]
self.mgxs_lib.domains = [mesh]
self.mgxs_lib.build_library()
# Initialize a tallies file
self._input_set.tallies = openmc.Tallies()
self.mgxs_lib.add_to_tallies_file(self._input_set.tallies, merge=False)
self._input_set.tallies.export_to_xml()
def _get_results(self, hash_output=False):
"""Digest info in the statepoint and return as a string."""
# Read the statepoint file.
statepoint = glob.glob(os.path.join(os.getcwd(), self._sp_name))[0]
sp = openmc.StatePoint(statepoint)
# Load the MGXS library from the statepoint
self.mgxs_lib.load_from_statepoint(sp)
# Build a string from Pandas Dataframe for each 1-group MGXS
outstr = ''
for domain in self.mgxs_lib.domains:
for mgxs_type in self.mgxs_lib.mgxs_types:
mgxs = self.mgxs_lib.get_mgxs(domain, mgxs_type)
df = mgxs.get_pandas_dataframe()
outstr += df.to_string() + '\n'
# Hash the results if necessary
if hash_output:
sha512 = hashlib.sha512()
sha512.update(outstr.encode('utf-8'))
outstr = sha512.hexdigest()
return outstr
def _cleanup(self):
super(MGXSTestHarness, self)._cleanup()
f = os.path.join(os.getcwd(), 'tallies.xml')
if os.path.exists(f): os.remove(f)
if __name__ == '__main__':
harness = MGXSTestHarness('statepoint.10.*', True)
harness.main()