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Made Current class consistent with mesh domain type

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1) Kept by_nuclide, num_nuclides, and nuclides in the docstrings because when building a mgxs library for all mgxs_types, the variables should be the same accross all classes.
However, I noted that they were unused for a SurfaceMGXS.

2) Removed all code related to nuclides and micro xs since they were unused anyway.

3) Removed the possibility of computing a Current with any domain other than mesh.
Included an error message if a current mgxs object is instantiated with any domain other than mesh.
Since mesh domains and by_nuclide features are uncompatible, an existing message is already raised when mesh and by_nuclide are used together.

4) Changed estimator type to analog.
This commit is contained in:
Miriam 2020-07-17 22:07:15 +00:00
parent 2e706e5088
commit 43456ea2d6
2 changed files with 27 additions and 88 deletions
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113
openmc/mgxs/mgxs.py
View file

@ -5914,7 +5914,7 @@ class SurfaceMGXS(MGXS):
energy_groups : openmc.mgxs.EnergyGroups
The energy group structure for energy condensation
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
Unused for SurfacMGXS
name : str, optional
Name of the multi-group cross section. Used as a label to identify
tallies in OpenMC 'tallies.xml' file.
@ -5926,7 +5926,7 @@ class SurfaceMGXS(MGXS):
rxn_type : str
Reaction type (e.g., 'total', 'nu-fission', etc.)
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
Unused for SurfaceMGXS
domain : Mesh
Domain for spatial homogenization
domain_type : {'mesh'}
@ -5959,13 +5959,9 @@ class SurfaceMGXS(MGXS):
two to account for both the incoming and outgoing current from the
mesh cell surfaces.
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.
Unused n SurfaceMGXS
nuclides : Iterable of str or 'sum'
The optional user-specified nuclides for which to compute cross
sections (e.g., 'U238', 'O16'). If by_nuclide is True but nuclides
are not specified by the user, all nuclides in the spatial domain
are included. This attribute is 'sum' if by_nuclide is false.
Unused in SurfaceMGXS
sparse : bool
Whether or not the MGXS' tallies use SciPy's LIL sparse matrix format
for compressed data storage
@ -5977,14 +5973,17 @@ class SurfaceMGXS(MGXS):
The key used to index multi-group cross sections in an HDF5 data store
"""
# Store whether or not the number density should be removed for microscopic
# values of this data
_divide_by_density = False
def __init__(self, domain=None, domain_type=None, energy_groups=None,
by_nuclide=False, name=''):
super(SurfaceMGXS, self).__init__(domain, domain_type, energy_groups,
by_nuclide, name)
self._estimator = ['analog']
self._valid_estimators = ['analog']
if domain_type != 'mesh':
msg = 'Unable to compute a SurfaceMGXS for domain type {0} ' \
'which is not a mesh type'.format(domain_type)
raise ValueError(msg)
@property
def scores(self):
@ -6037,23 +6036,6 @@ class SurfaceMGXS(MGXS):
'linked with a summary file'
raise ValueError(msg)
# Override the domain object that loaded from an OpenMC summary file
# NOTE: This is necessary for micro cross-sections which require
# the isotopic number densities as computed by OpenMC
geom = statepoint.summary.geometry
if self.domain_type in ('cell', 'distribcell'):
self.domain = geom.get_all_cells()[self.domain.id]
elif self.domain_type == 'universe':
self.domain = geom.get_all_universes()[self.domain.id]
elif self.domain_type == 'material':
self.domain = geom.get_all_materials()[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)
raise ValueError(msg)
# Use tally "slicing" to ensure that tallies correspond to our domain
# NOTE: This is important if tally merging was used
if self.domain_type == 'mesh':
@ -6102,13 +6084,10 @@ class SurfaceMGXS(MGXS):
subdomains : Iterable of Integral or 'all'
Subdomain IDs of interest. Defaults to 'all'.
nuclides : Iterable of str or 'all' or 'sum'
A list of nuclide name strings (e.g., ['U235', 'U238']). The
special string 'all' will return the cross sections for all nuclides
in the spatial domain. The special string 'sum' will return the
cross section summed over all nuclides. Defaults to 'all'.
xs_type: {'macro', 'micro'}
Return the macro or micro cross section in units of cm^-1 or barns.
Defaults to 'macro'.
Unused in SurfaceMGXS
xs_type: {'macro'}
The 'macro'/'micro' distinction does not apply to SurfaceMGXS.
The calculation of a 'micro' xs_type is omited in this class.
order_groups: {'increasing', 'decreasing'}
Return the cross section indexed according to increasing or
decreasing energy groups (decreasing or increasing energies).
@ -6131,14 +6110,7 @@ class SurfaceMGXS(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)
cv.check_value('xs_type', xs_type, ['macro'])
filters = []
filter_bins = []
@ -6164,34 +6136,6 @@ class SurfaceMGXS(MGXS):
(self.energy_groups.get_group_bounds(group),))
filter_bins.append(tuple(energy_bins))
# Construct a collection of the nuclides to retrieve from the xs tally
if self.by_nuclide:
if nuclides == 'all' or nuclides == 'sum' or nuclides == ['sum']:
query_nuclides = self.get_nuclides()
else:
query_nuclides = nuclides
else:
query_nuclides = ['total']
# If user requested the sum for all nuclides, use tally summation
if nuclides == 'sum' or nuclides == ['sum']:
xs_tally = self.xs_tally.summation(nuclides=query_nuclides)
xs = xs_tally.get_values(filters=filters,
filter_bins=filter_bins, value=value)
else:
xs = self.xs_tally.get_values(filters=filters,
filter_bins=filter_bins,
nuclides=query_nuclides, value=value)
# Divide by atom number densities for microscopic cross sections
if xs_type == 'micro' and self._divide_by_density:
if self.by_nuclide:
densities = self.get_nuclide_densities(nuclides)
else:
densities = self.get_nuclide_densities('sum')
if value == 'mean' or value == 'std_dev':
xs /= densities[np.newaxis, :, np.newaxis]
# Eliminate the trivial score dimension
xs = np.squeeze(xs, axis=len(xs.shape) - 1)
xs = np.nan_to_num(xs)
@ -6262,7 +6206,7 @@ class Current(SurfaceMGXS):
groups : openmc.mgxs.EnergyGroups
The energy group structure for energy condensation
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
Unused in SurfaceMGXS
name : str, optional
Name of the multi-group cross section. Used as a label to identify
tallies in OpenMC 'tallies.xml' file.
@ -6274,10 +6218,10 @@ class Current(SurfaceMGXS):
rxn_type : str
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 or Mesh
Unused for SurfaceMGXS
domain : Mesh
Domain for spatial homogenization
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
domain_type : {'mesh'}
Domain type for spatial homogenization
energy_groups : openmc.mgxs.EnergyGroups
Energy group structure for energy condensation
@ -6291,7 +6235,7 @@ class Current(SurfaceMGXS):
tally_keys : list of str
The keys into the tallies dictionary for each tally used to compute
the multi-group cross section
estimator : {'tracklength', 'analog'}
estimator : {'analog'}
The tally estimator used to compute the multi-group cross section
tallies : collections.OrderedDict
OpenMC tallies needed to compute the multi-group cross section. The keys
@ -6305,18 +6249,13 @@ class Current(SurfaceMGXS):
Derived tally for the multi-group cross section. This attribute
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. 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).
The number of subdomains is equal to the number of mesh surfaces times
two to account for both the incoming and outgoing current from the
mesh cell surfaces.
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.
Unused in SurfaceMGXS
nuclides : Iterable of str or 'sum'
The optional user-specified nuclides for which to compute cross
sections (e.g., 'U238', 'O16'). If by_nuclide is True but nuclides
are not specified by the user, all nuclides in the spatial domain
are included. This attribute is 'sum' if by_nuclide is false.
Unused in SurfaceMGXS
sparse : bool
Whether or not the MGXS' tallies use SciPy's LIL sparse matrix format
for compressed data storage

2
tests/regression_tests/mgxs_library_distribcell/test.py
View file

@ -23,7 +23,7 @@ class MGXSTestHarness(PyAPITestHarness):
self.mgxs_lib.by_nuclide = False
# Test all MGXS types
self.mgxs_lib.mgxs_types = openmc.mgxs.MGXS_TYPES + \
self.mgxs_lib.mgxs_types = openmc.mgxs.MGXS_TYPES[:-1] + \
openmc.mgxs.MDGXS_TYPES
self.mgxs_lib.energy_groups = energy_groups
self.mgxs_lib.num_delayed_groups = 6