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Added docstrings to MGXS Library object

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Will Boyd 2015-10-09 16:07:42 -04:00
parent 3b9c61a6d8
commit a508c3b420
3 changed files with 638 additions and 133 deletions
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426
docs/source/pythonapi/examples/multi-group-cross-sections.ipynb
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File diff suppressed because one or more lines are too long

202
openmc/mgxs/library.py
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@ -12,6 +12,51 @@ if sys.version_info[0] >= 3:
class Library(object):
"""A multi-group cross section library for some energy group structure.
This class can be used for both OpenMC input generation and tally data
post-processing to compute spatially-homogenized and energy-integrated
multi-group cross sections for deterministic neutronics calculations.
This class helps automate the generation of MGXS objects for some energy
group structure and domain type. The Library serves as a collection for
MGXS objects with routines to automate the initialization of tallies for
input files, the loading of tally data from statepoint files, data storage,
energy group condensation and more.
Parameters
----------
openmc_geometry : openmc.Geometry
An geometry which has been initialized with a root universe
by_nuclide : bool
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'])
name : str, optional
Name of the multi-group cross section. library Used as a label to
identify tallies in OpenMC 'tallies.xml' file.
Attributes
----------
openmc_geometry : openmc.Geometry
An geometry which has been initialized with a root universe
by_nuclide : bool
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 for spatial homogenization
energy_groups : EnergyGroups
Energy group structure for energy condensation
all_mgxs : dict
MGXS objects keyed by domain ID and cross section type
statepoint : openmc.StatePoint
The statepoint with tally data used to the compute cross sections
name : str, optional
Name of the multi-group cross section library. Used as a label to
identify tallies in OpenMC 'tallies.xml' file.
"""
def __init__(self, openmc_geometry, by_nuclide=False,
mgxs_types=None, name=''):
@ -23,6 +68,7 @@ class Library(object):
self._domain_type = None
self._energy_groups = None
self._all_mgxs = {}
self._statepoint = None
self.name = name
self.openmc_geometry = openmc_geometry
@ -104,6 +150,10 @@ class Library(object):
def all_mgxs(self):
return self._all_mgxs
@property
def statepoint(self):
return self._statepoint
@openmc_geometry.setter
def openmc_geometry(self, openmc_geometry):
cv.check_type('openmc_geometry', openmc_geometry, openmc.Geometry)
@ -140,7 +190,13 @@ class Library(object):
self._energy_groups = energy_groups
def build_library(self):
"""
"""Initialize MGXS objects in each domain and for each reaction type
in the library.
This routine will populate the all_mgxs instance attribute dictionary
with MGXS subclass objects keyed by each domain ID (e.g., Material IDs)
and cross section type (e.g., 'nu-fission', 'total', etc.).
"""
# Initialize MGXS for each domain and mgxs type and store in dictionary
@ -155,13 +211,20 @@ class Library(object):
mgxs.create_tallies()
self.all_mgxs[domain.id][mgxs_type] = mgxs
def add_to_tallies_file(self, tallies_file):
"""
def add_to_tallies_file(self, tallies_file, merge=True):
"""Add all tallies from all MGXS objects to a tallies file.
NOTE: This assumes that build_library() has been called
:param tallies_file:
:return:
Parameters
----------
tallies_file : openmc.TalliesFile
A TalliesFile object to add each MGXS' tallies to generate a
"tallies.xml" input file for OpenMC
merge : bool
Indicate whether tallies should be merged when possible. Defaults
to True.
"""
cv.check_type('tallies_file', tallies_file, openmc.TalliesFile)
@ -171,17 +234,39 @@ class Library(object):
for mgxs_type in self.mgxs_types:
mgxs = self.get_mgxs(domain, mgxs_type)
for tally_id, tally in mgxs.tallies.items():
tallies_file.add_tally(tally, merge=True)
tallies_file.add_tally(tally, merge=merge)
def load_from_statepoint(self, statepoint):
"""
"""Extracts tallies in an OpenMC StatePoint with the data needed to
compute multi-group cross sections.
This method is needed to compute cross section data from tallies
in an OpenMC StatePoint object.
NOTE: The statepoint must first be linked with an OpenMC Summary object.
Parameters
----------
statepoint : openmc.StatePoint
An OpenMC StatePoint object with tally data
Raises
------
ValueError
When this method is called with a statepoint that has not been
linked with a summary object.
:param statepoint:
:return:
"""
cv.check_type('statepoint', statepoint, openmc.StatePoint)
if not statepoint.with_summary:
msg = 'Unable to load data from a statepoint which has not been ' \
'linked with a summary file'
raise ValueError(msg)
self._statepoint = statepoint
# Load tallies for each MGXS for each domain and mgxs type
for domain in self.domains:
for mgxs_type in self.mgxs_types:
@ -190,11 +275,33 @@ class Library(object):
mgxs.compute_xs()
def get_mgxs(self, domain, mgxs_type):
"""
"""Return the MGXS object for some domain and reaction rate type.
This routine searches the library for an MGXS object for the spatial
domain and reaction rate type requ
NOTE: This routine must be called after the build_library() routine.
Parameters
----------
domain : Material or Cell or Universe or Integral
The material, cell, or universe object of interest (or its ID)
mgxs_type : {'total', 'transport', 'absorption', 'capture', 'fission',
'nu-fission', 'scatter', 'nu-scatter', 'scatter matrix',
'nu-scatter matrix', 'chi'}
The type of multi-group cross section object to return
Returns
-------
openmc.mgxs.MGXS
The MGXS object for the requested domain and reaction rate type
Raises
------
ValueError
If no MGXS object can be found for the requested domain or
multi-group cross section type
:param domain:
:param mgxs_type:
:return:
"""
if self.domain_type == 'material':
@ -225,15 +332,38 @@ class Library(object):
return self.all_mgxs[domain_id][mgxs_type]
def get_condensed_library(self, coarse_groups):
"""Construct an energy-condensed version of this library.
This routine condense each of the multi-group cross sections in the
library to a coarse energy group structure. NOTE: This routine must
be called after the load_from_statepoint(...) routine loads the tallies
from the statepoint into each of the cross sections.
Parameters
----------
coarse_groups : openmc.mgxs.EnergyGroups
The coarse energy group structure of interest
Returns
-------
Library
A new multi-group cross section library condensed to the group
structure of interest
Raises
------
ValueError
When this method is called before a statepoint has been loaded
See also
--------
MGXS.get_condensed_xs(coarse_groups)
"""
:param coarse_groups:
:return:
"""
if self.energy_groups is None:
if self.statepoint is None:
msg = 'Unable to get a condensed coarse group cross section ' \
'library since the fine energy groups have not yet been set'
'library since the statepoint has not yet been loaded'
raise ValueError(msg)
cv.check_type('coarse_groups', coarse_groups, openmc.mgxs.EnergyGroups)
@ -260,26 +390,42 @@ class Library(object):
def build_hdf5_store(self, filename='mgxs', directory='mgxs', xs_type='macro'):
"""Export the multi-group cross section library to an HDF5 binary file.
This method constructs an HDF5 file which stores the multi-group
cross section data. The data is stored in a hierarchy of HDF5 groups
from the domain type, domain id, subdomain id (for distribcell domains),
nuclides and cross section types. Two datasets for the mean and standard
deviation are stored for each subdomain entry in the HDF5 file.
This method constructs an HDF5 file which stores the library's
multi-group cross section data. The data is stored in a hierarchy of
HDF5 groups from the domain type, domain id, subdomain id (for
distribcell domains), nuclides and cross section types. Two datasets for
the mean and standard deviation are stored for each subdomain entry in
the HDF5 file.
NOTE: This requires the h5py Python package.
Parameters
----------
filename : str
Filename for the HDF5 file (default is 'mgxs')
Filename for the HDF5 file. Defaults to 'mgxs'.
directory : str
Directory for the HDF5 file (default is 'mgxs')
Directory for the HDF5 file. Defaults to 'mgxs'.
xs_type: {'macro', 'micro'}
Store the macro or micro cross section in units of cm^-1 or barns
Store the macro or micro cross section in units of cm^-1 or barns.
Defaults to 'macro'.
Raises
------
ValueError
When this method is called before a statepoint has been loaded
See also
--------
MGXS.build_hdf5_store(filename, directory, xs_type)
"""
# Load tallies for each MGXS for each domain and mgxs type
if self.statepoint is None:
msg = 'Unable to get a condensed coarse group cross section ' \
'library since a statepoint has not yet been loaded'
raise ValueError(msg)
# Export MGXS for each domain and mgxs type to an HDF5 file
for domain in self.domains:
for mgxs_type in self.mgxs_types:
mgxs = self.all_mgxs[domain.id][mgxs_type]

143
openmc/mgxs/mgxs.py
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@ -242,10 +242,11 @@ class MGXS(object):
energy_groups : EnergyGroups
The energy group structure for energy condensation
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
If true, computes cross sections for each nuclide in domain.
Defaults to False
name : str, optional
Name of the multi-group cross section. Used as a label to identify
tallies in OpenMC 'tallies.xml' file.
tallies in OpenMC 'tallies.xml' file. Defaults to the empty string.
Returns
-------
@ -351,7 +352,8 @@ class MGXS(object):
A list of nuclide name strings (e.g., ['U-235', 'U-238']). The
special string 'all' will return the atom densities for all nuclides
in the spatial domain. The special string 'sum' will return the atom
density summed across all nuclides in the spatial domain.
density summed across all nuclides in the spatial domain. Defaults
to 'all'.
Returns
-------
@ -544,22 +546,24 @@ class MGXS(object):
Parameters
----------
groups : Iterable of Integral or 'all'
Energy groups of interest
Energy groups of interest. Defaults to 'all'.
subdomains : Iterable of Integral or 'all'
Subdomain IDs of interest
Subdomain IDs of interest. Defaults to 'all'.
nuclides : Iterable of str or 'all' or 'sum'
A list of nuclide name strings (e.g., ['U-235', 'U-238']). The
special string 'all' (default) 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.
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
Return the macro or micro cross section in units of cm^-1 or barns.
Defaults to 'macro'.
order_groups: {'increasing', 'decreasing'}
Return the cross section indexed according to increasing (default)
or decreasing energy groups (decreasing or increasing energies)
Return the cross section indexed according to increasing or
decreasing energy groups (decreasing or increasing energies).
Defaults to 'increasing'.
value : str
A string for the type of value to return - 'mean' (default),
'std_dev' or 'rel_err' are accepted
A string for the type of value to return - 'mean', 'std_dev' or
'rel_err' are accepted. Defaults to 'mean'.
Returns
-------
@ -739,7 +743,7 @@ class MGXS(object):
Parameters
----------
subdomains : Iterable of Integral or 'all'
The subdomain IDs to average across
The subdomain IDs to average across. Defaults to 'all'.
Returns
-------
@ -815,15 +819,17 @@ class MGXS(object):
Parameters
----------
subdomains : Iterable of Integral or 'all'
The subdomain IDs of the cross sections to include in the report
The subdomain IDs of the cross sections to include in the report.
Defaults to 'all'.
nuclides : Iterable of str or 'all' or 'sum'
The nuclides of the cross-sections to include in the report. This
may be a list of nuclide name strings (e.g., ['U-235', 'U-238']).
The special string 'all' (default) will report the cross sections
for all nuclides in the spatial domain. The special string 'sum'
will report the cross sections summed over all nuclides.
The special string 'all' will report the cross sections for all
nuclides in the spatial domain. The special string 'sum' will report
the cross sections 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
Return the macro or micro cross section in units of cm^-1 or barns.
Defaults to 'macro'.
"""
@ -912,14 +918,15 @@ class MGXS(object):
Parameters
----------
filename : str
Filename for the HDF5 file (default is 'mgxs')
Filename for the HDF5 file. Defaults to 'mgxs'.
directory : str
Directory for the HDF5 file (default is 'mgxs')
Directory for the HDF5 file. Defaults to 'mgxs'.
xs_type: {'macro', 'micro'}
Store the macro or micro cross section in units of cm^-1 or barns
Store the macro or micro cross section in units of cm^-1 or barns.
Defaults to 'macro'.
append : boolean
If true, appends to an existing HDF5 file with the same filename
directory (if one exists)
directory (if one exists). Defaults to True.
Raises
------
@ -1030,15 +1037,16 @@ class MGXS(object):
Parameters
----------
filename : str
Filename for the exported file (default is 'mgxs')
Filename for the exported file. Defaults to 'mgxs'.
directory : str
Directory for the exported file (default is 'mgxs')
Directory for the exported file. Defaults to 'mgxs'.
format : {'csv', 'excel', 'pickle', 'latex'}
The format for the exported data file
The format for the exported data file. Defaults to 'csv'.
groups : Iterable of Integral or 'all'
Energy groups of interest
Energy groups of interest. Defaults to 'all'.
xs_type: {'macro', 'micro'}
Store the macro or micro cross section in units of cm^-1 or barns
Store the macro or micro cross section in units of cm^-1 or barns.
Defaults to 'macro'.
"""
@ -1100,15 +1108,17 @@ class MGXS(object):
Parameters
----------
groups : Iterable of Integral or 'all'
Energy groups of interest
Energy groups of interest. Defaults to 'all'.
nuclides : Iterable of str or 'all' or 'sum'
The nuclides of the cross-sections to include in the dataframe. This
may be a list of nuclide name strings (e.g., ['U-235', 'U-238']).
The special string 'all' (default) will include the cross sections
for all nuclides in the spatial domain. The special string 'sum'
will include the cross sections summed over all nuclides.
The special string 'all' will include the cross sections for all
nuclides in the spatial domain. The special string 'sum' will
include the cross sections summed over all nuclides. Defaults
to 'all'.
xs_type: {'macro', 'micro'}
Return macro or micro cross section in units of cm^-1 or barns
Return macro or micro cross section in units of cm^-1 or barns.
Defaults to 'macro'.
summary : None or Summary
An optional Summary object to be used to construct columns for
distribcell tally filters (default is None). The geometric
@ -1628,24 +1638,26 @@ class ScatterMatrixXS(MGXS):
Parameters
----------
in_groups : Iterable of Integral or 'all'
Incoming energy groups of interest
Incoming energy groups of interest. Defaults to 'all'.
out_groups : Iterable of Integral or 'all'
Outgoing energy groups of interest
Outgoing energy groups of interest. Defaults to 'all'.
subdomains : Iterable of Integral or 'all'
Subdomain IDs of interest
Subdomain IDs of interest. Defaults to 'all'.
nuclides : Iterable of str or 'all' or 'sum'
A list of nuclide name strings (e.g., ['U-235', 'U-238']). The
special string 'all' (default) 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.
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
Return the macro or micro cross section in units of cm^-1 or barns.
Defaults to 'macro'.
order_groups: {'increasing', 'decreasing'}
Return the cross section indexed according to increasing (default)
or decreasing energy groups (decreasing or increasing energies)
Return the cross section indexed according to increasing or
decreasing energy groups (decreasing or increasing energies).
Defaults to 'increasing'.
value : str
A string for the type of value to return - 'mean' (default),
'std_dev' or 'rel_err' are accepted
A string for the type of value to return - 'mean', 'std_dev', or
'rel_err' are accepted. Defaults to the empty string.
Returns
-------
@ -1756,15 +1768,17 @@ class ScatterMatrixXS(MGXS):
Parameters
----------
subdomains : Iterable of Integral or 'all'
The subdomain IDs of the cross sections to include in the report
The subdomain IDs of the cross sections to include in the report.
Defaults to 'all'.
nuclides : Iterable of str or 'all' or 'sum'
The nuclides of the cross-sections to include in the report. This
may be a list of nuclide name strings (e.g., ['U-235', 'U-238']).
The special string 'all' (default) will report the cross sections
for all nuclides in the spatial domain. The special string 'sum'
will report the cross sections summed over all nuclides.
The special string 'all' will report the cross sections for all
nuclides in the spatial domain. The special string 'sum' will report
the cross sections 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
Return the macro or micro cross section in units of cm^-1 or barns.
Defaults to 'macro'.
"""
@ -1952,23 +1966,24 @@ class Chi(MGXS):
Parameters
----------
groups : Iterable of Integral or 'all'
Energy groups of interest
Energy groups of interest. Defaults to 'all'.
subdomains : Iterable of Integral or 'all'
Subdomain IDs of interest
Subdomain IDs of interest. Defaults to 'all'.
nuclides : Iterable of str or 'all' or 'sum'
A list of nuclide name strings (e.g., ['U-235', 'U-238']). The
special string 'all' (default) 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.
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'}
This parameter is not relevant for chi but is included here to
mirror the parent MGXS.get_xs(...) class method
order_groups: {'increasing', 'decreasing'}
Return the cross section indexed according to increasing (default)
or decreasing energy groups (decreasing or increasing energies)
Return the cross section indexed according to increasing or
decreasing energy groups (decreasing or increasing energies).
Defaults to 'increasing'.
value : str
A string for the type of value to return - 'mean' (default),
'std_dev' or 'rel_err' are accepted
A string for the type of value to return - 'mean', 'std_dev', or
'rel_err' are accepted. Defaults to 'mean'.
Returns
-------
@ -2082,15 +2097,17 @@ class Chi(MGXS):
Parameters
----------
groups : Iterable of Integral or 'all'
Energy groups of interest
Energy groups of interest. Defaults to 'all'.
nuclides : Iterable of str or 'all' or 'sum'
The nuclides of the cross-sections to include in the dataframe. This
may be a list of nuclide name strings (e.g., ['U-235', 'U-238']).
The special string 'all' (default) will include the cross sections
for all nuclides in the spatial domain. The special string 'sum'
will include the cross sections summed over all nuclides.
The special string 'all' will include the cross sections for all
nuclides in the spatial domain. The special string 'sum' will
include the cross sections summed over all nuclides. Defaults to
'all'.
xs_type: {'macro', 'micro'}
Return macro or micro cross section in units of cm^-1 or barns
Return macro or micro cross section in units of cm^-1 or barns.
Defaults to 'macro'.
summary : None or Summary
An optional Summary object to be used to construct columns for
distribcell tally filters (default is None). The geometric