# openmc.mgxs.TotalXS¶

class openmc.mgxs.TotalXS(domain=None, domain_type=None, groups=None, by_nuclide=False, name='', num_polar=1, num_azimuthal=1)[source]

A total multi-group cross section.

This class can be used for both OpenMC input generation and tally data post-processing to compute spatially-homogenized and energy-integrated multi-group total cross sections for multi-group neutronics calculations. At a minimum, one needs to set the TotalXS.energy_groups and TotalXS.domain properties. Tallies for the flux and appropriate reaction rates over the specified domain are generated automatically via the TotalXS.tallies property, which can then be appended to a openmc.Tallies instance.

For post-processing, the MGXS.load_from_statepoint() will pull in the necessary data to compute multi-group cross sections from a openmc.StatePoint instance. The derived multi-group cross section can then be obtained from the TotalXS.xs_tally property.

For a spatial domain $$V$$ and energy group $$[E_g,E_{g-1}]$$, the total cross section is calculated as:

$\frac{\int_{r \in V} dr \int_{4\pi} d\Omega \int_{E_g}^{E_{g-1}} dE \; \sigma_t (r, E) \psi (r, E, \Omega)}{\int_{r \in V} dr \int_{4\pi} d\Omega \int_{E_g}^{E_{g-1}} dE \; \psi (r, E, \Omega)}.$
Parameters: domain (openmc.Material or openmc.Cell or openmc.Universe or openmc.RegularMesh) – The domain for spatial homogenization domain_type ({'material', 'cell', 'distribcell', 'universe', 'mesh'}) – The domain type for spatial homogenization groups (openmc.mgxs.EnergyGroups) – The energy group structure for energy condensation by_nuclide (bool) – If true, computes cross sections for each nuclide in domain name (str, optional) – Name of the multi-group cross section. Used as a label to identify tallies in OpenMC ‘tallies.xml’ file. num_polar (Integral, optional) – Number of equi-width polar angle bins for angle discretization; defaults to one bin num_azimuthal (Integral, optional) – Number of equi-width azimuthal angle bins for angle discretization; defaults to one bin name (str, optional) – Name of the multi-group cross section 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 (openmc.Material or openmc.Cell or openmc.Universe or openmc.RegularMesh) – Domain for spatial homogenization domain_type ({'material', 'cell', 'distribcell', 'universe', 'mesh'}) – Domain type for spatial homogenization energy_groups (openmc.mgxs.EnergyGroups) – Energy group structure for energy condensation num_polar (Integral) – Number of equi-width polar angle bins for angle discretization num_azimuthal (Integral) – Number of equi-width azimuthal angle bins for angle discretization tally_trigger (openmc.Trigger) – An (optional) tally precision trigger given to each tally used to compute the cross section scores (list of str) – The scores in each tally used to compute the multi-group cross section filters (list of openmc.Filter) – The filters in each tally used to compute the multi-group cross section tally_keys (list of str) – The keys into the tallies dictionary for each tally used to compute the multi-group cross section estimator ({'tracklength', 'collision', '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 are strings listed in the TotalXS.tally_keys property and values are instances of openmc.Tally. rxn_rate_tally (openmc.Tally) – Derived tally for the reaction rate tally used in the numerator to compute the multi-group cross section. This attribute is None unless the multi-group cross section has been computed. xs_tally (openmc.Tally) – 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). 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. 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. sparse (bool) – Whether or not the MGXS’ tallies use SciPy’s LIL sparse matrix format for compressed data storage loaded_sp (bool) – Whether or not a statepoint file has been loaded with tally data derived (bool) – Whether or not the MGXS is merged from one or more other MGXS hdf5_key (str) – The key used to index multi-group cross sections in an HDF5 data store
build_hdf5_store(filename='mgxs.h5', directory='mgxs', subdomains='all', nuclides='all', xs_type='macro', row_column='inout', append=True, libver='earliest')

Export the multi-group cross section data 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 type. 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. Defaults to ‘mgxs.h5’. directory (str) – Directory for the HDF5 file. Defaults to ‘mgxs’. subdomains (Iterable of Integral or 'all') – 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., [‘U235’, ‘U238’]). 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'}) – Store the macro or micro cross section in units of cm^-1 or barns. Defaults to ‘macro’. row_column ({'inout', 'outin'}) – Store scattering matrices indexed first by incoming group and second by outgoing group (‘inout’), or vice versa (‘outin’). Defaults to ‘inout’. append (bool) – If true, appends to an existing HDF5 file with the same filename directory (if one exists). Defaults to True. libver ({'earliest', 'latest'}) – Compatibility mode for the HDF5 file. ‘latest’ will produce files that are less backwards compatible but have performance benefits. ValueError – When this method is called before the multi-group cross section is computed from tally data.
can_merge(other)

Determine if another MGXS can be merged with this one

If results have been loaded from a statepoint, then MGXS are only mergeable along one and only one of enegy groups or nuclides.

Parameters: other (openmc.mgxs.MGXS) – MGXS to check for merging
export_xs_data(filename='mgxs', directory='mgxs', format='csv', groups='all', xs_type='macro')

Export the multi-group cross section data to a file.

This method leverages the functionality in the Pandas library to export the multi-group cross section data in a variety of output file formats for storage and/or post-processing.

Parameters: filename (str) – Filename for the exported file. Defaults to ‘mgxs’. directory (str) – Directory for the exported file. Defaults to ‘mgxs’. format ({'csv', 'excel', 'pickle', 'latex'}) – The format for the exported data file. Defaults to ‘csv’. groups (Iterable of Integral or 'all') – 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. Defaults to ‘macro’.
get_condensed_xs(coarse_groups)

Construct an energy-condensed version of this cross section.

Parameters: coarse_groups (openmc.mgxs.EnergyGroups) – The coarse energy group structure of interest A new MGXS condensed to the group structure of interest MGXS
get_flux(groups='all', subdomains='all', order_groups='increasing', value='mean', squeeze=True, **kwargs)

Returns an array of the fluxes used to weight the MGXS.

This method constructs a 2D NumPy array for the requested weighting flux for one or more subdomains (1st dimension), and energy groups (2nd dimension).

Parameters: groups (Iterable of Integral or 'all') – Energy groups of interest. Defaults to ‘all’. subdomains (Iterable of Integral or 'all') – Subdomain IDs of interest. Defaults to ‘all’. order_groups ({'increasing', 'decreasing'}) – Return the cross section indexed according to increasing or decreasing energy groups (decreasing or increasing energies). Defaults to ‘increasing’. value ({'mean', 'std_dev', 'rel_err'}) – A string for the type of value to return. Defaults to ‘mean’. squeeze (bool) – A boolean representing whether to eliminate the extra dimensions of the multi-dimensional array to be returned. Defaults to True. A NumPy array of the flux indexed in the order each group and subdomain is listed in the parameters. numpy.ndarray ValueError – When this method is called before the data is available from tally data, or, when this is used on an MGXS type without a flux score.
get_homogenized_mgxs(other_mgxs)

Construct a homogenized mgxs with other MGXS objects.

Parameters: other_mgxs (openmc.mgxs.MGXS or Iterable of openmc.mgxs.MGXS) – The MGXS to homogenize with this one. A new homogenized MGXS openmc.mgxs.MGXS ValueError – If the other_mgxs is of a different type.
static get_mgxs(mgxs_type, domain=None, domain_type=None, energy_groups=None, by_nuclide=False, name='', num_polar=1, num_azimuthal=1)

Return a MGXS subclass object for some energy group structure within some spatial domain for some reaction type.

This is a factory method which can be used to quickly create MGXS subclass objects for various reaction types.

Parameters: mgxs_type (str or Integral) – The type of multi-group cross section object to return; valid values are members of MGXS_TYPES, or the reaction types that are the keys of REACTION_MT. Note that if a reaction type from REACTION_MT is used, it can be appended with ‘ matrix’ to obtain a multigroup matrix (from incoming to outgoing energy groups) for reactions with a neutron in an outgoing channel. domain (openmc.Material or openmc.Cell or openmc.Universe or openmc.RegularMesh) – The domain for spatial homogenization 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 by_nuclide (bool) – 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. Defaults to the empty string. num_polar (Integral, optional) – Number of equi-width polar angles for angle discretization; defaults to no discretization num_azimuthal (Integral, optional) – Number of equi-width azimuthal angles for angle discretization; defaults to no discretization A subclass of the abstract MGXS class for the multi-group cross section type requested by the user openmc.mgxs.MGXS
get_nuclide_densities(nuclides='all')

Get an array of atomic number densities in units of atom/b-cm for all nuclides in the cross section’s spatial domain.

Parameters: 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 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. Defaults to ‘all’. An array of the atomic number densities (atom/b-cm) for each of the nuclides in the spatial domain numpy.ndarray of float ValueError – When this method is called before the spatial domain has been set.
get_nuclide_density(nuclide)

Get the atomic number density in units of atoms/b-cm for a nuclide in the cross section’s spatial domain.

Parameters: nuclide (str) – A nuclide name string (e.g., ‘U235’) The atomic number density (atom/b-cm) for the nuclide of interest float
get_nuclides()

Get all nuclides in the cross section’s spatial domain.

Returns: A list of the string names for each nuclide in the spatial domain (e.g., [‘U235’, ‘U238’, ‘O16’]) list of str ValueError – When this method is called before the spatial domain has been set.
get_pandas_dataframe(groups='all', nuclides='all', xs_type='macro', paths=True)

Build a Pandas DataFrame for the MGXS data.

This method leverages openmc.Tally.get_pandas_dataframe(), but renames the columns with terminology appropriate for cross section data.

Parameters: groups (Iterable of Integral or 'all') – 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., [‘U235’, ‘U238’]). 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. Defaults to ‘macro’. paths (bool, optional) – Construct columns for distribcell tally filters (default is True). The geometric information in the Summary object is embedded into a Multi-index column with a geometric “path” to each distribcell instance. A Pandas DataFrame for the cross section data. pandas.DataFrame ValueError – When this method is called before the multi-group cross section is computed from tally data.
get_slice(nuclides=[], groups=[])

Build a sliced MGXS for the specified nuclides and energy groups.

This method constructs a new MGXS to encapsulate a subset of the data represented by this MGXS. The subset of data to include in the tally slice is determined by the nuclides and energy groups specified in the input parameters.

Parameters: nuclides (list of str) – A list of nuclide name strings (e.g., [‘U235’, ‘U238’]; default is []) groups (list of int) – A list of energy group indices starting at 1 for the high energies (e.g., [1, 2, 3]; default is []) A new MGXS object which encapsulates the subset of data requested for the nuclide(s) and/or energy group(s) requested in the parameters. openmc.mgxs.MGXS
get_subdomain_avg_xs(subdomains='all')

Construct a subdomain-averaged version of this cross section.

This method is useful for averaging cross sections across distribcell instances. The method performs spatial homogenization to compute the scalar flux-weighted average cross section across the subdomains.

Parameters: subdomains (Iterable of Integral or 'all') – The subdomain IDs to average across. Defaults to ‘all’. A new MGXS averaged across the subdomains of interest openmc.mgxs.MGXS ValueError – When this method is called before the multi-group cross section is computed from tally data.
get_units(xs_type='macro')

This method returns the units of a MGXS based on a desired xs_type.

Parameters: xs_type ({'macro', 'micro'}) – Return the macro or micro cross section units. Defaults to ‘macro’. A string representing the units of the MGXS. str
get_xs(groups='all', subdomains='all', nuclides='all', xs_type='macro', order_groups='increasing', value='mean', squeeze=True, **kwargs)

Returns an array of multi-group cross sections.

This method constructs a 3D NumPy array for the requested multi-group cross section data for one or more subdomains (1st dimension), energy groups (2nd dimension), and nuclides (3rd dimension).

Parameters: groups (Iterable of Integral or 'all') – Energy groups of interest. Defaults to ‘all’. 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’. order_groups ({'increasing', 'decreasing'}) – Return the cross section indexed according to increasing or decreasing energy groups (decreasing or increasing energies). Defaults to ‘increasing’. value ({'mean', 'std_dev', 'rel_err'}) – A string for the type of value to return. Defaults to ‘mean’. squeeze (bool) – A boolean representing whether to eliminate the extra dimensions of the multi-dimensional array to be returned. Defaults to True. A NumPy array of the multi-group cross section indexed in the order each group, subdomain and nuclide is listed in the parameters. numpy.ndarray ValueError – When this method is called before the multi-group cross section is computed from tally data.
load_from_statepoint(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 be linked with an OpenMC Summary object.

Parameters: statepoint (openmc.StatePoint) – An OpenMC StatePoint object with tally data ValueError – When this method is called with a statepoint that has not been linked with a summary object.
merge(other)

Merge another MGXS with this one

MGXS are only mergeable if their energy groups and nuclides are either identical or mutually exclusive. If results have been loaded from a statepoint, then MGXS are only mergeable along one and only one of energy groups or nuclides.

Parameters: other (openmc.mgxs.MGXS) – MGXS to merge with this one merged_mgxs – Merged MGXS openmc.mgxs.MGXS
print_xs(subdomains='all', nuclides='all', xs_type='macro')

Print a string representation for the multi-group cross section.

Parameters: subdomains (Iterable of Integral or 'all') – 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., [‘U235’, ‘U238’]). 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. Defaults to ‘macro’.