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Added docstrings to all consistent scattering matrix MGXS classes in openmc.mgxs
This commit is contained in:
Will Boyd
parent
2369ff4443
commit
fe74423c48
2 changed files with 371 additions and 14 deletions
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@ -115,16 +115,6 @@ Many of the above classes are derived from several abstract classes:
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openmc.Region
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openmc.Lattice
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One function is also available to create a hexagonal region defined by the
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intersection of six surface half-spaces.
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.. autosummary::
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:toctree: generated
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:nosignatures:
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:template: myfunction.rst
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openmc.make_hexagon_region
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Constructing Tallies
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--------------------
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@ -5201,8 +5201,101 @@ class NuScatterProbabilityMatrix(ScatterProbabilityMatrix):
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@add_metaclass(ABCMeta)
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class ConvolvedMGXS(MGXS):
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"""An abstract convolution of multiple multi-group cross sections for some
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energy group structure within some spatial domain.
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# FIXME: Add docstring
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This class can be used for both OpenMC input generation and tally data
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post-processing to compute spatially-homogenized and energy-integrated
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multi-group cross sections for multi-group neutronics calculations.
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NOTE: Users should instantiate the subclasses of this abstract class.
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Parameters
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----------
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domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
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The domain for spatial homogenization
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domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
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The domain type for spatial homogenization
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energy_groups : openmc.mgxs.EnergyGroups
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The energy group structure for energy condensation
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by_nuclide : bool
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If true, computes cross sections for each nuclide in domain
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name : str, optional
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Name of the multi-group cross section. Used as a label to identify
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tallies in OpenMC 'tallies.xml' file.
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num_polar : Integral, optional
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Number of equi-width polar angle bins for angle discretization;
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defaults to one bin
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num_azimuthal : Integral, optional
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Number of equi-width azimuthal angle bins for angle discretization;
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defaults to one bin
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Attributes
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----------
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name : str, optional
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Name of the multi-group cross section
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rxn_type : str
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Reaction type (e.g., 'total', 'nu-fission', etc.)
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by_nuclide : bool
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If true, computes cross sections for each nuclide in domain
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domain : Material or Cell or Universe or Mesh
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Domain for spatial homogenization
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domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
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Domain type for spatial homogenization
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energy_groups : openmc.mgxs.EnergyGroups
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Energy group structure for energy condensation
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num_polar : Integral
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Number of equi-width polar angle bins for angle discretization
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num_azimuthal : Integral
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Number of equi-width azimuthal angle bins for angle discretization
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tally_trigger : openmc.Trigger
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An (optional) tally precision trigger given to each tally used to
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compute the cross section
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scores : list of str
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The scores in each tally used to compute the multi-group cross section
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filters : list of openmc.Filter
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The filters in each tally used to compute the multi-group cross section
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tally_keys : list of str
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The keys into the tallies dictionary for each tally used to compute
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the multi-group cross section
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estimator : {'tracklength', 'collision', 'analog'}
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The tally estimator used to compute the multi-group cross section
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tallies : collections.OrderedDict
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OpenMC tallies needed to compute the multi-group cross section
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rxn_rate_tally : openmc.Tally
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Derived tally for the reaction rate tally used in the numerator to
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compute the multi-group cross section. This attribute is None
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unless the multi-group cross section has been computed.
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xs_tally : openmc.Tally
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Derived tally for the multi-group cross section. This attribute
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is None unless the multi-group cross section has been computed.
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num_subdomains : int
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The number of subdomains is unity for 'material', 'cell' and 'universe'
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domain types. This is equal to the number of cell instances
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for 'distribcell' domain types (it is equal to unity prior to loading
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tally data from a statepoint file) and the number of mesh cells for
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'mesh' domain types.
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num_nuclides : int
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The number of nuclides for which the multi-group cross section is
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being tracked. This is unity if the by_nuclide attribute is False.
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nuclides : Iterable of str or 'sum'
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The optional user-specified nuclides for which to compute cross
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sections (e.g., 'U238', 'O16'). If by_nuclide is True but nuclides
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are not specified by the user, all nuclides in the spatial domain
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are included. This attribute is 'sum' if by_nuclide is false.
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sparse : bool
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Whether or not the MGXS' tallies use SciPy's LIL sparse matrix format
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for compressed data storage
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loaded_sp : bool
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Whether or not a statepoint file has been loaded with tally data
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derived : bool
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Whether or not the MGXS is merged from one or more other MGXS
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hdf5_key : str
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The key used to index multi-group cross sections in an HDF5 data store
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mgxs : list of openmc.mgxs.MGXS
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A list of MGXS to combine to compute this multi-group cross section
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"""
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def __init__(self, domain=None, domain_type=None, groups=None,
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by_nuclide=False, name='', num_polar=1, num_azimuthal=1):
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@ -5461,8 +5554,150 @@ class ConvolvedMGXS(MGXS):
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class ConsistentScatterMatrixXS(ConvolvedMGXS, ScatterMatrixXS):
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r"""A scattering matrix multi-group cross section computed as the product
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of the scatter cross section and group-to-group scattering probabilities.
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# FIXME: Add docstring
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This class is a variation of the :class:`ScatterMatrixXS` which computes
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the scattering matrix as the convolution product of :class:`ScatterXS` and
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:class:`ScatterProbabilityMatrix`. Unlike the :class:`ScatterMatrixXS`,
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this scattering matrix is computed from the scattering cross section which
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uses a tracklength estimator. This ensures that reaction rate balance is
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exactly preserved with a :class:`TotalXS` computed using a tracklength
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estimator.
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This class can be used for both OpenMC input generation and tally data
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post-processing to compute spatially-homogenized and energy-integrated
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multi-group cross sections for multi-group neutronics calculations. At a
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minimum, one needs to set the
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:attr:`ConsistentScatterMatrixXS.energy_groups` and
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:attr:`ConsistentScatterMatrixXS.domain` properties. Tallies for the flux
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and appropriate reaction rates over the specified domain are generated
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automatically via the :attr:`ConsistentScatterMatrixXS.tallies` property,
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which can then be appended to a :class:`openmc.Tallies` instance.
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For post-processing, the :meth:`MGXS.load_from_statepoint` will pull in the
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necessary data to compute multi-group cross sections from a
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:class:`openmc.StatePoint` instance. The derived multi-group cross section
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can then be obtained from the :attr:`ConsistentScatterMatrixXS.xs_tally`
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property.
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For a spatial domain :math:`V`, incoming energy group
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:math:`[E_{g'},E_{g'-1}]`, and outgoing energy group :math:`[E_g,E_{g-1}]`,
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the Legendre scattering moments are calculated as:
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.. math::
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\langle \sigma_{s,g'\rightarrow g} \phi \rangle &= \int_{r \in V} dr
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\int_{4\pi} d\Omega' \int_{E_{g'}}^{E_{g'-1}} dE' \int_{4\pi} d\Omega
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\int_{E_g}^{E_{g-1}} dE \; \sigma_s (r, E'
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\rightarrow E, \Omega' \cdot \Omega) \psi(r, E', \Omega')\\
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\langle \phi \rangle &= \int_{r \in V} dr \int_{4\pi} d\Omega
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\int_{E_g}^{E_{g-1}} dE \; \psi (r, E, \Omega) \\
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\sigma_{s,g'\rightarrow g} &= \frac{\langle
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\sigma_{s,,g'\rightarrow g} \phi \rangle}{\langle \phi \rangle}
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Parameters
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----------
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domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
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The domain for spatial homogenization
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domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
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The domain type for spatial homogenization
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groups : openmc.mgxs.EnergyGroups
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The energy group structure for energy condensation
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by_nuclide : bool
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If true, computes cross sections for each nuclide in domain
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name : str, optional
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Name of the multi-group cross section. Used as a label to identify
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tallies in OpenMC 'tallies.xml' file.
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num_polar : Integral, optional
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Number of equi-width polar angle bins for angle discretization;
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defaults to one bin
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num_azimuthal : Integral, optional
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Number of equi-width azimuthal angle bins for angle discretization;
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defaults to one bin
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Attributes
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----------
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correction : 'P0' or None
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Apply the P0 correction to scattering matrices if set to 'P0'; this is
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used only if :attr:`ConsistentScatterMatrixXS.scatter_format` is
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'legendre'
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scatter_format : {'legendre', or 'histogram'}
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Representation of the angular scattering distribution (default is
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'legendre')
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legendre_order : int
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The highest Legendre moment in the scattering matrix; this is used if
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:attr:`ConsistentScatterMatrixXS.scatter_format` is 'legendre'.
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(default is 0)
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histogram_bins : int
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The number of equally-spaced bins for the histogram representation of
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the angular scattering distribution; this is used if
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:attr:`ConsistentScatterMatrixXS.scatter_format` is 'histogram'.
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(default is 16)
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name : str, optional
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Name of the multi-group cross section
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rxn_type : str
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Reaction type (e.g., 'total', 'nu-fission', etc.)
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by_nuclide : bool
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If true, computes cross sections for each nuclide in domain
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domain : Material or Cell or Universe or Mesh
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Domain for spatial homogenization
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domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
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Domain type for spatial homogenization
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energy_groups : openmc.mgxs.EnergyGroups
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Energy group structure for energy condensation
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num_polar : Integral
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Number of equi-width polar angle bins for angle discretization
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num_azimuthal : Integral
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Number of equi-width azimuthal angle bins for angle discretization
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tally_trigger : openmc.Trigger
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An (optional) tally precision trigger given to each tally used to
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compute the cross section
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scores : list of str
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The scores in each tally used to compute the multi-group cross section
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filters : list of openmc.Filter
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The filters in each tally used to compute the multi-group cross section
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tally_keys : list of str
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The keys into the tallies dictionary for each tally used to compute
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the multi-group cross section
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estimator : 'analog'
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The tally estimator used to compute the multi-group cross section
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tallies : collections.OrderedDict
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OpenMC tallies needed to compute the multi-group cross section. The keys
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are strings listed in the :attr:`ConsistentScatterMatrixXS.tally_keys`
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property
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and values are instances of :class:`openmc.Tally`.
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rxn_rate_tally : openmc.Tally
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Derived tally for the reaction rate tally used in the numerator to
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compute the multi-group cross section. This attribute is None
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unless the multi-group cross section has been computed.
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xs_tally : openmc.Tally
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Derived tally for the multi-group cross section. This attribute
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is None unless the multi-group cross section has been computed.
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num_subdomains : int
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The number of subdomains is unity for 'material', 'cell' and 'universe'
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domain types. This is equal to the number of cell instances
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for 'distribcell' domain types (it is equal to unity prior to loading
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tally data from a statepoint file).
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num_nuclides : int
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The number of nuclides for which the multi-group cross section is
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being tracked. This is unity if the by_nuclide attribute is False.
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nuclides : Iterable of str or 'sum'
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The optional user-specified nuclides for which to compute cross
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sections (e.g., 'U238', 'O16'). If by_nuclide is True but nuclides
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are not specified by the user, all nuclides in the spatial domain
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are included. This attribute is 'sum' if by_nuclide is false.
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sparse : bool
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Whether or not the MGXS' tallies use SciPy's LIL sparse matrix format
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for compressed data storage
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loaded_sp : bool
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Whether or not a statepoint file has been loaded with tally data
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derived : bool
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Whether or not the MGXS is merged from one or more other MGXS
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hdf5_key : str
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The key used to index multi-group cross sections in an HDF5 data store
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"""
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def __init__(self, domain=None, domain_type=None, groups=None,
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by_nuclide=False, name='', num_polar=1, num_azimuthal=1):
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@ -5619,8 +5854,140 @@ class ConsistentScatterMatrixXS(ConvolvedMGXS, ScatterMatrixXS):
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class ConsistentNuScatterMatrixXS(ConsistentScatterMatrixXS):
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r"""A scattering-production matrix multi-group cross section computed as
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the product of the scattering-production cross section and group-to-group
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scattering-production probabilities.
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# FIXME: Add docstring
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This class is a variation of the :class:`NuScatterMatrixXS` which computes
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the scattering-production matrix as the convolution product of
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:class:`NuScatterXS` and :class:`NuScatterProbabilityMatrix`. Unlike the
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:class:`NuScatterMatrixXS`, this scattering-production matrix is computed
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from the scattering-production cross section which uses a tracklength
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estimator. This ensures that reaction rate balance is exactly preserved
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with a :class:`TotalXS` computed using a tracklength estimator.
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This class can be used for both OpenMC input generation and tally data
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post-processing to compute spatially-homogenized and energy-integrated
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multi-group cross sections for multi-group neutronics calculations. At a
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minimum, one needs to set the
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:attr:`ConsistentNuScatterMatrixXS.energy_groups` and
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:attr:`ConsistentNuScatterMatrixXS.domain` properties. Tallies for the flux
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and appropriate reaction rates over the specified domain are generated
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automatically via the :attr:`ConsistentNuScatterMatrixXS.tallies` property,
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which can then be appended to a :class:`openmc.Tallies` instance.
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For post-processing, the :meth:`MGXS.load_from_statepoint` will pull in the
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necessary data to compute multi-group cross sections from a
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:class:`openmc.StatePoint` instance. The derived multi-group cross section
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can then be obtained from the :attr:`ConsistentNuScatterMatrixXS.xs_tally`
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property.
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The calculation of the scattering-production matrix is the same as that for
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:class:`ConsistentScatterMatrixXS` except that the scattering multiplicity
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is accounted for.
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Parameters
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----------
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domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
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The domain for spatial homogenization
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domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
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The domain type for spatial homogenization
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groups : openmc.mgxs.EnergyGroups
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The energy group structure for energy condensation
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by_nuclide : bool
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If true, computes cross sections for each nuclide in domain
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name : str, optional
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Name of the multi-group cross section. Used as a label to identify
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tallies in OpenMC 'tallies.xml' file.
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num_polar : Integral, optional
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Number of equi-width polar angle bins for angle discretization;
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defaults to one bin
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num_azimuthal : Integral, optional
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Number of equi-width azimuthal angle bins for angle discretization;
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defaults to one bin
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Attributes
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----------
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correction : 'P0' or None
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Apply the P0 correction to scattering matrices if set to 'P0'; this is
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used only if :attr:`ConsistentNuScatterMatrixXS.scatter_format` is
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'legendre'
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scatter_format : {'legendre', or 'histogram'}
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Representation of the angular scattering distribution (default is
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'legendre')
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legendre_order : int
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The highest Legendre moment in the scattering matrix; this is used if
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:attr:`ConsistentScatterNuMatrixXS.scatter_format` is 'legendre'.
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(default is 0)
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histogram_bins : int
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The number of equally-spaced bins for the histogram representation of
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the angular scattering distribution; this is used if
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:attr:`ConsistentScatterNuMatrixXS.scatter_format` is 'histogram'.
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(default is 16)
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name : str, optional
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Name of the multi-group cross section
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rxn_type : str
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Reaction type (e.g., 'total', 'nu-fission', etc.)
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by_nuclide : bool
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If true, computes cross sections for each nuclide in domain
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domain : Material or Cell or Universe or Mesh
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Domain for spatial homogenization
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domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
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Domain type for spatial homogenization
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energy_groups : openmc.mgxs.EnergyGroups
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Energy group structure for energy condensation
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num_polar : Integral
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Number of equi-width polar angle bins for angle discretization
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num_azimuthal : Integral
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Number of equi-width azimuthal angle bins for angle discretization
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tally_trigger : openmc.Trigger
|
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An (optional) tally precision trigger given to each tally used to
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compute the cross section
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scores : list of str
|
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The scores in each tally used to compute the multi-group cross section
|
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filters : list of openmc.Filter
|
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The filters in each tally used to compute the multi-group cross section
|
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tally_keys : list of str
|
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The keys into the tallies dictionary for each tally used to compute
|
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the multi-group cross section
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estimator : 'analog'
|
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The tally estimator used to compute the multi-group cross section
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tallies : collections.OrderedDict
|
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OpenMC tallies needed to compute the multi-group cross section. The keys
|
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are strings listed in the :attr:`ConsistentScatterNuMatrixXS.tally_keys`
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property
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and values are instances of :class:`openmc.Tally`.
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rxn_rate_tally : openmc.Tally
|
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Derived tally for the reaction rate tally used in the numerator to
|
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compute the multi-group cross section. This attribute is None
|
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unless the multi-group cross section has been computed.
|
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xs_tally : openmc.Tally
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Derived tally for the multi-group cross section. This attribute
|
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is None unless the multi-group cross section has been computed.
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num_subdomains : int
|
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The number of subdomains is unity for 'material', 'cell' and 'universe'
|
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domain types. This is equal to the number of cell instances
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for 'distribcell' domain types (it is equal to unity prior to loading
|
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tally data from a statepoint file).
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num_nuclides : int
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The number of nuclides for which the multi-group cross section is
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being tracked. This is unity if the by_nuclide attribute is False.
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nuclides : Iterable of str or 'sum'
|
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The optional user-specified nuclides for which to compute cross
|
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sections (e.g., 'U238', 'O16'). If by_nuclide is True but nuclides
|
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are not specified by the user, all nuclides in the spatial domain
|
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are included. This attribute is 'sum' if by_nuclide is false.
|
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sparse : bool
|
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Whether or not the MGXS' tallies use SciPy's LIL sparse matrix format
|
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for compressed data storage
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loaded_sp : bool
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Whether or not a statepoint file has been loaded with tally data
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derived : bool
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Whether or not the MGXS is merged from one or more other MGXS
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hdf5_key : str
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The key used to index multi-group cross sections in an HDF5 data store
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"""
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def __init__(self, domain=None, domain_type=None, groups=None,
|
||||
by_nuclide=False, name='', num_polar=1, num_azimuthal=1):
|
||||
|
|
@ -5630,7 +5997,7 @@ class ConsistentNuScatterMatrixXS(ConsistentScatterMatrixXS):
|
|||
|
||||
self._rxn_type = 'nu-scatter'
|
||||
self._hdf5_key = 'consistent nu-scatter matrix'
|
||||
self._mgxs = [NuScatterXS(), ScatterProbabilityMatrix()]
|
||||
self._mgxs = [NuScatterXS(), NuScatterProbabilityMatrix()]
|
||||
|
||||
# Assign parameters to each MGXS in the convlution
|
||||
for mgxs in self.mgxs:
|
||||
|
|
|
|||
Loading…
Add table
Add a link
Reference in a new issue