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Fixed a few docstrings in openmc.mgxs
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Will Boyd
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commit
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2 changed files with 43 additions and 34 deletions
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@ -127,7 +127,6 @@ respectively.
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openmc.get_hexagonal_prism
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openmc.get_rectangular_prism
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>>>>>>> upstream/develop
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Constructing Tallies
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--------------------
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@ -288,20 +287,13 @@ Multi-group Cross Sections
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openmc.mgxs.CaptureXS
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openmc.mgxs.Chi
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openmc.mgxs.ChiPrompt
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openmc.mgxs.ConsistentNuScatterMatrixXS
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openmc.mgxs.ConsistentScatterMatrixXS
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openmc.mgxs.ConvolvedMGXS
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openmc.mgxs.FissionXS
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openmc.mgxs.InverseVelocity
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openmc.mgxs.KappaFissionXS
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openmc.mgxs.MultiplicityMatrixXS
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openmc.mgxs.NuFissionXS
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openmc.mgxs.NuFissionMatrixXS
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openmc.mgxs.NuScatterXS
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openmc.mgxs.NuScatterMatrixXS
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openmc.mgxs.NuScatterProbabilityMatrix
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openmc.mgxs.PromptNuFissionXS
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openmc.mgxs.PromptNuFissionMatrixXS
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openmc.mgxs.ScatterXS
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openmc.mgxs.ScatterMatrixXS
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openmc.mgxs.ScatterProbabilityMatrix
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@ -125,7 +125,7 @@ class MGXS(object):
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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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.. note:: Users should instantiate the subclasses of this abstract class.
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Parameters
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----------
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@ -894,7 +894,7 @@ class MGXS(object):
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This method is needed to compute cross section data from tallies
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in an OpenMC StatePoint object.
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NOTE: The statepoint must first be linked with an OpenMC Summary object.
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.. note:: The statepoint must be linked with an OpenMC Summary object.
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Parameters
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----------
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@ -1642,7 +1642,7 @@ class MGXS(object):
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nuclides and cross section type. Two datasets for the mean and standard
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deviation are stored for each subdomain entry in the HDF5 file.
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NOTE: This requires the h5py Python package.
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.. note:: This requires the h5py Python package.
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Parameters
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----------
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@ -1988,7 +1988,7 @@ class MatrixMGXS(MGXS):
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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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.. note:: Users should instantiate the subclasses of this abstract class.
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Parameters
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----------
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@ -3389,9 +3389,6 @@ class ScatterXS(MGXS):
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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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nu : bool
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If True, the cross section data will include neutron multiplication;
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defaults to False
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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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@ -3403,6 +3400,9 @@ class ScatterXS(MGXS):
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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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nu : bool
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If True, the cross section data will include neutron multiplication;
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defaults to False
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Attributes
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----------
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@ -3553,9 +3553,6 @@ class ScatterMatrixXS(MatrixMGXS):
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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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nu : bool
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If True, the cross section data will include neutron multiplication;
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defaults to False
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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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@ -3567,6 +3564,9 @@ class ScatterMatrixXS(MatrixMGXS):
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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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nu : bool
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If True, the cross section data will include neutron multiplication;
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defaults to False
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Attributes
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----------
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@ -3837,7 +3837,7 @@ class ScatterMatrixXS(MatrixMGXS):
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This method is needed to compute cross section data from tallies
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in an OpenMC StatePoint object.
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NOTE: The statepoint must first be linked with an OpenMC Summary object.
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.. note:: The statepoint must be linked with an OpenMC Summary object.
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Parameters
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----------
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@ -3959,9 +3959,9 @@ class ScatterMatrixXS(MatrixMGXS):
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(3rd dimension), nuclides (4th dimension), and moments/histograms
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(5th dimension).
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NOTE: The scattering moments are not multiplied by the :math:`(2l+1)/2`
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prefactor in the expansion of the scattering source into Legendre
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moments in the neutron transport equation.
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.. note:: The scattering moments are not multiplied by the
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:math:`(2l+1)/2` prefactor in the expansion of the scattering source
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into Legendre moments in the neutron transport equation.
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Parameters
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----------
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@ -4599,9 +4599,6 @@ class ScatterProbabilityMatrix(MatrixMGXS):
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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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nu : bool
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If True, the cross section data will include neutron multiplication;
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defaults to False
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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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@ -4613,6 +4610,9 @@ class ScatterProbabilityMatrix(MatrixMGXS):
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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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nu : bool
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If True, the cross section data will include neutron multiplication;
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defaults to False
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Attributes
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----------
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@ -4693,7 +4693,10 @@ class ScatterProbabilityMatrix(MatrixMGXS):
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self._valid_estimators = ['analog']
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self.nu = nu
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# FIXME: Add __deepcopy__
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def __deepcopy__(self, memo):
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clone = super(ScatterProbabilityMatrixXS, self).__deepcopy__(memo)
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clone._nu = self.nu
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return clone
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@property
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def nu(self):
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@ -4761,7 +4764,7 @@ class ConvolvedMGXS(MGXS):
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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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.. note:: Users should instantiate the subclasses of this abstract class.
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Parameters
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----------
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@ -5029,7 +5032,7 @@ class ConvolvedMGXS(MGXS):
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This method is needed to compute cross section data from tallies
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in an OpenMC StatePoint object.
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NOTE: The statepoint must first be linked with an OpenMC Summary object.
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.. note:: The statepoint must be linked with an OpenMC Summary object.
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Parameters
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----------
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@ -5171,6 +5174,9 @@ class ConsistentScatterMatrixXS(ConvolvedMGXS, ScatterMatrixXS):
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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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nu : bool
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If True, the cross section data will include neutron multiplication;
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defaults to False
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Attributes
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----------
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@ -5194,6 +5200,8 @@ class ConsistentScatterMatrixXS(ConvolvedMGXS, ScatterMatrixXS):
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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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nu : bool
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If True, the cross section data will include neutron multiplication
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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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@ -5262,8 +5270,6 @@ class ConsistentScatterMatrixXS(ConvolvedMGXS, ScatterMatrixXS):
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domain, domain_type, groups, by_nuclide=by_nuclide,
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name=name, num_polar=num_polar, num_azimuthal=num_azimuthal)
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self._rxn_type = 'h'
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self._rxn_type
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self.nu = nu
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# Initialize each MGXS used by the convolution
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@ -5284,7 +5290,7 @@ class ConsistentScatterMatrixXS(ConvolvedMGXS, ScatterMatrixXS):
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mgxs.energy_groups = groups
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mgxs.num_polar = num_polar
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mgxs.num_azimuthal = num_azimuthal
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@property
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def scores(self):
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scores = super(ConsistentScatterMatrixXS, self).scores
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@ -5486,6 +5492,9 @@ class NuFissionMatrixXS(MatrixMGXS):
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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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prompt : bool
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If true, computes cross sections which only includes prompt neutrons;
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defaults to False which includes prompt and delayed in total
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Attributes
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----------
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@ -5493,6 +5502,8 @@ class NuFissionMatrixXS(MatrixMGXS):
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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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prompt : bool
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If true, computes cross sections which only includes prompt neutrons
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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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@ -5553,9 +5564,9 @@ class NuFissionMatrixXS(MatrixMGXS):
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"""
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def __init__(self, domain=None, domain_type=None,
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groups=None, by_nuclide=False, name='', num_polar=1,
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num_azimuthal=1):
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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,
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num_azimuthal=1, prompt=False):
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super(NuFissionMatrixXS, self).__init__(domain, domain_type,
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groups, by_nuclide, name,
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num_polar, num_azimuthal)
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@ -5792,6 +5803,12 @@ class Chi(MGXS):
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def prompt(self, prompt):
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cv.check_type('prompt', prompt, bool)
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self._prompt = prompt
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if not self.prompt:
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self._rxn_type = 'nu-fission'
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self._hdf5_key = 'nu-fission matrix'
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else:
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self._rxn_type = 'prompt-nu-fission'
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self._hdf5_key = 'prompt-nu-fission matrix'
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def get_homogenized_mgxs(self, other_mgxs):
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"""Construct a homogenized mgxs with other MGXS objects.
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