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updates per @wbinventor comments
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Adam Nelson
parent
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commit
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3 changed files with 86 additions and 100 deletions
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@ -57,7 +57,7 @@ class MGXS(object):
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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 deterministic neutronics calculations.
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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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@ -727,9 +727,8 @@ class MGXS(object):
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Return the cross section indexed according to increasing or
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decreasing energy groups (decreasing or increasing energies).
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Defaults to 'increasing'.
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value : str
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A string for the type of value to return - 'mean', 'std_dev' or
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'rel_err' are accepted. Defaults to 'mean'.
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value : {'mean', 'std_dev', 'rel_err'}
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A string for the type of value to return. Defaults to 'mean'.
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Returns
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-------
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@ -963,8 +962,9 @@ class MGXS(object):
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Returns
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-------
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openmc.mgxs.MGXS
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A new tally which encapsulates the subset of data requested for the
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nuclide(s) and/or energy group(s) requested in the parameters.
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A new MGXS object which encapsulates the subset of data requested
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for the nuclide(s) and/or energy group(s) requested in the
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parameters.
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"""
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@ -1536,7 +1536,7 @@ class MatrixMGXS(MGXS):
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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 deterministic neutronics calculations.
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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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@ -1624,9 +1624,7 @@ class MatrixMGXS(MGXS):
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energy = openmc.Filter('energy', group_edges)
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energyout = openmc.Filter('energyout', group_edges)
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filters = [[energy], [energy, energyout]]
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return filters
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return [[energy], [energy, energyout]]
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@property
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def estimator(self):
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@ -1636,10 +1634,10 @@ class MatrixMGXS(MGXS):
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subdomains='all', nuclides='all',
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xs_type='macro', order_groups='increasing',
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row_column='inout', value='mean', **kwargs):
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r"""Returns an array of multi-group cross sections.
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"""Returns an array of multi-group cross sections.
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This method constructs a 2D NumPy array for the requested multiplicity
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matrix data data for one or more energy groups and subdomains.
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This method constructs a 2D NumPy array for the requested multi-group
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matrix data for one or more energy groups and subdomains.
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Parameters
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----------
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@ -1666,9 +1664,8 @@ class MatrixMGXS(MGXS):
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Return the cross section indexed first by incoming group and
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second by outgoing group ('inout'), or vice versa ('outin').
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Defaults to 'inout'.
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value : str
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A string for the type of value to return - 'mean', 'std_dev', or
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'rel_err' are accepted. Defaults to the empty string.
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value : {'mean', 'std_dev', 'rel_err'}
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A string for the type of value to return. Defaults to 'mean'.
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Returns
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-------
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@ -1777,7 +1774,7 @@ class MatrixMGXS(MGXS):
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return xs
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def get_slice(self, nuclides=[], in_groups=[], out_groups=[]):
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"""Build a sliced NuFissionMatrix for the specified nuclides and
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"""Build a sliced matrixMGXS object for the specified nuclides and
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energy groups.
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This method constructs a new MGXS to encapsulate a subset of the data
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@ -1799,9 +1796,10 @@ class MatrixMGXS(MGXS):
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Returns
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-------
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openmc.mgxs.MGXS
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A new tally which encapsulates the subset of data requested for the
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nuclide(s) and/or energy group(s) requested in the parameters.
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openmc.mgxs.MatrixMGXS
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A new MatrixMGXS object which encapsulates the subset of data
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requested for the nuclide(s) and/or energy group(s) requested in
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the parameters.
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"""
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@ -1940,7 +1938,7 @@ class TotalXS(MGXS):
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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 deterministic neutronics calculations.
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multi-group cross sections for multi-group neutronics calculations.
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Parameters
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----------
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@ -2028,7 +2026,7 @@ class TransportXS(MGXS):
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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 deterministic neutronics calculations.
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multi-group cross sections for multi-group neutronics calculations.
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Parameters
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----------
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@ -2142,7 +2140,7 @@ class NuTransportXS(TransportXS):
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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 deterministic neutronics calculations.
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multi-group cross sections for multi-group neutronics calculations.
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Parameters
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----------
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@ -2238,7 +2236,7 @@ class AbsorptionXS(MGXS):
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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 deterministic neutronics calculations.
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multi-group cross sections for multi-group neutronics calculations.
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Parameters
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----------
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@ -2331,7 +2329,7 @@ class CaptureXS(MGXS):
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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 deterministic neutronics calculations.
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multi-group cross sections for multi-group neutronics calculations.
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Parameters
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----------
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@ -2431,7 +2429,7 @@ class FissionXS(MGXS):
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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 deterministic neutronics calculations.
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multi-group cross sections for multi-group neutronics calculations.
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Parameters
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----------
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@ -2519,7 +2517,7 @@ class NuFissionXS(MGXS):
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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 deterministic neutronics calculations.
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multi-group cross sections for multi-group neutronics calculations.
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Parameters
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----------
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@ -2607,7 +2605,7 @@ class KappaFissionXS(MGXS):
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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 deterministic neutronics calculations.
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multi-group cross sections for multi-group neutronics calculations.
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Parameters
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----------
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@ -2695,7 +2693,7 @@ class ScatterXS(MGXS):
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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 deterministic neutronics calculations.
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multi-group cross sections for multi-group neutronics calculations.
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Parameters
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----------
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@ -2783,7 +2781,7 @@ class NuScatterXS(MGXS):
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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 deterministic neutronics calculations.
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multi-group cross sections for multi-group neutronics calculations.
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Parameters
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----------
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@ -2872,7 +2870,7 @@ class ScatterMatrixXS(MatrixMGXS):
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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 deterministic neutronics calculations.
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multi-group cross sections for multi-group neutronics calculations.
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Parameters
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----------
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@ -3116,9 +3114,10 @@ class ScatterMatrixXS(MatrixMGXS):
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Returns
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-------
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openmc.mgxs.MGXS
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A new tally which encapsulates the subset of data requested for the
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nuclide(s) and/or energy group(s) requested in the parameters.
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openmc.mgxs.MatrixMGXS
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A new MatrixMGXS which encapsulates the subset of data requested
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for the nuclide(s) and/or energy group(s) requested in the
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parameters.
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"""
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@ -3200,9 +3199,8 @@ class ScatterMatrixXS(MatrixMGXS):
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Return the cross section indexed first by incoming group and
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second by outgoing group ('inout'), or vice versa ('outin').
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Defaults to 'inout'.
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value : str
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A string for the type of value to return - 'mean', 'std_dev', or
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'rel_err' are accepted. Defaults to the empty string.
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value : {'mean', 'std_dev', 'rel_err'}
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A string for the type of value to return. Defaults to 'mean'.
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Returns
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-------
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@ -3430,7 +3428,7 @@ class ScatterMatrixXS(MatrixMGXS):
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cv.check_value('xs_type', xs_type, ['macro', 'micro'])
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if self.correction != 'P0':
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rxn_type= '{0} (P{1})'.format(self.rxn_type, moment)
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rxn_type = '{0} (P{1})'.format(self.rxn_type, moment)
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else:
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rxn_type = self.rxn_type
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@ -3449,7 +3447,7 @@ class ScatterMatrixXS(MatrixMGXS):
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template = '{0: <12}Group {1} [{2: <10} - {3: <10}MeV]\n'
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# Loop over energy groups ranges
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for group in range(1, self.num_groups+1):
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for group in range(1, self.num_groups + 1):
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bounds = self.energy_groups.get_group_bounds(group)
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string += template.format('', group, bounds[0], bounds[1])
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@ -3476,8 +3474,8 @@ class ScatterMatrixXS(MatrixMGXS):
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template = '{0: <12}Group {1} -> Group {2}:\t\t'
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# Loop over incoming/outgoing energy groups ranges
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for in_group in range(1, self.num_groups+1):
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for out_group in range(1, self.num_groups+1):
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for in_group in range(1, self.num_groups + 1):
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for out_group in range(1, self.num_groups + 1):
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string += template.format('', in_group, out_group)
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average = \
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self.get_xs([in_group], [out_group],
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@ -3489,7 +3487,8 @@ class ScatterMatrixXS(MatrixMGXS):
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xs_type=xs_type, value='rel_err')
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average = average.flatten()[0]
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rel_err = rel_err.flatten()[0] * 100.
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string += '{:1.2e} +/- {:1.2e}%'.format(average, rel_err)
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string += '{:1.2e} +/- {:1.2e}%'.format(average,
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rel_err)
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string += '\n'
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string += '\n'
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string += '\n'
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@ -3504,7 +3503,7 @@ class NuScatterMatrixXS(ScatterMatrixXS):
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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 deterministic neutronics calculations.
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multi-group cross sections for multi-group neutronics calculations.
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Parameters
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----------
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@ -3597,7 +3596,7 @@ class MultiplicityMatrixXS(MatrixMGXS):
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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 deterministic neutronics calculations.
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multi-group cross sections for multi-group neutronics calculations.
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Parameters
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----------
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@ -3691,9 +3690,7 @@ class MultiplicityMatrixXS(MatrixMGXS):
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energy = openmc.Filter('energy', group_edges)
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energyout = openmc.Filter('energyout', group_edges)
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filters = [[energy, energyout], [energy, energyout]]
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return filters
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return [[energy, energyout], [energy, energyout]]
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@property
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def rxn_rate_tally(self):
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@ -3720,7 +3717,7 @@ class NuFissionMatrixXS(MatrixMGXS):
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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 deterministic neutronics calculations.
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multi-group cross sections for multi-group neutronics calculations.
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Parameters
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----------
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@ -3800,19 +3797,8 @@ class NuFissionMatrixXS(MatrixMGXS):
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groups=None, by_nuclide=False, name=''):
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super(NuFissionMatrixXS, self).__init__(domain, domain_type,
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groups, by_nuclide, name)
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self._rxn_type = 'nu-fission matrix'
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@property
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def scores(self):
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scores = ['flux', 'nu-fission']
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return scores
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@property
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def rxn_rate_tally(self):
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if self._rxn_rate_tally is None:
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self._rxn_rate_tally = self.tallies['nu-fission']
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self._rxn_rate_tally.sparse = self.sparse
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return self._rxn_rate_tally
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self._rxn_type = 'nu-fission'
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self._hdf5_key = 'nu-fission matrix'
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class Chi(MGXS):
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@ -3820,7 +3806,7 @@ class Chi(MGXS):
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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 deterministic neutronics calculations.
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multi-group cross sections for multi-group neutronics calculations.
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Parameters
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----------
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@ -3966,9 +3952,10 @@ class Chi(MGXS):
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Returns
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-------
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MGXS
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A new tally which encapsulates the subset of data requested for the
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nuclide(s) and/or energy group(s) requested in the parameters.
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openmc.mgxs.MGXS
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A new MGXS which encapsulates the subset of data requested
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for the nuclide(s) and/or energy group(s) requested in the
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parameters.
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"""
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@ -4080,9 +4067,8 @@ class Chi(MGXS):
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Return the cross section indexed according to increasing or
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decreasing energy groups (decreasing or increasing energies).
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Defaults to 'increasing'.
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value : str
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A string for the type of value to return - 'mean', 'std_dev', or
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'rel_err' are accepted. Defaults to 'mean'.
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value : {'mean', 'std_dev', 'rel_err'}
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A string for the type of value to return. Defaults to 'mean'.
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Returns
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-------
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@ -525,7 +525,7 @@ class XSdata(object):
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def chi(self, chi):
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if self.use_chi is not None:
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if not self.use_chi:
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msg = 'Providing chi when nu_fission already provided as a' \
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msg = 'Providing "chi" when "nu-fission" already provided as a' \
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'matrix'
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raise ValueError(msg)
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@ -753,7 +753,7 @@ class XSdata(object):
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msg = 'Angular-Dependent MGXS have not yet been implemented'
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raise ValueError(msg)
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if type(nu_fission) is openmc.mgxs.NuFissionMatrixXS:
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if isinstance(nu_fission, openmc.mgxs.NuFissionMatrixXS):
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self.use_chi = False
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else:
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self.use_chi = True
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