Saving changes before looking at num_polar/azimuthal defaulting to 1

docs/source/io_formats/mgxs_library.rst

@@ -90,7 +90,7 @@ Temperature-dependent data, provided for temperature <TTT>K.
 :Datasets: - **total** (*double[]* or *double[][][]*) -- Total cross section.
              This is a 1-D vector if `representation` is "isotropic", or a 3-D
              vector if `representation` is "angle" with dimensions of
-             [groups][azimuthal][polar].
+             [polar][azimuthal][groups].
            - **absorption** (*double[]* or *double[][][]*) -- Absorption
              cross section.
              This is a 1-D vector if `representation` is "isotropic", or a 3-D
@@ -100,19 +100,19 @@ Temperature-dependent data, provided for temperature <TTT>K.
              cross section.
              This is a 1-D vector if `representation` is "isotropic", or a 3-D
              vector if `representation` is "angle" with dimensions of
-             [groups][azimuthal][polar].  This is only required if the dataset
+             [polar][azimuthal][groups].  This is only required if the dataset
              is fissionable and fission-tallies are expected to be used.
            - **kappa-fission** (*double[]* or *double[][][]*) -- Kappa-Fission
              (energy-release from fission) cross section.
              This is a 1-D vector if `representation` is "isotropic", or a 3-D
              vector if `representation` is "angle" with dimensions of
-             [groups][azimuthal][polar].  This is only required if the dataset
+             [polar][azimuthal][groups].  This is only required if the dataset
              is fissionable and fission-tallies are expected to be used.
            - **chi** (*double[]* or *double[][][]*) -- Fission neutron energy
              spectra.
              This is a 1-D vector if `representation` is "isotropic", or a 3-D
              vector if `representation` is "angle" with dimensions of
-             [groups][azimuthal][polar].  This is only required if the dataset
+             [polar][azimuthal][groups].  This is only required if the dataset
              is fissionable and fission-tallies are expected to be used.
            - **nu-fission** (*double[]* to *double[][][][]*) -- Nu-Fission
              cross section.
@@ -124,7 +124,9 @@ Temperature-dependent data, provided for temperature <TTT>K.
              same dimensionality as `multiplicity matrix`.
            - **inverse-velocity** (*double[]* or *double[][][]*) --
              Average inverse velocity for each of the groups in the library.
-             This dataset is optional.
+             This dataset is optional. This is a 1-D vector if `representation`
+             is "isotropic", or a 3-D vector if `representation` is "angle"
+             with dimensions of [polar][azimuthal][groups].
 
 **/<library name>/<TTT>K/scatter_data/**
 

openmc/mgxs/mdgxs.py

@@ -408,9 +408,9 @@ class MDGXS(MGXS):
         else:
             num_delayed_groups = len(delayed_groups)
 
-        # Reshape tally data array with separate axes for domain, energy groups,
-        # delayed groups, and nuclides
-        # Accomodate the polar and azimuthal bins if needed
+        # Reshape tally data array with separate axes for domain,
+        # energy groups, delayed groups, and nuclides
+        # Accommodate the polar and azimuthal bins if needed
         if self.num_polar or self.num_azimuthal:
             if self.num_polar:
                 num_pol = self.num_polar
@@ -852,7 +852,6 @@ class MDGXS(MGXS):
         if self.by_nuclide and nuclides == 'sum':
 
             # Use tally summation to sum across all nuclides
-            query_nuclides = [nuclides]
             xs_tally = self.xs_tally.summation(nuclides=self.get_nuclides())
             df = xs_tally.get_pandas_dataframe(
                 distribcell_paths=distribcell_paths)
@@ -865,14 +864,12 @@ class MDGXS(MGXS):
 
         # If the user requested a specific set of nuclides
         elif self.by_nuclide and nuclides != 'all':
-            query_nuclides = nuclides
             xs_tally = self.xs_tally.get_slice(nuclides=nuclides)
             df = xs_tally.get_pandas_dataframe(
                 distribcell_paths=distribcell_paths)
 
         # If the user requested all nuclides, keep nuclide column in dataframe
         else:
-            query_nuclides = self.nuclides
             df = self.xs_tally.get_pandas_dataframe(
                 distribcell_paths=distribcell_paths)
 

openmc/mgxs/mgxs.py

@@ -1760,7 +1760,6 @@ class MGXS(object):
         if self.by_nuclide and nuclides == 'sum':
 
             # Use tally summation to sum across all nuclides
-            query_nuclides = [nuclides]
             xs_tally = self.xs_tally.summation(nuclides=self.get_nuclides())
             df = xs_tally.get_pandas_dataframe(
                 distribcell_paths=distribcell_paths)
@@ -4332,6 +4331,7 @@ class ScatterMatrixXS(MatrixMGXS):
                     dont_squeeze = (1, 2, 3)
                 else:
                     dont_squeeze = (1, 2)
+
         # Squeeze will return a ValueError if the axis has a size
         # greater than 1, so try each axis in axes one at a time,
         # and do our own check to preclude the ValueError