From 9245ebda77509ba4bc2123e62a8c07fe6eb696d6 Mon Sep 17 00:00:00 2001 From: Adam Nelson Date: Wed, 30 Nov 2016 20:16:05 -0500 Subject: [PATCH] Docstring changes (forgot optional on some optional attribs) --- openmc/plotter.py | 12 ++++++------ 1 file changed, 6 insertions(+), 6 deletions(-) diff --git a/openmc/plotter.py b/openmc/plotter.py index ac11e4073e..b3febb5250 100644 --- a/openmc/plotter.py +++ b/openmc/plotter.py @@ -98,15 +98,15 @@ def plot_xs(this, types, divisor_types=None, temperature=294., axis=None, Enrichment for U235 in weight percent. For example, input 4.95 for 4.95 weight percent enriched U. Default is None. This is only used for items which are instances of openmc.Element - plot_CE : bool + plot_CE : bool, optional Denotes whether or not continuous-energy will be plotted. Defaults to plotting the continuous-energy data. - orders : Iterable of Integral + orders : Iterable of Integral, optional The scattering order or delayed group index to use for the corresponding entry in types. Defaults to the 0th order for scattering and the total delayed neutron data. This only applies to plots of multi-group data. - divisor_orders : Iterable of Integral + divisor_orders : Iterable of Integral, optional Same as orders, but for divisor_types **kwargs All keyword arguments are passed to @@ -619,7 +619,7 @@ def calculate_mgxs(this, types, orders=None, temperature=294., Object to source data from types : Iterable of values of PLOT_TYPES The type of cross sections to calculate - orders : Iterable of Integral + orders : Iterable of Integral, optional The scattering order or delayed group index to use for the corresponding entry in types. Defaults to the 0th order for scattering and the total delayed neutron data. @@ -705,7 +705,7 @@ def _calculate_mgxs_nuc_macro(this, types, library, orders=None, in openmc.PLOT_TYPES_MGXS library : openmc.MGXSLibrary MGXS Library containing the data of interest - orders : Iterable of Integral + orders : Iterable of Integral, optional The scattering order or delayed group index to use for the corresponding entry in types. Defaults to the 0th order for scattering and the total delayed neutron data. @@ -814,7 +814,7 @@ def _calculate_mgxs_elem_mat(this, types, library, orders=None, in openmc.PLOT_TYPES_MGXS library : openmc.MGXSLibrary MGXS Library containing the data of interest - orders : Iterable of Integral + orders : Iterable of Integral, optional The scattering order or delayed group index to use for the corresponding entry in types. Defaults to the 0th order for scattering and the total delayed neutron data.