crp/OpenMC
1
0
Fork 0
mirror of https://github.com/openmc-dev/openmc.git synced 2026-07-28 14:15:42 -04:00
Code Issues Projects Releases Packages Wiki Activity Actions 5

[skip ci] review improvments to docstring from @paulromano

Browse source 
Co-authored-by: Paul Romano <paul.k.romano@gmail.com>
This commit is contained in:
Jonathan Shimwell 2023-03-22 13:11:33 +00:00 • committed by GitHub
parent 17e9918f5d
commit 83a124c492
No known key found for this signature in database
GPG key ID: 4AEE18F83AFDEB23
1 changed files with 6 additions and 6 deletions
Download patch file Download diff file Expand all files Collapse all files

12
openmc/plotter.py
View file

@ -63,8 +63,8 @@ def plot_xs(this, types, divisor_types=None, temperature=294., axis=None,
Parameters
----------
this : {str, openmc.Material}
Object to source data from. Nuclides and Elements can be input as a str
this : str or openmc.Material
Object to source data from. Nuclides and elements can be input as a str
types : Iterable of values of PLOT_TYPES
The type of cross sections to include in the plot.
divisor_types : Iterable of values of PLOT_TYPES, optional
@ -217,8 +217,8 @@ def calculate_cexs(this, types, temperature=294., sab_name=None,
Parameters
----------
this : {str, openmc.Material}
Object to source data from. Nuclides and Elements should be input as a
this : str or openmc.Material
Object to source data from. Nuclides and elements should be input as a
str
types : Iterable of values of PLOT_TYPES
The type of cross sections to calculate
@ -603,8 +603,8 @@ def calculate_mgxs(this, types, orders=None, temperature=294.,
Parameters
----------
this : {str, openmc.Material}
Object to source data from. Nuclides and Elements can be input as a str
this : str or openmc.Material
Object to source data from. Nuclides and elements can be input as a str
types : Iterable of values of PLOT_TYPES_MGXS
The type of cross sections to calculate
orders : Iterable of Integral, optional