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Cleaned up docstring for nuclides attribute of MGXS class for @paulromano

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Will Boyd 2016-03-28 17:30:19 -04:00
parent ad79ba214e
commit f90c46d2f3
1 changed files with 5 additions and 7 deletions
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12
openmc/mgxs/mgxs.py
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@ -67,10 +67,6 @@ class MGXS(object):
The energy group structure for energy condensation
by_nuclide : bool
If true, computes cross sections for each nuclide in domain
nuclides : Iterable of basestring
The user-specified nuclides to compute cross sections. If by_nuclide
is True but nuclides are not specified by the user, all nuclides in the
spatial domain will be used.
name : str, optional
Name of the multi-group cross section. Used as a label to identify
tallies in OpenMC 'tallies.xml' file.
@ -109,9 +105,11 @@ class MGXS(object):
num_nuclides : Integral
The number of nuclides for which the multi-group cross section is
being tracked. This is unity if the by_nuclide attribute is False.
nuclides : list of str or 'sum'
A list of nuclide string names (e.g., 'U-238', 'O-16') when by_nuclide
is True and 'sum' when by_nuclide is False.
nuclides : Iterable of str or 'sum'
The optional user-specified nuclides for which to compute cross
sections (e.g., 'U-238', 'O-16'). If by_nuclide is True but nuclides
are not specified by the user, all nuclides in the spatial domain
are included. This attribute is 'sum' if by_nuclide is false.
sparse : bool
Whether or not the MGXS' tallies use SciPy's LIL sparse matrix format
for compressed data storage