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address most of paulromano's comments

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yardasol 2022-07-21 11:00:45 -05:00
parent 61ad9c977c
commit 06f34b9f23
5 changed files with 159 additions and 189 deletions
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8
docs/source/usersguide/depletion.rst
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@ -113,7 +113,7 @@ should be, including indirect components. Some examples are provided below::
fission_q = {"U235": 202e+6} # energy in eV
# create a Model object
model = openmc.model.Model(geometry, settings)
model = openmc.Model(geometry, settings)
# create a modified chain and write it to a new file
chain = openmc.deplete.Chain.from_xml("chain.xml", fission_q)
@ -180,7 +180,7 @@ across all material instances.
number of tallies and material definitions.
Transport-independent depletion
-------------------------------
===============================
.. note::
@ -188,10 +188,10 @@ Transport-independent depletion
possible and likely in the near future.
OpenMC supports running depletion calculations independent of the OpenMC
transport solver using the :class:`FluxDepletionOperator` class. This class
transport solver using the :class:`~openmc.deplete.FluxDepletionOperator` class. This class
has two ways to initalize it; the default constructor accepts an
:class:`openmc.Materials` object, a flux spectra, and one-group microscopic
cross sections as a :class:`pandas.Dataframe`, while the `from_nuclides`
cross sections as a :class:`pandas.DataFrame`, while the `from_nuclides`
method accepts a volume and dictionary of nuclide concentrations in place of
the :class:`openmc.Materials` object in addition to the other parameters.
The class includes helper functions to construct the dataframe from a csv file