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Typo and syntax fixes

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Co-authored-by: Paul Romano <paul.k.romano@gmail.com>
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Olek 2022-09-08 14:02:47 -05:00 • committed by GitHub
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docs/source/usersguide/depletion.rst
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@ -259,12 +259,12 @@ transport-depletion calculation and follow the same steps from there.
will not be simulated.
Loading and Generating Microscopic Cross Sections
-------------------------------------
-------------------------------------------------
As mentioned earlier, any transport code could be used to calculate one-group
microscopic cross sections. The :mod:`openmc.deplete` module provides the
:class:`~openmc.deplete.MicroXS` class, which contains methods to read in
pre-caluclated cross sections from a ``.csv`` file or from data arrays::
pre-calculated cross sections from a ``.csv`` file or from data arrays::
micro_xs = MicroXS.from_csv(micro_xs_path)
@ -352,7 +352,7 @@ normalizing reaction rates:
Multiple Materials
~~~~~~~~~~~~~~~~~~
A transport-independent depletion simulation using ``source-race`` normalization
A transport-independent depletion simulation using ``source-rate`` normalization
will calculate reaction rates for each material independently. This can be
useful for running many different cases of a particular scenario. A depletion
simulation using ``fission-q`` normalization will sum the energy values from