From ca12e3656ee94147bbc8409a452acc6a204c5f97 Mon Sep 17 00:00:00 2001 From: Olek <45364492+yardasol@users.noreply.github.com> Date: Thu, 8 Sep 2022 14:02:47 -0500 Subject: [PATCH] Typo and syntax fixes Co-authored-by: Paul Romano --- docs/source/usersguide/depletion.rst | 6 +++--- 1 file changed, 3 insertions(+), 3 deletions(-) diff --git a/docs/source/usersguide/depletion.rst b/docs/source/usersguide/depletion.rst index 245070237..c1f9bf94a 100644 --- a/docs/source/usersguide/depletion.rst +++ b/docs/source/usersguide/depletion.rst @@ -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