From a414f5a32ad82272c449d0157f6aeee72a95ce4b Mon Sep 17 00:00:00 2001 From: Olek <45364492+yardasol@users.noreply.github.com> Date: Thu, 8 Sep 2022 12:49:54 -0500 Subject: [PATCH 1/2] Apply suggestions from code review Co-authored-by: Paul Romano --- docs/source/usersguide/depletion.rst | 8 ++++---- 1 file changed, 4 insertions(+), 4 deletions(-) diff --git a/docs/source/usersguide/depletion.rst b/docs/source/usersguide/depletion.rst index b0d8ff4541..2450702374 100644 --- a/docs/source/usersguide/depletion.rst +++ b/docs/source/usersguide/depletion.rst @@ -296,12 +296,12 @@ units of barns:: model.materials[0], chain_file) -If you are runnnig :meth:`~openmc.deplete.MicroXS.from_model()` on a cluster -that does not share local filesystems across nodes, you'll need to set an +If you are running :meth:`~openmc.deplete.MicroXS.from_model()` on a cluster +where temporary files are created on a local filesystem that is not shared across nodes, you'll need to set an environment variable so that each MPI process knows where to store output files used to calculate the microscopic cross sections. In order of priority, they -are `TMPDIR`. `TEMP`, and `TMP`. Users interested in further details can read -the `relevant docpage on the tempfile pacakge `_ +are :envvar:`TMPDIR`. :envvar:`TEMP`, and :envvar:`TMP`. Users interested in further details can read +the documentation for the `tempfile `_ module. Caveats 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 2/2] 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 2450702374..c1f9bf94a3 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