From 4f9bbca7f338ceedf97199b23724e021c0f20385 Mon Sep 17 00:00:00 2001 From: Paul Romano Date: Mon, 8 Mar 2021 09:20:51 -0600 Subject: [PATCH] Add note about specifying volume for depletion --- docs/source/usersguide/depletion.rst | 6 ++++++ 1 file changed, 6 insertions(+) diff --git a/docs/source/usersguide/depletion.rst b/docs/source/usersguide/depletion.rst index d550222cf..6fa2b89c9 100644 --- a/docs/source/usersguide/depletion.rst +++ b/docs/source/usersguide/depletion.rst @@ -30,6 +30,12 @@ operator class requires a :class:`openmc.Geometry` instance and a Any material that contains a fissionable nuclide is depleted by default, but this can behavior can be changed with the :attr:`Material.depletable` attribute. + +.. important:: The volume must be specified for each material that is depleted by + setting the :attr:`Material.volume` attribute. This is necessary + in order to calculate the proper normalization of tally results + based on the source rate. + :mod:`openmc.deplete` supports multiple time-integration methods for determining material compositions over time. Each method appears as a different class. For example, :class:`openmc.deplete.CECMIntegrator` runs a depletion calculation