From efbd375fcee23c6095fb5e63f578e27d65944b3d Mon Sep 17 00:00:00 2001 From: Paul Romano Date: Mon, 10 Aug 2020 08:03:23 -0500 Subject: [PATCH] Add note about depletable attribute in user's guide documentation --- docs/source/usersguide/depletion.rst | 8 ++++++++ 1 file changed, 8 insertions(+) diff --git a/docs/source/usersguide/depletion.rst b/docs/source/usersguide/depletion.rst index ad1ba779e..d550222cf 100644 --- a/docs/source/usersguide/depletion.rst +++ b/docs/source/usersguide/depletion.rst @@ -28,6 +28,8 @@ operator class requires a :class:`openmc.Geometry` instance and a op = openmc.deplete.Operator(geom, settings) +Any material that contains a fissionable nuclide is depleted by default, but +this can behavior can be changed with the :attr:`Material.depletable` attribute. :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 @@ -63,6 +65,12 @@ need to set the run mode:: settings.run_mode = 'fixed source' +Additionally, all materials that you wish to deplete need to be marked as such +using the :attr:`Material.depletable` attribute:: + + mat = openmc.Material() + mat.depletable = True + When constructing the :class:`~openmc.deplete.Operator`, you should indicate that normalization of tally results will be done based on the source rate rather than a power or power density::