crp/OpenMC
1
0
Fork 0
mirror of https://github.com/openmc-dev/openmc.git synced 2026-07-27 05:35:49 -04:00
Code Issues Projects Releases Packages Wiki Activity Actions

Add note about depletable attribute in user's guide documentation

Browse source
This commit is contained in:
Paul Romano 2020-08-10 08:03:23 -05:00
parent c3968ff9a9
commit efbd375fce
1 changed files with 8 additions and 0 deletions
Download patch file Download diff file Expand all files Collapse all files

8
docs/source/usersguide/depletion.rst
View file

@ -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::