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update user guide

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yardasol 2022-07-21 12:27:36 -05:00
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18
docs/source/usersguide/depletion.rst
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@ -188,9 +188,19 @@ Transport-independent depletion
possible and likely in the near future.
OpenMC supports running depletion calculations independent of the OpenMC
transport solver using the :class:`~openmc.deplete.FluxDepletionOperator` class. This class
has two ways to initalize it; the default constructor accepts an
:class:`openmc.Materials` object, a flux spectra, and one-group microscopic
transport solver using the :class:`~openmc.deplete.FluxDepletionOperator` class.
This class supports both constant-flux and constant-power depletion.
.. important::
Make sure you set the correct parameter in the :class:`openmc.abc.Integrator` class. Use the ``flux`` parameter when ``normalization_mode == constant-flux``, and use ``power`` or ``power_density`` when ``normalization_mode == constant-power``.
.. warning::
The accuracy of results when using ``constant-power`` is entirely dependent on your depletion chain. Make sure it has sufficient data to resolve the dynamics of your particular scenario.
This class has two ways to initalize it; the default constructor accepts an
:class:`openmc.Materials` object and one-group microscopic
cross sections as a :class:`pandas.DataFrame`, while the `from_nuclides`
method accepts a volume and dictionary of nuclide concentrations in place of
the :class:`openmc.Materials` object in addition to the other parameters.
@ -201,7 +211,7 @@ or from data arrays::
# load in the microscopic cross sections
micro_xs = FluxDepletionOperator.create_micro_xs_from_csv(micro_xs_path)
flux = 1.16e15
op = FluxDepletionOperator(materials, micro_xs, flux, chain_file)
op = FluxDepletionOperator(materials, micro_xs, chain_file)
# alternate construtor
nuclides = {'U234': 8.92e18,