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Made changes from paulromano's 2nd round of comments

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- Syntax improvements and fixes
- removed `flux` parameter from Integrator
- Moved generate_1g_cross_sections to a top level function in
  flux_operator.py
- changed normalization modes: constant-flux -> source-rate;
  constant-power -> fission-q
- fixed regression tests (fission-q reference solution was bad before,
  but is now much more reasonable and comparable to the source-rate
  reference solution)
- added `nuc_units` parameter to the `from_nuclides` method.
- docstring fixes
- RST doc fixes
- spelling fixes
This commit is contained in:
yardasol 2022-07-27 12:27:16 -05:00
parent 8102226e70
commit 4818a296d4
11 changed files with 178 additions and 175 deletions
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16
docs/source/usersguide/depletion.rst
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@ -19,10 +19,10 @@ transmutation equations and the method used for advancing time. At present, the
:class:`openmc.deplete.Operator` (which uses the OpenMC transport solver), but
in principle additional operator classes based on other transport codes could be
implemented and no changes to the depletion solver itself would be needed. The
operator class requires a :class:`openmc.model.Model` instance containing
operator class requires a :class:`~openmc.Model` instance containing
material, geometry, and settings information::
model = openmc.model.Model()
model = openmc.Model()
...
op = openmc.deplete.Operator(model)
@ -189,17 +189,18 @@ Transport-independent depletion
OpenMC supports running depletion calculations independent of the OpenMC
transport solver using the :class:`~openmc.deplete.FluxDepletionOperator` class.
This class supports both constant-flux and constant-power depletion.
This class supports both constant-flux (``source-rate`` normalization) and
constant-power depletion (``fission-q`` normalization).
.. 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``.
Make sure you set the correct parameter in the :class:`openmc.abc.Integrator` class. Use the ``source_rates`` parameter when ``normalization_mode == source-rate``, and use ``power`` or ``power_density`` when ``normalization_mode == fission-q``.
.. 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.
The accuracy of results when using ``fission-q`` 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
This class has two ways to initialize 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
@ -221,7 +222,8 @@ or from data arrays::
'O16': 4.64e22,
'O17': 1.76e19}
volume = 0.5
op = FluxDepletionOperator.from_nuclides(volume, nuclides, micro_xs, flux, chain_file)
op = FluxDepletionOperator.from_nuclides(volume, nuclides, 'atom/cm3',
micro_xs, flux, chain_file)
A user can then define an integrator class as they would for a coupled
transport-depletion calculation and follow the same steps from there.