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Address paulromano's comments

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- various syntax adjustments and cleanups
- remove unneeded import statements
- add more information to new section in user's guide
- typo fixes
- remove dilute_initial
- move _validate_micro_xs_inputs back into the class as a static method
- update tests to reflect changes
This commit is contained in:
yardasol 2022-07-12 11:44:12 -05:00
parent af3d460665
commit 5f8e1ff9d9
4 changed files with 123 additions and 66 deletions
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34
docs/source/usersguide/depletion.rst
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@ -181,25 +181,25 @@ Transport-independent depletion
.. note::
This is a brand-new feature and is under heavy development. API changes are
This feature is still under heavy development. API changes are
possible and likely in the near future.
OpenMC also supports transport-independent depletion calculations using the
:class:`FluxDepletionOperator` class. Rather than taking a
:class:`openmc.model.Model` object, this class accepts a volume,
OpenMC supports running depletion calculations independent of the OpenMC
transport solver using the :class:`FluxDepletionOperator` class. Rather than
taking a :class:`openmc.model.Model` object, this class accepts a volume,
a dictionary of nuclide concentrations, a flux spectra, and one-group
microscopic cross sections as a pandas dataframe. The class includes
helper functions to constructe the dataframe from a csv file or from
microscopic cross sections as a :class:`pandas.DataFrame`. The class includes
helper functions to construct the dataframe from a csv file or from
data arrays::
...
micro_xs = FluxDepletionOperator.create_micro_xs_from_csv(micro_xs_path)
nuclides = {'U234':8.92e18,
'U235':9.98e20,
'U238':2.22e22,
'U236':4.57e18,
'O16':4.64e22,
'O17':1.76e19}
nuclides = {'U234': 8.92e18,
'U235': 9.98e20,
'U238': 2.22e22,
'U236': 4.57e18,
'O16': 4.64e22,
'O17': 1.76e19}
volume = 0.5
flux = 1.16e15
@ -207,8 +207,10 @@ data arrays::
A user can then define an integrator class as they would for a coupled
transport-depletion calculation and follow the steps from there.
present in the depletion chain.
transport-depletion calculation and follow the same steps from there.
.. note:: Ideally, one-group cross section data should be available for every reaction
in the depletion chain.
.. note:: Ideally, one-group cross section data should be available for every
reaction in the depletion chain. If a nuclide that has a reaction
associated with it in the depletion chain is present in the `nuclides`
parameter but not the cross section data, that reaction will not be
simulated.