crp/OpenMC
1
0
Fork 0
mirror of https://github.com/openmc-dev/openmc.git synced 2026-07-28 14:15:42 -04:00
Code Issues Projects Releases Packages Wiki Activity Actions 5

Improve documentation for FissionYieldHelpers

Browse source 
Add a separate section in the documentation describing
the helpers.

Reword purpose of mat_indexes in generate_tallies
This commit is contained in:
Andrew Johnson 2019-08-14 18:02:27 -05:00
parent 454e6cc6b6
commit aa86aa5262
No known key found for this signature in database
GPG key ID: 253418E91B7F6FEB
5 changed files with 44 additions and 25 deletions
Download patch file Download diff file Expand all files Collapse all files

19
docs/source/pythonapi/deplete.rst
View file

@ -77,16 +77,25 @@ data, such as number densities and reaction rates for each material.
:template: myclass.rst
AtomNumber
AveragedFissionYieldHelper
ChainFissionHelper
ConstantFissionYieldHelper
DirectReactionRateHelper
FissionYieldCutoffHelper
OperatorResult
ReactionRates
Results
ResultsList
The following classes are used to help the :class:`openmc.deplete.Operator`
compute quantities like effective fission yields, reaction rates, and
total system energy.
.. autosummary::
:toctree: generated
:nosignatures:
:template: myclass.rst
helpers.AveragedFissionYieldHelper
helpers.ChainFissionHelper
helpers.ConstantFissionYieldHelper
helpers.DirectReactionRateHelper
helpers.FissionYieldCutoffHelper
The following classes are abstract classes that can be used to extend the
:mod:`openmc.deplete` capabilities:

14
openmc/deplete/abc.py
View file

@ -443,8 +443,11 @@ class FissionYieldHelper(ABC):
materials : iterable of C-API materials
Materials to be used in :class:`openmc.capi.MaterialFilter`
mat_indexes : iterable of int
Indexes for materials in ``materials`` tracked on this
process
Indices of tallied materials that will have their fission
yields computed by this helper. Necessary as the
:class:`openmc.deplete.Operator` that uses this helper
may only burn a subset of all materials when running
in parallel mode.
"""
def update_nuclides_from_operator(self, nuclides):
@ -526,8 +529,11 @@ class TalliedFissionYieldHelper(FissionYieldHelper):
materials : iterable of :class:`openmc.capi.Material`
Materials to be used in :class:`openmc.capi.MaterialFilter`
mat_indexes : iterable of int
Indexes for materials in ``materials`` tracked on this
process
Indices of tallied materials that will have their fission
yields computed by this helper. Necessary as the
:class:`openmc.deplete.Operator` that uses this helper
may only burn a subset of all materials when running
in parallel mode.
"""
self._local_indexes = asarray(mat_indexes)

19
openmc/deplete/helpers.py
View file

@ -376,8 +376,11 @@ class FissionYieldCutoffHelper(TalliedFissionYieldHelper):
materials : iterable of :class:`openmc.capi.Material`
Materials to be used in :class:`openmc.capi.MaterialFilter`
mat_indexes : iterable of int
Indexes for materials in ``materials`` tracked on this
process
Indices of tallied materials that will have their fission
yields computed by this helper. Necessary as the
:class:`openmc.deplete.Operator` that uses this helper
may only burn a subset of all materials when running
in parallel mode.
"""
super().generate_tallies(materials, mat_indexes)
energy_filter = EnergyFilter()
@ -465,8 +468,9 @@ class AveragedFissionYieldHelper(TalliedFissionYieldHelper):
Conversely, if the average energy is above the highest energy
with yield data, that set of fission yields is used.
For the case where the average energy is between two sets
of yields, a geometric mean of the yield distributions is
used.
of yields, the effective fission yield computed by
linearly interpolating between yields provided at the
nearest energies above and below the average.
Parameters
----------
@ -502,8 +506,11 @@ class AveragedFissionYieldHelper(TalliedFissionYieldHelper):
materials : iterable of :class:`openmc.capi.Material`
Materials to be used in :class:`openmc.capi.MaterialFilter`
mat_indexes : iterable of int
Indexes for materials in ``materials`` tracked on this
process
Indices of tallied materials that will have their fission
yields computed by this helper. Necessary as the
:class:`openmc.deplete.Operator` that uses this helper
may only burn a subset of all materials when running
in parallel mode.
"""
super().generate_tallies(materials, mat_indexes)
fission_tally = self._fission_rate_tally

2
openmc/deplete/nuclide.py
View file

@ -236,7 +236,7 @@ class Nuclide(object):
class FissionYieldDistribution(Mapping):
"""Class for storing energy-dependent fission yields for a single nuclide
"""Energy-dependent fission product yields for a single nuclide
Can be used as a dictionary mapping energies and products to fission
yields::

15
openmc/deplete/operator.py
View file

@ -86,19 +86,16 @@ class Operator(TransportOperator):
in initial condition to ensure they exist in the decay chain.
Only done for nuclides with reaction rates.
Defaults to 1.0e3.
fission_yield_mode : str, {"constant", "cutoff"}
fission_yield_mode : ("constant", "cutoff", "average")
Key indicating what fission product yield scheme to use. The
key determines what fission energy helper is used::
key determines what fission energy helper is used:
* "constant": :class:`openmc.deplete.ConstantFissionYieldHelper`
* "cutoff": :class:`openmc.deplete.FissionYieldCutoffHelper`
* "average": :class:`openmc.deplete.AveragedFissionYieldHelper`
* "constant": :class:`~openmc.deplete.helpers.ConstantFissionYieldHelper`
* "cutoff": :class:`~openmc.deplete.helpers.FissionYieldCutoffHelper`
* "average": :class:`~openmc.deplete.helpers.AveragedFissionYieldHelper`
The documentation on these classes describe their methodology
and differences. ``"constant"`` will treat fission yields as
constant with respect to energy, while ``"cutoff"`` will
compute effective yields based on the number of fission events
above and below a cutoff. Default: ``"constant"``
and differences. Default: ``"constant"``
fission_yield_opts : dict of str to option, optional
Optional arguments to pass to the helper determined by
``fission_yield_mode``. Will be passed directly on to the