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doc adjustments from @paulromano's comments

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yardasol 2022-08-01 12:20:46 -05:00
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2
docs/source/methods/depletion.rst
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@ -231,7 +231,7 @@ of FPY:
to transport-coupled depletion.
3. Compute the average energy at which fission events occur and use an effective
FPY by linearly interpolating between FPY provided at neighboring energies.
Only applicable to transport-coupled depletion
Only applicable to transport-coupled depletion.
The method for transport-coupled depletion can be selected through the
``fission_yield_mode`` argument to the :class:`openmc.deplete.CoupledOperator`

6
docs/source/pythonapi/deplete.rst
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@ -15,7 +15,7 @@ are:
1) A depletion operator
2) A time-integration scheme
The former is responsible for calcuating retaining important information required for depletion. The most common examples are reaction rates and power
The former is responsible for calcuating and retaining important information required for depletion. The most common examples are reaction rates and power
normalization data. The latter is responsible for projecting reaction rates and
compositions forward in calendar time across some step size :math:`\Delta t`,
and obtaining new compositions given a power or power density. The
@ -194,8 +194,8 @@ total system energy.
helpers.FissionYieldCutoffHelper
helpers.FluxCollapseHelper
The :class:`openmc.deplete.IndependentOperator` uses inner class subclassed from
those listed to perform similar calculations.
The :class:`openmc.deplete.IndependentOperator` uses inner classes subclassed
from those listed above to perform similar calculations.
Intermediate Classes
--------------------

21
docs/source/usersguide/depletion.rst
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@ -8,7 +8,7 @@ OpenMC supports transport-coupled and transport-independent depletion, or
burnup, calculations through the :mod:`openmc.deplete` Python module. OpenMC
uses transmutation reaction rates to solve a set of transmutation equations
that determine the evolution of nuclide densities within a material. The
nuclide densities predicted as some future time are then used to determine
nuclide densities predicted at some future time are then used to determine
updated reaction rates, and the process is repeated for as many timesteps as
are requested.
@ -197,8 +197,8 @@ Transport-independent depletion
.. note::
This feature is still under heavy development and has yet to be verifed
code-to-code . API changes and feature additions are possible and likely in
This feature is still under heavy development and has yet to be rigorously
verified. API changes and feature additions are possible and likely in
the near future.
This category of operator uses pre-calculated one-group microscopic cross
@ -282,7 +282,7 @@ expects the units to be. The :class:`~openmc.deplete.MicroXS` class also include
[0.01, 0.5]])
micro_xs = MicroXS.from_array(nuclides, reactions, data)
.. important ::
.. important::
Both :meth:`~openmc.deplete.MicroXS.from_csv()` and
:meth:`~openmc.deplete.MicroXS.from_array()` assume the cross section values
@ -306,12 +306,12 @@ normalizing reaction rates:
``normalization_mode == source-rate``, and use ``power`` or ``power_density``
when ``normalization_mode == fission-q``.
1. ``soure-rate`` normalization, which assumes the ``source-rate`` provided by
1. ``source-rate`` normalization, which assumes the ``source_rate`` provided by
the time integrator is a flux, and obtains the reaction rates by multiplying
the cross-sections by the ``source-rate``.
2. ``fission-q`` normalization, which assumes the ``source-rate`` provided by
the time integrator is a power, and obtains the reaction rates by computing a
value for the flux based on this power. The general equation for the flux is
2. ``fission-q`` normalization, which uses the ``power`` or ``power_density``
provided by the time integrator to obtain reaction rates by computing a value
for the flux based on this power. The general equation for the flux is
.. math::
@ -324,8 +324,7 @@ normalizing reaction rates:
However, there is a method to converge to a more accurate value for flux by
using substeps during time integration.
`This paper <https://doi.org/10.1016/j.anucene.2016.05.031>`_ provides a
good discussion of this method. Hopefully such a method will be implemented
in OpenMC in the near future.
good discussion of this method.
.. warning::
@ -342,7 +341,7 @@ separate with respect to reaction rates. This can be useful for running many
different cases of a particular scenario. However, running a depletion
simulation with multiple materials using the ``fission-q`` normalization method
treats each material as part of the same "reactor" due to how ``fission-q``
normalization conglomerates energy values from each material to a single value.
normalization accumulates energy values from each material to a single value.
This behavior may change in the future.
Time integration