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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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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