diff --git a/docs/source/methods/depletion.rst b/docs/source/methods/depletion.rst index 66ee50397..308a62e19 100644 --- a/docs/source/methods/depletion.rst +++ b/docs/source/methods/depletion.rst @@ -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` diff --git a/docs/source/pythonapi/deplete.rst b/docs/source/pythonapi/deplete.rst index 9419244fe..a57ec915d 100644 --- a/docs/source/pythonapi/deplete.rst +++ b/docs/source/pythonapi/deplete.rst @@ -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 -------------------- diff --git a/docs/source/usersguide/depletion.rst b/docs/source/usersguide/depletion.rst index cc59efa6c..5052f7263 100644 --- a/docs/source/usersguide/depletion.rst +++ b/docs/source/usersguide/depletion.rst @@ -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 `_ 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