Inclusion of spontaneous fission yields in the Bateman equation

docs/source/methods/depletion.rst

@@ -176,14 +176,16 @@ only solving the transport equation to estimate transmutation reaction rates
 sections (in the case of transport-independent depletion), but also a series of
 choices about what data to include. In OpenMC, the burnup matrix is constructed
 based on data inside of a *depletion chain* file, which includes fundamental
-data gathered from ENDF incident neutron, decay, and fission product yield
-sublibraries. For each nuclide, this file includes:
+data gathered from ENDF incident neutron, decay, fission product yield and
+spontaneous fission yield sublibraries. For each nuclide, this file includes:
 
 - What transmutation reactions are possible, their Q values, and their products;
 - If a nuclide is not stable, what decay modes are possible, their branching
-  ratios, and their products; and
+  ratios, and their products;
 - If a nuclide is fissionable, the fission products yields at any number of
-  incident neutron energies.
+  incident neutron energies; and
+- If a nuclide can undergo spontanous fission, the products yields for
+  spontaneous fission.
 
 Transmutation Reactions
 -----------------------

docs/source/pythonapi/deplete.rst

@@ -135,6 +135,19 @@ The :class:`Chain` class uses information from the following module variable:
 
    :type: dict
 
+It contains the following switch:
+
+.. data:: chain.INCLUDE_SPONT_FISSION
+
+   Boolean switch to enable or disable the use of spontaneous fission
+   product yields when solving the Bateman equations. The default is to 
+   include spontaneous fission product yields if they are present in
+   the Chain object. 
+
+   :type: bool
+
+
+
 The following classes are used during a depletion simulation and store auxiliary
 data, such as number densities and reaction rates for each material.
 

docs/source/usersguide/depletion.rst

@@ -107,14 +107,15 @@ Caveats
 
 .. _energy-deposition:
 
-Energy Deposition
+Energy Deposition and Spontaneous Fission
 ~~~~~~~~~~~~~~~~~
 
 The default energy deposition mode, ``"fission-q"``, instructs the
 :class:`~openmc.deplete.CoupledOperator` to normalize reaction rates using the
 product of fission reaction rates and fission Q values taken from the depletion
 chain. This approach does not consider indirect contributions to energy
-deposition, such as neutron heating and energy from secondary photons. In doing
+deposition, such as neutron heating and energy from secondary photons, nor does it 
+consider energy released in spontaneous fission. In doing
 this, the energy deposited during a transport calculation will be lower than
 expected. This causes the reaction rates to be over-adjusted to hit the
 user-specific power, or power density, leading to an over-depletion of burnable
@@ -151,6 +152,13 @@ These modified heating libraries can be generated by running the latest version
 of :meth:`openmc.data.IncidentNeutron.from_njoy()`, and will eventually be bundled
 into the distributed libraries.
 
+By default, depletion calculations include fission yields from spontaneous fission
+decay. Such decay should result in the emission of neutrons but a spontaneous fission 
+source term is not included in the Boltzmann equation in current version of OpenMC.
+In practice, this should be a minor issue as neutron-induced fission is much more 
+important than spontaneous fission for the sourcing of neutrons. 
+
+
 Local Spectra and Repeated Materials
 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~