Documentation updates

docs/source/pythonapi/data.rst

@@ -63,6 +63,7 @@ Core Functions
     atomic_weight
     combine_distributions
     decay_constant
+    decay_photon_source
     dose_coefficients
     gnd_name
     half_life

openmc/data/decay.py

@@ -579,13 +579,15 @@ class Decay(EqualityMixin):
 _DECAY_PHOTON_SOURCE = {}
 
 
-def decay_photon_source(nuclide):
+def decay_photon_source(nuclide: str) -> Univariate:
     """Get photon source from the decay of a nuclide
 
     This function relies on data stored in a depletion chain. Before calling it
     for the first time, you need to ensure that a depletion chain has been
     specified in openmc.config['chain_file'].
 
+    .. versionadded:: 0.14.0
+
     Parameters
     ----------
     nuclide : str

openmc/deplete/stepresult.py

@@ -202,6 +202,8 @@ class StepResult:
     def get_material(self, mat_id):
         """Return material object for given depleted composition
 
+        .. versionadded:: 0.14.0
+
         Parameters
         ----------
         mat_id : str

openmc/material.py

@@ -18,6 +18,7 @@ import openmc.checkvalue as cv
 from ._xml import clean_indentation, reorder_attributes
 from .mixin import IDManagerMixin
 from openmc.checkvalue import PathLike
+from openmc.stats import Univariate
 
 
 # Units for density supported by OpenMC
@@ -34,10 +35,10 @@ class Material(IDManagerMixin):
     To create a material, one should create an instance of this class, add
     nuclides or elements with :meth:`Material.add_nuclide` or
     :meth:`Material.add_element`, respectively, and set the total material
-    density with :meth:`Material.set_density()`. Alternatively, you can
-    use :meth:`Material.add_components()` to pass a dictionary
-    containing all the component information. The material can then be
-    assigned to a cell using the :attr:`Cell.fill` attribute.
+    density with :meth:`Material.set_density()`. Alternatively, you can use
+    :meth:`Material.add_components()` to pass a dictionary containing all the
+    component information. The material can then be assigned to a cell using the
+    :attr:`Cell.fill` attribute.
 
     Parameters
     ----------
@@ -91,6 +92,13 @@ class Material(IDManagerMixin):
     fissionable_mass : float
         Mass of fissionable nuclides in the material in [g]. Requires that the
         :attr:`volume` attribute is set.
+    decay_photon_source : openmc.stats.Univariate
+        Energy distribution of photons emitted from decay of unstable nuclides
+        within the material. The integral of this distribution is the total
+        intensity of the photon source.
+
+        .. versionadded:: 0.14.0
+
     """
 
     next_id = 1
@@ -256,7 +264,7 @@ class Material(IDManagerMixin):
         return density*self.volume
 
     @property
-    def decay_photon_source(self):
+    def decay_photon_source(self) -> Univariate:
         atoms = self.get_nuclide_atoms()
         dists = []
         probs = []
@@ -832,7 +840,7 @@ class Material(IDManagerMixin):
             for nuclide in self._nuclides:
                 if nuclide.name not in matching_nuclides:
                     matching_nuclides.append(nuclide.name)
-                        
+
         return matching_nuclides
 
     def get_nuclide_densities(self):