diff --git a/examples/python/pincell_depletion/run_depletion.py b/examples/python/pincell_depletion/run_depletion.py
index 8c12211bc..a17bf5fd9 100644
--- a/examples/python/pincell_depletion/run_depletion.py
+++ b/examples/python/pincell_depletion/run_depletion.py
@@ -49,7 +49,7 @@ borated_water.add_element('O', 2.4e-2)
 borated_water.add_s_alpha_beta('c_H_in_H2O')
 
 ###############################################################################
-#                 Exporting to OpenMC geometry.xml file
+#                             Create geometry
 ###############################################################################
 
 # Instantiate ZCylinder surfaces
@@ -94,7 +94,7 @@ root.add_cells([fuel, gap, clad, water])
 geometry = openmc.Geometry(root)
 
 ###############################################################################
-#                   Exporting to OpenMC materials.xml file
+#                     Volumes of depletable materials
 ###############################################################################
 
 # Compute cell areas
@@ -105,7 +105,7 @@ area[fuel] = np.pi * fuel_or.coefficients['R'] ** 2
 uo2.volume = area[fuel]
 
 ###############################################################################
-#                   Exporting to OpenMC settings.xml file
+#                  Transport calculation settings
 ###############################################################################
 
 # Instantiate a Settings object, set all runtime parameters, and export to XML
diff --git a/openmc/deplete/results_list.py b/openmc/deplete/results_list.py
index 58e4b66dc..a12b16ab4 100644
--- a/openmc/deplete/results_list.py
+++ b/openmc/deplete/results_list.py
@@ -103,3 +103,18 @@ class ResultsList(list):
             eigenvalue[i] = result.k[0]
 
         return time, eigenvalue
+
+    def transfer_volumes(self, geometry):
+        """Transfers volumes from depletion results to geometry
+
+        Parameters
+        ----------
+        geometry : OpenMC geometry to be used in a depletion restart
+            calculation
+
+        """
+        cells = geometry.get_all_material_cells()
+        for c in cells:
+            material = next(iter(cells[c].get_all_materials().values()))
+            if material.depletable == True:
+                material.volume = self[0].volume[str(material.id)]
diff --git a/openmc/material.py b/openmc/material.py
index cfa3eba3c..dac2fe14a 100644
--- a/openmc/material.py
+++ b/openmc/material.py
@@ -467,7 +467,7 @@ class Material(IDManagerMixin):
             raise ValueError(msg)
 
         # Generally speaking, the density for a macroscopic object will
-        # be 1.0.  Therefore, lets set density to 1.0 so that the user
+        # be 1.0. Therefore, lets set density to 1.0 so that the user
         # doesnt need to set it unless its needed.
         # Of course, if the user has already set a value of density,
         # then we will not override it.
diff --git a/src/summary.F90 b/src/summary.F90
index bd6ef7158..de3d20a17 100644
--- a/src/summary.F90
+++ b/src/summary.F90
@@ -440,6 +440,7 @@ contains
 
       if (m % depletable) then
         call write_attribute(material_group, "depletable", 1)
+        call write_attribute(material_group, "volume", m % volume)
       else
         call write_attribute(material_group, "depletable", 0)
       end if
@@ -500,7 +501,7 @@ contains
         call write_dataset(material_group, "nuclides", nuc_names)
         ! Deallocate temporary array
         deallocate(nuc_names)
-        ! Write atom densities
+        ! Write nuclide atom densities
         call write_dataset(material_group, "nuclide_densities", nuc_densities)
         deallocate(nuc_densities)
       end if