diff --git a/openmc/cell.py b/openmc/cell.py
index 9055f10e59..23246a0f68 100644
--- a/openmc/cell.py
+++ b/openmc/cell.py
@@ -85,6 +85,10 @@ class Cell(object):
         Array of offsets used for distributed cell searches
     distribcell_index : int
         Index of this cell in distribcell arrays
+    volume_information : dict
+        Estimate of the volume and total number of atoms of each nuclide from a
+        stochastic volume calculation. This information is set with the
+        :meth:`Cell.add_volume_information` method.
 
     """
 
@@ -100,6 +104,7 @@ class Cell(object):
         self._translation = None
         self._offsets = None
         self._distribcell_index = None
+        self._volume_information = None
 
     def __contains__(self, point):
         if self.region is None:
@@ -212,6 +217,10 @@ class Cell(object):
     def distribcell_index(self):
         return self._distribcell_index
 
+    @property
+    def volume_information(self):
+        return self._volume_information
+
     @id.setter
     def id(self, cell_id):
         if cell_id is None:
@@ -351,6 +360,22 @@ class Cell(object):
             else:
                 self.region = Intersection(self.region, region)
 
+    def add_volume_information(self, volume_calc):
+        """Add volume information to a cell.
+
+        Parameters
+        ----------
+        volume_calc : openmc.VolumeCalculation
+            Results from a stochastic volume calculation
+
+        """
+        for cell_id in volume_calc.results:
+            if cell_id == self.id:
+                self._volume_information = volume_calc.results[cell_id]
+                break
+        else:
+            raise ValueError('No volume information found for this cell.')
+
     def get_cell_instance(self, path, distribcell_index):
 
         # If the Cell is filled by a Material
@@ -368,8 +393,19 @@ class Cell(object):
 
         return offset
 
-    def get_all_nuclides(self):
-        """Return all nuclides contained in the cell
+    def get_nuclides(self):
+        """Returns all nuclides in the cell
+
+        Returns
+        -------
+        nuclides : list
+            List of nuclide names
+
+        """
+        return self.fill.get_nuclides() if self.fill_type != 'void' else []
+
+    def get_nuclide_densities(self):
+        """Return all nuclides contained in the cell and their densities
 
         Returns
         -------
@@ -381,8 +417,24 @@ class Cell(object):
 
         nuclides = OrderedDict()
 
-        if self.fill_type != 'void':
-            nuclides.update(self.fill.get_all_nuclides())
+        if self.fill_type == 'material':
+            nuclides.update(self.fill.get_nuclide_densities())
+        elif self.fill_type == 'void':
+            pass
+        else:
+            if self.volume_information is not None:
+                volume = self.volume_information['volume'][0]
+                for full_name, atoms in self.volume_information['atoms']:
+                    name, xs = full_name.split('.')
+                    nuclide = openmc.Nuclide(name, xs)
+                    density = 1.0e-24 * atoms[0]/volume  # density in atoms/b-cm
+                    nuclides[name] = (nuclide, density)
+            else:
+                raise RuntimeError(
+                    'Volume information is needed to calculate microscopic cross '
+                    'sections for cell {}. This can be done by running a '
+                    'stochastic volume calculation via the '
+                    'openmc.VolumeCalculation object'.format(self.id))
 
         return nuclides
 
diff --git a/openmc/geometry.py b/openmc/geometry.py
index 14fd48fb1e..8314a9049c 100644
--- a/openmc/geometry.py
+++ b/openmc/geometry.py
@@ -50,6 +50,19 @@ class Geometry(object):
 
         self._root_universe = root_universe
 
+    def add_volume_information(self, volume_calc):
+        """Add volume information to from a stochastic volume calculation.
+
+        Parameters
+        ----------
+        volume_calc : openmc.VolumeCalculation
+            Results from a stochastic volume calculation
+
+        """
+        for cell in self.get_all_cells():
+            if cell.id in volume_calc.results:
+                cell.add_volume_information(volume_calc)
+
     def export_to_xml(self):
         """Create a geometry.xml file that can be used for a simulation.
 
@@ -166,24 +179,6 @@ class Geometry(object):
         universes.sort(key=lambda x: x.id)
         return universes
 
-    def get_all_nuclides(self):
-        """Return all nuclides assigned to a material in the geometry
-
-        Returns
-        -------
-        list of openmc.Nuclide
-            Nuclides in the geometry
-
-        """
-
-        nuclides = OrderedDict()
-        materials = self.get_all_materials()
-
-        for material in materials:
-            nuclides.update(material.get_all_nuclides())
-
-        return nuclides
-
     def get_all_materials(self):
         """Return all materials assigned to a cell
 
diff --git a/openmc/lattice.py b/openmc/lattice.py
index 81144e4d81..b26aab4c67 100644
--- a/openmc/lattice.py
+++ b/openmc/lattice.py
@@ -144,25 +144,26 @@ class Lattice(object):
 
         return univs
 
-    def get_all_nuclides(self):
-        """Return all nuclides contained in the lattice
+    def get_nuclides(self):
+        """Returns all nuclides in the lattice
 
         Returns
         -------
-        nuclides : collections.OrderedDict
-            Dictionary whose keys are nuclide names and values are 2-tuples of
-            (nuclide, density)
+        nuclides : list
+            List of nuclide names
 
         """
 
-        nuclides = OrderedDict()
+        nuclides = []
 
         # Get all unique Universes contained in each of the lattice cells
         unique_universes = self.get_unique_universes()
 
         # Append all Universes containing each cell to the dictionary
-        for universe_id, universe in unique_universes.items():
-            nuclides.update(universe.get_all_nuclides())
+        for universe in unique_universes.values():
+            for nuclide in universe.get_nuclides():
+                if nuclide not in nuclides:
+                    nuclides.append(nuclide)
 
         return nuclides
 
diff --git a/openmc/material.py b/openmc/material.py
index d7ffd04e94..5603d9d25c 100644
--- a/openmc/material.py
+++ b/openmc/material.py
@@ -492,9 +492,31 @@ class Material(object):
         for element, percent, percent_type in self._elements:
             element.scattering = 'iso-in-lab'
 
-    def get_all_nuclides(self):
+    def get_nuclides(self):
         """Returns all nuclides in the material
 
+        Returns
+        -------
+        nuclides : list
+            List of nuclide names
+
+        """
+
+        nuclides = []
+
+        for nuclide, density, density_type in self._nuclides:
+            nuclides.append(nuclide.name)
+
+        for element, density, density_type in self._elements:
+            # Expand natural element into isotopes
+            for isotope, abundance in element.expand():
+                nuclides.append(isotope.name)
+
+        return nuclides
+
+    def get_nuclide_densities(self):
+        """Returns all nuclides in the material and their densities
+
         Returns
         -------
         nuclides : dict
diff --git a/openmc/mgxs/library.py b/openmc/mgxs/library.py
index 84571ce236..ee11d0ef64 100644
--- a/openmc/mgxs/library.py
+++ b/openmc/mgxs/library.py
@@ -1008,7 +1008,7 @@ class Library(object):
         # Create the xsdata object and add it to the mgxs_file
         for i, domain in enumerate(self.domains):
             if self.by_nuclide:
-                nuclides = list(domain.get_all_nuclides().keys())
+                nuclides = domain.get_nuclides()
             else:
                 nuclides = ['total']
             for nuclide in nuclides:
diff --git a/openmc/mgxs/mgxs.py b/openmc/mgxs/mgxs.py
index e1c2e62217..7aa12d0418 100644
--- a/openmc/mgxs/mgxs.py
+++ b/openmc/mgxs/mgxs.py
@@ -288,9 +288,7 @@ class MGXS(object):
 
                 # If this is a by-nuclide cross-section, add nuclides to Tally
                 if self.by_nuclide and score != 'flux':
-                    all_nuclides = self.get_all_nuclides()
-                    for nuclide in all_nuclides:
-                        self._tallies[key].nuclides.append(nuclide)
+                    self._tallies[key].nuclides += self.get_nuclides()
                 else:
                     self._tallies[key].nuclides.append('total')
 
@@ -329,14 +327,14 @@ class MGXS(object):
     @property
     def num_nuclides(self):
         if self.by_nuclide:
-            return len(self.get_all_nuclides())
+            return len(self.get_nuclides())
         else:
             return 1
 
     @property
     def nuclides(self):
         if self.by_nuclide:
-            return self.get_all_nuclides()
+            return self.get_nuclides()
         else:
             return 'sum'
 
@@ -503,7 +501,7 @@ class MGXS(object):
         mgxs.name = name
         return mgxs
 
-    def get_all_nuclides(self):
+    def get_nuclides(self):
         """Get all nuclides in the cross section's spatial domain.
 
         Returns
@@ -528,8 +526,7 @@ class MGXS(object):
 
         # Otherwise, return all nuclides in the spatial domain
         else:
-            nuclides = self.domain.get_all_nuclides()
-            return list(nuclides.keys())
+            return self.domain.get_nuclides()
 
     def get_nuclide_density(self, nuclide):
         """Get the atomic number density in units of atoms/b-cm for a nuclide
@@ -556,7 +553,7 @@ class MGXS(object):
         cv.check_type('nuclide', nuclide, basestring)
 
         # Get list of all nuclides in the spatial domain
-        nuclides = self.domain.get_all_nuclides()
+        nuclides = self.domain.get_nuclide_densities()
 
         if nuclide not in nuclides:
             msg = 'Unable to get density for nuclide "{0}" which is not in ' \
@@ -597,14 +594,14 @@ class MGXS(object):
 
         # Sum the atomic number densities for all nuclides
         if nuclides == 'sum':
-            nuclides = self.get_all_nuclides()
+            nuclides = self.get_nuclides()
             densities = np.zeros(1, dtype=np.float)
             for nuclide in nuclides:
                 densities[0] += self.get_nuclide_density(nuclide)
 
         # Tabulate the atomic number densities for all nuclides
         elif nuclides == 'all':
-            nuclides = self.get_all_nuclides()
+            nuclides = self.get_nuclides()
             densities = np.zeros(self.num_nuclides, dtype=np.float)
             for i, nuclide in enumerate(nuclides):
                 densities[i] += self.get_nuclide_density(nuclide)
@@ -635,7 +632,7 @@ class MGXS(object):
         # If computing xs for each nuclide, replace CrossNuclides with originals
         if self.by_nuclide:
             self.xs_tally._nuclides = []
-            nuclides = self.get_all_nuclides()
+            nuclides = self.get_nuclides()
             for nuclide in nuclides:
                 self.xs_tally.nuclides.append(openmc.Nuclide(nuclide))
 
@@ -796,7 +793,7 @@ class MGXS(object):
         # Construct a collection of the nuclides to retrieve from the xs tally
         if self.by_nuclide:
             if nuclides == 'all' or nuclides == 'sum' or nuclides == ['sum']:
-                query_nuclides = self.get_all_nuclides()
+                query_nuclides = self.get_nuclides()
             else:
                 query_nuclides = nuclides
         else:
@@ -1164,7 +1161,7 @@ class MGXS(object):
         # Construct a collection of the nuclides to report
         if self.by_nuclide:
             if nuclides == 'all':
-                nuclides = self.get_all_nuclides()
+                nuclides = self.get_nuclides()
             elif nuclides == 'sum':
                 nuclides = ['sum']
             else:
@@ -1302,7 +1299,7 @@ class MGXS(object):
         # Construct a collection of the nuclides to report
         if self.by_nuclide:
             if nuclides == 'all':
-                nuclides = self.get_all_nuclides()
+                nuclides = self.get_nuclides()
                 densities = np.zeros(len(nuclides), dtype=np.float)
             elif nuclides == 'sum':
                 nuclides = ['sum']
@@ -1484,7 +1481,7 @@ class MGXS(object):
         if self.by_nuclide and nuclides == 'sum':
 
             # Use tally summation to sum across all nuclides
-            query_nuclides = self.get_all_nuclides()
+            query_nuclides = self.get_nuclides()
             xs_tally = self.xs_tally.summation(nuclides=query_nuclides)
             df = xs_tally.get_pandas_dataframe(
                 distribcell_paths=distribcell_paths)
@@ -1790,7 +1787,7 @@ class MatrixMGXS(MGXS):
         # Construct a collection of the nuclides to retrieve from the xs tally
         if self.by_nuclide:
             if nuclides == 'all' or nuclides == 'sum' or nuclides == ['sum']:
-                query_nuclides = self.get_all_nuclides()
+                query_nuclides = self.get_nuclides()
             else:
                 query_nuclides = nuclides
         else:
@@ -1937,7 +1934,7 @@ class MatrixMGXS(MGXS):
         # Construct a collection of the nuclides to report
         if self.by_nuclide:
             if nuclides == 'all':
-                nuclides = self.get_all_nuclides()
+                nuclides = self.get_nuclides()
             if nuclides == 'sum':
                 nuclides = ['sum']
             else:
@@ -3622,7 +3619,7 @@ class ScatterMatrixXS(MatrixMGXS):
         # Construct a collection of the nuclides to retrieve from the xs tally
         if self.by_nuclide:
             if nuclides == 'all' or nuclides == 'sum' or nuclides == ['sum']:
-                query_nuclides = self.get_all_nuclides()
+                query_nuclides = self.get_nuclides()
             else:
                 query_nuclides = nuclides
         else:
@@ -3789,7 +3786,7 @@ class ScatterMatrixXS(MatrixMGXS):
         # Construct a collection of the nuclides to report
         if self.by_nuclide:
             if nuclides == 'all':
-                nuclides = self.get_all_nuclides()
+                nuclides = self.get_nuclides()
             if nuclides == 'sum':
                 nuclides = ['sum']
             else:
@@ -4595,7 +4592,7 @@ class Chi(MGXS):
                 nu_fission_out = self.tallies['nu-fission-out']
 
                 # Sum out all nuclides
-                nuclides = self.get_all_nuclides()
+                nuclides = self.get_nuclides()
                 nu_fission_in = nu_fission_in.summation(nuclides=nuclides)
                 nu_fission_out = nu_fission_out.summation(nuclides=nuclides)
 
@@ -4615,7 +4612,7 @@ class Chi(MGXS):
 
             # Get chi for all nuclides in the domain
             elif nuclides == 'all':
-                nuclides = self.get_all_nuclides()
+                nuclides = self.get_nuclides()
                 xs = self.xs_tally.get_values(filters=filters,
                                               filter_bins=filter_bins,
                                               nuclides=nuclides, value=value)
diff --git a/openmc/statepoint.py b/openmc/statepoint.py
index 6337746650..58a79fa219 100644
--- a/openmc/statepoint.py
+++ b/openmc/statepoint.py
@@ -2,6 +2,7 @@ import sys
 import re
 import os
 import warnings
+import glob
 
 import numpy as np
 
@@ -22,8 +23,9 @@ class StatePoint(object):
     filename : str
         Path to file to load
     autolink : bool, optional
-        Whether to automatically link in metadata from a summary.h5
-        file. Defaults to True.
+        Whether to automatically link in metadata from a summary.h5 file and
+        stochastic volume calculation results from volume_*.h5 files. Defaults
+        to True.
 
     Attributes
     ----------
@@ -143,6 +145,12 @@ class StatePoint(object):
                 su = openmc.Summary(path_summary)
                 self.link_with_summary(su)
 
+            path_volume = os.path.join(os.path.dirname(filename), 'volume_*.h5')
+            for path_i in glob.glob(path_volume):
+                if re.search(r'volume_\d+\.h5', path_i):
+                    vol = openmc.VolumeCalculation.from_hdf5(path_i)
+                    self.add_volume_information(vol)
+
     def close(self):
         self._f.close()
 
@@ -501,6 +509,19 @@ class StatePoint(object):
             for tally_id in self.tallies:
                 self.tallies[tally_id].sparse = self.sparse
 
+    def add_volume_information(self, volume_calc):
+        """Add volume information to the geometry within the file
+
+        Parameters
+        ----------
+        volume_calc : openmc.VolumeCalculation
+            Results from a stochastic volume calculation
+
+        """
+        if self.summary is not None:
+            self.summary.add_volume_information(volume_calc)
+
+
     def get_tally(self, scores=[], filters=[], nuclides=[],
                   name=None, id=None, estimator=None, exact_filters=False,
                   exact_nuclides=False, exact_scores=False):
diff --git a/openmc/summary.py b/openmc/summary.py
index 8c626940e5..6fbace72ea 100644
--- a/openmc/summary.py
+++ b/openmc/summary.py
@@ -594,6 +594,17 @@ class Summary(object):
             # Add Tally to the global dictionary of all Tallies
             self.tallies[tally_id] = tally
 
+    def add_volume_information(self, volume_calc):
+        """Add volume information to the geometry within the summary file
+
+        Parameters
+        ----------
+        volume_calc : openmc.VolumeCalculation
+            Results from a stochastic volume calculation
+
+        """
+        self.openmc_geometry.add_volume_information(volume_calc)
+
     def get_material_by_id(self, material_id):
         """Return a Material object given the material id
 
diff --git a/openmc/universe.py b/openmc/universe.py
index c8e7fcab1e..f2b8641699 100644
--- a/openmc/universe.py
+++ b/openmc/universe.py
@@ -349,7 +349,27 @@ class Universe(object):
         # Return the offset computed at all nested Universe levels
         return offset
 
-    def get_all_nuclides(self):
+    def get_nuclides(self):
+        """Returns all nuclides in the universe
+
+        Returns
+        -------
+        nuclides : list
+            List of nuclide names
+
+        """
+
+        nuclides = []
+
+        # Append all Nuclides in each Cell in the Universe to the dictionary
+        for cell in self.cells.values():
+            for nuclide in cell.get_nuclides():
+                if nuclide not in nuclides:
+                    nuclides.append(nuclide)
+
+        return nuclides
+
+    def get_nuclide_densities(self):
         """Return all nuclides contained in the universe
 
         Returns
@@ -360,13 +380,8 @@ class Universe(object):
 
         """
 
-        nuclides = OrderedDict()
-
-        # Append all Nuclides in each Cell in the Universe to the dictionary
-        for cell in self._cells.values():
-            nuclides.update(cell.get_all_nuclides())
-
-        return nuclides
+        raise NotImplementedError('Determining average nuclide densities over '
+                                  'an entire universe not yet supported.')
 
     def get_all_cells(self):
         """Return all cells that are contained within the universe