diff --git a/docs/source/pythonapi/examples/mgxs-part-iii.ipynb b/docs/source/pythonapi/examples/mgxs-part-iii.ipynb
index 39980a8fd7..a1eaa7ad82 100644
--- a/docs/source/pythonapi/examples/mgxs-part-iii.ipynb
+++ b/docs/source/pythonapi/examples/mgxs-part-iii.ipynb
@@ -574,7 +574,7 @@
    "cell_type": "markdown",
    "metadata": {},
    "source": [
-    "Now we must specify the type of domain over which we would like the `Library` to compute multi-group cross sections. The domain type corresponds to the type of tally filter to be used in the tallies created to compute multi-group cross sections. At the present time, the `Library` supports `\"material,\"` `\"cell,\"` and `\"universe\"` domain types. We will use a `\"cell\"` domain type here to compute cross sections in each of the cells in the fuel assembly geometry.\n",
+    "Now we must specify the type of domain over which we would like the `Library` to compute multi-group cross sections. The domain type corresponds to the type of tally filter to be used in the tallies created to compute multi-group cross sections. At the present time, the `Library` supports `\"material\"`, `\"cell\"`, `\"universe\"`, and `\"mesh\"` domain types. We will use a `\"cell\"` domain type here to compute cross sections in each of the cells in the fuel assembly geometry.\n",
     "\n",
     "**Note:** By default, the `Library` class will instantiate `MGXS` objects for each and every domain (material, cell or universe) in the geometry of interest. However, one may specify a subset of these domains to the `Library.domains` property. In our case, we wish to compute multi-group cross sections in each and every cell since they will be needed in our downstream OpenMOC calculation on the identical combinatorial geometry mesh."
    ]
diff --git a/docs/source/pythonapi/examples/mgxs-part-iv.ipynb b/docs/source/pythonapi/examples/mgxs-part-iv.ipynb
index ae015b0a64..eaae92f7c9 100644
--- a/docs/source/pythonapi/examples/mgxs-part-iv.ipynb
+++ b/docs/source/pythonapi/examples/mgxs-part-iv.ipynb
@@ -519,9 +519,9 @@
    "cell_type": "markdown",
    "metadata": {},
    "source": [
-    "Now we must specify the type of domain over which we would like the `Library` to compute multi-group cross sections. The domain type corresponds to the type of tally filter to be used in the tallies created to compute multi-group cross sections. At the present time, the `Library` supports \"material,\" \"cell,\" and \"universe\" domain types. In this simple example, we wish to compute multi-group cross sections only for each material andtherefore will use a \"material\" domain type.\n",
+    "Now we must specify the type of domain over which we would like the `Library` to compute multi-group cross sections. The domain type corresponds to the type of tally filter to be used in the tallies created to compute multi-group cross sections. At the present time, the `Library` supports \"material\" \"cell\", \"universe\", and \"mesh\" domain types. In this simple example, we wish to compute multi-group cross sections only for each material and therefore will use a \"material\" domain type.\n",
     "\n",
-    "**Note:** By default, the `Library` class will instantiate `MGXS` objects for each and every domain (material, cell or universe) in the geometry of interest. However, one may specify a subset of these domains to the `Library.domains` property."
+    "**Note:** By default, the `Library` class will instantiate `MGXS` objects for each and every domain (material, cell, universe, or mesh) in the geometry of interest. However, one may specify a subset of these domains to the `Library.domains` property."
    ]
   },
   {
@@ -1437,21 +1437,21 @@
  ],
  "metadata": {
   "kernelspec": {
-   "display_name": "Python 3",
+   "display_name": "Python 2",
    "language": "python",
-   "name": "python3"
+   "name": "python2"
   },
   "language_info": {
    "codemirror_mode": {
     "name": "ipython",
-    "version": 3
+    "version": 2
    },
    "file_extension": ".py",
    "mimetype": "text/x-python",
    "name": "python",
    "nbconvert_exporter": "python",
-   "pygments_lexer": "ipython3",
-   "version": "3.5.2"
+   "pygments_lexer": "ipython2",
+   "version": "2.7.11"
   }
  },
  "nbformat": 4,
diff --git a/openmc/__init__.py b/openmc/__init__.py
index 0bde0f5843..557e13039f 100644
--- a/openmc/__init__.py
+++ b/openmc/__init__.py
@@ -9,8 +9,8 @@ from openmc.plots import *
 from openmc.settings import *
 from openmc.surface import *
 from openmc.universe import *
-from openmc.mgxs_library import *
 from openmc.mesh import *
+from openmc.mgxs_library import *
 from openmc.filter import *
 from openmc.trigger import *
 from openmc.tallies import *
diff --git a/openmc/mgxs/library.py b/openmc/mgxs/library.py
index bf822017c1..84571ce236 100644
--- a/openmc/mgxs/library.py
+++ b/openmc/mgxs/library.py
@@ -54,9 +54,9 @@ class Library(object):
         If true, computes cross sections for each nuclide in each domain
     mgxs_types : Iterable of str
         The types of cross sections in the library (e.g., ['total', 'scatter'])
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         Domain type for spatial homogenization
-    domains : Iterable of openmc.Material, openmc.Cell or openmc.Universe
+    domains : Iterable of openmc.Material, openmc.Cell, openmc.Universe or openmc.Mesh
         The spatial domain(s) for which MGXS in the Library are computed
     correction : {'P0', None}
         Apply the P0 correction to scattering matrices if set to 'P0'
@@ -183,6 +183,8 @@ class Library(object):
                 return self.openmc_geometry.get_all_material_cells()
             elif self.domain_type == 'universe':
                 return self.openmc_geometry.get_all_universes()
+            elif self.domain_type == 'mesh':
+                raise ValueError('Unable to get domains for Mesh domain type')
             else:
                 raise ValueError('Unable to get domains without a domain type')
         else:
@@ -273,6 +275,12 @@ class Library(object):
             elif self.domain_type == 'universe':
                 cv.check_iterable_type('domain', domains, openmc.Universe)
                 all_domains = self.openmc_geometry.get_all_universes()
+            elif self.domain_type == 'mesh':
+                cv.check_iterable_type('domain', domains, openmc.Mesh)
+
+                # The mesh and geometry are independent, so set all_domains
+                # to the input domains
+                all_domains = domains
             else:
                 raise ValueError('Unable to set domains with domain '
                                  'type "{}"'.format(self.domain_type))
@@ -474,6 +482,8 @@ class Library(object):
             cv.check_type('domain', domain, (openmc.Cell, Integral))
         elif self.domain_type == 'universe':
             cv.check_type('domain', domain, (openmc.Universe, Integral))
+        elif self.domain_type == 'mesh':
+            cv.check_type('domain', domain, (openmc.Mesh, Integral))
 
         # Check that requested domain is included in library
         if isinstance(domain, Integral):
diff --git a/openmc/mgxs/mgxs.py b/openmc/mgxs/mgxs.py
index 36d5a00be1..e1c2e62217 100644
--- a/openmc/mgxs/mgxs.py
+++ b/openmc/mgxs/mgxs.py
@@ -7,6 +7,7 @@ import os
 import sys
 import copy
 import abc
+import itertools
 
 import numpy as np
 
@@ -14,7 +15,6 @@ import openmc
 import openmc.checkvalue as cv
 from openmc.mgxs import EnergyGroups
 
-
 if sys.version_info[0] >= 3:
     basestring = str
 
@@ -41,17 +41,17 @@ MGXS_TYPES = ['total',
 
 
 # Supported domain types
-# TODO: Implement Mesh domains
 DOMAIN_TYPES = ['cell',
                 'distribcell',
                 'universe',
-                'material']
+                'material',
+                'mesh']
 
 # Supported domain classes
-# TODO: Implement Mesh domains
 _DOMAINS = (openmc.Cell,
             openmc.Universe,
-            openmc.Material)
+            openmc.Material,
+            openmc.Mesh)
 
 
 class MGXS(object):
@@ -66,9 +66,9 @@ class MGXS(object):
 
     Parameters
     ----------
-    domain : openmc.Material or openmc.Cell or openmc.Universe
+    domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
         The domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         The domain type for spatial homogenization
     energy_groups : openmc.mgxs.EnergyGroups
         The energy group structure for energy condensation
@@ -86,9 +86,9 @@ class MGXS(object):
         Reaction type (e.g., 'total', 'nu-fission', etc.)
     by_nuclide : bool
         If true, computes cross sections for each nuclide in domain
-    domain : Material or Cell or Universe
+    domain : Material or Cell or Universe or Mesh
         Domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         Domain type for spatial homogenization
     energy_groups : openmc.mgxs.EnergyGroups
         Energy group structure for energy condensation
@@ -115,9 +115,10 @@ class MGXS(object):
         is None unless the multi-group cross section has been computed.
     num_subdomains : int
         The number of subdomains is unity for 'material', 'cell' and 'universe'
-        domain types. When the  This is equal to the number of cell instances
+        domain types. This is equal to the number of cell instances
         for 'distribcell' domain types (it is equal to unity prior to loading
-        tally data from a statepoint file).
+        tally data from a statepoint file) and the number of mesh cells for
+        'mesh' domain types.
     num_nuclides : int
         The number of nuclides for which the multi-group cross section is
         being tracked. This is unity if the by_nuclide attribute is False.
@@ -263,6 +264,10 @@ class MGXS(object):
             # Create a domain Filter object
             domain_filter = openmc.Filter(self.domain_type, self.domain.id)
 
+            # If a mesh domain, give the mesh to the domain filter
+            if self.domain_type == 'mesh':
+                domain_filter.mesh = self.domain
+
             # Create each Tally needed to compute the multi group cross section
             tally_metadata = zip(self.scores, self.tally_keys, self.filters)
             for score, key, filters in tally_metadata:
@@ -378,6 +383,8 @@ class MGXS(object):
                 self._domain_type = 'cell'
             elif isinstance(domain, openmc.Universe):
                 self._domain_type = 'universe'
+            elif isinstance(domain, openmc.Mesh):
+                self._domain_type = 'mesh'
 
     @domain_type.setter
     def domain_type(self, domain_type):
@@ -432,9 +439,9 @@ class MGXS(object):
         ----------
         mgxs_type : {'total', 'transport', 'nu-transport', 'absorption', 'capture', 'fission', 'nu-fission', 'kappa-fission', 'scatter', 'nu-scatter', 'scatter matrix', 'nu-scatter matrix', 'multiplicity matrix', 'nu-fission matrix', 'chi', 'chi-prompt', 'inverse-velocity', 'prompt-nu-fission'}
             The type of multi-group cross section object to return
-        domain : openmc.Material or openmc.Cell or openmc.Universe
+        domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
             The domain for spatial homogenization
-        domain_type : {'material', 'cell', 'distribcell', 'universe'}
+        domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
             The domain type for spatial homogenization
         energy_groups : openmc.mgxs.EnergyGroups
             The energy group structure for energy condensation
@@ -675,6 +682,8 @@ class MGXS(object):
             self.domain = statepoint.summary.get_universe_by_id(self.domain.id)
         elif self.domain_type == 'material':
             self.domain = statepoint.summary.get_material_by_id(self.domain.id)
+        elif self.domain_type == 'mesh':
+            self.domain = statepoint.meshes[self.domain.id]
         else:
             msg = 'Unable to load data from a statepoint for domain type {0} ' \
                   'which is not yet supported'.format(self.domain_type)
@@ -682,7 +691,11 @@ class MGXS(object):
 
         # Use tally "slicing" to ensure that tallies correspond to our domain
         # NOTE: This is important if tally merging was used
-        if self.domain_type != 'distribcell':
+        if self.domain_type == 'mesh':
+            filters = [self.domain_type]
+            xyz = [range(1, x+1) for x in self.domain.dimension]
+            filter_bins = [tuple(itertools.product(*xyz))]
+        elif self.domain_type != 'distribcell':
             filters = [self.domain_type]
             filter_bins = [(self.domain.id,)]
         # Distribcell filters only accept single cell - neglect it when slicing
@@ -756,12 +769,19 @@ class MGXS(object):
         cv.check_value('value', value, ['mean', 'std_dev', 'rel_err'])
         cv.check_value('xs_type', xs_type, ['macro', 'micro'])
 
+        # FIXME: Unable to get microscopic xs for mesh domain because the mesh
+        # cells do not know the nuclide densities in each mesh cell.
+        if self.domain_type == 'mesh' and xs_type == 'micro':
+            msg = 'Unable to get micro xs for mesh domain since the mesh ' \
+                  'cells do not know the nuclide densities in each mesh cell.'
+            raise ValueError(msg)
+
         filters = []
         filter_bins = []
 
         # Construct a collection of the domain filter bins
         if not isinstance(subdomains, basestring):
-            cv.check_iterable_type('subdomains', subdomains, Integral, max_depth=2)
+            cv.check_iterable_type('subdomains', subdomains, Integral, max_depth=3)
             for subdomain in subdomains:
                 filters.append(self.domain_type)
                 filter_bins.append((subdomain,))
@@ -981,16 +1001,16 @@ class MGXS(object):
         cv.check_iterable_type('energy_groups', groups, Integral)
 
         # Build lists of filters and filter bins to slice
-        if len(groups) == 0:
-            filters = []
-            filter_bins = []
-        else:
-            filter_bins = []
+        filters = []
+        filter_bins = []
+
+        if len(groups) != 0:
+            energy_bins = []
             for group in groups:
                 group_bounds = self.energy_groups.get_group_bounds(group)
-                filter_bins.append(group_bounds)
-            filter_bins = [tuple(filter_bins)]
-            filters = ['energy']
+                energy_bins.append(group_bounds)
+            filter_bins.append(tuple(energy_bins))
+            filters.append('energy')
 
         # Clone this MGXS to initialize the sliced version
         slice_xs = copy.deepcopy(self)
@@ -1135,6 +1155,9 @@ class MGXS(object):
             cv.check_iterable_type('subdomains', subdomains, Integral)
         elif self.domain_type == 'distribcell':
             subdomains = np.arange(self.num_subdomains, dtype=np.int)
+        elif self.domain_type == 'mesh':
+            xyz = [range(1, x+1) for x in self.domain.dimension]
+            subdomains = list(itertools.product(*xyz))
         else:
             subdomains = [self.domain.id]
 
@@ -1168,7 +1191,7 @@ class MGXS(object):
         # Loop over all subdomains
         for subdomain in subdomains:
 
-            if self.domain_type == 'distribcell':
+            if self.domain_type == 'distribcell' or self.domain_type == 'mesh':
                 string += '{0: <16}=\t{1}\n'.format('\tSubdomain', subdomain)
 
             # Loop over all Nuclides
@@ -1270,6 +1293,9 @@ class MGXS(object):
         elif self.domain_type == 'avg(distribcell)':
             domain_filter = self.xs_tally.find_filter('avg(distribcell)')
             subdomains = domain_filter.bins
+        elif self.domain_type == 'mesh':
+            xyz = [range(1, x+1) for x in self.domain.dimension]
+            subdomains = list(itertools.product(*xyz))
         else:
             subdomains = [self.domain.id]
 
@@ -1376,15 +1402,14 @@ class MGXS(object):
         # Get a Pandas DataFrame for the data
         df = self.get_pandas_dataframe(groups=groups, xs_type=xs_type)
 
-        # Capitalize column label strings
-        df.columns = df.columns.astype(str)
-        df.columns = map(str.title, df.columns)
-
         # Export the data using Pandas IO API
         if format == 'csv':
             df.to_csv(filename + '.csv', index=False)
         elif format == 'excel':
-            df.to_excel(filename + '.xls', index=False)
+            if self.domain_type == 'mesh':
+                df.to_excel(filename + '.xls')
+            else:
+                df.to_excel(filename + '.xls', index=False)
         elif format == 'pickle':
             df.to_pickle(filename + '.pkl')
         elif format == 'latex':
@@ -1465,7 +1490,10 @@ class MGXS(object):
                 distribcell_paths=distribcell_paths)
 
             # Remove nuclide column since it is homogeneous and redundant
-            df.drop('nuclide', axis=1, inplace=True)
+            if self.domain_type == 'mesh':
+                df.drop('nuclide', axis=1, level=0, inplace=True)
+            else:
+                df.drop('nuclide', axis=1, inplace=True)
 
         # If the user requested a specific set of nuclides
         elif self.by_nuclide and nuclides != 'all':
@@ -1479,7 +1507,10 @@ class MGXS(object):
                 distribcell_paths=distribcell_paths)
 
         # Remove the score column since it is homogeneous and redundant
-        df = df.drop('score', axis=1)
+        if self.domain_type == 'mesh':
+            df = df.drop('score', axis=1, level=0)
+        else:
+            df = df.drop('score', axis=1)
 
         # Override energy groups bounds with indices
         all_groups = np.arange(self.num_groups, 0, -1, dtype=np.int)
@@ -1535,7 +1566,12 @@ class MGXS(object):
 
         # Sort the dataframe by domain type id (e.g., distribcell id) and
         # energy groups such that data is from fast to thermal
-        df.sort_values(by=[self.domain_type] + columns, inplace=True)
+        if self.domain_type == 'mesh':
+            mesh_str = 'mesh {0}'.format(self.domain.id)
+            df.sort_values(by=[(mesh_str, 'x'), (mesh_str, 'y'), \
+                               (mesh_str, 'z')] + columns, inplace=True)
+        else:
+            df.sort_values(by=[self.domain_type] + columns, inplace=True)
         return df
 
     def get_units(self, xs_type='macro'):
@@ -1574,9 +1610,9 @@ class MatrixMGXS(MGXS):
 
     Parameters
     ----------
-    domain : openmc.Material or openmc.Cell or openmc.Universe
+    domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
         The domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         The domain type for spatial homogenization
     energy_groups : openmc.mgxs.EnergyGroups
         The energy group structure for energy condensation
@@ -1594,9 +1630,9 @@ class MatrixMGXS(MGXS):
         Reaction type (e.g., 'total', 'nu-fission', etc.)
     by_nuclide : bool
         If true, computes cross sections for each nuclide in domain
-    domain : Material or Cell or Universe
+    domain : Material or Cell or Universe or Mesh
         Domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         Domain type for spatial homogenization
     energy_groups : openmc.mgxs.EnergyGroups
         Energy group structure for energy condensation
@@ -1623,9 +1659,10 @@ class MatrixMGXS(MGXS):
         is None unless the multi-group cross section has been computed.
     num_subdomains : int
         The number of subdomains is unity for 'material', 'cell' and 'universe'
-        domain types. When the  This is equal to the number of cell instances
+        domain types. This is equal to the number of cell instances
         for 'distribcell' domain types (it is equal to unity prior to loading
-        tally data from a statepoint file).
+        tally data from a statepoint file) and the number of mesh cells for
+        'mesh' domain types.
     num_nuclides : int
         The number of nuclides for which the multi-group cross section is
         being tracked. This is unity if the by_nuclide attribute is False.
@@ -1716,13 +1753,20 @@ class MatrixMGXS(MGXS):
         cv.check_value('value', value, ['mean', 'std_dev', 'rel_err'])
         cv.check_value('xs_type', xs_type, ['macro', 'micro'])
 
+        # FIXME: Unable to get microscopic xs for mesh domain because the mesh
+        # cells do not know the nuclide densities in each mesh cell.
+        if self.domain_type == 'mesh' and xs_type == 'micro':
+            msg = 'Unable to get micro xs for mesh domain since the mesh ' \
+                  'cells do not know the nuclide densities in each mesh cell.'
+            raise ValueError(msg)
+
         filters = []
         filter_bins = []
 
         # Construct a collection of the domain filter bins
         if not isinstance(subdomains, basestring):
             cv.check_iterable_type('subdomains', subdomains, Integral,
-                                   max_depth=2)
+                                   max_depth=3)
             for subdomain in subdomains:
                 filters.append(self.domain_type)
                 filter_bins.append((subdomain,))
@@ -1884,6 +1928,9 @@ class MatrixMGXS(MGXS):
             cv.check_iterable_type('subdomains', subdomains, Integral)
         elif self.domain_type == 'distribcell':
             subdomains = np.arange(self.num_subdomains, dtype=np.int)
+        elif self.domain_type == 'mesh':
+            xyz = [range(1, x+1) for x in self.domain.dimension]
+            subdomains = list(itertools.product(*xyz))
         else:
             subdomains = [self.domain.id]
 
@@ -1992,9 +2039,9 @@ class TotalXS(MGXS):
 
     Parameters
     ----------
-    domain : openmc.Material or openmc.Cell or openmc.Universe
+    domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
         The domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         The domain type for spatial homogenization
     groups : openmc.mgxs.EnergyGroups
         The energy group structure for energy condensation
@@ -2012,9 +2059,9 @@ class TotalXS(MGXS):
         Reaction type (e.g., 'total', 'nu-fission', etc.)
     by_nuclide : bool
         If true, computes cross sections for each nuclide in domain
-    domain : Material or Cell or Universe
+    domain : Material or Cell or Universe or Mesh
         Domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         Domain type for spatial homogenization
     energy_groups : openmc.mgxs.EnergyGroups
         Energy group structure for energy condensation
@@ -2043,7 +2090,7 @@ class TotalXS(MGXS):
         is None unless the multi-group cross section has been computed.
     num_subdomains : int
         The number of subdomains is unity for 'material', 'cell' and 'universe'
-        domain types. When the  This is equal to the number of cell instances
+        domain types. This is equal to the number of cell instances
         for 'distribcell' domain types (it is equal to unity prior to loading
         tally data from a statepoint file).
     num_nuclides : int
@@ -2110,9 +2157,9 @@ class TransportXS(MGXS):
 
     Parameters
     ----------
-    domain : openmc.Material or openmc.Cell or openmc.Universe
+    domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
         The domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         The domain type for spatial homogenization
     groups : openmc.mgxs.EnergyGroups
         The energy group structure for energy condensation
@@ -2130,9 +2177,9 @@ class TransportXS(MGXS):
         Reaction type (e.g., 'total', 'nu-fission', etc.)
     by_nuclide : bool
         If true, computes cross sections for each nuclide in domain
-    domain : Material or Cell or Universe
+    domain : Material or Cell or Universe or Mesh
         Domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         Domain type for spatial homogenization
     energy_groups : openmc.mgxs.EnergyGroups
         Energy group structure for energy condensation
@@ -2161,7 +2208,7 @@ class TransportXS(MGXS):
         is None unless the multi-group cross section has been computed.
     num_subdomains : int
         The number of subdomains is unity for 'material', 'cell' and 'universe'
-        domain types. When the  This is equal to the number of cell instances
+        domain types. This is equal to the number of cell instances
         for 'distribcell' domain types (it is equal to unity prior to loading
         tally data from a statepoint file).
     num_nuclides : int
@@ -2240,9 +2287,9 @@ class NuTransportXS(TransportXS):
 
     Parameters
     ----------
-    domain : openmc.Material or openmc.Cell or openmc.Universe
+    domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
         The domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         The domain type for spatial homogenization
     groups : openmc.mgxs.EnergyGroups
         The energy group structure for energy condensation
@@ -2260,9 +2307,9 @@ class NuTransportXS(TransportXS):
         Reaction type (e.g., 'total', 'nu-fission', etc.)
     by_nuclide : bool
         If true, computes cross sections for each nuclide in domain
-    domain : Material or Cell or Universe
+    domain : Material or Cell or Universe or Mesh
         Domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         Domain type for spatial homogenization
     energy_groups : openmc.mgxs.EnergyGroups
         Energy group structure for energy condensation
@@ -2291,7 +2338,7 @@ class NuTransportXS(TransportXS):
         is None unless the multi-group cross section has been computed.
     num_subdomains : int
         The number of subdomains is unity for 'material', 'cell' and 'universe'
-        domain types. When the  This is equal to the number of cell instances
+        domain types. This is equal to the number of cell instances
         for 'distribcell' domain types (it is equal to unity prior to loading
         tally data from a statepoint file).
     num_nuclides : int
@@ -2361,9 +2408,9 @@ class AbsorptionXS(MGXS):
 
     Parameters
     ----------
-    domain : openmc.Material or openmc.Cell or openmc.Universe
+    domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
         The domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         The domain type for spatial homogenization
     groups : openmc.mgxs.EnergyGroups
         The energy group structure for energy condensation
@@ -2381,9 +2428,9 @@ class AbsorptionXS(MGXS):
         Reaction type (e.g., 'total', 'nu-fission', etc.)
     by_nuclide : bool
         If true, computes cross sections for each nuclide in domain
-    domain : Material or Cell or Universe
+    domain : Material or Cell or Universe or Mesh
         Domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         Domain type for spatial homogenization
     energy_groups : openmc.mgxs.EnergyGroups
         Energy group structure for energy condensation
@@ -2412,7 +2459,7 @@ class AbsorptionXS(MGXS):
         is None unless the multi-group cross section has been computed.
     num_subdomains : int
         The number of subdomains is unity for 'material', 'cell' and 'universe'
-        domain types. When the  This is equal to the number of cell instances
+        domain types. This is equal to the number of cell instances
         for 'distribcell' domain types (it is equal to unity prior to loading
         tally data from a statepoint file).
     num_nuclides : int
@@ -2477,9 +2524,9 @@ class CaptureXS(MGXS):
 
     Parameters
     ----------
-    domain : openmc.Material or openmc.Cell or openmc.Universe
+    domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
         The domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         The domain type for spatial homogenization
     groups : openmc.mgxs.EnergyGroups
         The energy group structure for energy condensation
@@ -2497,9 +2544,9 @@ class CaptureXS(MGXS):
         Reaction type (e.g., 'total', 'nu-fission', etc.)
     by_nuclide : bool
         If true, computes cross sections for each nuclide in domain
-    domain : Material or Cell or Universe
+    domain : Material or Cell or Universe or Mesh
         Domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         Domain type for spatial homogenization
     energy_groups : openmc.mgxs.EnergyGroups
         Energy group structure for energy condensation
@@ -2528,7 +2575,7 @@ class CaptureXS(MGXS):
         is None unless the multi-group cross section has been computed.
     num_subdomains : int
         The number of subdomains is unity for 'material', 'cell' and 'universe'
-        domain types. When the  This is equal to the number of cell instances
+        domain types. This is equal to the number of cell instances
         for 'distribcell' domain types (it is equal to unity prior to loading
         tally data from a statepoint file).
     num_nuclides : int
@@ -2599,9 +2646,9 @@ class FissionXS(MGXS):
 
     Parameters
     ----------
-    domain : openmc.Material or openmc.Cell or openmc.Universe
+    domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
         The domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         The domain type for spatial homogenization
     groups : openmc.mgxs.EnergyGroups
         The energy group structure for energy condensation
@@ -2619,9 +2666,9 @@ class FissionXS(MGXS):
         Reaction type (e.g., 'total', 'nu-fission', etc.)
     by_nuclide : bool
         If true, computes cross sections for each nuclide in domain
-    domain : Material or Cell or Universe
+    domain : Material or Cell or Universe or Mesh
         Domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         Domain type for spatial homogenization
     energy_groups : openmc.mgxs.EnergyGroups
         Energy group structure for energy condensation
@@ -2650,7 +2697,7 @@ class FissionXS(MGXS):
         is None unless the multi-group cross section has been computed.
     num_subdomains : int
         The number of subdomains is unity for 'material', 'cell' and 'universe'
-        domain types. When the  This is equal to the number of cell instances
+        domain types. This is equal to the number of cell instances
         for 'distribcell' domain types (it is equal to unity prior to loading
         tally data from a statepoint file).
     num_nuclides : int
@@ -2710,9 +2757,9 @@ class NuFissionXS(MGXS):
 
     Parameters
     ----------
-    domain : openmc.Material or openmc.Cell or openmc.Universe
+    domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
         The domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         The domain type for spatial homogenization
     groups : openmc.mgxs.EnergyGroups
         The energy group structure for energy condensation
@@ -2730,9 +2777,9 @@ class NuFissionXS(MGXS):
         Reaction type (e.g., 'total', 'nu-fission', etc.)
     by_nuclide : bool
         If true, computes cross sections for each nuclide in domain
-    domain : Material or Cell or Universe
+    domain : Material or Cell or Universe or Mesh
         Domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         Domain type for spatial homogenization
     energy_groups : openmc.mgxs.EnergyGroups
         Energy group structure for energy condensation
@@ -2761,7 +2808,7 @@ class NuFissionXS(MGXS):
         is None unless the multi-group cross section has been computed.
     num_subdomains : int
         The number of subdomains is unity for 'material', 'cell' and 'universe'
-        domain types. When the  This is equal to the number of cell instances
+        domain types. This is equal to the number of cell instances
         for 'distribcell' domain types (it is equal to unity prior to loading
         tally data from a statepoint file).
     num_nuclides : int
@@ -2826,9 +2873,9 @@ class KappaFissionXS(MGXS):
 
     Parameters
     ----------
-    domain : openmc.Material or openmc.Cell or openmc.Universe
+    domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
         The domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         The domain type for spatial homogenization
     groups : openmc.mgxs.EnergyGroups
         The energy group structure for energy condensation
@@ -2846,9 +2893,9 @@ class KappaFissionXS(MGXS):
         Reaction type (e.g., 'total', 'nu-fission', etc.)
     by_nuclide : bool
         If true, computes cross sections for each nuclide in domain
-    domain : Material or Cell or Universe
+    domain : Material or Cell or Universe or Mesh
         Domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         Domain type for spatial homogenization
     energy_groups : openmc.mgxs.EnergyGroups
         Energy group structure for energy condensation
@@ -2877,7 +2924,7 @@ class KappaFissionXS(MGXS):
         is None unless the multi-group cross section has been computed.
     num_subdomains : int
         The number of subdomains is unity for 'material', 'cell' and 'universe'
-        domain types. When the  This is equal to the number of cell instances
+        domain types. This is equal to the number of cell instances
         for 'distribcell' domain types (it is equal to unity prior to loading
         tally data from a statepoint file).
     num_nuclides : int
@@ -2939,9 +2986,9 @@ class ScatterXS(MGXS):
 
     Parameters
     ----------
-    domain : openmc.Material or openmc.Cell or openmc.Universe
+    domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
         The domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         The domain type for spatial homogenization
     groups : openmc.mgxs.EnergyGroups
         The energy group structure for energy condensation
@@ -2959,9 +3006,9 @@ class ScatterXS(MGXS):
         Reaction type (e.g., 'total', 'nu-fission', etc.)
     by_nuclide : bool
         If true, computes cross sections for each nuclide in domain
-    domain : Material or Cell or Universe
+    domain : Material or Cell or Universe or Mesh
         Domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         Domain type for spatial homogenization
     energy_groups : openmc.mgxs.EnergyGroups
         Energy group structure for energy condensation
@@ -2990,7 +3037,7 @@ class ScatterXS(MGXS):
         is None unless the multi-group cross section has been computed.
     num_subdomains : int
         The number of subdomains is unity for 'material', 'cell' and 'universe'
-        domain types. When the  This is equal to the number of cell instances
+        domain types. This is equal to the number of cell instances
         for 'distribcell' domain types (it is equal to unity prior to loading
         tally data from a statepoint file).
     num_nuclides : int
@@ -3054,9 +3101,9 @@ class NuScatterXS(MGXS):
 
     Parameters
     ----------
-    domain : openmc.Material or openmc.Cell or openmc.Universe
+    domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
         The domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         The domain type for spatial homogenization
     groups : openmc.mgxs.EnergyGroups
         The energy group structure for energy condensation
@@ -3074,9 +3121,9 @@ class NuScatterXS(MGXS):
         Reaction type (e.g., 'total', 'nu-fission', etc.)
     by_nuclide : bool
         If true, computes cross sections for each nuclide in domain
-    domain : Material or Cell or Universe
+    domain : Material or Cell or Universe or Mesh
         Domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         Domain type for spatial homogenization
     energy_groups : openmc.mgxs.EnergyGroups
         Energy group structure for energy condensation
@@ -3105,7 +3152,7 @@ class NuScatterXS(MGXS):
         is None unless the multi-group cross section has been computed.
     num_subdomains : int
         The number of subdomains is unity for 'material', 'cell' and 'universe'
-        domain types. When the  This is equal to the number of cell instances
+        domain types. This is equal to the number of cell instances
         for 'distribcell' domain types (it is equal to unity prior to loading
         tally data from a statepoint file).
     num_nuclides : int
@@ -3184,9 +3231,9 @@ class ScatterMatrixXS(MatrixMGXS):
 
     Parameters
     ----------
-    domain : openmc.Material or openmc.Cell or openmc.Universe
+    domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
         The domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         The domain type for spatial homogenization
     groups : openmc.mgxs.EnergyGroups
         The energy group structure for energy condensation
@@ -3208,9 +3255,9 @@ class ScatterMatrixXS(MatrixMGXS):
         Reaction type (e.g., 'total', 'nu-fission', etc.)
     by_nuclide : bool
         If true, computes cross sections for each nuclide in domain
-    domain : Material or Cell or Universe
+    domain : Material or Cell or Universe or Mesh
         Domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         Domain type for spatial homogenization
     energy_groups : openmc.mgxs.EnergyGroups
         Energy group structure for energy condensation
@@ -3239,7 +3286,7 @@ class ScatterMatrixXS(MatrixMGXS):
         is None unless the multi-group cross section has been computed.
     num_subdomains : int
         The number of subdomains is unity for 'material', 'cell' and 'universe'
-        domain types. When the  This is equal to the number of cell instances
+        domain types. This is equal to the number of cell instances
         for 'distribcell' domain types (it is equal to unity prior to loading
         tally data from a statepoint file).
     num_nuclides : int
@@ -3531,12 +3578,19 @@ class ScatterMatrixXS(MatrixMGXS):
         cv.check_value('value', value, ['mean', 'std_dev', 'rel_err'])
         cv.check_value('xs_type', xs_type, ['macro', 'micro'])
 
+        # FIXME: Unable to get microscopic xs for mesh domain because the mesh
+        # cells do not know the nuclide densities in each mesh cell.
+        if self.domain_type == 'mesh' and xs_type == 'micro':
+            msg = 'Unable to get micro xs for mesh domain since the mesh ' \
+                  'cells do not know the nuclide densities in each mesh cell.'
+            raise ValueError(msg)
+
         filters = []
         filter_bins = []
 
         # Construct a collection of the domain filter bins
         if not isinstance(subdomains, basestring):
-            cv.check_iterable_type('subdomains', subdomains, Integral, max_depth=2)
+            cv.check_iterable_type('subdomains', subdomains, Integral, max_depth=3)
             for subdomain in subdomains:
                 filters.append(self.domain_type)
                 filter_bins.append((subdomain,))
@@ -3681,9 +3735,12 @@ class ScatterMatrixXS(MatrixMGXS):
             df['moment'] = moments
 
             # Place the moment column before the mean column
-            mean_index = df.columns.get_loc('mean')
             columns = df.columns.tolist()
-            df = df[columns[:mean_index] + ['moment'] + columns[mean_index:-1]]
+            mean_index = [i for i, s in enumerate(columns) if 'mean' in s][0]
+            if self.domain_type == 'mesh':
+                df = df[columns[:mean_index] + [('moment', '')] + columns[mean_index:-1]]
+            else:
+                df = df[columns[:mean_index] + ['moment'] + columns[mean_index:-1]]
 
         # Select rows corresponding to requested scattering moment
         if moment != 'all':
@@ -3723,6 +3780,9 @@ class ScatterMatrixXS(MatrixMGXS):
             cv.check_iterable_type('subdomains', subdomains, Integral)
         elif self.domain_type == 'distribcell':
             subdomains = np.arange(self.num_subdomains, dtype=np.int)
+        elif self.domain_type == 'mesh':
+            xyz = [range(1, x+1) for x in self.domain.dimension]
+            subdomains = list(itertools.product(*xyz))
         else:
             subdomains = [self.domain.id]
 
@@ -3832,9 +3892,9 @@ class NuScatterMatrixXS(ScatterMatrixXS):
 
     Parameters
     ----------
-    domain : openmc.Material or openmc.Cell or openmc.Universe
+    domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
         The domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         The domain type for spatial homogenization
     groups : openmc.mgxs.EnergyGroups
         The energy group structure for energy condensation
@@ -3856,9 +3916,9 @@ class NuScatterMatrixXS(ScatterMatrixXS):
         Reaction type (e.g., 'total', 'nu-fission', etc.)
     by_nuclide : bool
         If true, computes cross sections for each nuclide in domain
-    domain : Material or Cell or Universe
+    domain : Material or Cell or Universe or Mesh
         Domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         Domain type for spatial homogenization
     energy_groups : openmc.mgxs.EnergyGroups
         Energy group structure for energy condensation
@@ -3887,7 +3947,7 @@ class NuScatterMatrixXS(ScatterMatrixXS):
         is None unless the multi-group cross section has been computed.
     num_subdomains : int
         The number of subdomains is unity for 'material', 'cell' and 'universe'
-        domain types. When the  This is equal to the number of cell instances
+        domain types. This is equal to the number of cell instances
         for 'distribcell' domain types (it is equal to unity prior to loading
         tally data from a statepoint file).
     num_nuclides : int
@@ -3958,9 +4018,9 @@ class MultiplicityMatrixXS(MatrixMGXS):
 
     Parameters
     ----------
-    domain : openmc.Material or openmc.Cell or openmc.Universe
+    domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
         The domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         The domain type for spatial homogenization
     groups : openmc.mgxs.EnergyGroups
         The energy group structure for energy condensation
@@ -3978,9 +4038,9 @@ class MultiplicityMatrixXS(MatrixMGXS):
         Reaction type (e.g., 'total', 'nu-fission', etc.)
     by_nuclide : bool
         If true, computes cross sections for each nuclide in domain
-    domain : Material or Cell or Universe
+    domain : Material or Cell or Universe or Mesh
         Domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         Domain type for spatial homogenization
     energy_groups : openmc.mgxs.EnergyGroups
         Energy group structure for energy condensation
@@ -4009,7 +4069,7 @@ class MultiplicityMatrixXS(MatrixMGXS):
         is None unless the multi-group cross section has been computed.
     num_subdomains : int
         The number of subdomains is unity for 'material', 'cell' and 'universe'
-        domain types. When the  This is equal to the number of cell instances
+        domain types. This is equal to the number of cell instances
         for 'distribcell' domain types (it is equal to unity prior to loading
         tally data from a statepoint file).
     num_nuclides : int
@@ -4105,9 +4165,9 @@ class NuFissionMatrixXS(MatrixMGXS):
 
     Parameters
     ----------
-    domain : openmc.Material or openmc.Cell or openmc.Universe
+    domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
         The domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         The domain type for spatial homogenization
     groups : openmc.mgxs.EnergyGroups
         The energy group structure for energy condensation
@@ -4125,9 +4185,9 @@ class NuFissionMatrixXS(MatrixMGXS):
         Reaction type (e.g., 'total', 'nu-fission', etc.)
     by_nuclide : bool
         If true, computes cross sections for each nuclide in domain
-    domain : Material or Cell or Universe
+    domain : Material or Cell or Universe or Mesh
         Domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         Domain type for spatial homogenization
     energy_groups : openmc.mgxs.EnergyGroups
         Energy group structure for energy condensation
@@ -4156,7 +4216,7 @@ class NuFissionMatrixXS(MatrixMGXS):
         is None unless the multi-group cross section has been computed.
     num_subdomains : int
         The number of subdomains is unity for 'material', 'cell' and 'universe'
-        domain types. When the  This is equal to the number of cell instances
+        domain types. This is equal to the number of cell instances
         for 'distribcell' domain types (it is equal to unity prior to loading
         tally data from a statepoint file).
     num_nuclides : int
@@ -4220,9 +4280,9 @@ class Chi(MGXS):
 
     Parameters
     ----------
-    domain : openmc.Material or openmc.Cell or openmc.Universe
+    domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
         The domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         The domain type for spatial homogenization
     groups : openmc.mgxs.EnergyGroups
         The energy group structure for energy condensation
@@ -4240,9 +4300,9 @@ class Chi(MGXS):
         Reaction type (e.g., 'total', 'nu-fission', etc.)
     by_nuclide : bool
         If true, computes cross sections for each nuclide in domain
-    domain : Material or Cell or Universe
+    domain : Material or Cell or Universe or Mesh
         Domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         Domain type for spatial homogenization
     energy_groups : openmc.mgxs.EnergyGroups
         Energy group structure for energy condensation
@@ -4271,7 +4331,7 @@ class Chi(MGXS):
         is None unless the multi-group cross section has been computed.
     num_subdomains : int
         The number of subdomains is unity for 'material', 'cell' and 'universe'
-        domain types. When the  This is equal to the number of cell instances
+        domain types. This is equal to the number of cell instances
         for 'distribcell' domain types (it is equal to unity prior to loading
         tally data from a statepoint file).
     num_nuclides : int
@@ -4499,12 +4559,19 @@ class Chi(MGXS):
         cv.check_value('value', value, ['mean', 'std_dev', 'rel_err'])
         cv.check_value('xs_type', xs_type, ['macro', 'micro'])
 
+        # FIXME: Unable to get microscopic xs for mesh domain because the mesh
+        # cells do not know the nuclide densities in each mesh cell.
+        if self.domain_type == 'mesh' and xs_type == 'micro':
+            msg = 'Unable to get micro xs for mesh domain since the mesh ' \
+                  'cells do not know the nuclide densities in each mesh cell.'
+            raise ValueError(msg)
+
         filters = []
         filter_bins = []
 
         # Construct a collection of the domain filter bins
         if not isinstance(subdomains, basestring):
-            cv.check_iterable_type('subdomains', subdomains, Integral, max_depth=2)
+            cv.check_iterable_type('subdomains', subdomains, Integral, max_depth=3)
             for subdomain in subdomains:
                 filters.append(self.domain_type)
                 filter_bins.append((subdomain,))
@@ -4703,9 +4770,9 @@ class ChiPrompt(Chi):
 
     Parameters
     ----------
-    domain : openmc.Material or openmc.Cell or openmc.Universe
+    domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
         The domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         The domain type for spatial homogenization
     groups : openmc.mgxs.EnergyGroups
         The energy group structure for energy condensation
@@ -4723,9 +4790,9 @@ class ChiPrompt(Chi):
         Reaction type (e.g., 'total', 'nu-fission', etc.)
     by_nuclide : bool
         If true, computes cross sections for each nuclide in domain
-    domain : Material or Cell or Universe
+    domain : Material or Cell or Universe or Mesh
         Domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         Domain type for spatial homogenization
     energy_groups : openmc.mgxs.EnergyGroups
         Energy group structure for energy condensation
@@ -4754,7 +4821,7 @@ class ChiPrompt(Chi):
         is None unless the multi-group cross section has been computed.
     num_subdomains : int
         The number of subdomains is unity for 'material', 'cell' and 'universe'
-        domain types. When the  This is equal to the number of cell instances
+        domain types. This is equal to the number of cell instances
         for 'distribcell' domain types (it is equal to unity prior to loading
         tally data from a statepoint file).
     num_nuclides : int
@@ -4818,9 +4885,9 @@ class InverseVelocity(MGXS):
 
     Parameters
     ----------
-    domain : openmc.Material or openmc.Cell or openmc.Universe
+    domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
         The domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         The domain type for spatial homogenization
     groups : openmc.mgxs.EnergyGroups
         The energy group structure for energy condensation
@@ -4838,9 +4905,9 @@ class InverseVelocity(MGXS):
         Reaction type (e.g., 'total', 'nu-fission', etc.)
     by_nuclide : bool
         If true, computes cross sections for each nuclide in domain
-    domain : Material or Cell or Universe
+    domain : Material or Cell or Universe or Mesh
         Domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         Domain type for spatial homogenization
     energy_groups : openmc.mgxs.EnergyGroups
         Energy group structure for energy condensation
@@ -4917,8 +4984,6 @@ class InverseVelocity(MGXS):
 
         """
 
-        cv.check_value('xs_type', xs_type, ['macro', 'micro'])
-
         if xs_type == 'macro':
             return 'second/cm'
         else:
@@ -4954,9 +5019,9 @@ class PromptNuFissionXS(MGXS):
 
     Parameters
     ----------
-    domain : openmc.Material or openmc.Cell or openmc.Universe
+    domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
         The domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         The domain type for spatial homogenization
     groups : openmc.mgxs.EnergyGroups
         The energy group structure for energy condensation
@@ -4974,9 +5039,9 @@ class PromptNuFissionXS(MGXS):
         Reaction type (e.g., 'total', 'nu-fission', etc.)
     by_nuclide : bool
         If true, computes cross sections for each nuclide in domain
-    domain : Material or Cell or Universe
+    domain : Material or Cell or Universe or Mesh
         Domain for spatial homogenization
-    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+    domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
         Domain type for spatial homogenization
     energy_groups : openmc.mgxs.EnergyGroups
         Energy group structure for energy condensation
@@ -5005,7 +5070,7 @@ class PromptNuFissionXS(MGXS):
         is None unless the multi-group cross section has been computed.
     num_subdomains : int
         The number of subdomains is unity for 'material', 'cell' and 'universe'
-        domain types. When the  This is equal to the number of cell instances
+        domain types. This is equal to the number of cell instances
         for 'distribcell' domain types (it is equal to unity prior to loading
         tally data from a statepoint file).
     num_nuclides : int
diff --git a/openmc/tallies.py b/openmc/tallies.py
index 79e7c56bcb..c68b0faacb 100644
--- a/openmc/tallies.py
+++ b/openmc/tallies.py
@@ -1293,7 +1293,7 @@ class Tally(object):
                     # Create list of 2- or 3-tuples tuples for mesh cell bins
                     if self_filter.type == 'mesh':
                         dimension = self_filter.mesh.dimension
-                        xyz = map(lambda x: np.arange(1, x+1), dimension)
+                        xyz = [range(1, x+1) for x in dimension]
                         bins = list(itertools.product(*xyz))
 
                     # Create list of 2-tuples for energy boundary bins
diff --git a/tests/test_mgxs_library_mesh/inputs_true.dat b/tests/test_mgxs_library_mesh/inputs_true.dat
new file mode 100644
index 0000000000..e036b49a26
--- /dev/null
+++ b/tests/test_mgxs_library_mesh/inputs_true.dat
@@ -0,0 +1 @@
+a4cd030bea212e45fdb159e75a7fb3d1947e9bf3d0384ac5d37a72298d67dcfdd1b9eb5c6af8ac6e5983bd5b47de9c17a2ea472b467b7222a4909ee070bf1ca3
\ No newline at end of file
diff --git a/tests/test_mgxs_library_mesh/results_true.dat b/tests/test_mgxs_library_mesh/results_true.dat
new file mode 100644
index 0000000000..e3d7ca6473
--- /dev/null
+++ b/tests/test_mgxs_library_mesh/results_true.dat
@@ -0,0 +1,132 @@
+  mesh 1       group in nuclide      mean std. dev.
+       x  y  z                                     
+0      1  1  1        1   total  0.640786  0.044177
+1      1  2  1        1   total  0.660597  0.128423
+2      2  1  1        1   total  0.615276  0.104046
+3      2  2  1        1   total  0.646999  0.186709
+  mesh 1       group in nuclide     mean std. dev.
+       x  y  z                                    
+0      1  1  1        1   total  0.36665  0.048814
+1      1  2  1        1   total  0.40784  0.096486
+2      2  1  1        1   total  0.36356  0.074111
+3      2  2  1        1   total  0.41456  0.160443
+  mesh 1       group in nuclide      mean std. dev.
+       x  y  z                                     
+0      1  1  1        1   total  0.366650  0.048814
+1      1  2  1        1   total  0.407840  0.096486
+2      2  1  1        1   total  0.363560  0.074111
+3      2  2  1        1   total  0.414593  0.160436
+  mesh 1       group in nuclide      mean std. dev.
+       x  y  z                                     
+0      1  1  1        1   total  0.025749  0.002863
+1      1  2  1        1   total  0.028400  0.005275
+2      2  1  1        1   total  0.022988  0.004099
+3      2  2  1        1   total  0.027589  0.010350
+  mesh 1       group in nuclide      mean std. dev.
+       x  y  z                                     
+0      1  1  1        1   total  0.015861  0.002876
+1      1  2  1        1   total  0.017280  0.004371
+2      2  1  1        1   total  0.014403  0.003542
+3      2  2  1        1   total  0.018061  0.010110
+  mesh 1       group in nuclide      mean std. dev.
+       x  y  z                                     
+0      1  1  1        1   total  0.009888  0.001077
+1      1  2  1        1   total  0.011121  0.002456
+2      2  1  1        1   total  0.008585  0.001552
+3      2  2  1        1   total  0.009527  0.003659
+  mesh 1       group in nuclide      mean std. dev.
+       x  y  z                                     
+0      1  1  1        1   total  0.026065  0.002907
+1      1  2  1        1   total  0.029084  0.006430
+2      2  1  1        1   total  0.022596  0.004062
+3      2  2  1        1   total  0.025066  0.009687
+  mesh 1       group in nuclide      mean std. dev.
+       x  y  z                                     
+0      1  1  1        1   total  1.938476  0.211550
+1      1  2  1        1   total  2.177360  0.480780
+2      2  1  1        1   total  1.682799  0.303764
+3      2  2  1        1   total  1.864890  0.715661
+  mesh 1       group in nuclide      mean std. dev.
+       x  y  z                                     
+0      1  1  1        1   total  0.615037  0.041754
+1      1  2  1        1   total  0.632196  0.123878
+2      2  1  1        1   total  0.592288  0.100439
+3      2  2  1        1   total  0.619410  0.177190
+  mesh 1       group in nuclide      mean std. dev.
+       x  y  z                                     
+0      1  1  1        1   total  0.584014  0.054315
+1      1  2  1        1   total  0.622514  0.111323
+2      2  1  1        1   total  0.587256  0.084833
+3      2  2  1        1   total  0.613792  0.168612
+   mesh 1       group in group out nuclide moment      mean std. dev.
+        x  y  z                                                      
+0       1  1  1        1         1   total     P0  0.584014  0.054315
+1       1  1  1        1         1   total     P1  0.243427  0.025488
+2       1  1  1        1         1   total     P2  0.089236  0.007357
+3       1  1  1        1         1   total     P3  0.008994  0.005768
+4       1  2  1        1         1   total     P0  0.622514  0.111323
+5       1  2  1        1         1   total     P1  0.239376  0.042594
+6       1  2  1        1         1   total     P2  0.088386  0.017200
+7       1  2  1        1         1   total     P3 -0.001243  0.005639
+8       2  1  1        1         1   total     P0  0.587256  0.084833
+9       2  1  1        1         1   total     P1  0.245120  0.041033
+10      2  1  1        1         1   total     P2  0.086784  0.016255
+11      2  1  1        1         1   total     P3  0.008660  0.004755
+12      2  2  1        1         1   total     P0  0.612950  0.167940
+13      2  2  1        1         1   total     P1  0.226176  0.061882
+14      2  2  1        1         1   total     P2  0.086593  0.026126
+15      2  2  1        1         1   total     P3  0.009672  0.011995
+   mesh 1       group in group out nuclide moment      mean std. dev.
+        x  y  z                                                      
+0       1  1  1        1         1   total     P0  0.584014  0.054315
+1       1  1  1        1         1   total     P1  0.243427  0.025488
+2       1  1  1        1         1   total     P2  0.089236  0.007357
+3       1  1  1        1         1   total     P3  0.008994  0.005768
+4       1  2  1        1         1   total     P0  0.622514  0.111323
+5       1  2  1        1         1   total     P1  0.239376  0.042594
+6       1  2  1        1         1   total     P2  0.088386  0.017200
+7       1  2  1        1         1   total     P3 -0.001243  0.005639
+8       2  1  1        1         1   total     P0  0.587256  0.084833
+9       2  1  1        1         1   total     P1  0.245120  0.041033
+10      2  1  1        1         1   total     P2  0.086784  0.016255
+11      2  1  1        1         1   total     P3  0.008660  0.004755
+12      2  2  1        1         1   total     P0  0.613792  0.168612
+13      2  2  1        1         1   total     P1  0.226142  0.061856
+14      2  2  1        1         1   total     P2  0.086174  0.025979
+15      2  2  1        1         1   total     P3  0.009721  0.012027
+  mesh 1       group in group out nuclide      mean std. dev.
+       x  y  z                                               
+0      1  1  1        1         1   total  1.000000  0.088094
+1      1  2  1        1         1   total  1.000000  0.160891
+2      2  1  1        1         1   total  1.000000  0.126864
+3      2  2  1        1         1   total  1.001374  0.305883
+  mesh 1       group in group out nuclide      mean std. dev.
+       x  y  z                                               
+0      1  1  1        1         1   total  0.027395  0.004680
+1      1  2  1        1         1   total  0.022914  0.006025
+2      2  1  1        1         1   total  0.019384  0.002846
+3      2  2  1        1         1   total  0.029629  0.006292
+  mesh 1       group out nuclide mean std. dev.
+       x  y  z                                 
+0      1  1  1         1   total  1.0  0.220956
+1      1  2  1         1   total  1.0  0.316565
+2      2  1  1         1   total  1.0  0.132140
+3      2  2  1         1   total  1.0  0.181577
+  mesh 1       group out nuclide mean std. dev.
+       x  y  z                                 
+0      1  1  1         1   total  1.0  0.222246
+1      1  2  1         1   total  1.0  0.316565
+2      2  1  1         1   total  1.0  0.132140
+3      2  2  1         1   total  1.0  0.181577
+  mesh 1       group in nuclide          mean     std. dev.
+       x  y  z                                             
+0      1  1  1        1   total  3.610522e-07  3.169931e-08
+1      1  2  1        1   total  3.942353e-07  8.459167e-08
+2      2  1  1        1   total  3.097784e-07  5.252025e-08
+3      2  2  1        1   total  3.799163e-07  1.806470e-07
+  mesh 1       group in nuclide      mean std. dev.
+       x  y  z                                     
+0      1  1  1        1   total  0.025735  0.002840
+1      1  2  1        1   total  0.028773  0.006349
+2      2  1  1        1   total  0.022306  0.004010
+3      2  2  1        1   total  0.024549  0.009379
diff --git a/tests/test_mgxs_library_mesh/test_mgxs_library_mesh.py b/tests/test_mgxs_library_mesh/test_mgxs_library_mesh.py
new file mode 100644
index 0000000000..df7a0a5ae8
--- /dev/null
+++ b/tests/test_mgxs_library_mesh/test_mgxs_library_mesh.py
@@ -0,0 +1,81 @@
+#!/usr/bin/env python
+
+import os
+import sys
+import glob
+import hashlib
+sys.path.insert(0, os.pardir)
+from testing_harness import PyAPITestHarness
+import openmc
+import openmc.mgxs
+
+
+class MGXSTestHarness(PyAPITestHarness):
+    def _build_inputs(self):
+        # Generate inputs using parent class routine
+        super(MGXSTestHarness, self)._build_inputs()
+
+        # Initialize a one-group structure
+        energy_groups = openmc.mgxs.EnergyGroups(group_edges=[0, 20.])
+
+        # Initialize MGXS Library for a few cross section types
+        # for one material-filled cell in the geometry
+        self.mgxs_lib = openmc.mgxs.Library(self._input_set.geometry)
+        self.mgxs_lib.by_nuclide = False
+
+        # Test all MGXS types
+        self.mgxs_lib.mgxs_types = openmc.mgxs.MGXS_TYPES
+        self.mgxs_lib.energy_groups = energy_groups
+        self.mgxs_lib.legendre_order = 3
+        self.mgxs_lib.domain_type = 'mesh'
+
+        # Instantiate a tally mesh
+        mesh = openmc.Mesh(mesh_id=1)
+        mesh.type = 'regular'
+        mesh.dimension = [2, 2]
+        mesh.lower_left = [-100., -100.]
+        mesh.width = [100., 100.]
+
+        self.mgxs_lib.domains = [mesh]
+        self.mgxs_lib.build_library()
+
+        # Initialize a tallies file
+        self._input_set.tallies = openmc.Tallies()
+        self.mgxs_lib.add_to_tallies_file(self._input_set.tallies, merge=False)
+        self._input_set.tallies.export_to_xml()
+
+    def _get_results(self, hash_output=False):
+        """Digest info in the statepoint and return as a string."""
+
+        # Read the statepoint file.
+        statepoint = glob.glob(os.path.join(os.getcwd(), self._sp_name))[0]
+        sp = openmc.StatePoint(statepoint)
+
+        # Load the MGXS library from the statepoint
+        self.mgxs_lib.load_from_statepoint(sp)
+
+        # Build a string from Pandas Dataframe for each 1-group MGXS
+        outstr = ''
+        for domain in self.mgxs_lib.domains:
+            for mgxs_type in self.mgxs_lib.mgxs_types:
+                mgxs = self.mgxs_lib.get_mgxs(domain, mgxs_type)
+                df = mgxs.get_pandas_dataframe()
+                outstr += df.to_string() + '\n'
+
+        # Hash the results if necessary
+        if hash_output:
+            sha512 = hashlib.sha512()
+            sha512.update(outstr.encode('utf-8'))
+            outstr = sha512.hexdigest()
+
+        return outstr
+
+    def _cleanup(self):
+        super(MGXSTestHarness, self)._cleanup()
+        f = os.path.join(os.getcwd(), 'tallies.xml')
+        if os.path.exists(f): os.remove(f)
+
+
+if __name__ == '__main__':
+    harness = MGXSTestHarness('statepoint.10.*', True)
+    harness.main()