diff --git a/docs/source/pythonapi/examples/mgxs-part-i.ipynb b/docs/source/pythonapi/examples/mgxs-part-i.ipynb
index b95bea4627..7c5132100e 100644
--- a/docs/source/pythonapi/examples/mgxs-part-i.ipynb
+++ b/docs/source/pythonapi/examples/mgxs-part-i.ipynb
@@ -383,6 +383,9 @@
     "* `ScatterMatrixXS`\n",
     "* `NuScatterMatrixXS`\n",
     "* `Chi`\n",
+    "* `ChiPrompt`\n",
+    "* `InverseVelocity`\n",
+    "* `PromptNuFissionXS`\n",
     "\n",
     "These classes provide us with an interface to generate the tally inputs as well as perform post-processing of OpenMC's tally data to compute the respective multi-group cross sections. In this case, let's create the multi-group total, absorption and scattering cross sections with our 2-group structure."
    ]
@@ -1164,21 +1167,21 @@
  ],
  "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/docs/source/pythonapi/examples/mgxs-part-iii.ipynb b/docs/source/pythonapi/examples/mgxs-part-iii.ipynb
index 789366d3aa..39980a8fd7 100644
--- a/docs/source/pythonapi/examples/mgxs-part-iii.ipynb
+++ b/docs/source/pythonapi/examples/mgxs-part-iii.ipynb
@@ -549,6 +549,9 @@
     "* `ScatterMatrixXS` (`\"scatter matrix\"`)\n",
     "* `NuScatterMatrixXS` (`\"nu-scatter matrix\"`)\n",
     "* `Chi` (`\"chi\"`)\n",
+    "* `ChiPrompt` (`\"chi prompt\"`)\n",
+    "* `InverseVelocity` (`\"inverse-velocity\"`)\n",
+    "* `PromptNuFissionXS` (`\"prompt-nu-fission\"`)\n",
     "\n",
     "In this case, let's create the multi-group cross sections needed to run an OpenMOC simulation to verify the accuracy of our cross sections. In particular, we will define `\"transport\"`, `\"nu-fission\"`, `'\"fission\"`, `\"nu-scatter matrix\"` and `\"chi\"` cross sections for our `Library`.\n",
     "\n",
@@ -1596,21 +1599,21 @@
  ],
  "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/docs/source/pythonapi/index.rst b/docs/source/pythonapi/index.rst
index 36f161b3c3..df50eee0ba 100644
--- a/docs/source/pythonapi/index.rst
+++ b/docs/source/pythonapi/index.rst
@@ -269,13 +269,16 @@ Multi-group Cross Sections
     openmc.mgxs.AbsorptionXS
     openmc.mgxs.CaptureXS
     openmc.mgxs.Chi
+    openmc.mgxs.ChiPrompt
     openmc.mgxs.FissionXS
+    openmc.mgxs.InverseVelocity
     openmc.mgxs.KappaFissionXS
     openmc.mgxs.MultiplicityMatrixXS
     openmc.mgxs.NuFissionXS
     openmc.mgxs.NuFissionMatrixXS
     openmc.mgxs.NuScatterXS
     openmc.mgxs.NuScatterMatrixXS
+    openmc.mgxs.PromptNuFissionXS
     openmc.mgxs.ScatterXS
     openmc.mgxs.ScatterMatrixXS
     openmc.mgxs.TotalXS
diff --git a/docs/source/usersguide/input.rst b/docs/source/usersguide/input.rst
index d6802075e5..e530097ded 100644
--- a/docs/source/usersguide/input.rst
+++ b/docs/source/usersguide/input.rst
@@ -1764,6 +1764,10 @@ The ``<tally>`` element accepts the following sub-elements:
         |                      |fission. This score type is not used in the        |
         |                      |multi-group :ref:`energy_mode`.                    |
         +----------------------+---------------------------------------------------+
+        |prompt-nu-fission     |Total production of prompt neutrons due to         |
+        |                      |fission. This score type is not used in the        |
+        |                      |multi-group :ref:`energy_mode`.                    |
+        +----------------------+---------------------------------------------------+
         |nu-fission            |Total production of neutrons due to fission.       |
         +----------------------+---------------------------------------------------+
         |nu-scatter,           |These scores are similar in functionality to their |
diff --git a/openmc/checkvalue.py b/openmc/checkvalue.py
index 7c0dff99dd..fe5683c06a 100644
--- a/openmc/checkvalue.py
+++ b/openmc/checkvalue.py
@@ -212,7 +212,7 @@ def check_less_than(name, value, maximum, equality=False):
             raise ValueError(msg)
 
 def check_greater_than(name, value, minimum, equality=False):
-    """Ensure that an object's value is less than a given value.
+    """Ensure that an object's value is greater than a given value.
 
     Parameters
     ----------
diff --git a/openmc/filter.py b/openmc/filter.py
index bc8b5bdf77..d6ee70f148 100644
--- a/openmc/filter.py
+++ b/openmc/filter.py
@@ -563,7 +563,7 @@ class Filter(object):
             # Initialize dictionary to build Pandas Multi-index column
             filter_dict = {}
 
-            # Append Mesh ID as outermost index of mult-index
+            # Append Mesh ID as outermost index of multi-index
             mesh_key = 'mesh {0}'.format(self.mesh.id)
 
             # Find mesh dimensions - use 3D indices for simplicity
diff --git a/openmc/mgxs/library.py b/openmc/mgxs/library.py
index 179b473301..bf822017c1 100644
--- a/openmc/mgxs/library.py
+++ b/openmc/mgxs/library.py
@@ -452,7 +452,7 @@ class Library(object):
         ----------
         domain : Material or Cell or Universe or Integral
             The material, cell, or universe object of interest (or its ID)
-        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'}
+        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
 
         Returns
diff --git a/openmc/mgxs/mgxs.py b/openmc/mgxs/mgxs.py
index 8e6d123f72..36d5a00be1 100644
--- a/openmc/mgxs/mgxs.py
+++ b/openmc/mgxs/mgxs.py
@@ -34,7 +34,10 @@ MGXS_TYPES = ['total',
               'nu-scatter matrix',
               'multiplicity matrix',
               'nu-fission matrix',
-              'chi']
+              'chi',
+              'chi-prompt',
+              'inverse-velocity',
+              'prompt-nu-fission']
 
 
 # Supported domain types
@@ -427,7 +430,7 @@ class MGXS(object):
 
         Parameters
         ----------
-        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'}
+        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
             The domain for spatial homogenization
@@ -482,6 +485,12 @@ class MGXS(object):
             mgxs = NuFissionMatrixXS(domain, domain_type, energy_groups)
         elif mgxs_type == 'chi':
             mgxs = Chi(domain, domain_type, energy_groups)
+        elif mgxs_type == 'chi-prompt':
+            mgxs = ChiPrompt(domain, domain_type, energy_groups)
+        elif mgxs_type == 'inverse-velocity':
+            mgxs = InverseVelocity(domain, domain_type, energy_groups)
+        elif mgxs_type == 'prompt-nu-fission':
+            mgxs = PromptNuFissionXS(domain, domain_type, energy_groups)
 
         mgxs.by_nuclide = by_nuclide
         mgxs.name = name
@@ -704,7 +713,7 @@ class MGXS(object):
     def get_xs(self, groups='all', subdomains='all', nuclides='all',
                xs_type='macro', order_groups='increasing',
                value='mean', **kwargs):
-        """Returns an array of multi-group cross sections.
+        r"""Returns an array of multi-group cross sections.
 
         This method constructs a 2D NumPy array for the requested multi-group
         cross section data data for one or more energy groups and subdomains.
@@ -1148,6 +1157,9 @@ class MGXS(object):
         string += '{0: <16}=\t{1}\n'.format('\tDomain Type', self.domain_type)
         string += '{0: <16}=\t{1}\n'.format('\tDomain ID', self.domain.id)
 
+        # Generate the header for an individual XS
+        xs_header = '\tCross Sections [{0}]:'.format(self.get_units(xs_type))
+
         # If cross section data has not been computed, only print string header
         if self.tallies is None:
             print(string)
@@ -1167,11 +1179,7 @@ class MGXS(object):
                     string += '{0: <16}=\t{1}\n'.format('\tNuclide', nuclide)
 
                 # Build header for cross section type
-                if xs_type == 'macro':
-                    string += '{0: <16}\n'.format('\tCross Sections [cm^-1]:')
-                else:
-                    string += '{0: <16}\n'.format('\tCross Sections [barns]:')
-
+                string += '{0: <16}\n'.format(xs_header)
                 template = '{0: <12}Group {1} [{2: <10} - {3: <10}MeV]:\t'
 
                 # Loop over energy groups ranges
@@ -1530,6 +1538,26 @@ class MGXS(object):
         df.sort_values(by=[self.domain_type] + columns, inplace=True)
         return df
 
+    def get_units(self, xs_type='macro'):
+        """This method returns the units of a MGXS based on a desired xs_type.
+
+        Parameters
+        ----------
+        xs_type: {'macro', 'micro'}
+            Return the macro or micro cross section units.
+            Defaults to 'macro'.
+
+        Returns
+        -------
+        str
+            A string representing the units of the MGXS.
+
+        """
+
+        cv.check_value('xs_type', xs_type, ['macro', 'micro'])
+
+        return 'cm^-1' if xs_type == 'macro' else 'barns'
+
 
 class MatrixMGXS(MGXS):
     """An abstract multi-group cross section for some energy group structure
@@ -1878,6 +1906,9 @@ class MatrixMGXS(MGXS):
         string += '{0: <16}=\t{1}\n'.format('\tDomain Type', self.domain_type)
         string += '{0: <16}=\t{1}\n'.format('\tDomain ID', self.domain.id)
 
+        # Generate the header for an individual XS
+        xs_header = '\tCross Sections [{0}]:'.format(self.get_units(xs_type))
+
         # If cross section data has not been computed, only print string header
         if self.tallies is None:
             print(string)
@@ -1906,11 +1937,7 @@ class MatrixMGXS(MGXS):
                     string += '{0: <16}=\t{1}\n'.format('\tNuclide', nuclide)
 
                 # Build header for cross section type
-                if xs_type == 'macro':
-                    string += '{0: <16}\n'.format('\tCross Sections [cm^-1]:')
-                else:
-                    string += '{0: <16}\n'.format('\tCross Sections [barns]:')
-
+                string += '{0: <16}\n'.format(xs_header)
                 template = '{0: <12}Group {1} -> Group {2}:\t\t'
 
                 # Loop over incoming/outgoing energy groups ranges
@@ -3723,6 +3750,9 @@ class ScatterMatrixXS(MatrixMGXS):
         string += '{0: <16}=\t{1}\n'.format('\tDomain Type', self.domain_type)
         string += '{0: <16}=\t{1}\n'.format('\tDomain ID', self.domain.id)
 
+        # Generate the header for an individual XS
+        xs_header = '\tCross Sections [{0}]:'.format(self.get_units(xs_type))
+
         # If cross section data has not been computed, only print string header
         if self.tallies is None:
             print(string)
@@ -3751,11 +3781,7 @@ class ScatterMatrixXS(MatrixMGXS):
                     string += '{0: <16}=\t{1}\n'.format('\tNuclide', nuclide)
 
                 # Build header for cross section type
-                if xs_type == 'macro':
-                    string += '{0: <16}\n'.format('\tCross Sections [cm^-1]:')
-                else:
-                    string += '{0: <16}\n'.format('\tCross Sections [barns]:')
-
+                string += '{0: <16}\n'.format(xs_header)
                 template = '{0: <12}Group {1} -> Group {2}:\t\t'
 
                 # Loop over incoming/outgoing energy groups ranges
@@ -4183,14 +4209,14 @@ class Chi(MGXS):
 
     .. math::
 
-       \langle \nu\sigma_{f,\rightarrow g} \phi \rangle &= \int_{r \in V} dr
+       \langle \nu\sigma_{f,g' \rightarrow g} \phi \rangle &= \int_{r \in V} dr
        \int_{4\pi} d\Omega' \int_0^\infty dE' \int_{E_g}^{E_{g-1}} dE \; \chi(E)
        \nu\sigma_f (r, E') \psi(r, E', \Omega')\\
        \langle \nu\sigma_f \phi \rangle &= \int_{r \in V} dr \int_{4\pi}
        d\Omega' \int_0^\infty dE' \int_0^\infty dE \; \chi(E) \nu\sigma_f (r,
        E') \psi(r, E', \Omega') \\
-       \chi_g &= \frac{\langle \nu\sigma_{f,\rightarrow g} \phi \rangle}{\langle
-       \nu\sigma_f \phi \rangle}
+       \chi_g &= \frac{\langle \nu\sigma_{f,g' \rightarrow g} \phi \rangle}
+       {\langle \nu\sigma_f \phi \rangle}
 
     Parameters
     ----------
@@ -4395,7 +4421,7 @@ class Chi(MGXS):
         """
 
         if not self.can_merge(other):
-            raise ValueError('Unable to merge Chi')
+            raise ValueError('Unable to merge a Chi MGXS')
 
         # Create deep copy of tally to return as merged tally
         merged_mgxs = copy.deepcopy(self)
@@ -4414,7 +4440,7 @@ class Chi(MGXS):
             # The nuclides must be mutually exclusive
             for nuclide in self.nuclides:
                 if nuclide in other.nuclides:
-                    msg = 'Unable to merge Chi with shared nuclides'
+                    msg = 'Unable to merge a Chi MGXS with shared nuclides'
                     raise ValueError(msg)
 
             # Concatenate lists of nuclides for the merged MGXS
@@ -4618,3 +4644,392 @@ class Chi(MGXS):
             df['std. dev.'] *= np.tile(densities, tile_factor)
 
         return df
+
+    def get_units(self, xs_type='macro'):
+        """Returns the units of Chi.
+
+        This method returns the units of Chi, which is "%" for both macro
+        and micro xs types.
+
+        Parameters
+        ----------
+        xs_type: {'macro', 'micro'}
+            Return the macro or micro cross section units.
+            Defaults to 'macro'.
+
+        Returns
+        -------
+        str
+            A string representing the units of Chi.
+
+        """
+
+        cv.check_value('xs_type', xs_type, ['macro', 'micro'])
+
+        # Chi has the same units (%) for both macro and micro
+        return '%'
+
+
+class ChiPrompt(Chi):
+    r"""The prompt fission spectrum.
+
+    This class can be used for both OpenMC input generation and tally data
+    post-processing to compute spatially-homogenized and energy-integrated
+    multi-group cross sections for multi-group neutronics calculations. At a
+    minimum, one needs to set the :attr:`ChiPrompt.energy_groups` and
+    :attr:`ChiPrompt.domain` properties. Tallies for the flux and appropriate
+    reaction rates over the specified domain are generated automatically via the
+    :attr:`ChiPrompt.tallies` property, which can then be appended to a
+    :class:`openmc.Tallies` instance.
+
+    For post-processing, the :meth:`MGXS.load_from_statepoint` will pull in the
+    necessary data to compute multi-group cross sections from a
+    :class:`openmc.StatePoint` instance. The derived multi-group cross section
+    can then be obtained from the :attr:`ChiPrompt.xs_tally` property.
+
+    For a spatial domain :math:`V` and energy group :math:`[E_g,E_{g-1}]`, the
+    fission spectrum is calculated as:
+
+    .. math::
+
+       \langle \nu^p \sigma_{f,g' \rightarrow g} \phi \rangle &= \int_{r \in V}
+       dr \int_{4\pi} d\Omega' \int_0^\infty dE' \int_{E_g}^{E_{g-1}} dE \;
+       \chi(E) \nu^p \sigma_f (r, E') \psi(r, E', \Omega')\\
+       \langle \nu^p \sigma_f \phi \rangle &= \int_{r \in V} dr \int_{4\pi}
+       d\Omega' \int_0^\infty dE' \int_0^\infty dE \; \chi(E) \nu^p \sigma_f (r,
+       E') \psi(r, E', \Omega') \\
+       \chi_g^p &= \frac{\langle \nu^p \sigma_{f,g' \rightarrow g} \phi \rangle}
+       {\langle \nu^p \sigma_f \phi \rangle}
+
+    Parameters
+    ----------
+    domain : openmc.Material or openmc.Cell or openmc.Universe
+        The domain for spatial homogenization
+    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+        The domain type for spatial homogenization
+    groups : openmc.mgxs.EnergyGroups
+        The energy group structure for energy condensation
+    by_nuclide : bool
+        If true, computes cross sections for each nuclide in domain
+    name : str, optional
+        Name of the multi-group cross section. Used as a label to identify
+        tallies in OpenMC 'tallies.xml' file.
+
+    Attributes
+    ----------
+    name : str, optional
+        Name of the multi-group cross section
+    rxn_type : str
+        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 for spatial homogenization
+    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+        Domain type for spatial homogenization
+    energy_groups : openmc.mgxs.EnergyGroups
+        Energy group structure for energy condensation
+    tally_trigger : openmc.Trigger
+        An (optional) tally precision trigger given to each tally used to
+        compute the cross section
+    scores : list of str
+        The scores in each tally used to compute the multi-group cross section
+    filters : list of openmc.Filter
+        The filters in each tally used to compute the multi-group cross section
+    tally_keys : list of str
+        The keys into the tallies dictionary for each tally used to compute
+        the multi-group cross section
+    estimator : {'tracklength', 'analog'}
+        The tally estimator used to compute the multi-group cross section
+    tallies : collections.OrderedDict
+        OpenMC tallies needed to compute the multi-group cross section. The keys
+        are strings listed in the :attr:`ChiPrompt.tally_keys` property and
+        values are instances of :class:`openmc.Tally`.
+    rxn_rate_tally : openmc.Tally
+        Derived tally for the reaction rate tally used in the numerator to
+        compute the multi-group cross section. This attribute is None
+        unless the multi-group cross section has been computed.
+    xs_tally : openmc.Tally
+        Derived tally for the multi-group cross section. This attribute
+        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
+        for 'distribcell' domain types (it is equal to unity prior to loading
+        tally data from a statepoint file).
+    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.
+    nuclides : Iterable of str or 'sum'
+        The optional user-specified nuclides for which to compute cross
+        sections (e.g., 'U-238', 'O-16'). If by_nuclide is True but nuclides
+        are not specified by the user, all nuclides in the spatial domain
+        are included. This attribute is 'sum' if by_nuclide is false.
+    sparse : bool
+        Whether or not the MGXS' tallies use SciPy's LIL sparse matrix format
+        for compressed data storage
+    loaded_sp : bool
+        Whether or not a statepoint file has been loaded with tally data
+    derived : bool
+        Whether or not the MGXS is merged from one or more other MGXS
+    hdf5_key : str
+        The key used to index multi-group cross sections in an HDF5 data store
+
+    """
+
+    def __init__(self, domain=None, domain_type=None,
+                 groups=None, by_nuclide=False, name=''):
+        super(ChiPrompt, self).__init__(domain, domain_type, groups, by_nuclide, name)
+        self._rxn_type = 'chi-prompt'
+
+    @property
+    def scores(self):
+        return ['prompt-nu-fission', 'prompt-nu-fission']
+
+
+class InverseVelocity(MGXS):
+    r"""An inverse velocity multi-group cross section.
+
+    This class can be used for both OpenMC input generation and tally data
+    post-processing to compute spatially-homogenized and energy-integrated
+    multi-group neutron inverse velocities for multi-group neutronics
+    calculations. The units of inverse velocity are seconds per centimeter. At a
+    minimum, one needs to set the :attr:`InverseVelocity.energy_groups` and
+    :attr:`InverseVelocity.domain` properties. Tallies for the flux and
+    appropriate reaction rates over the specified domain are generated
+    automatically via the :attr:`InverseVelocity.tallies` property, which can
+    then be appended to a :class:`openmc.Tallies` instance.
+
+    For post-processing, the :meth:`MGXS.load_from_statepoint` will pull in the
+    necessary data to compute multi-group cross sections from a
+    :class:`openmc.StatePoint` instance. The derived multi-group cross section
+    can then be obtained from the :attr:`InverseVelocity.xs_tally` property.
+
+    For a spatial domain :math:`V` and energy group :math:`[E_g,E_{g-1}]`, the
+    neutron inverse velocities are calculated by tallying the flux-weighted
+    inverse velocity and the flux. The inverse velocity is then the
+    flux-weighted inverse velocity divided by the flux:
+
+    .. math::
+
+       \frac{\int_{r \in V} dr \int_{4\pi} d\Omega \int_{E_g}^{E_{g-1}} dE \;
+       \frac{\psi (r, E, \Omega)}{v (r, E)}}{\int_{r \in V} dr \int_{4\pi}
+       d\Omega \int_{E_g}^{E_{g-1}} dE \; \psi (r, E, \Omega)}
+
+    Parameters
+    ----------
+    domain : openmc.Material or openmc.Cell or openmc.Universe
+        The domain for spatial homogenization
+    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+        The domain type for spatial homogenization
+    groups : openmc.mgxs.EnergyGroups
+        The energy group structure for energy condensation
+    by_nuclide : bool
+        If true, computes cross sections for each nuclide in domain
+    name : str, optional
+        Name of the multi-group cross section. Used as a label to identify
+        tallies in OpenMC 'tallies.xml' file.
+
+    Attributes
+    ----------
+    name : str, optional
+        Name of the multi-group cross section
+    rxn_type : str
+        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 for spatial homogenization
+    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+        Domain type for spatial homogenization
+    energy_groups : openmc.mgxs.EnergyGroups
+        Energy group structure for energy condensation
+    tally_trigger : openmc.Trigger
+        An (optional) tally precision trigger given to each tally used to
+        compute the cross section
+    scores : list of str
+        The scores in each tally used to compute the multi-group cross section
+    filters : list of openmc.Filter
+        The filters in each tally used to compute the multi-group cross section
+    tally_keys : list of str
+        The keys into the tallies dictionary for each tally used to compute
+        the multi-group cross section
+    estimator : {'tracklength', 'analog'}
+        The tally estimator used to compute the multi-group cross section
+    tallies : collections.OrderedDict
+        OpenMC tallies needed to compute the multi-group cross section. The keys
+        are strings listed in the :attr:`InverseVelocity.tally_keys` property
+        and values are instances of :class:`openmc.Tally`.
+    rxn_rate_tally : openmc.Tally
+        Derived tally for the reaction rate tally used in the numerator to
+        compute the multi-group cross section. This attribute is None
+        unless the multi-group cross section has been computed.
+    xs_tally : openmc.Tally
+        Derived tally for the multi-group cross section. This attribute
+        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. 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
+        The number of nuclides for which the multi-group cross section is
+        being tracked. This is unity if the by_nuclide attribute is False.
+    nuclides : Iterable of str or 'sum'
+        The optional user-specified nuclides for which to compute cross
+        sections (e.g., 'U-238', 'O-16'). If by_nuclide is True but nuclides
+        are not specified by the user, all nuclides in the spatial domain
+        are included. This attribute is 'sum' if by_nuclide is false.
+    sparse : bool
+        Whether or not the MGXS' tallies use SciPy's LIL sparse matrix format
+        for compressed data storage
+    loaded_sp : bool
+        Whether or not a statepoint file has been loaded with tally data
+    derived : bool
+        Whether or not the MGXS is merged from one or more other MGXS
+    hdf5_key : str
+        The key used to index multi-group cross sections in an HDF5 data store
+
+    """
+
+    def __init__(self, domain=None, domain_type=None,
+                 groups=None, by_nuclide=False, name=''):
+        super(InverseVelocity, self).__init__(domain, domain_type,
+                                              groups, by_nuclide, name)
+        self._rxn_type = 'inverse-velocity'
+
+    def get_units(self, xs_type='macro'):
+        """Returns the units of InverseVelocity.
+
+        This method returns the units of an InverseVelocity based on a desired
+        xs_type.
+
+        Parameters
+        ----------
+        xs_type: {'macro', 'micro'}
+            Return the macro or micro cross section units.
+            Defaults to 'macro'.
+
+        Returns
+        -------
+        str
+            A string representing the units of the InverseVelocity.
+
+        """
+
+        cv.check_value('xs_type', xs_type, ['macro', 'micro'])
+
+        if xs_type == 'macro':
+            return 'second/cm'
+        else:
+            raise ValueError('Unable to return the units of InverseVelocity for'
+                             ' xs_type other than "macro"')
+
+
+class PromptNuFissionXS(MGXS):
+    r"""A prompt fission neutron production multi-group cross section.
+
+    This class can be used for both OpenMC input generation and tally data
+    post-processing to compute spatially-homogenized and energy-integrated
+    multi-group cross sections for multi-group neutronics calculations. At a
+    minimum, one needs to set the :attr:`PromptNuFissionXS.energy_groups` and
+    :attr:`PromptNuFissionXS.domain` properties. Tallies for the flux and
+    appropriate reaction rates over the specified domain are generated
+    automatically via the :attr:`PromptNuFissionXS.tallies` property, which can
+    then be appended to a :class:`openmc.Tallies` instance.
+
+    For post-processing, the :meth:`MGXS.load_from_statepoint` will pull in the
+    necessary data to compute multi-group cross sections from a
+    :class:`openmc.StatePoint` instance. The derived multi-group cross section
+    can then be obtained from the :attr:`PromptNuFissionXS.xs_tally` property.
+
+    For a spatial domain :math:`V` and energy group :math:`[E_g,E_{g-1}]`, the
+    fission spectrum is calculated as:
+
+    .. math::
+
+       \frac{\int_{r \in V} dr \int_{4\pi} d\Omega \int_{E_g}^{E_{g-1}} dE \;
+       \nu\sigma_f^p (r, E) \psi (r, E, \Omega)}{\int_{r \in V} dr \int_{4\pi}
+       d\Omega \int_{E_g}^{E_{g-1}} dE \; \psi (r, E, \Omega)}.
+
+    Parameters
+    ----------
+    domain : openmc.Material or openmc.Cell or openmc.Universe
+        The domain for spatial homogenization
+    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+        The domain type for spatial homogenization
+    groups : openmc.mgxs.EnergyGroups
+        The energy group structure for energy condensation
+    by_nuclide : bool
+        If true, computes cross sections for each nuclide in domain
+    name : str, optional
+        Name of the multi-group cross section. Used as a label to identify
+        tallies in OpenMC 'tallies.xml' file.
+
+    Attributes
+    ----------
+    name : str, optional
+        Name of the multi-group cross section
+    rxn_type : str
+        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 for spatial homogenization
+    domain_type : {'material', 'cell', 'distribcell', 'universe'}
+        Domain type for spatial homogenization
+    energy_groups : openmc.mgxs.EnergyGroups
+        Energy group structure for energy condensation
+    tally_trigger : openmc.Trigger
+        An (optional) tally precision trigger given to each tally used to
+        compute the cross section
+    scores : list of str
+        The scores in each tally used to compute the multi-group cross section
+    filters : list of openmc.Filter
+        The filters in each tally used to compute the multi-group cross section
+    tally_keys : list of str
+        The keys into the tallies dictionary for each tally used to compute
+        the multi-group cross section
+    estimator : {'tracklength', 'analog'}
+        The tally estimator used to compute the multi-group cross section
+    tallies : collections.OrderedDict
+        OpenMC tallies needed to compute the multi-group cross section. The keys
+        are strings listed in the :attr:`PromptNuFissionXS.tally_keys` property
+        and values are instances of :class:`openmc.Tally`.
+    rxn_rate_tally : openmc.Tally
+        Derived tally for the reaction rate tally used in the numerator to
+        compute the multi-group cross section. This attribute is None
+        unless the multi-group cross section has been computed.
+    xs_tally : openmc.Tally
+        Derived tally for the multi-group cross section. This attribute
+        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
+        for 'distribcell' domain types (it is equal to unity prior to loading
+        tally data from a statepoint file).
+    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.
+    nuclides : Iterable of str or 'sum'
+        The optional user-specified nuclides for which to compute cross
+        sections (e.g., 'U-238', 'O-16'). If by_nuclide is True but nuclides
+        are not specified by the user, all nuclides in the spatial domain
+        are included. This attribute is 'sum' if by_nuclide is false.
+    sparse : bool
+        Whether or not the MGXS' tallies use SciPy's LIL sparse matrix format
+        for compressed data storage
+    loaded_sp : bool
+        Whether or not a statepoint file has been loaded with tally data
+    derived : bool
+        Whether or not the MGXS is merged from one or more other MGXS
+    hdf5_key : str
+        The key used to index multi-group cross sections in an HDF5 data store
+
+    """
+
+    def __init__(self, domain=None, domain_type=None,
+                 groups=None, by_nuclide=False, name=''):
+        super(PromptNuFissionXS, self).__init__(domain, domain_type, groups,
+                                                by_nuclide, name)
+        self._rxn_type = 'prompt-nu-fission'
diff --git a/src/constants.F90 b/src/constants.F90
index d2bb6e1cf6..a22c9ac050 100644
--- a/src/constants.F90
+++ b/src/constants.F90
@@ -289,7 +289,7 @@ module constants
        EVENT_ABSORB  =  2
 
   ! Tally score type
-  integer, parameter :: N_SCORE_TYPES = 20
+  integer, parameter :: N_SCORE_TYPES = 21
   integer, parameter :: &
        SCORE_FLUX               = -1,  & ! flux
        SCORE_TOTAL              = -2,  & ! total reaction rate
@@ -310,7 +310,8 @@ module constants
        SCORE_NU_SCATTER_YN      = -17, & ! angular flux-weighted nu-scattering moment (0:N)
        SCORE_EVENTS             = -18, & ! number of events
        SCORE_DELAYED_NU_FISSION = -19, & ! delayed neutron production rate
-       SCORE_INVERSE_VELOCITY   = -20    ! flux-weighted inverse velocity
+       SCORE_PROMPT_NU_FISSION  = -20, & ! prompt neutron production rate
+       SCORE_INVERSE_VELOCITY   = -21    ! flux-weighted inverse velocity
 
   ! Maximum scattering order supported
   integer, parameter :: MAX_ANG_ORDER = 10
diff --git a/src/endf.F90 b/src/endf.F90
index ad5e97a033..a836a54397 100644
--- a/src/endf.F90
+++ b/src/endf.F90
@@ -42,6 +42,8 @@ contains
       string = "nu-fission"
     case (SCORE_DELAYED_NU_FISSION)
       string = "delayed-nu-fission"
+    case (SCORE_PROMPT_NU_FISSION)
+      string = "prompt-nu-fission"
     case (SCORE_KAPPA_FISSION)
       string = "kappa-fission"
     case (SCORE_CURRENT)
diff --git a/src/input_xml.F90 b/src/input_xml.F90
index 3c8291e9c8..3e5a687124 100644
--- a/src/input_xml.F90
+++ b/src/input_xml.F90
@@ -3619,6 +3619,12 @@ contains
               ! Set tally estimator to analog
               t % estimator = ESTIMATOR_ANALOG
             end if
+          case ('prompt-nu-fission')
+            t % score_bins(j) = SCORE_PROMPT_NU_FISSION
+            if (t % find_filter(FILTER_ENERGYOUT) > 0) then
+              ! Set tally estimator to analog
+              t % estimator = ESTIMATOR_ANALOG
+            end if
 
             ! Disallow for MG mode since data not present
             if (.not. run_CE) then
diff --git a/src/output.F90 b/src/output.F90
index 09df23f0d5..22733a3051 100644
--- a/src/output.F90
+++ b/src/output.F90
@@ -790,6 +790,7 @@ contains
     score_names(abs(SCORE_NU_SCATTER_PN))      = "Scattering Prod. Rate Moment"
     score_names(abs(SCORE_NU_SCATTER_YN))      = "Scattering Prod. Rate Moment"
     score_names(abs(SCORE_DELAYED_NU_FISSION)) = "Delayed-Nu-Fission Rate"
+    score_names(abs(SCORE_PROMPT_NU_FISSION))  = "Prompt-Nu-Fission Rate"
     score_names(abs(SCORE_INVERSE_VELOCITY))   = "Flux-Weighted Inverse Velocity"
 
     ! Create filename for tally output
diff --git a/src/tally.F90 b/src/tally.F90
index b40a39ce0c..94143184ae 100644
--- a/src/tally.F90
+++ b/src/tally.F90
@@ -407,7 +407,7 @@ contains
               ! neutrons were emitted with different energies, multiple
               ! outgoing energy bins may have been scored to. The following
               ! logic treats this special case and results to multiple bins
-              call score_fission_eout_ce(p, t, score_index)
+              call score_fission_eout_ce(p, t, score_index, score_bin)
               cycle SCORE_LOOP
             end if
           end if
@@ -441,6 +441,67 @@ contains
         end if
 
 
+      case (SCORE_PROMPT_NU_FISSION)
+        if (t % estimator == ESTIMATOR_ANALOG) then
+          if (survival_biasing .or. p % fission) then
+            if (t % find_filter(FILTER_ENERGYOUT) > 0) then
+              ! Normally, we only need to make contributions to one scoring
+              ! bin. However, in the case of fission, since multiple fission
+              ! neutrons were emitted with different energies, multiple
+              ! outgoing energy bins may have been scored to. The following
+              ! logic treats this special case and results to multiple bins
+              call score_fission_eout_ce(p, t, score_index, score_bin)
+              cycle SCORE_LOOP
+            end if
+          end if
+          if (survival_biasing) then
+            ! No fission events occur if survival biasing is on -- need to
+            ! calculate fraction of absorptions that would have resulted in
+            ! prompt-nu-fission
+            if (micro_xs(p % event_nuclide) % absorption > ZERO) then
+                score = p % absorb_wgt * micro_xs(p % event_nuclide) % fission &
+                     * nuclides(p % event_nuclide) % nu(E, EMISSION_PROMPT) &
+                     / micro_xs(p % event_nuclide) % absorption
+            else
+              score = ZERO
+            end if
+          else
+            ! Skip any non-fission events
+            if (.not. p % fission) cycle SCORE_LOOP
+            ! If there is no outgoing energy filter, than we only need to
+            ! score to one bin. For the score to be 'analog', we need to
+            ! score the number of particles that were banked in the fission
+            ! bank as prompt neutrons. Since this was weighted by 1/keff, we
+            ! multiply by keff to get the proper score.
+            score = keff * p % wgt_bank * (ONE - sum(p % n_delayed_bank) &
+                 / real(p % n_bank, 8))
+          end if
+
+        else
+          if (i_nuclide > 0) then
+              score = micro_xs(i_nuclide) % fission * nuclides(i_nuclide) % &
+                   nu(E, EMISSION_PROMPT) * atom_density * flux
+          else
+
+            score = ZERO
+
+            ! Loop over all nuclides in the current material
+            do l = 1, materials(p % material) % n_nuclides
+
+              ! Get atom density
+              atom_density_ = materials(p % material) % atom_density(l)
+
+              ! Get index in nuclides array
+              i_nuc = materials(p % material) % nuclide(l)
+
+              ! Accumulate the contribution from each nuclide
+              score = score + micro_xs(i_nuc) % fission * nuclides(i_nuc) % &
+                   nu(E, EMISSION_PROMPT) * atom_density_ * flux
+            end do
+          end if
+        end if
+
+
       case (SCORE_DELAYED_NU_FISSION)
 
         ! make sure the correct energy is used
@@ -461,7 +522,7 @@ contains
               ! neutrons were emitted with different energies, multiple
               ! outgoing energy bins may have been scored to. The following
               ! logic treats this special case and results to multiple bins
-              call score_fission_delayed_eout(p, t, score_index)
+              call score_fission_eout_ce(p, t, score_index, score_bin)
               cycle SCORE_LOOP
             end if
           end if
@@ -1515,12 +1576,18 @@ contains
 ! neutrons produced with different energies.
 !===============================================================================
 
-  subroutine score_fission_eout_ce(p, t, i_score)
-    type(Particle), intent(in) :: p
+  subroutine score_fission_eout_ce(p, t, i_score, score_bin)
+
+    type(Particle), intent(in)       :: p
     type(TallyObject), intent(inout) :: t
-    integer, intent(in)        :: i_score ! index for score
+    integer, intent(in)              :: i_score ! index for score
+    integer, intent(in)              :: score_bin
 
     integer :: i             ! index of outgoing energy filter
+    integer :: j             ! index of delayedgroup filter
+    integer :: d             ! delayed group
+    integer :: g             ! another delayed group
+    integer :: d_bin         ! delayed group bin index
     integer :: n             ! number of energies on filter
     integer :: k             ! loop index for bank sites
     integer :: bin_energyout ! original outgoing energy bin
@@ -1542,6 +1609,10 @@ contains
 
     ! loop over number of particles banked
     do k = 1, p % n_bank
+
+      ! get the delayed group
+      g = fission_bank(n_bank - p % n_bank + k) % delayed_group
+
       ! determine score based on bank site weight and keff
       score = keff * fission_bank(n_bank - p % n_bank + k) % wgt
 
@@ -1555,13 +1626,49 @@ contains
       ! change outgoing energy bin
       matching_bins(i) = binary_search(t % filters(i) % real_bins, n, E_out)
 
-      ! determine scoring index
-      i_filter = sum((matching_bins(1:t%n_filters) - 1) * t % stride) + 1
+      ! Case for tallying prompt neutrons
+      if (score_bin == SCORE_NU_FISSION .or. &
+           (score_bin == SCORE_PROMPT_NU_FISSION .and. g == 0)) then
 
-      ! Add score to tally
+        ! determine scoring index
+        i_filter = sum((matching_bins(1:t%n_filters) - 1) * t % stride) + 1
+
+        ! Add score to tally
 !$omp atomic
-      t % results(i_score, i_filter) % value = &
-           t % results(i_score, i_filter) % value + score
+        t % results(i_score, i_filter) % value = &
+             t % results(i_score, i_filter) % value + score
+
+      ! Case for tallying delayed emissions
+      else if (score_bin == SCORE_DELAYED_NU_FISSION .and. g /= 0) then
+
+        ! Get the index of delayed group filter
+        j = t % find_filter(FILTER_DELAYEDGROUP)
+
+        ! if the delayed group filter is present, tally to corresponding
+        ! delayed group bin if it exists
+        if (j > 0) then
+
+          ! loop over delayed group bins until the corresponding bin is found
+          do d_bin = 1, t % filters(j) % n_bins
+            d = t % filters(j) % int_bins(d_bin)
+
+            ! check whether the delayed group of the particle is equal to the
+            ! delayed group of this bin
+            if (d == g) call score_fission_delayed_dg(t, d_bin, score, i_score)
+          end do
+
+        ! if the delayed group filter is not present, add score to tally
+        else
+
+          ! determine scoring index
+          i_filter = sum((matching_bins(1:t%n_filters) - 1) * t % stride) + 1
+
+          ! Add score to tally
+!$omp atomic
+          t % results(i_score, i_filter) % value = &
+               t % results(i_score, i_filter) % value + score
+        end if
+      end if
     end do
 
     ! reset outgoing energy bin and score index
@@ -1647,101 +1754,6 @@ contains
 
   end subroutine score_fission_eout_mg
 
-!===============================================================================
-! SCORE_FISSION_DELAYED_EOUT handles a special case where we need to store
-! delayed neutron production rate with an outgoing energy filter (think of a
-! fission matrix). In this case, we may need to score to multiple bins if there
-! were multiple neutrons produced with different energies.
-!===============================================================================
-
-  subroutine score_fission_delayed_eout(p, t, i_score)
-
-    type(Particle), intent(in)       :: p
-    type(TallyObject), intent(inout) :: t
-    integer, intent(in)              :: i_score ! index for score
-
-    integer :: i             ! index of outgoing energy filter
-    integer :: j             ! index of delayedgroup filter
-    integer :: d             ! delayed group
-    integer :: g             ! another delayed group
-    integer :: d_bin         ! delayed group bin index
-    integer :: n             ! number of energies on filter
-    integer :: k             ! loop index for bank sites
-    integer :: bin_energyout ! original outgoing energy bin
-    integer :: i_filter      ! index for matching filter bin combination
-    real(8) :: score         ! actual score
-    real(8) :: E_out         ! energy of fission bank site
-
-    ! Save original outgoing energy bin
-    i = t % find_filter(FILTER_ENERGYOUT)
-    bin_energyout = matching_bins(i)
-
-    ! Get the index of delayed group filter
-    j = t % find_filter(FILTER_DELAYEDGROUP)
-
-    ! Get number of energies on filter
-    n = size(t % filters(i) % real_bins)
-
-    ! Since the creation of fission sites is weighted such that it is
-    ! expected to create n_particles sites, we need to multiply the
-    ! score by keff to get the true delayed-nu-fission rate.
-
-    ! loop over number of particles banked
-    do k = 1, p % n_bank
-
-      ! get the delayed group
-      g = fission_bank(n_bank - p % n_bank + k) % delayed_group
-
-      ! check if the particle was born delayed
-      if (g /= 0) then
-
-        ! determine score based on bank site weight and keff
-        score = keff * fission_bank(n_bank - p % n_bank + k) % wgt
-
-        ! determine outgoing energy from fission bank
-        E_out = fission_bank(n_bank - p % n_bank + k) % E
-
-        ! check if outgoing energy is within specified range on filter
-        if (E_out < t % filters(i) % real_bins(1) .or. &
-             E_out > t % filters(i) % real_bins(n)) cycle
-
-        ! change outgoing energy bin
-        matching_bins(i) = binary_search(t % filters(i) % real_bins, n, E_out)
-
-        ! if the delayed group filter is present, tally to corresponding
-        ! delayed group bin if it exists
-        if (j > 0) then
-
-          ! loop over delayed group bins until the corresponding bin is found
-          do d_bin = 1, t % filters(j) % n_bins
-            d = t % filters(j) % int_bins(d_bin)
-
-            ! check whether the delayed group of the particle is equal to the
-            ! delayed group of this bin
-            if (d == g) then
-              call score_fission_delayed_dg(t, d_bin, score, i_score)
-            end if
-          end do
-
-        ! if the delayed group filter is not present, add score to tally
-        else
-
-          ! determine scoring index
-          i_filter = sum((matching_bins(1:t%n_filters) - 1) * t % stride) + 1
-
-          ! Add score to tally
-!$omp atomic
-          t % results(i_score, i_filter) % value = &
-               t % results(i_score, i_filter) % value + score
-        end if
-      end if
-    end do
-
-    ! reset outgoing energy bin
-    matching_bins(i) = bin_energyout
-
-  end subroutine score_fission_delayed_eout
-
 !===============================================================================
 ! SCORE_FISSION_DELAYED_DG helper function used to increment the tally when a
 ! delayed group filter is present.
diff --git a/tests/test_mgxs_library_condense/inputs_true.dat b/tests/test_mgxs_library_condense/inputs_true.dat
index d92cc888a7..0c648376e8 100644
--- a/tests/test_mgxs_library_condense/inputs_true.dat
+++ b/tests/test_mgxs_library_condense/inputs_true.dat
@@ -1 +1 @@
-855919f7a333acff6423527b82656d6a472ea8416002fb475c2d21c676e2f75f5c040d209a3e9a3a6118cf944f2c1bfb8cf2318273b890861d790b1927791a35
\ No newline at end of file
+e2cdca7ea5b3532050af5b12fac26d7ef212d2696bb1b73cdd00929b2243c40d100ad02438c7b090555b49815d0de6c48cf1b4ebf437a48bc80c2d2b4bad292e
\ No newline at end of file
diff --git a/tests/test_mgxs_library_condense/results_true.dat b/tests/test_mgxs_library_condense/results_true.dat
index 13c277b15f..ae768cbe67 100644
--- a/tests/test_mgxs_library_condense/results_true.dat
+++ b/tests/test_mgxs_library_condense/results_true.dat
@@ -34,6 +34,12 @@
 0     10000         1          1   total  0.085835   0.005592
    material  group out nuclide  mean  std. dev.
 0     10000          1   total   1.0   0.046071
+   material  group out nuclide  mean  std. dev.
+0     10000          1   total   1.0   0.051471
+   material  group in nuclide          mean     std. dev.
+0     10000         1   total  4.996730e-07  3.650635e-08
+   material  group in nuclide      mean  std. dev.
+0     10000         1   total  0.090004   0.006367
    material  group in nuclide      mean  std. dev.
 0     10001         1   total  0.311594   0.013793
    material  group in nuclide      mean  std. dev.
@@ -70,6 +76,12 @@
 0     10001         1          1   total   0.0        0.0
    material  group out nuclide  mean  std. dev.
 0     10001          1   total   0.0        0.0
+   material  group out nuclide  mean  std. dev.
+0     10001          1   total   0.0        0.0
+   material  group in nuclide          mean     std. dev.
+0     10001         1   total  5.454760e-07  4.949800e-08
+   material  group in nuclide  mean  std. dev.
+0     10001         1   total   0.0        0.0
    material  group in nuclide      mean  std. dev.
 0     10002         1   total  0.904999   0.043964
    material  group in nuclide      mean  std. dev.
@@ -106,3 +118,9 @@
 0     10002         1          1   total   0.0        0.0
    material  group out nuclide  mean  std. dev.
 0     10002          1   total   0.0        0.0
+   material  group out nuclide  mean  std. dev.
+0     10002          1   total   0.0        0.0
+   material  group in nuclide          mean     std. dev.
+0     10002         1   total  5.773006e-07  5.322132e-08
+   material  group in nuclide  mean  std. dev.
+0     10002         1   total   0.0        0.0
diff --git a/tests/test_mgxs_library_distribcell/inputs_true.dat b/tests/test_mgxs_library_distribcell/inputs_true.dat
index 9c33eeab44..055ce35a57 100644
--- a/tests/test_mgxs_library_distribcell/inputs_true.dat
+++ b/tests/test_mgxs_library_distribcell/inputs_true.dat
@@ -1 +1 @@
-3a3b7f75b326c94a8e5c7efe3046b2fdb887e9f75ecf6eb27587f9450c77cf8fd6acc4198c15bffb4e7ceead6d7b4327c19536bbf9cc35dfaae3f4ce4c26cc1a
\ No newline at end of file
+2d948f3b12293294eaeca231a3df9d51195379e8bb38dd3e68d3bc512a7d08ed52a1109054ca381684ec127268710f6d6e9210ac8154c9b379608e996627624a
\ No newline at end of file
diff --git a/tests/test_mgxs_library_distribcell/results_true.dat b/tests/test_mgxs_library_distribcell/results_true.dat
index 5000d60c3b..c21ca09e99 100644
--- a/tests/test_mgxs_library_distribcell/results_true.dat
+++ b/tests/test_mgxs_library_distribcell/results_true.dat
@@ -34,3 +34,9 @@
 0  (0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13,...         1          1   total   0.0        0.0
                                     avg(distribcell)  group out nuclide  mean  std. dev.
 0  (0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13,...          1   total   0.0        0.0
+                                    avg(distribcell)  group out nuclide  mean  std. dev.
+0  (0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13,...          1   total   0.0        0.0
+                                    avg(distribcell)  group in nuclide      mean     std. dev.
+0  (0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13,...         1   total  0.000001  6.946255e-07
+                                    avg(distribcell)  group in nuclide  mean  std. dev.
+0  (0, 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13,...         1   total   0.0        0.0
diff --git a/tests/test_mgxs_library_hdf5/inputs_true.dat b/tests/test_mgxs_library_hdf5/inputs_true.dat
index d92cc888a7..0c648376e8 100644
--- a/tests/test_mgxs_library_hdf5/inputs_true.dat
+++ b/tests/test_mgxs_library_hdf5/inputs_true.dat
@@ -1 +1 @@
-855919f7a333acff6423527b82656d6a472ea8416002fb475c2d21c676e2f75f5c040d209a3e9a3a6118cf944f2c1bfb8cf2318273b890861d790b1927791a35
\ No newline at end of file
+e2cdca7ea5b3532050af5b12fac26d7ef212d2696bb1b73cdd00929b2243c40d100ad02438c7b090555b49815d0de6c48cf1b4ebf437a48bc80c2d2b4bad292e
\ No newline at end of file
diff --git a/tests/test_mgxs_library_hdf5/results_true.dat b/tests/test_mgxs_library_hdf5/results_true.dat
index 7391b2e427..b2bd28f279 100644
--- a/tests/test_mgxs_library_hdf5/results_true.dat
+++ b/tests/test_mgxs_library_hdf5/results_true.dat
@@ -63,6 +63,15 @@ domain=10000 type=nu-fission matrix
 domain=10000 type=chi
 [ 1.  0.]
 [ 0.04607052  0.        ]
+domain=10000 type=chi-prompt
+[ 1.  0.]
+[ 0.05147146  0.        ]
+domain=10000 type=inverse-velocity
+[  5.70932329e-08   2.85573950e-06]
+[  4.68792969e-09   2.44216503e-07]
+domain=10000 type=prompt-nu-fission
+[ 0.01923922  0.46671903]
+[ 0.00130951  0.04141087]
 domain=10001 type=total
 [ 0.31373767  0.3008214 ]
 [ 0.0155819   0.02805245]
@@ -128,6 +137,15 @@ domain=10001 type=nu-fission matrix
 domain=10001 type=chi
 [ 0.  0.]
 [ 0.  0.]
+domain=10001 type=chi-prompt
+[ 0.  0.]
+[ 0.  0.]
+domain=10001 type=inverse-velocity
+[  5.99598048e-08   2.98549021e-06]
+[  4.55309296e-09   3.41701554e-07]
+domain=10001 type=prompt-nu-fission
+[ 0.  0.]
+[ 0.  0.]
 domain=10002 type=total
 [ 0.66457226  2.05238401]
 [ 0.03121475  0.22434291]
@@ -193,3 +211,12 @@ domain=10002 type=nu-fission matrix
 domain=10002 type=chi
 [ 0.  0.]
 [ 0.  0.]
+domain=10002 type=chi-prompt
+[ 0.  0.]
+[ 0.  0.]
+domain=10002 type=inverse-velocity
+[  6.02207831e-08   3.04495537e-06]
+[  3.78043696e-09   3.60007673e-07]
+domain=10002 type=prompt-nu-fission
+[ 0.  0.]
+[ 0.  0.]
diff --git a/tests/test_mgxs_library_no_nuclides/inputs_true.dat b/tests/test_mgxs_library_no_nuclides/inputs_true.dat
index d92cc888a7..0c648376e8 100644
--- a/tests/test_mgxs_library_no_nuclides/inputs_true.dat
+++ b/tests/test_mgxs_library_no_nuclides/inputs_true.dat
@@ -1 +1 @@
-855919f7a333acff6423527b82656d6a472ea8416002fb475c2d21c676e2f75f5c040d209a3e9a3a6118cf944f2c1bfb8cf2318273b890861d790b1927791a35
\ No newline at end of file
+e2cdca7ea5b3532050af5b12fac26d7ef212d2696bb1b73cdd00929b2243c40d100ad02438c7b090555b49815d0de6c48cf1b4ebf437a48bc80c2d2b4bad292e
\ No newline at end of file
diff --git a/tests/test_mgxs_library_no_nuclides/results_true.dat b/tests/test_mgxs_library_no_nuclides/results_true.dat
index 599cee6c49..141143c8c3 100644
--- a/tests/test_mgxs_library_no_nuclides/results_true.dat
+++ b/tests/test_mgxs_library_no_nuclides/results_true.dat
@@ -75,6 +75,15 @@
    material  group out nuclide  mean  std. dev.
 1     10000          1   total   1.0   0.046071
 0     10000          2   total   0.0   0.000000
+   material  group out nuclide  mean  std. dev.
+1     10000          1   total   1.0   0.051471
+0     10000          2   total   0.0   0.000000
+   material  group in nuclide          mean     std. dev.
+1     10000         1   total  5.709323e-08  4.687930e-09
+0     10000         2   total  2.855740e-06  2.442165e-07
+   material  group in nuclide      mean  std. dev.
+1     10000         1   total  0.019239   0.001310
+0     10000         2   total  0.466719   0.041411
    material  group in nuclide      mean  std. dev.
 1     10001         1   total  0.313738   0.015582
 0     10001         2   total  0.300821   0.028052
@@ -152,6 +161,15 @@
    material  group out nuclide  mean  std. dev.
 1     10001          1   total   0.0        0.0
 0     10001          2   total   0.0        0.0
+   material  group out nuclide  mean  std. dev.
+1     10001          1   total   0.0        0.0
+0     10001          2   total   0.0        0.0
+   material  group in nuclide          mean     std. dev.
+1     10001         1   total  5.995980e-08  4.553093e-09
+0     10001         2   total  2.985490e-06  3.417016e-07
+   material  group in nuclide  mean  std. dev.
+1     10001         1   total   0.0        0.0
+0     10001         2   total   0.0        0.0
    material  group in nuclide      mean  std. dev.
 1     10002         1   total  0.664572   0.031215
 0     10002         2   total  2.052384   0.224343
@@ -229,3 +247,12 @@
    material  group out nuclide  mean  std. dev.
 1     10002          1   total   0.0        0.0
 0     10002          2   total   0.0        0.0
+   material  group out nuclide  mean  std. dev.
+1     10002          1   total   0.0        0.0
+0     10002          2   total   0.0        0.0
+   material  group in nuclide          mean     std. dev.
+1     10002         1   total  6.022078e-08  3.780437e-09
+0     10002         2   total  3.044955e-06  3.600077e-07
+   material  group in nuclide  mean  std. dev.
+1     10002         1   total   0.0        0.0
+0     10002         2   total   0.0        0.0
diff --git a/tests/test_mgxs_library_nuclides/inputs_true.dat b/tests/test_mgxs_library_nuclides/inputs_true.dat
index ea7655c916..a15bbee4c8 100644
--- a/tests/test_mgxs_library_nuclides/inputs_true.dat
+++ b/tests/test_mgxs_library_nuclides/inputs_true.dat
@@ -1 +1 @@
-739796983940a1bad601998cf9ea2f90453a994477c7f675c2fd404d1864fe04a7fbfb5a15c5fe7cf9bad016b78432ba0910baba6f9cc026143761e9f526b62a
\ No newline at end of file
+e4a5f03ab6167e96462c4ef537533fe33b98d7878ae00824c5619356bda8d548b3c71af01ba8c88d5a9b46dd1471d331e6f678a164af922200f2ee3642be6340
\ No newline at end of file
diff --git a/tests/test_mgxs_library_nuclides/results_true.dat b/tests/test_mgxs_library_nuclides/results_true.dat
index d8ebe19a1b..3da8146042 100644
--- a/tests/test_mgxs_library_nuclides/results_true.dat
+++ b/tests/test_mgxs_library_nuclides/results_true.dat
@@ -1 +1 @@
-d56c6bae6bf3cd8950d3f50f089458c1c6c807be780fc97570532c6af6eb7e3057968d3345bd3c363f01315271129ef7b2ca028b05767353e601dc37b035f8a7
\ No newline at end of file
+e494320a213b5704a2ac915a2ba504857be91961ceb6735b6ad05d81eb31c44c9584d5bd9d40baececf1dcb5b030e6ecec63cfbd20639baf69bcb596c5c46591
\ No newline at end of file
diff --git a/tests/test_tallies/inputs_true.dat b/tests/test_tallies/inputs_true.dat
index ddab630f25..17f04238c2 100644
--- a/tests/test_tallies/inputs_true.dat
+++ b/tests/test_tallies/inputs_true.dat
@@ -1 +1 @@
-a392a7a8f27fd2f959b06a6809df29b3482e215da175b19846c248af44dc2ee7e2b05269b802dd9d6ed434506b05092f349436a0048411adab6b296dba0bd683
\ No newline at end of file
+ca47172a42f6c13b244a763c990cbe4811662708ee03307d810a4542ee34bb5db7cc29d66aea313dad95b9f38a4ff7943ded527cfd0c7c8825372fec40cfc0d0
\ No newline at end of file
diff --git a/tests/test_tallies/results_true.dat b/tests/test_tallies/results_true.dat
index ff3a828454..76ad489ca1 100644
--- a/tests/test_tallies/results_true.dat
+++ b/tests/test_tallies/results_true.dat
@@ -1 +1 @@
-a0c7d6ca246ecd7dd5fed06373af142390971401c4e97744f29e55810ab9c231c97c4d8947cdf0b3d2df0ae829a9ddf768e5b2d889bbea34f2b6db0e567db884
\ No newline at end of file
+7d085e38f331083a8c3f7814eb6025f9457a1a8e3899e7d9a65af1ba92ea75c0182fcc605d4eadd3eb7edfc0949df688d39a3f45683e10cbdcfb6d7954f34129
\ No newline at end of file
diff --git a/tests/test_tallies/test_tallies.py b/tests/test_tallies/test_tallies.py
index e51e4616f5..0e8d622680 100644
--- a/tests/test_tallies/test_tallies.py
+++ b/tests/test_tallies/test_tallies.py
@@ -122,7 +122,8 @@ class TalliesTestHarness(PyAPITestHarness):
             t.filters = [cell_filter]
             t.scores = ['absorption', 'delayed-nu-fission', 'events', 'fission',
                         'inverse-velocity', 'kappa-fission', '(n,2n)', '(n,n1)',
-                        '(n,gamma)', 'nu-fission', 'scatter', 'elastic', 'total']
+                        '(n,gamma)', 'nu-fission', 'scatter', 'elastic', 'total',
+                        'prompt-nu-fission']
         score_tallies[0].estimator = 'tracklength'
         score_tallies[1].estimator = 'analog'
         score_tallies[2].estimator = 'collision'