openmc-dev · GuySten · Mar 18, 2026 · Mar 18, 2026 · Mar 18, 2026 · Mar 18, 2026
diff --git a/docs/source/methods/energy_deposition.rst b/docs/source/methods/energy_deposition.rst
@@ -137,7 +137,7 @@ run with :math:`N918` reflecting fission heating computed from NJOY.
       + E_{i, \gamma, d}\right]\sigma_{i, f}(E).
 
 This modified heating data is stored as the MT=901 reaction and will be scored
-if ``heating-local`` is included in :attr:`openmc.Tally.scores`.
+if ``heating-local`` is included in :attr:`openmc.TallyBase.scores`.
 
 Coupled Neutron-Photon Transport
 --------------------------------
@@ -155,7 +155,7 @@ Let :math:`N301` represent the total heating number returned from this
       + \left[E_{i, fr}(E) + E_{i, \beta}(E)\right]\sigma_{i, f}(E).
 
 This modified heating data is stored as the MT=301 reaction and will be scored
-if ``heating`` is included in :attr:`openmc.Tally.scores`.
+if ``heating`` is included in :attr:`openmc.TallyBase.scores`.
 
 Photons and Charged Particles
 -----------------------------

diff --git a/docs/source/pythonapi/base.rst b/docs/source/pythonapi/base.rst
@@ -153,7 +153,10 @@ Constructing Tallies
    openmc.MeshMaterialVolumes
    openmc.Trigger
    openmc.TallyDerivative
-   openmc.Tally
+   openmc.TallyBase
+   openmc.VolumeTally
+   openmc.SurfaceTally
+   openmc.PulseHeightTally
    openmc.Tallies
 
 Meshes

diff --git a/docs/source/usersguide/decay_sources.rst b/docs/source/usersguide/decay_sources.rst
@@ -100,7 +100,7 @@ create a shallow copy of the original neutron model (available as the
 ``neutron_model`` attribute) and store it in the ``photon_model`` attribute. We
 can use this to define tallies specific to the photon model::
 
-    dose_tally = openmc.Tally()
+    dose_tally = openmc.VolumeTally()
     ...
     r2s.photon_model.tallies = [dose_tally]
 

diff --git a/docs/source/usersguide/kinetics.rst b/docs/source/usersguide/kinetics.rst
@@ -78,7 +78,7 @@ Alternatively, each of the tallies can be manually defined using group-wise or
 total :math:`\beta_{\text{eff}}` specified by providing a 6-group
 :class:`openmc.DelayedGroupFilter`::
 
-    beta_tally = openmc.Tally(name="group-beta-score")
+    beta_tally = openmc.VolumeTally(name="group-beta-score")
     beta_tally.scores = ["ifp-beta-numerator"]
 
     # Add DelayedGroupFilter to enable group-wise tallies
@@ -87,7 +87,7 @@ total :math:`\beta_{\text{eff}}` specified by providing a 6-group
 Here is an example showing how to declare the three available IFP scores in a
 single tally::
 
-    tally = openmc.Tally(name="ifp-scores")
+    tally = openmc.VolumeTally(name="ifp-scores")
     tally.scores = [
         "ifp-time-numerator",
         "ifp-beta-numerator",

diff --git a/docs/source/usersguide/processing.rst b/docs/source/usersguide/processing.rst
@@ -126,7 +126,7 @@ a flat array for a given dataset can be passed directly to this method.
     mesh.lower_left = [-5, -10, -15]
     mesh.upper_right = [5, 10, 15]
     mesh_filter = openmc.MeshFilter(mesh)
-    tally = openmc.Tally()
+    tally = openmc.VolumeTally()
     tally.filters = [mesh_filter]
     tally.scores = ['flux']
 
@@ -136,10 +136,10 @@ a flat array for a given dataset can be passed directly to this method.
     # provide the data as-is to the method
     mesh.write_data_to_vtk('flux.vtk', {'flux-mean': tally.mean})
 
-The :class:`~openmc.Tally` object also provides a way to expand the dimensions
+The :class:`~openmc.TallyBase` object also provides a way to expand the dimensions
 of the mesh filter into a meaningful form where indexing the mesh filter
 dimensions results in intuitive slicing of structured meshes by setting
-``expand_dims=True`` when using :meth:`openmc.Tally.get_reshaped_data`. This
+``expand_dims=True`` when using :meth:`openmc.TallyBase.get_reshaped_data`. This
 reshaping does cause flat indexing of the data to change, however. As noted
 above, provided datasets are allowed to be shaped so long as such datasets have
 shapes that match the mesh dimensions. The ability to pass datasets in this way
@@ -158,7 +158,7 @@ demonstrates such a case for tally with both a :class:`~openmc.MeshFilter` and
     mesh.upper_right = [5, 10, 15]
     mesh_filter = openmc.MeshFilter(mesh)
     energy_filter = openmc.EnergyFilter([0.0, 1.0, 20.0e6])
-    tally = openmc.Tally()
+    tally = openmc.VolumeTally()
     tally.filters = [mesh_filter, energy_filter]
     tally.scores = ['flux']
 

diff --git a/docs/source/usersguide/random_ray.rst b/docs/source/usersguide/random_ray.rst
@@ -1139,7 +1139,7 @@ given below:
     energy_filter = openmc.EnergyFilter(group_edges)
 
     # Create tally using our two filters and add scores
-    tally = openmc.Tally()
+    tally = openmc.VolumeTally()
     tally.filters = [mesh_filter, energy_filter]
     tally.scores = ['flux', 'fission', 'nu-fission']
 
@@ -1222,7 +1222,7 @@ given below:
     energy_filter = openmc.EnergyFilter(ebins)
 
     # Create tally using our two filters and add scores
-    tally = openmc.Tally()
+    tally = openmc.VolumeTally()
     tally.filters = [mesh_filter, energy_filter]
     tally.scores = ['flux']
 

diff --git a/docs/source/usersguide/tallies.rst b/docs/source/usersguide/tallies.rst
@@ -7,7 +7,8 @@ Specifying Tallies
 .. currentmodule:: openmc
 
 In order to obtain estimates of physical quantities in your simulation, you need
-to create one or more tallies using the :class:`openmc.Tally` class. As
+to create one or more tallies using some of the :class:`openmc.TallyBase` subclasses. 
+The most used :class:`openmc.TallyBase` subclass is the :class:`openmc.VolumeTally` class. As
 explained in detail in the :ref:`theory manual <methods_tallies>`, tallies
 provide estimates of a scoring function times the flux integrated over some
 region of phase space, as in:
@@ -53,8 +54,8 @@ Energies are specified in eV and need to be monotonically increasing.
 .. caution:: An energy bin between zero and the lowest energy specified is not
              included by default as it is in MCNP.
 
-Once you have created a filter, it should be assigned to a :class:`openmc.Tally`
-instance through the :attr:`Tally.filters` attribute::
+Once you have created a filter, it should be assigned to a :class:`openmc.TallyBase`
+instance through the :attr:`TallyBase.filters` attribute::
 
   tally.filters.append(cell_filter)
   tally.filters.append(energy_filter)
@@ -74,15 +75,15 @@ Scores
 ------
 
 To specify the scoring functions, a list of strings needs to be given to the
-:attr:`Tally.scores` attribute. You can score the flux ('flux'), or a reaction
+:attr:`TallyBase.scores` attribute. You can score the flux ('flux'), or a reaction
 rate ('total', 'fission', etc.). For example, to tally the elastic scattering
 rate and the fission neutron production, you'd assign::
 
   tally.scores = ['elastic', 'nu-fission']
 
 With no further specification, you will get the total elastic scattering rate
 and the total fission neutron production. If you want reaction rates for a
-particular nuclide or set of nuclides, you can set the :attr:`Tally.nuclides`
+particular nuclide or set of nuclides, you can set the :attr:`TallyBase.nuclides`
 attribute to a list of strings indicating which nuclides. The nuclide names
 should follow the same :ref:`naming convention <usersguide_naming>` as that used
 for material specification. If we wanted the reaction rates only for U235 and

diff --git a/examples/custom_source/build_xml.py b/examples/custom_source/build_xml.py
@@ -29,7 +29,7 @@
 mesh.upper_right = (5.0, 5.0)
 mesh.dimension = (50, 50)
 
-tally = openmc.Tally()
+tally = openmc.VolumeTally()
 tally.filters = [openmc.MeshFilter(mesh)]
 tally.scores = ['flux']
 tallies = openmc.Tallies([tally])

diff --git a/examples/lattice/hexagonal/build_xml.py b/examples/lattice/hexagonal/build_xml.py
@@ -153,7 +153,7 @@
 ###############################################################################
 
 # Instantiate a distribcell Tally
-tally = openmc.Tally(tally_id=1)
+tally = openmc.VolumeTally(tally_id=1)
 tally.filters = [openmc.DistribcellFilter(cell2)]
 tally.scores = ['total']
 

diff --git a/examples/lattice/nested/build_xml.py b/examples/lattice/nested/build_xml.py
@@ -160,7 +160,7 @@
 mesh_filter = openmc.MeshFilter(mesh)
 
 # Instantiate the Tally
-tally = openmc.Tally(tally_id=1)
+tally = openmc.VolumeTally(tally_id=1)
 tally.filters = [mesh_filter]
 tally.scores = ['total']
 

diff --git a/examples/lattice/simple/build_xml.py b/examples/lattice/simple/build_xml.py
@@ -157,7 +157,7 @@
 trigger.scores = ['all']
 
 # Instantiate the Tally
-tally = openmc.Tally(tally_id=1)
+tally = openmc.VolumeTally(tally_id=1)
 tally.filters = [mesh_filter]
 tally.scores = ['total']
 tally.triggers = [trigger]

diff --git a/examples/parameterized_custom_source/build_xml.py b/examples/parameterized_custom_source/build_xml.py
@@ -30,7 +30,7 @@
 mesh.upper_right = (5.0, 5.0)
 mesh.dimension = (50, 50)
 
-tally = openmc.Tally()
+tally = openmc.VolumeTally()
 tally.filters = [openmc.MeshFilter(mesh)]
 tally.scores = ['flux']
 tallies = openmc.Tallies([tally])

diff --git a/examples/pincell/build_xml.py b/examples/pincell/build_xml.py
@@ -93,7 +93,7 @@
 mesh_filter = openmc.MeshFilter(mesh)
 
 # Now use the mesh filter in a tally and indicate what scores are desired
-mesh_tally = openmc.Tally(name="Mesh tally")
+mesh_tally = openmc.VolumeTally(name="Mesh tally")
 mesh_tally.filters = [mesh_filter]
 mesh_tally.scores = ['flux', 'fission', 'nu-fission']
 
@@ -104,7 +104,7 @@
 energies = np.logspace(log10(e_min), log10(e_max), groups + 1)
 energy_filter = openmc.EnergyFilter(energies)
 
-spectrum_tally = openmc.Tally(name="Flux spectrum")
+spectrum_tally = openmc.VolumeTally(name="Flux spectrum")
 spectrum_tally.filters = [energy_filter]
 spectrum_tally.scores = ['flux']
 

diff --git a/examples/pincell_multigroup/build_xml.py b/examples/pincell_multigroup/build_xml.py
@@ -131,7 +131,7 @@
 mesh_filter = openmc.MeshFilter(mesh)
 
 # Now use the mesh filter in a tally and indicate what scores are desired
-mesh_tally = openmc.Tally(name="Mesh tally")
+mesh_tally = openmc.VolumeTally(name="Mesh tally")
 mesh_tally.filters = [mesh_filter]
 mesh_tally.scores = ['flux', 'fission', 'nu-fission']
 
@@ -142,7 +142,7 @@
 energies = np.logspace(log10(e_min), log10(e_max), groups + 1)
 energy_filter = openmc.EnergyFilter(energies)
 
-spectrum_tally = openmc.Tally(name="Flux spectrum")
+spectrum_tally = openmc.VolumeTally(name="Flux spectrum")
 spectrum_tally.filters = [energy_filter]
 spectrum_tally.scores = ['flux']
 

diff --git a/examples/pincell_pulsed/run_pulse.py b/examples/pincell_pulsed/run_pulse.py
@@ -76,7 +76,7 @@
 time_filter = openmc.TimeFilter(t_grid)
 
 # Tally for total neutron density in time
-density_tally = openmc.Tally(name="Density")
+density_tally = openmc.VolumeTally(name="Density")
 density_tally.filters = [time_filter]
 density_tally.scores = ["inverse-velocity"]
 

diff --git a/examples/pincell_random_ray/build_xml.py b/examples/pincell_random_ray/build_xml.py
@@ -180,7 +180,7 @@
 energy_filter = openmc.EnergyFilter(group_edges)
 
 # Now use the mesh filter in a tally and indicate what scores are desired
-tally = openmc.Tally(name="Mesh and Energy tally")
+tally = openmc.VolumeTally(name="Mesh and Energy tally")
 tally.filters = [mesh_filter, energy_filter]
 tally.scores = ['flux', 'fission', 'nu-fission']