diff --git a/docs/source/methods/depletion.rst b/docs/source/methods/depletion.rst
index 0e06b3a50..66ee50397 100644
--- a/docs/source/methods/depletion.rst
+++ b/docs/source/methods/depletion.rst
@@ -234,7 +234,7 @@ of FPY:
    Only applicable to transport-coupled depletion
 
 The method for transport-coupled depletion can be selected through the
-``fission_yield_mode`` argument to the :class:`openmc.deplete.Operator`
+``fission_yield_mode`` argument to the :class:`openmc.deplete.CoupledOperator`
 constructor.
 
 Power Normalization
@@ -246,8 +246,8 @@ compute an absolute reaction rate in reactions per second. To do so, the
 reaction rates are normalized based on a specified power. A complete
 description of how this normalization can be performed is described in
 :ref:`usersguide_tally_normalization`. Here, we simply note that the main
-depletion class, :class:`openmc.deplete.Operator`, allows the user to choose
-one of two methods for estimating the heating rate, including:
+depletion class, :class:`openmc.deplete.CoupledOperator`, allows the user to
+choose one of two methods for estimating the heating rate, including:
 
 1. Using fixed Q values from a depletion chain file (useful for comparisons to
    other codes that use fixed Q values), or
@@ -255,4 +255,4 @@ one of two methods for estimating the heating rate, including:
    energy-dependent estimate of the true heating rate.
 
 The method for normalization can be chosen through the ``normalization_mode``
-argument to the :class:`openmc.deplete.Operator` class.
+argument to the :class:`openmc.deplete.CoupledOperator` class.
diff --git a/docs/source/pythonapi/deplete.rst b/docs/source/pythonapi/deplete.rst
index cc36a17c7..bc07ed1b0 100644
--- a/docs/source/pythonapi/deplete.rst
+++ b/docs/source/pythonapi/deplete.rst
@@ -18,8 +18,8 @@ are:
 The former is responsible for obtaining transmuation reaction rates. The latter
 is responsible for projecting reaction rates and compositions forward in
 calendar time across some step size :math:`\Delta t`, and obtaining new
-compositions given a power or power density. The :class:`Operator` class is
-provided to obtain reaction rates via tallies through OpenMC's transport
+compositions given a power or power density. The :class:`CoupledOperator` class
+is provided to obtain reaction rates via tallies through OpenMC's transport
 solver, and the :class:`IndependentOperator` class is provided to obtain
 reaction rates from cross-section data. Several classes are provided that
 implement different time-integration algorithms for depletion calculations,
@@ -49,11 +49,11 @@ provided by OpenMC are available using the following classes:
    :nosignatures:
    :template: mycallable.rst
 
-   Operator
+   CoupledOperator
    IndependentOperator
 
-The :class:`Operator` and :class:`IndependentOperator` classes must also have
-some knowledge of how nuclides transmute and decay. This is handled by the 
+The :class:`CoupledOperator` and :class:`IndependentOperator` classes must also
+have some knowledge of how nuclides transmute and decay. This is handled by the 
 :class:`Chain`.
 
 Minimal Example
@@ -71,7 +71,7 @@ A minimal example for performing depletion would be:
 
     # Representation of a depletion chain
     >>> chain_file = "chain_casl.xml"
-    >>> operator = openmc.deplete.Operator(
+    >>> operator = openmc.deplete.CoupledOperator(
     ...     model, chain_file)
 
     # Set up 5 time steps of one day each
@@ -177,7 +177,7 @@ with :func:`cram.CRAM48` being the default.
    :class:`multiprocessing.pool.Pool` class. If set to ``None`` (default), the
    number returned by :func:`os.cpu_count` is used.
 
-The following classes are used to help the :class:`openmc.deplete.Operator`
+The following classes are used to help the :class:`openmc.deplete.CoupledOperator`
 compute quantities like effective fission yields, reaction rates, and
 total system energy.
 
@@ -194,6 +194,8 @@ total system energy.
    helpers.FissionYieldCutoffHelper
    helpers.FluxCollapseHelper
 
+The :class:`openmc.deplete.IndependentOperator` uses inner class subclassed from
+those listed to perform similar calculations.
 
 Intermediate Classes
 --------------------
diff --git a/docs/source/usersguide/depletion.rst b/docs/source/usersguide/depletion.rst
index 7409c6610..990031c40 100644
--- a/docs/source/usersguide/depletion.rst
+++ b/docs/source/usersguide/depletion.rst
@@ -47,17 +47,17 @@ Transport-coupled depletion
 
 This category of operator solves the transport equation to obtain transmutation
 reaction rates. At present, the :mod:`openmc.deplete` module offers a single
-transport-coupled operator, :class:`openmc.deplete.Operator` (which uses the
-OpenMC transport solver), but in principle additional transport-coupled operator
-classes based on other transport codes could be implemented and no changes to
-the depletion solver itself would be needed. The
-:class:`openmc.deplete.Operator` class requires a :class:`~openmc.Model`
+transport-coupled operator, :class:`openmc.deplete.CoupledOperator` (which uses
+the OpenMC transport solver), but in principle additional transport-coupled
+operator classes based on other transport codes could be implemented and no
+changes to the depletion solver itself would be needed. The
+:class:`openmc.deplete.CoupledOperator` class requires a :class:`~openmc.Model`
 instance containing material, geometry, and settings information::
 
     model = openmc.Model()
     ...
 
-    op = openmc.deplete.Operator(model)
+    op = openmc.deplete.CoupledOperator(model)
 
 Any material that contains a fissionable nuclide is depleted by default, but
 this can behavior can be changed with the :attr:`Material.depletable` attribute.
@@ -86,11 +86,11 @@ using the :attr:`Material.depletable` attribute::
     mat = openmc.Material()
     mat.depletable = True
 
-When constructing the :class:`~openmc.deplete.Operator`, you should indicate
-that normalization of tally results will be done based on the source rate rather
-than a power or power density::
+When constructing the :class:`~openmc.deplete.CoupledOperator`, you should
+indicate that normalization of tally results will be done based on the source
+rate rather than a power or power density::
 
-    op = openmc.deplete.Operator(model, normalization_mode='source-rate')
+    op = openmc.deplete.CoupledOperator(model, normalization_mode='source-rate')
 
 Finally, when creating a depletion integrator, use the ``source_rates`` argument::
 
@@ -109,13 +109,14 @@ Energy Deposition
 ~~~~~~~~~~~~~~~~~
 
 The default energy deposition mode, ``"fission-q"``, instructs the
-:class:`~openmc.deplete.Operator` to normalize reaction rates using the product
-of fission reaction rates and fission Q values taken from the depletion chain.
-This approach does not consider indirect contributions to energy deposition,
-such as neutron heating and energy from secondary photons. In doing this, the
-energy deposited during a transport calculation will be lower than expected.
-This causes the reaction rates to be over-adjusted to hit the user-specific
-power, or power density, leading to an over-depletion of burnable materials.
+:class:`~openmc.deplete.CoupledOperator` to normalize reaction rates using the
+product of fission reaction rates and fission Q values taken from the depletion
+chain. This approach does not consider indirect contributions to energy
+deposition, such as neutron heating and energy from secondary photons. In doing
+this, the energy deposited during a transport calculation will be lower than
+expected. This causes the reaction rates to be over-adjusted to hit the
+user-specific power, or power density, leading to an over-depletion of burnable
+materials.
 
 There are some remedies. First, the fission Q values can be directly set in a
 variety of ways. This requires knowing what the total fission energy release
@@ -130,10 +131,10 @@ should be, including indirect components. Some examples are provided below::
     # create a modified chain and write it to a new file
     chain = openmc.deplete.Chain.from_xml("chain.xml", fission_q)
     chain.export_to_xml("chain_mod_q.xml")
-    op = openmc.deplete.Operator(model, "chain_mod_q.xml")
+    op = openmc.deplete.CoupledOperator(model, "chain_mod_q.xml")
 
     # alternatively, pass the modified fission Q directly to the operator
-    op = openmc.deplete.Operator(model, "chain.xml",
+    op = openmc.deplete.CoupledOperator(model, "chain.xml",
         fission_q=fission_q)
 
 
@@ -141,7 +142,7 @@ A more complete way to model the energy deposition is to use the modified
 heating reactions described in :ref:`methods_heating`. These values can be used
 to normalize reaction rates instead of using the fission reaction rates with::
 
-    op = openmc.deplete.Operator(model, "chain.xml",
+    op = openmc.deplete.CoupledOperator(model, "chain.xml",
         normalization_mode="energy-deposition")
 
 These modified heating libraries can be generated by running the latest version
@@ -174,7 +175,7 @@ the next transport step.
 This can be countered by instructing the operator to treat repeated instances
 of the same material as a unique material definition with::
 
-    op = openmc.deplete.Operator(model, chain_file,
+    op = openmc.deplete.CoupledOperator(model, chain_file,
         diff_burnable_mats=True)
 
 For our example problem, this would deplete fuel on the outer region of the
diff --git a/openmc/deplete/abc.py b/openmc/deplete/abc.py
index 5cb3b254a..dd1fd5b04 100644
--- a/openmc/deplete/abc.py
+++ b/openmc/deplete/abc.py
@@ -82,7 +82,8 @@ class TransportOperator(ABC):
     operator that takes a vector of material compositions and returns an
     eigenvalue and reaction rates. This abstract class sets the requirements
     for such a transport operator. Users should instantiate
-    :class:`openmc.deplete.Operator` rather than this class.
+    :class:`openmc.deplete.CoupledOperator` or
+    :class:`openmc.deplete.IndependentOperator` rather than this class.
 
     Parameters
     ----------
@@ -220,9 +221,11 @@ class ReactionRateHelper(ABC):
     Parameters
     ----------
     n_nucs : int
-        Number of burnable nuclides tracked by :class:`openmc.deplete.Operator`
+        Number of burnable nuclides tracked by
+        :class:`openmc.deplete.abc.TransportOperator`
     n_react : int
-        Number of reactions tracked by :class:`openmc.deplete.Operator`
+        Number of reactions tracked by
+        :class:`openmc.deplete.abc.TransportOperator`
 
     Attributes
     ----------
@@ -291,9 +294,9 @@ class NormalizationHelper(ABC):
     """Abstract class for obtaining normalization factor on tallies
 
     This helper class determines how reaction rates calculated by an instance of
-    :class:`openmc.deplete.Operator` should be normalized for the purpose of
-    constructing a burnup matrix. Based on the method chosen, the power or
-    source rate provided by the user, and reaction rates from a
+    :class:`openmc.deplete.abc.TransportOperator` should be normalized for the
+    purpose of constructing a burnup matrix. Based on the method chosen, the
+    power or source rate provided by the user, and reaction rates from a
     :class:`ReactionRateHelper`, this class will scale reaction rates to the
     correct values.
 
@@ -301,7 +304,7 @@ class NormalizationHelper(ABC):
     ----------
     nuclides : list of str
         All nuclides with desired reaction rates. Ordered to be
-        consistent with :class:`openmc.deplete.Operator`
+        consistent with :class:`openmc.deplete.abc.TransportOperator`
 
     """
 
@@ -315,9 +318,9 @@ class NormalizationHelper(ABC):
     def prepare(self, chain_nucs, rate_index):
         """Perform work needed to obtain energy produced
 
-        This method is called prior to the transport simulations
-        in :meth:`openmc.deplete.Operator.initial_condition`. Only used for
-        energy-based normalization.
+        This method is called prior to calculating the reaction rates
+        in :meth:`openmc.deplete.abc.TransportOperator.initial_condition`. Only
+        used for energy-based normalization.
 
         Parameters
         ----------
@@ -430,7 +433,7 @@ class FissionYieldHelper(ABC):
     def unpack():
         """Unpack tally data prior to compute fission yields.
 
-        Called after a :meth:`openmc.deplete.Operator.__call__`
+        Called after a :meth:`openmc.deplete.abc.TransportOperator.__call__`
         routine during the normalization of reaction rates.
 
         Not necessary for all subclasses to implement, unless tallies
@@ -451,7 +454,7 @@ class FissionYieldHelper(ABC):
         mat_indexes : iterable of int
             Indices of tallied materials that will have their fission
             yields computed by this helper. Necessary as the
-            :class:`openmc.deplete.Operator` that uses this helper
+            :class:`openmc.deplete.CoupledOperator` that uses this helper
             may only burn a subset of all materials when running
             in parallel mode.
         """
@@ -463,14 +466,15 @@ class FissionYieldHelper(ABC):
         ----------
         nuclides : iterable of str
             Nuclides with non-zero densities from the
-            :class:`openmc.deplete.Operator`
+            :class:`openmc.deplete.abc.TransportOperator`
 
         Returns
         -------
         nuclides : list of str
-            Union of nuclides that the :class:`openmc.deplete.Operator`
-            says have non-zero densities at this stage and those that
-            have yield data. Sorted by nuclide name
+            Union of nuclides that the
+            :class:`openmc.deplete.abc.TransportOperator` says have non-zero
+            densities at this stage and those that have yield data. Sorted by
+            nuclide name
 
         """
         return sorted(self._chain_set & set(nuclides))
@@ -484,7 +488,7 @@ class FissionYieldHelper(ABC):
 
         Parameters
         ----------
-        operator : openmc.deplete.TransportOperator
+        operator : openmc.deplete.abc.TransportOperator
             Operator with a depletion chain
         kwargs: optional
             Additional keyword arguments to be used in constuction
@@ -505,7 +509,7 @@ class Integrator(ABC):
     _params = r"""
     Parameters
     ----------
-    operator : openmc.deplete.TransportOperator
+    operator : openmc.deplete.abc.TransportOperator
         Operator to perform transport simulations
     timesteps : iterable of float or iterable of tuple
         Array of timesteps. Note that values are not cumulative. The units are
@@ -548,7 +552,7 @@ class Integrator(ABC):
 
     Attributes
     ----------
-    operator : openmc.deplete.TransportOperator
+    operator : openmc.deplete.abc.TransportOperator
         Operator to perform transport simulations
     chain : openmc.deplete.Chain
         Depletion chain
@@ -843,7 +847,7 @@ class SIIntegrator(Integrator):
     _params = r"""
     Parameters
     ----------
-    operator : openmc.deplete.TransportOperator
+    operator : openmc.deplete.abc.TransportOperator
         The operator object to simulate on.
     timesteps : iterable of float or iterable of tuple
         Array of timesteps. Note that values are not cumulative. The units are
@@ -889,7 +893,7 @@ class SIIntegrator(Integrator):
 
     Attributes
     ----------
-    operator : openmc.deplete.TransportOperator
+    operator : openmc.deplete.abc.TransportOperator
         Operator to perform transport simulations
     chain : openmc.deplete.Chain
         Depletion chain
diff --git a/openmc/deplete/chain.py b/openmc/deplete/chain.py
index 0c1c34cf7..d63bc6bfb 100644
--- a/openmc/deplete/chain.py
+++ b/openmc/deplete/chain.py
@@ -264,7 +264,8 @@ class Chain:
     requires a list of ENDF incident neutron, decay, and neutron fission product
     yield sublibrary files. The depletion chain used during a depletion
     simulation is indicated by either an argument to
-    :class:`openmc.deplete.Operator` or through the
+    :class:`openmc.deplete.CoupledOperator` or
+    :class:`openmc.deplete.IndependentOperator`, or through the
     ``depletion_chain`` item in the :envvar:`OPENMC_CROSS_SECTIONS`
     environment variable.
 
diff --git a/openmc/deplete/helpers.py b/openmc/deplete/helpers.py
index 4fcde4f6b..262e61b35 100644
--- a/openmc/deplete/helpers.py
+++ b/openmc/deplete/helpers.py
@@ -68,7 +68,7 @@ class TalliedFissionYieldHelper(FissionYieldHelper):
         mat_indexes : iterable of int
             Indices of tallied materials that will have their fission
             yields computed by this helper. Necessary as the
-            :class:`openmc.deplete.Operator` that uses this helper
+            :class:`openmc.deplete.CoupledOperator` that uses this helper
             may only burn a subset of all materials when running
             in parallel mode.
         """
@@ -133,9 +133,11 @@ class DirectReactionRateHelper(ReactionRateHelper):
     Parameters
     ----------
     n_nucs : int
-        Number of burnable nuclides tracked by :class:`openmc.deplete.Operator`
+        Number of burnable nuclides tracked by
+        :class:`openmc.deplete.CoupledOperator`
     n_react : int
-        Number of reactions tracked by :class:`openmc.deplete.Operator`
+        Number of reactions tracked by an instance of
+        :class:`openmc.deplete.CoupledOperator`
 
     Attributes
     ----------
@@ -217,9 +219,10 @@ class FluxCollapseHelper(ReactionRateHelper):
     Parameters
     ----------
     n_nucs : int
-        Number of burnable nuclides tracked by :class:`openmc.deplete.Operator`
+        Number of burnable nuclides tracked by
+        :class:`openmc.deplete.CoupledOperator`
     n_react : int
-        Number of reactions tracked by :class:`openmc.deplete.Operator`
+        Number of reactions tracked by :class:`openmc.deplete.CoupledOperator`
     energies : iterable of float
         Energy group boundaries for flux spectrum in [eV]
     reactions : iterable of str
@@ -385,7 +388,7 @@ class ChainFissionHelper(EnergyNormalizationHelper):
     ----------
     nuclides : list of str
         All nuclides with desired reaction rates. Ordered to be
-        consistent with :class:`openmc.deplete.Operator`
+        consistent with :class:`openmc.deplete.CoupledOperator`
     energy : float
         Total energy [J/s/source neutron] produced in a transport simulation.
         Updated in the material iteration with :meth:`update`.
@@ -637,7 +640,7 @@ class FissionYieldCutoffHelper(TalliedFissionYieldHelper):
         that occurred below ``cutoff``. The number
         of materials in the first axis corresponds
         to the number of materials burned by the
-        :class:`openmc.deplete.Operator`
+        :class:`openmc.deplete.CoupledOperator`
     """
 
     def __init__(self, chain_nuclides, n_bmats, cutoff=112.0,
@@ -694,7 +697,7 @@ class FissionYieldCutoffHelper(TalliedFissionYieldHelper):
 
         Parameters
         ----------
-        operator : openmc.deplete.Operator
+        operator : openmc.deplete.CoupledOperator
             Operator with a chain and burnable materials
         kwargs:
             Additional keyword arguments to be used in construction
@@ -720,7 +723,7 @@ class FissionYieldCutoffHelper(TalliedFissionYieldHelper):
         mat_indexes : iterable of int
             Indices of tallied materials that will have their fission
             yields computed by this helper. Necessary as the
-            :class:`openmc.deplete.Operator` that uses this helper
+            :class:`openmc.deplete.CoupledOperator` that uses this helper
             may only burn a subset of all materials when running
             in parallel mode.
         """
@@ -845,7 +848,7 @@ class AveragedFissionYieldHelper(TalliedFissionYieldHelper):
         mat_indexes : iterable of int
             Indices of tallied materials that will have their fission
             yields computed by this helper. Necessary as the
-            :class:`openmc.deplete.Operator` that uses this helper
+            :class:`openmc.deplete.CoupledOperator` that uses this helper
             may only burn a subset of all materials when running
             in parallel mode.
         """
@@ -958,7 +961,7 @@ class AveragedFissionYieldHelper(TalliedFissionYieldHelper):
 
         Parameters
         ----------
-        operator : openmc.deplete.TransportOperator
+        operator : openmc.deplete.CoupledOperator
             Operator with a depletion chain
         kwargs :
             Additional keyword arguments to be used in construction
diff --git a/openmc/deplete/operator.py b/openmc/deplete/operator.py
index 2f8dcfba2..8bf031453 100644
--- a/openmc/deplete/operator.py
+++ b/openmc/deplete/operator.py
@@ -30,7 +30,7 @@ from .helpers import (
     SourceRateHelper, FluxCollapseHelper)
 
 
-__all__ = ["Operator", "OperatorResult"]
+__all__ = ["CoupledOperator", "Operator", "OperatorResult"]
 
 
 def _find_cross_sections(model):
@@ -56,13 +56,13 @@ def _find_cross_sections(model):
     return cross_sections
 
 
-class Operator(OpenMCOperator):
-    """OpenMC transport operator for depletion.
+class CoupledOperator(OpenMCOperator):
+    """Transport-coupled operator for depletion.
 
-    Instances of this class can be used to perform depletion using OpenMC as the
-    transport operator. Normally, a user needn't call methods of this class
-    directly. Instead, an instance of this class is passed to an integrator
-    class, such as :class:`openmc.deplete.CECMIntegrator`.
+    Instances of this class can be used to perform transport-coupled depletion
+    using OpenMC's transport solver. Normally, a user needn't call methods of
+    this class directly. Instead, an instance of this class is passed to an
+    integrator class, such as :class:`openmc.deplete.CECMIntegrator`.
 
     .. versionchanged:: 0.13.0
         The geometry and settings parameters have been replaced with a
@@ -193,11 +193,11 @@ class Operator(OpenMCOperator):
                  reduce_chain=False, reduce_chain_level=None):
         # check for old call to constructor
         if isinstance(model, openmc.Geometry):
-            msg = "As of version 0.13.0 openmc.deplete.Operator requires an " \
-                "openmc.Model object rather than the openmc.Geometry and " \
-                "openmc.Settings parameters. Please use the geometry and " \
-                "settings objects passed here to create a model with which " \
-                "to generate the depletion Operator."
+            msg = "As of version 0.13.0 openmc.deplete.CoupledOperator " \
+                "requires an openmc.Model object rather than the " \
+                "openmc.Geometry and openmc.Settings parameters. Please use " \
+                "the geometry and settings objects passed here to create a " \
+                " model with which to generate the depletion Operator."
             raise TypeError(msg)
 
         # Determine cross sections / depletion chain
@@ -516,3 +516,6 @@ class Operator(OpenMCOperator):
         """Finalize a depletion simulation and release resources."""
         if self.cleanup_when_done:
             openmc.lib.finalize()
+
+# Retain deprecated name for the time being
+Operator = CoupledOperator
diff --git a/openmc/deplete/results.py b/openmc/deplete/results.py
index c238002fe..1c5535601 100644
--- a/openmc/deplete/results.py
+++ b/openmc/deplete/results.py
@@ -81,9 +81,9 @@ class Results(list):
         .. note::
             Initial values for some isotopes that do not appear in
             initial concentrations may be non-zero, depending on the
-            value of the :attr:`openmc.deplete.Operator.dilute_initial`
-            attribute. The :class:`openmc.deplete.Operator` class adds isotopes
-            according to this setting, which can be set to zero.
+            value of the :attr:`openmc.deplete.CoupledOperator.dilute_initial`
+            attribute. The :class:`openmc.deplete.CoupledOperator` class adds
+            isotopes according to this setting, which can be set to zero.
 
         Parameters
         ----------
@@ -145,8 +145,8 @@ class Results(list):
 
             Initial values for some isotopes that do not appear in
             initial concentrations may be non-zero, depending on the
-            value of :class:`openmc.deplete.Operator` ``dilute_initial``
-            The :class:`openmc.deplete.Operator` adds isotopes according
+            value of :class:`openmc.deplete.CoupledOperator` ``dilute_initial``
+            The :class:`openmc.deplete.CoupledOperator` adds isotopes according
             to this setting, which can be set to zero.
 
         Parameters