diff --git a/docs/source/usersguide/cross_sections.rst b/docs/source/usersguide/data.rst
similarity index 64%
rename from docs/source/usersguide/cross_sections.rst
rename to docs/source/usersguide/data.rst
index 0cbb581bd..48003ecb7 100644
--- a/docs/source/usersguide/cross_sections.rst
+++ b/docs/source/usersguide/data.rst
@@ -1,55 +1,73 @@
-.. _usersguide_cross_sections:
+.. _usersguide_data:
 
-===========================
-Cross Section Configuration
-===========================
+==================
+Data Configuration
+==================
 
-In order to run a simulation with OpenMC, you will need cross section data for
-each nuclide or material in your problem. OpenMC can be run in continuous-energy
-or multi-group mode.
+OpenMC relies on a variety of physical data in order to carry out transport
+simulations, depletion simulations, and other common tasks. As a user, you are
+responsible for specifying one or more of the following:
 
-In continuous-energy mode, OpenMC uses a native `HDF5
-<https://support.hdfgroup.org/HDF5/>`_ format (see :ref:`io_nuclear_data`) to
-store all nuclear data. Pregenerated HDF5 libraries can be found at
-https://openmc.org; unless you have specific data needs, it is highly
-recommended to use one of the pregenerated libraries. Alternatively, if you have
-ACE format data that was produced with NJOY_, such as that distributed with
-MCNP_ or Serpent_, it can be converted to the HDF5 format using the :ref:`using
-the Python API <create_xs_library>`. Several sources provide openly available
-ACE data including the `ENDF/B`_, JEFF_, and TENDL_ libraries as well as the
-`LANL Nuclear Data Team <https://nucleardata.lanl.gov/>`_. In addition to
-tabulated cross sections in the HDF5 files, OpenMC relies on :ref:`windowed
-multipole <windowed_multipole>` data to perform on-the-fly Doppler broadening.
+- **Cross sections (XML)** -- A :ref:`cross sections XML <io_cross_sections>`
+  file (commonly named ``cross_sections.xml``) contains a listing of other data
+  files, in particular neutron cross sections, photon cross sections, and
+  windowed multipole data. Each of those files, in turn, uses a `HDF5
+  <https://support.hdfgroup.org/HDF5/>`_ format (see :ref:`io_nuclear_data`). In
+  order to run transport simulations with continuous-energy cross sections, you
+  need to specify this file.
 
-In multi-group mode, OpenMC utilizes an HDF5-based library format which can be
-used to describe nuclide- or material-specific quantities.
+- **Depletion chain (XML)** -- A :ref:`depletion chain XML <io_depletion_chain>`
+  file contains decay data, fission product yields, and information on what
+  neutron reactions can result in transmutation. This file is needed for
+  depletion/activation calculations as well as some basic functions in the
+  :mod:`openmc.data` module.
+
+- **Multigroup cross sections (HDF5)** -- OpenMC can also perform transport
+  simulations using multigroup data. In this case, multigroup cross sections are
+  stored in a single :ref:`HDF5 file <io_mgxs_library>`. Thus, in order to run a
+  multigroup transport simulation, this file needs to be specified.
+
+Each of the above files can specified in several ways. In the Python API, a
+:ref:`runtime configuration variable <usersguide_data_runtime>`
+:data:`openmc.config` can be used to specify any of the above and is initialized
+using a set of environment variables.
+
+.. _usersguide_data_runtime:
 
 ---------------------
-Environment Variables
+Runtime Configuration
 ---------------------
 
-When :ref:`scripts_openmc` is run, it will look for several environment
-variables that indicate where cross sections can be found. While the location of
-cross sections can also be indicated through the
-:attr:`openmc.Materials.cross_sections` attribute (or in the :ref:`materials.xml
-<io_materials>` file), if you always use the same set of cross section data, it
-is often easier to just set an environment variable that will be picked up by
-default every time OpenMC is run. The following environment variables are used:
+Data sources for OpenMC can be specified at runtime in Python using the
+:data:`openmc.config` variable. This variable acts like a dictionary and stores
+key-values pairs, where the values are file paths (strings or path-like objects)
+and the key can be one of the following:
 
-:envvar:`OPENMC_CROSS_SECTIONS`
-  Indicates the path to the :ref:`cross_sections.xml <io_cross_sections>`
-  summary file that is used to locate HDF5 format cross section libraries if the
-  user has not specified :attr:`openmc.Materials.cross_sections` (equivalently,
-  the :ref:`cross_sections` in :ref:`materials.xml <io_materials>`).
+``"cross_sections"``
+  Indicates the path to the :ref:`cross sections XML <io_cross_sections>` file
+  that lists HDF5 format neutron cross sections, photon cross sections, and
+  windowed multipole data. At startup, this is initialized with the value of the
+  :envvar:`OPENMC_CROSS_SECTIONS` environment variable. Note that the
+  :attr:`openmc.Materials.cross_sections` attribute will override this, if
+  specified.
 
-:envvar:`OPENMC_MG_CROSS_SECTIONS`
+``"chain_file"``
+  Indicates the path to the :ref:`depletion chain XML <io_depletion_chain>` file
+  that contains decay data, fission product yields, and what neutron reactions
+  may result in transmutation of a target nuclide. At startup, this is
+  initialized with the value of the :envvar:`OPENMC_CHAIN_FILE` environment
+  variable.
+
+``"mg_cross_sections"``
   Indicates the path to an :ref:`HDF5 file <io_mgxs_library>` that contains
-  multi-group cross sections if the user has not specified
-  :attr:`openmc.Materials.cross_sections` (equivalently, the
-  :ref:`cross_sections` in :ref:`materials.xml <io_materials>`).
+  multigroup cross sections. At startup, this is initialized with the value of
+  the :envvar:`OPENMC_MG_CROSS_SECTIONS` environment variable. Note that the
+  :attr:`openmc.Materials.cross_sections` attribute will override this if
+  specified.
 
-To set these environment variables persistently, export them from your shell
-profile (``.profile`` or ``.bashrc`` in bash_).
+If you want to persistently set the environment variables used to initialized
+the configuration, export them from your shell profile (``.profile`` or
+``.bashrc`` in bash_).
 
 .. _bash: http://www.linuxfromscratch.org/blfs/view/6.3/postlfs/profile.html
 
@@ -61,12 +79,13 @@ Using Pregenerated Libraries
 ----------------------------
 
 Various evaluated nuclear data libraries have been processed into the HDF5
-format required by OpenMC and can be found at https://openmc.org. You
-can find both libraries generated by the OpenMC development team as well as
-libraries based on ACE files distributed elsewhere. To use these libraries,
-download the archive file, unpack it, and then set your
-:envvar:`OPENMC_CROSS_SECTIONS` environment variable to the absolute path of
-the ``cross_sections.xml`` file contained in the unpacked directory.
+format required by OpenMC and can be found at https://openmc.org. Unless you
+have specific data needs, it is highly recommended to use one of the
+pregenerated libraries. You can find both libraries generated by the OpenMC
+development team as well as libraries based on ACE files distributed elsewhere.
+To use these libraries, download the archive file, unpack it, and then specify
+the path of the ``cross_sections.xml`` file contained in the unpacked directory
+as described in :ref:`usersguide_data_runtime`.
 
 .. _create_xs_library:
 
@@ -75,6 +94,12 @@ Manually Creating a Library from ACE files
 
 .. currentmodule:: openmc.data
 
+If you have ACE format data that was produced with NJOY_, such as that
+distributed with MCNP_ or Serpent_, it can be converted to the HDF5 format using
+the using the Python API. Several sources provide openly available ACE data
+including the `ENDF/B`_, JEFF_, and TENDL_ libraries as well as the `LANL
+Nuclear Data Team <https://nucleardata.lanl.gov/>`_.
+
 The :mod:`openmc.data` module in the Python API enables users to directly
 convert ACE data to OpenMC's HDF5 format and create a corresponding
 :ref:`cross_sections.xml <io_cross_sections>` file. For those who prefer to use
@@ -224,6 +249,19 @@ relaxation sublibrary files are required:
 Once the HDF5 files have been generated, a library can be created using the
 :class:`DataLibrary` class as described in :ref:`create_xs_library`.
 
+-----------
+Chain Files
+-----------
+
+Pregenerated depletion chain XML files can be found at https://openmc.org.
+Additionally, depletion chains can be generated using the
+:class:`openmc.deplete.Chain` class. In particular, the
+:meth:`~openmc.deplete.Chain.from_endf` method allows a chain to be generated
+starting from a set of ENDF incident neutron, decay, and fission product yield
+sublibrary files. Once you've downloaded or generated a depletion chain XML
+file, make sure to specify its path as described in
+:ref:`usersguide_data_runtime`.
+
 -----------------------
 Windowed Multipole Data
 -----------------------
@@ -241,18 +279,16 @@ The `official ENDF/B-VII.1 HDF5 library
 multipole library, so if you are using this library, the windowed multipole data
 will already be available to you.
 
---------------------------
-Multi-Group Cross Sections
---------------------------
+-------------------------
+Multigroup Cross Sections
+-------------------------
 
-Multi-group cross section libraries are generally tailored to the specific
+Multigroup cross section libraries are generally tailored to the specific
 calculation to be performed.  Therefore, at this point in time, OpenMC is not
-distributed with any pre-existing multi-group cross section libraries.
-However, if  obtained or generated their own library, the user
-should set the :envvar:`OPENMC_MG_CROSS_SECTIONS` environment variable
-to the absolute path of the file library expected to used most frequently.
-
-For an example of how to create a multi-group library, see the `example notebook
+distributed with any pre-existing multigroup cross section libraries. However,
+if a multigroup library file is downloaded or generated, the path to the file
+needs to be specified as described in :ref:`usersguide_data_runtime`. For an
+example of how to create a multigroup library, see the `example notebook
 <../examples/mg-mode-part-i.ipynb>`__.
 
 .. _NJOY: http://www.njoy21.io/
diff --git a/docs/source/usersguide/index.rst b/docs/source/usersguide/index.rst
index fab353c77..511bdc6ce 100644
--- a/docs/source/usersguide/index.rst
+++ b/docs/source/usersguide/index.rst
@@ -13,7 +13,7 @@ essential aspects of using OpenMC to perform simulations.
 
     beginners
     install
-    cross_sections
+    data
     basics
     materials
     geometry
diff --git a/openmc/__init__.py b/openmc/__init__.py
index abcac899f..214695e67 100644
--- a/openmc/__init__.py
+++ b/openmc/__init__.py
@@ -32,6 +32,7 @@ from openmc.search import *
 from openmc.polynomial import *
 from openmc.tracks import *
 from . import examples
+from .config import *
 
 # Import a few names from the model module
 from openmc.model import rectangular_prism, hexagonal_prism, Model
diff --git a/openmc/config.py b/openmc/config.py
new file mode 100644
index 000000000..628df7edc
--- /dev/null
+++ b/openmc/config.py
@@ -0,0 +1,83 @@
+from collections.abc import MutableMapping
+import os
+from pathlib import Path
+import warnings
+
+from openmc.data import DataLibrary
+
+__all__ = ["config"]
+
+
+class _Config(MutableMapping):
+    def __init__(self, data=()):
+        self._mapping = {}
+        self.update(data)
+
+    def __getitem__(self, key):
+        return self._mapping[key]
+
+    def __delitem__(self, key):
+        del self._mapping[key]
+        if key == 'cross_sections':
+            del os.environ['OPENMC_CROSS_SECTIONS']
+        elif key == 'mg_cross_sections':
+            del os.environ['OPENMC_MG_CROSS_SECTIONS']
+
+    def __setitem__(self, key, value):
+        if key == 'cross_sections':
+            # Force environment variable to match
+            self._set_path(key, value)
+            os.environ['OPENMC_CROSS_SECTIONS'] = str(value)
+        elif key == 'mg_cross_sections':
+            self._set_path(key, value)
+            os.environ['OPENMC_MG_CROSS_SECTIONS'] = str(value)
+        elif key == 'chain_file':
+            self._set_path(key, value)
+            os.environ['OPENMC_CHAIN_FILE'] = str(value)
+        else:
+            raise KeyError(f'Unrecognized config key: {key}')
+
+    def __iter__(self):
+        return iter(self._mapping)
+
+    def __len__(self):
+        return len(self._mapping)
+
+    def __repr__(self):
+        return repr(self._mapping)
+
+    def _set_path(self, key, value):
+        self._mapping[key] = p = Path(value)
+        if not p.exists():
+            warnings.warn(f"'{value}' does not exist.")
+
+
+def _default_config():
+    """Return default configuration"""
+    config = _Config()
+
+    # Set cross sections using environment variable
+    if "OPENMC_CROSS_SECTIONS" in os.environ:
+        config['cross_sections'] = os.environ["OPENMC_CROSS_SECTIONS"]
+    if "OPENMC_MG_CROSS_SECTIONS" in os.environ:
+        config['mg_cross_sections'] = os.environ["OPENMC_MG_CROSS_SECTIONS"]
+
+    # Set depletion chain
+    chain_file = os.environ.get("OPENMC_CHAIN_FILE")
+    if (chain_file is None and
+        config.get('cross_sections') is not None and
+        config['cross_sections'].exists()
+    ):
+        # Check for depletion chain in cross_sections.xml
+        data = DataLibrary.from_xml(config['cross_sections'])
+        for lib in reversed(data.libraries):
+            if lib['type'] == 'depletion_chain':
+                chain_file = lib['path']
+                break
+    if chain_file is not None:
+        config['chain_file'] = chain_file
+
+    return config
+
+
+config = _default_config()
diff --git a/openmc/data/library.py b/openmc/data/library.py
index bf937e57a..ad6c65fb6 100644
--- a/openmc/data/library.py
+++ b/openmc/data/library.py
@@ -4,6 +4,7 @@ import pathlib
 
 import h5py
 
+import openmc
 from openmc.mixin import EqualityMixin
 from openmc._xml import clean_indentation, reorder_attributes
 
@@ -124,8 +125,8 @@ class DataLibrary(EqualityMixin):
         Parameters
         ----------
         path : str, optional
-            Path to XML file to read. If not provided, the
-            :envvar:`OPENMC_CROSS_SECTIONS` environment variable will be  used.
+            Path to XML file to read. If not provided,
+            openmc.config['cross_sections'] will be used.
 
         Returns
         -------
@@ -136,15 +137,14 @@ class DataLibrary(EqualityMixin):
 
         data = cls()
 
-        # If path is None, get the cross sections from the
-        # OPENMC_CROSS_SECTIONS environment variable
+        # If path is None, get the cross sections from the global configuration
         if path is None:
-            path = os.environ.get('OPENMC_CROSS_SECTIONS')
+            path = openmc.config.get('cross_sections')
 
-        # Check to make sure there was an environmental variable.
+        # Check to make sure we picked up cross sections
         if path is None:
-            raise ValueError("Either path or OPENMC_CROSS_SECTIONS "
-                             "environmental variable must be set")
+            raise ValueError("Either path or openmc.config['cross_sections'] "
+                             "must be set")
 
         tree = ET.parse(path)
         root = tree.getroot()
@@ -162,7 +162,6 @@ class DataLibrary(EqualityMixin):
             data.libraries.append(library)
 
         # get depletion chain data
-
         dep_node = root.find("depletion_chain")
         if dep_node is not None:
             filename = os.path.join(directory, dep_node.attrib['path'])
diff --git a/openmc/deplete/chain.py b/openmc/deplete/chain.py
index d63bc6bfb..217c9cac0 100644
--- a/openmc/deplete/chain.py
+++ b/openmc/deplete/chain.py
@@ -239,24 +239,6 @@ def replace_missing_fpy(actinide, fpy_data, decay_data):
     return 'U235'
 
 
-def _find_chain_file(cross_sections=None):
-    # First check deprecated OPENMC_DEPLETE_CHAIN environment variable
-    chain_file = os.environ.get("OPENMC_DEPLETE_CHAIN")
-    if chain_file is not None:
-        warn("Use of OPENMC_DEPLETE_CHAIN is deprecated in favor of adding "
-             "depletion_chain to OPENMC_CROSS_SECTIONS", FutureWarning)
-        return chain_file
-
-    # Check for depletion chain in cross_sections.xml
-    data = DataLibrary.from_xml(cross_sections)
-    for lib in reversed(data.libraries):
-        if lib['type'] == 'depletion_chain':
-            return lib['path']
-
-    raise DataError("No depletion chain specified and could not find depletion "
-                    f"chain in {cross_sections}")
-
-
 class Chain:
     """Full representation of a depletion chain.
 
@@ -265,9 +247,8 @@ class Chain:
     yield sublibrary files. The depletion chain used during a depletion
     simulation is indicated by either an argument to
     :class:`openmc.deplete.CoupledOperator` or
-    :class:`openmc.deplete.IndependentOperator`, or through the
-    ``depletion_chain`` item in the :envvar:`OPENMC_CROSS_SECTIONS`
-    environment variable.
+    :class:`openmc.deplete.IndependentOperator`, or through
+    openmc.config['chain_file'].
 
     Attributes
     ----------
diff --git a/openmc/deplete/coupled_operator.py b/openmc/deplete/coupled_operator.py
index dd14f282f..a8c1fa156 100644
--- a/openmc/deplete/coupled_operator.py
+++ b/openmc/deplete/coupled_operator.py
@@ -9,7 +9,6 @@ filesystem.
 """
 
 import copy
-import os
 from warnings import warn
 
 import numpy as np
@@ -22,7 +21,6 @@ from openmc.exceptions import DataError
 import openmc.lib
 from openmc.mpi import comm
 from .abc import OperatorResult
-from .chain import _find_chain_file
 from .openmc_operator import OpenMCOperator, _distribute
 from .results import Results
 from .helpers import (
@@ -48,11 +46,11 @@ def _find_cross_sections(model):
         return model.materials.cross_sections
 
     # otherwise fallback to environment variable
-    cross_sections = os.environ.get("OPENMC_CROSS_SECTIONS")
+    cross_sections = openmc.config.get("cross_sections")
     if cross_sections is None:
         raise DataError(
             "Cross sections were not specified in Model.materials and "
-            "the OPENMC_CROSS_SECTIONS environment variable is not set."
+            "openmc.config['cross_sections'] is not set."
         )
     return cross_sections
 
@@ -103,9 +101,8 @@ class CoupledOperator(OpenMCOperator):
     model : openmc.model.Model
         OpenMC model object
     chain_file : str, optional
-        Path to the depletion chain XML file.  Defaults to the file
-        listed under ``depletion_chain`` in
-        :envvar:`OPENMC_CROSS_SECTIONS` environment variable.
+        Path to the depletion chain XML file. Defaults to
+        ``openmc.config['chain_file']``.
     prev_results : Results, optional
         Results from a previous depletion calculation. If this argument is
         specified, the depletion calculation will start from the latest state
@@ -231,10 +228,8 @@ class CoupledOperator(OpenMCOperator):
                 " model with which to generate the transport Operator."
             raise TypeError(msg)
 
-        # Determine cross sections / depletion chain
+        # Determine cross sections
         cross_sections = _find_cross_sections(model)
-        if chain_file is None:
-            chain_file = _find_chain_file(cross_sections)
 
         check_value('fission yield mode', fission_yield_mode,
                     self._fission_helpers.keys())
diff --git a/openmc/deplete/independent_operator.py b/openmc/deplete/independent_operator.py
index 7f0bf5fb2..f628a7db1 100644
--- a/openmc/deplete/independent_operator.py
+++ b/openmc/deplete/independent_operator.py
@@ -40,7 +40,8 @@ class IndependentOperator(OpenMCOperator):
     micro_xs : MicroXS
         One-group microscopic cross sections in [b] .
     chain_file : str
-        Path to the depletion chain XML file.
+        Path to the depletion chain XML file.  Defaults to
+        ``openmc.config['chain_file']``.
     keff : 2-tuple of float, optional
        keff eigenvalue and uncertainty from transport calculation.
        Default is None.
@@ -111,7 +112,7 @@ class IndependentOperator(OpenMCOperator):
     def __init__(self,
                  materials,
                  micro_xs,
-                 chain_file,
+                 chain_file=None,
                  keff=None,
                  normalization_mode='fission-q',
                  fission_q=None,
diff --git a/openmc/deplete/microxs.py b/openmc/deplete/microxs.py
index 72d87d3b7..7f3a84bf6 100644
--- a/openmc/deplete/microxs.py
+++ b/openmc/deplete/microxs.py
@@ -5,18 +5,16 @@ nuclide names as row indices and reaction names as column indices.
 """
 
 import tempfile
-from pathlib import Path
 from copy import deepcopy
 
-from pandas import DataFrame, read_csv, concat
+from pandas import DataFrame, read_csv
 import numpy as np
 
 from openmc.checkvalue import check_type, check_value, check_iterable_type
+from openmc.exceptions import DataError
 from openmc.mgxs import EnergyGroups, ArbitraryXS, FissionXS
-from openmc.data import DataLibrary
-from openmc import Tallies, StatePoint, Materials, Material
-
-
+from openmc import Tallies, StatePoint, Materials
+import openmc
 from .chain import Chain, REACTIONS
 from .coupled_operator import _find_cross_sections, _get_nuclides_with_data
 
@@ -35,7 +33,7 @@ class MicroXS(DataFrame):
     def from_model(cls,
                    model,
                    reaction_domain,
-                   chain_file,
+                   chain_file=None,
                    dilute_initial=1.0e3,
                    energy_bounds=(0, 20e6),
                    run_kwargs=None):
@@ -48,11 +46,12 @@ class MicroXS(DataFrame):
             OpenMC model object. Must contain geometry, materials, and settings.
         reaction_domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.RegularMesh
             Domain in which to tally reaction rates.
-        chain_file : str
+        chain_file : str, optional
             Path to the depletion chain XML file that will be used in depletion
             simulation. Used to determine cross sections for materials not
-            present in the inital composition.
-        dilute_initial : float
+            present in the inital composition. Defaults to
+            ``openmc.config['chain_file']``.
+        dilute_initial : float, optional
             Initial atom density [atoms/cm^3] to add for nuclides that
             are zero in initial condition to ensure they exist in the cross
             section data. Only done for nuclides with reaction rates.
@@ -144,6 +143,13 @@ class MicroXS(DataFrame):
             :class:`openmc.Materials` object with nuclides added to burnable
             materials.
         """
+        if chain_file is None:
+            chain_file = openmc.config.get('chain_file')
+            if chain_file is None:
+                raise DataError(
+                    "No depletion chain specified and could not find depletion "
+                    "chain in openmc.config['chain_file']"
+                )
         chain = Chain.from_xml(chain_file)
         reactions = chain.reactions
         cross_sections = _find_cross_sections(model)
diff --git a/openmc/deplete/openmc_operator.py b/openmc/deplete/openmc_operator.py
index 76f8e964e..ca1516803 100644
--- a/openmc/deplete/openmc_operator.py
+++ b/openmc/deplete/openmc_operator.py
@@ -11,6 +11,7 @@ from collections import OrderedDict
 import numpy as np
 
 import openmc
+from openmc.exceptions import DataError
 from openmc.mpi import comm
 from .abc import TransportOperator, OperatorResult
 from .atom_number import AtomNumber
@@ -57,9 +58,8 @@ class OpenMCOperator(TransportOperator):
         Path to continuous energy cross section library, or object containing
         one-group cross-sections.
     chain_file : str, optional
-        Path to the depletion chain XML file.  Defaults to the file
-        listed under ``depletion_chain`` in
-        :envvar:`OPENMC_CROSS_SECTIONS` environment variable.
+        Path to the depletion chain XML file. Defaults to
+        openmc.config['chain_file'].
     prev_results : Results, optional
         Results from a previous depletion calculation. If this argument is
         specified, the depletion calculation will start from the latest state
@@ -83,7 +83,6 @@ class OpenMCOperator(TransportOperator):
         if ``reduce_chain`` evaluates to true. The default value of
         ``None`` implies no limit on the depth.
 
-
     Attributes
     ----------
     materials : openmc.Materials
@@ -133,6 +132,15 @@ class OpenMCOperator(TransportOperator):
             reduce_chain=False,
             reduce_chain_level=None):
 
+        # If chain file was not specified, try to get it from global config
+        if chain_file is None:
+            chain_file = openmc.config.get('chain_file')
+            if chain_file is None:
+                raise DataError(
+                    "No depletion chain specified and could not find depletion "
+                    "chain in openmc.config['chain_file']"
+                )
+
         super().__init__(chain_file, fission_q, dilute_initial, prev_results)
         self.round_number = False
         self.materials = materials
diff --git a/openmc/deplete/results.py b/openmc/deplete/results.py
index 15a65e98a..b2c8e7733 100644
--- a/openmc/deplete/results.py
+++ b/openmc/deplete/results.py
@@ -392,7 +392,7 @@ class Results(list):
             as such cannot be used in subsequent transport calculations.
             If not provided, nuclides from the cross_sections element of
             materials.xml will be used. If that element is not present,
-            nuclides from OPENMC_CROSS_SECTIONS will be used.
+            nuclides from openmc.config['cross_sections'] will be used.
 
         Returns
         -------
@@ -412,7 +412,7 @@ class Results(list):
         # the new materials XML file. The precedence of nuclides to select
         # is first ones provided as a kwarg here, then ones specified
         # in the materials.xml file if provided, then finally from
-        # the environment variable OPENMC_CROSS_SECTIONS.
+        # openmc.config['cross_sections'].
         if nuc_with_data:
             cv.check_iterable_type('nuclide names', nuc_with_data, str)
             available_cross_sections = nuc_with_data
diff --git a/openmc/element.py b/openmc/element.py
index 1473bd63e..49aeaf464 100644
--- a/openmc/element.py
+++ b/openmc/element.py
@@ -1,9 +1,9 @@
 from collections import OrderedDict
-import os
 import re
 from xml.etree import ElementTree as ET
 
 import openmc.checkvalue as cv
+import openmc
 from openmc.data import NATURAL_ABUNDANCE, atomic_mass, \
     isotopes as natural_isotopes
 
@@ -40,10 +40,10 @@ class Element(str):
                cross_sections=None):
         """Expand natural element into its naturally-occurring isotopes.
 
-        An optional cross_sections argument or the :envvar:`OPENMC_CROSS_SECTIONS`
-        environment variable is used to specify a cross_sections.xml file.
-        If the cross_sections.xml file is found, the element is expanded only
-        into the isotopes/nuclides present in cross_sections.xml. If no
+        An optional cross_sections argument or the ``cross_sections``
+        configuration value is used to specify a cross_sections.xml file. If the
+        cross_sections.xml file is found, the element is expanded only into the
+        isotopes/nuclides present in cross_sections.xml. If no
         cross_sections.xml file is found, the element is expanded based on its
         naturally occurring isotopes.
 
@@ -54,12 +54,13 @@ class Element(str):
         percent_type : {'ao', 'wo'}
             'ao' for atom percent and 'wo' for weight percent
         enrichment : float, optional
-            Enrichment of an enrichment_target nuclide in percent (ao or wo).
-            If enrichment_target is not supplied then it is enrichment for U235
-            in weight percent. For example, input 4.95 for 4.95 weight percent
+            Enrichment of an enrichment_target nuclide in percent (ao or wo). If
+            enrichment_target is not supplied then it is enrichment for U235 in
+            weight percent. For example, input 4.95 for 4.95 weight percent
             enriched U. Default is None (natural composition).
         enrichment_target: str, optional
-            Single nuclide name to enrich from a natural composition (e.g., 'O16')
+            Single nuclide name to enrich from a natural composition (e.g.,
+            'O16')
 
             .. versionadded:: 0.12
         enrichment_type: {'ao', 'wo'}, optional
@@ -82,8 +83,8 @@ class Element(str):
         ValueError
             No data is available for any of natural isotopes of the element
         ValueError
-            If only some natural isotopes are available in the cross-section data
-            library and the element is not O, W, or Ta
+            If only some natural isotopes are available in the cross-section
+            data library and the element is not O, W, or Ta
         ValueError
             If a non-naturally-occurring isotope is requested
         ValueError
@@ -101,8 +102,8 @@ class Element(str):
         `ORNL/CSD/TM-244 <https://doi.org/10.2172/5561567>`_ is used to
         calculate the weight fractions of U234, U235, U236, and U238. Namely,
         the weight fraction of U234 and U236 are taken to be 0.89% and 0.46%,
-        respectively, of the U235 weight fraction. The remainder of the
-        isotopic weight is assigned to U238.
+        respectively, of the U235 weight fraction. The remainder of the isotopic
+        weight is assigned to U238.
 
         When the `enrichment` argument is specified with `enrichment_target`, a
         general enrichment procedure is used for elements composed of exactly
@@ -125,10 +126,10 @@ class Element(str):
         # Create dict to store the expanded nuclides and abundances
         abundances = OrderedDict()
 
-        # If cross_sections is None, get the cross sections from the
-        # OPENMC_CROSS_SECTIONS environment variable
+        # If cross_sections is None, get the cross sections from the global
+        # configuration
         if cross_sections is None:
-            cross_sections = os.environ.get('OPENMC_CROSS_SECTIONS')
+            cross_sections = openmc.config.get('cross_sections')
 
         # If a cross_sections library is present, check natural nuclides
         # against the nuclides in the library
diff --git a/openmc/material.py b/openmc/material.py
index ece70df88..1b2b46a73 100644
--- a/openmc/material.py
+++ b/openmc/material.py
@@ -1310,9 +1310,8 @@ class Materials(cv.CheckedList):
     """Collection of Materials used for an OpenMC simulation.
 
     This class corresponds directly to the materials.xml input file. It can be
-    thought of as a normal Python list where each member is a
-    :class:`Material`. It behaves like a list as the following example
-    demonstrates:
+    thought of as a normal Python list where each member is a :class:`Material`.
+    It behaves like a list as the following example demonstrates:
 
     >>> fuel = openmc.Material()
     >>> clad = openmc.Material()
@@ -1332,9 +1331,9 @@ class Materials(cv.CheckedList):
         Indicates the path to an XML cross section listing file (usually named
         cross_sections.xml). If it is not set, the
         :envvar:`OPENMC_CROSS_SECTIONS` environment variable will be used for
-        continuous-energy calculations and
-        :envvar:`OPENMC_MG_CROSS_SECTIONS` will be used for multi-group
-        calculations to find the path to the HDF5 cross section file.
+        continuous-energy calculations and :envvar:`OPENMC_MG_CROSS_SECTIONS`
+        will be used for multi-group calculations to find the path to the HDF5
+        cross section file.
 
     """
 
diff --git a/openmc/mgxs_library.py b/openmc/mgxs_library.py
index 67ad151cc..98a99196e 100644
--- a/openmc/mgxs_library.py
+++ b/openmc/mgxs_library.py
@@ -1,6 +1,5 @@
 import copy
 from numbers import Real, Integral
-import os
 
 import h5py
 import numpy as np
@@ -2536,8 +2535,7 @@ class MGXSLibrary:
         ----------
         filename : str, optional
             Name of HDF5 file containing MGXS data. Default is None.
-            If not provided, the value of the OPENMC_MG_CROSS_SECTIONS
-            environmental variable will be used
+            If not provided, openmc.config['mg_cross_sections'] will be used.
 
         Returns
         -------
@@ -2545,15 +2543,14 @@ class MGXSLibrary:
             Multi-group cross section data object.
 
         """
-        # If filename is None, get the cross sections from the
-        # OPENMC_CROSS_SECTIONS environment variable
+        # If filename is None, get the cross sections from openmc.config
         if filename is None:
-            filename = os.environ.get('OPENMC_MG_CROSS_SECTIONS')
+            filename = openmc.config.get('mg_cross_sections')
 
         # Check to make sure there was an environmental variable.
         if filename is None:
-            raise ValueError("Either path or OPENMC_MG_CROSS_SECTIONS "
-                             "environmental variable must be set")
+            raise ValueError("Either path or openmc.config['mg_cross_sections']"
+                             "must be set")
 
         check_type('filename', filename, str)
         file = h5py.File(filename, 'r')
diff --git a/tests/unit_tests/test_config.py b/tests/unit_tests/test_config.py
new file mode 100644
index 000000000..1d3c0f173
--- /dev/null
+++ b/tests/unit_tests/test_config.py
@@ -0,0 +1,43 @@
+from collections.abc import Mapping
+import os
+
+import openmc
+import pytest
+
+
+@pytest.fixture(autouse=True, scope='module')
+def reset_config():
+    config = dict(openmc.config)
+    try:
+        yield
+    finally:
+        openmc.config.clear()
+        openmc.config.update(config)
+
+
+def test_config_basics():
+    assert isinstance(openmc.config, Mapping)
+    for key, value in openmc.config.items():
+        assert isinstance(key, str)
+        assert isinstance(value, os.PathLike)
+
+    # Set and delete
+    openmc.config['cross_sections'] = '/path/to/cross_sections.xml'
+    del openmc.config['cross_sections']
+    assert 'cross_sections' not in openmc.config
+    assert 'OPENMC_CROSS_SECTIONS' not in os.environ
+
+    # Can't use any key
+    with pytest.raises(KeyError):
+        openmc.config['🐖'] = '/like/to/eat/bacon'
+
+
+def test_config_set_envvar():
+    openmc.config['cross_sections'] = '/path/to/cross_sections.xml'
+    assert os.environ['OPENMC_CROSS_SECTIONS'] == '/path/to/cross_sections.xml'
+
+    openmc.config['mg_cross_sections'] = '/path/to/mg_cross_sections.h5'
+    assert os.environ['OPENMC_MG_CROSS_SECTIONS'] == '/path/to/mg_cross_sections.h5'
+
+    openmc.config['chain_file'] = '/path/to/chain_file.xml'
+    assert os.environ['OPENMC_CHAIN_FILE'] == '/path/to/chain_file.xml'
diff --git a/tests/unit_tests/test_deplete_operator.py b/tests/unit_tests/test_deplete_operator.py
index 5fe8715ac..6ea89fc4a 100644
--- a/tests/unit_tests/test_deplete_operator.py
+++ b/tests/unit_tests/test_deplete_operator.py
@@ -1,37 +1,15 @@
 """Basic unit tests for openmc.deplete.Operator instantiation
 
-Modifies and resets environment variable OPENMC_CROSS_SECTIONS
-to a custom file with new depletion_chain node
 """
 
 from pathlib import Path
 
-import pytest
 from openmc.deplete.abc import TransportOperator
-from openmc.deplete.chain import Chain, _find_chain_file
+from openmc.deplete.chain import Chain
 
-BARE_XS_FILE = "bare_cross_sections.xml"
 CHAIN_PATH = Path(__file__).parents[1] / "chain_simple.xml"
 
 
-@pytest.fixture()
-def bare_xs(run_in_tmpdir):
-    """Create a very basic cross_sections file, return simple Chain.
-
-    """
-
-    bare_xs_contents = """<?xml version="1.0"?>
-<cross_sections>
-  <depletion_chain path="{}" />
-</cross_sections>
-""".format(CHAIN_PATH)
-
-    with open(BARE_XS_FILE, "w") as out:
-        out.write(bare_xs_contents)
-
-    yield BARE_XS_FILE
-
-
 class BareDepleteOperator(TransportOperator):
     """Very basic class for testing the initialization."""
 
@@ -52,10 +30,10 @@ class BareDepleteOperator(TransportOperator):
         pass
 
 
-def test_operator_init(bare_xs):
+def test_operator_init():
     """The test uses a temporary dummy chain. This file will be removed
     at the end of the test, and only contains a depletion_chain node."""
-    bare_op = BareDepleteOperator(_find_chain_file(bare_xs))
+    bare_op = BareDepleteOperator(CHAIN_PATH)
     act_chain = bare_op.chain
     ref_chain = Chain.from_xml(CHAIN_PATH)
     assert len(act_chain) == len(ref_chain)
@@ -73,8 +51,7 @@ def test_operator_init(bare_xs):
 def test_operator_fiss_q():
     """Make sure fission q values can be set"""
     new_q = {"U235": 2.0E8, "U238": 2.0E8, "U234": 5.0E7}
-    chain_file = Path(__file__).parents[1] / "chain_simple.xml"
-    operator = BareDepleteOperator(chain_file=chain_file, fission_q=new_q)
+    operator = BareDepleteOperator(chain_file=CHAIN_PATH, fission_q=new_q)
     mod_chain = operator.chain
     for name, q in new_q.items():
         chain_nuc = mod_chain[name]