diff --git a/docs/source/usersguide/install.rst b/docs/source/usersguide/install.rst
index 164b0151be..ff33d9a6c6 100644
--- a/docs/source/usersguide/install.rst
+++ b/docs/source/usersguide/install.rst
@@ -264,10 +264,27 @@ Cross Section Configuration
 
 In order to run a simulation with OpenMC, you will need cross section data for
 each nuclide in your problem. Since OpenMC uses ACE format cross sections, you
-can use nuclear data that was processed with NJOY, such as that distributed with
-MCNP_ or Serpent_. The TALYS-based evaluated nuclear data library, TENDL_, is
+can use nuclear data that was processed with NJOY_, such as that distributed
+with MCNP_ or Serpent_. Several sources provide free processed ACE data as
+described below. The TALYS-based evaluated nuclear data library, TENDL_, is also
 openly available in ACE format.
 
+Using ENDF/B-VII.1 Cross Sections from NNDC
+-------------------------------------------
+
+The NNDC_ provides ACE data from the ENDF/B-VII.1 neutron and thermal scattering
+sublibraries at four temperatures processed using NJOY_. To use this data with
+OpenMC, a script is provided with OpenMC that will automatically download,
+extract, and set up a confiuration file:
+
+.. code-block:: sh
+
+    cd openmc/data
+    python get_nndc_data.py
+
+At this point, you should set the :envvar:`CROSS_SECTIONS` environment variable
+to the absolute path of the file ``openmc/data/nndc/cross_sections.xml``.
+
 Using JEFF Cross Sections from OECD/NEA
 ---------------------------------------
 
@@ -314,6 +331,8 @@ distribution to the location of the Serpent cross sections. Then, either set the
 environment variable to the absolute path of the ``cross_sections_serpent.xml``
 file.
 
+.. _NJOY: http://t2.lanl.gov/nis/codes.shtml
+.. _NNDC: http://www.nndc.bnl.gov/endf/b7.1/acefiles.html
 .. _NEA: http://www.oecd-nea.org
 .. _JEFF: http://www.oecd-nea.org/dbdata/jeff/
 .. _here: http://www.oecd-nea.org/dbdata/pubs/jeff312-cd.html