diff --git a/docs/source/usersguide/install.rst b/docs/source/usersguide/install.rst
index 077befb5b..ef01aeff7 100644
--- a/docs/source/usersguide/install.rst
+++ b/docs/source/usersguide/install.rst
@@ -239,7 +239,8 @@ 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_.
+MCNP_ or Serpent_. The TALYS-based evaluated nuclear data library, TENDL_, is
+openly available in ACE format.
 
 Using JEFF Cross Sections from OECD/NEA
 ---------------------------------------
@@ -292,6 +293,7 @@ file.
 .. _here: http://www.oecd-nea.org/dbdata/pubs/jeff312-cd.html
 .. _MCNP: http://mcnp.lanl.gov
 .. _Serpent: http://montecarlo.vtt.fi
+.. _TENDL: ftp://ftp.nrg.eu/pub/www/talys/tendl2012/tendl2012.html
 
 --------------
 Running OpenMC