diff --git a/docs/source/usersguide/cross_sections.rst b/docs/source/usersguide/data.rst
similarity index 99%
rename from docs/source/usersguide/cross_sections.rst
rename to docs/source/usersguide/data.rst
index 0cbb581bd4..6a135df76a 100644
--- a/docs/source/usersguide/cross_sections.rst
+++ b/docs/source/usersguide/data.rst
@@ -1,8 +1,8 @@
-.. _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
diff --git a/docs/source/usersguide/index.rst b/docs/source/usersguide/index.rst
index fab353c773..511bdc6cef 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