Moved cross_sections.xml to data directory and added a readme file.

data/readme.rst

@@ -0,0 +1,33 @@
+========================
+cross_sections.xml Files
+========================
+
+As a reminder, in order to run a simulation with OpenMC, you will need cross
+section data for each nuclide in your problem. OpenMC is not currently
+distributed with cross section data, so you will have to obtain cross section
+data by other means. The `user's guide`_ offers some helpful advice on how you
+can obtain cross sections.
+
+When OpenMC starts up, it needs a cross_sections.xml file that tells it where to
+find ACE format cross sections. The files in this directory are configured to
+work with a few common cross section sources.
+
+- **cross_sections_ascii.xml** -- This file matches ENDF/B-VII.0 cross sections
+    distributed with MCNP5 / MCNP6 beta.
+
+- **cross_sections_serpent.xml** -- This file matches ENDF/B-VII.0 cross
+    sections distributed with Serpent 1.1.7.
+
+- **cross_sections.xml** - This file matches ENDF/B-VII.0 cross sections
+    distributed with MCNP5 / MCNP6 beta *that have been converted to binary*.
+
+To use any of these files, you need to follow two steps:
+
+1. Change the path on the ``<directory>`` element in the cross_sections.xml file
+to the directory containing the ACE files.
+
+2. Enter the absolute path of the cross_sections.xml on the ``<cross_sections>``
+element in your settings.xml, or set the CROSS_SECTIONS environment variable to
+the full path of the cross_sections.xml file.
+
+.. _user's guide: http://mit-crpg.github.io/openmc/usersguide/install.html#cross-section-configuration