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Update installation guide with information about NNDC data.

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Paul Romano 2014-02-17 17:46:10 -05:00
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docs/source/usersguide/install.rst
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@ -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