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Added note about TENDL in documentation.

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Paul Romano 2013-06-21 16:11:17 -04:00
parent 182ebdba45
commit 76ddc858d6
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docs/source/usersguide/install.rst
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@ -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
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Running OpenMC