From 76ddc858d6331a296c6ddc672174ff1e2348223b Mon Sep 17 00:00:00 2001 From: Paul Romano Date: Fri, 21 Jun 2013 16:11:17 -0400 Subject: [PATCH] Added note about TENDL in documentation. --- docs/source/usersguide/install.rst | 4 +++- 1 file changed, 3 insertions(+), 1 deletion(-) 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