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37 lines
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37 lines
1.7 KiB
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========================
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cross_sections.xml Files
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========================
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As a reminder, in order to run a simulation with OpenMC, you will need cross
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section data for each nuclide in your problem. OpenMC is not currently
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distributed with cross section data, so you will have to obtain cross section
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data by other means. The `user's guide`_ offers some helpful advice on how you
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can obtain cross sections.
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When OpenMC starts up, it needs a cross_sections.xml file that tells it where to
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find ACE format cross sections. The files in this directory are configured to
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work with a few common cross section sources.
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- **cross_sections_ascii.xml** -- This file matches ENDF/B-VII.0 cross sections
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distributed with MCNP5 / MCNP6 beta.
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- **cross_sections_nndc.xml** -- This file matches ENDF/B-VII.1 cross sections
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distributed from the `NNDC website`_.
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- **cross_sections_serpent.xml** -- This file matches ENDF/B-VII.0 cross
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sections distributed with Serpent 1.1.7.
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- **cross_sections.xml** - This file matches ENDF/B-VII.0 cross sections
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distributed with MCNP5 / MCNP6 beta *that have been converted to binary*.
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To use any of these files, you need to follow two steps:
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1. Change the path on the ``<directory>`` element in the cross_sections.xml file
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to the directory containing the ACE files.
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2. Enter the absolute path of the cross_sections.xml on the ``<cross_sections>``
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element in your settings.xml, or set the CROSS_SECTIONS environment variable to
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the full path of the cross_sections.xml file.
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.. _user's guide: http://mit-crpg.github.io/openmc/usersguide/install.html#cross-section-configuration
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.. _NNDC website: http://www.nndc.bnl.gov/endf/b7.1/acefiles.html
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