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.