I'm using the MCNP5 cross section data.

User Support
1

I’m using the MCNP5 cross section data by setting the path in the ‘cross_sections_ascii.xml’ file as follows.

E:\LANL\MCNP_DATA</directory>

I checked the MCNP data files, and I found there are various temperatures in the MCNP data. (e.g. 293.6K & 600K)
Then, does OpenMC use cross sections at 296.3K? Or those at 600K?
I’m pretty sure it uses only one temperature.

Thanks,
Chidong

2

The <default_xs> entry (http://mit-crpg.github.io/openmc/usersguide/input.html#default-xs-element) of materials.xml, or the ‘xs=’ attribute in each nuclide’s entry of materials.xml points to the different ENDF libraries which are those evaluated/processed at different temperatures. The cross_sections.xml file lists the temperatures associated with each of the .7c libraries.
If you want 293, use .70c, if you want 600, use 71c.