# I'm using the MCNP5 cross section data.

**URL:** <https://openmc.discourse.group/t/im-using-the-mcnp5-cross-section-data/23>\
**Category:** User Support\
**Created:** [June 11, 2013, 9:26am UTC](https://openmc.discourse.group/t/im-using-the-mcnp5-cross-section-data/23 "2013-06-11T09:26:23Z")\
**Posts on this page:** 2\
**Page:** 1

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**Author:** ![kcd1006](https://avatars.discourse-cdn.com/v4/letter/k/258eb7/32.png) [@kcd1006](https://openmc.discourse.group/u/kcd1006)\
**Post date:** [June 11, 2013, 9:26am UTC](https://openmc.discourse.group/t/im-using-the-mcnp5-cross-section-data/23/1 "2013-06-11T09:26:23Z")

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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

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**Author:** ![Adam\_Nelson](https://avatars.discourse-cdn.com/v4/letter/a/8edcca/32.png) [@Adam\_Nelson](https://openmc.discourse.group/u/Adam_Nelson)\
**Post date:** [June 11, 2013, 1:42pm UTC](https://openmc.discourse.group/t/im-using-the-mcnp5-cross-section-data/23/2 "2013-06-11T13:42:51Z")

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The \<default\_xs\> entry ([http://mit-crpg.github.io/openmc/usersguide/input.html#default-xs-element](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 _.7_c libraries.  
If you want 293, use .70c, if you want 600, use 71c.