# Creating Cross Section Files at 959K and 900K

**URL:** <https://openmc.discourse.group/t/creating-cross-section-files-at-959k-and-900k/5286>\
**Category:** User Support\
**Created:** [March 3, 2025, 10:49pm UTC](https://openmc.discourse.group/t/creating-cross-section-files-at-959k-and-900k/5286 "2025-03-03T22:49:01Z")\
**Posts on this page:** 3\
**Page:** 1

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**Author:** ![rogejose](https://avatars.discourse-cdn.com/v4/letter/r/7c8e57/32.png) [@rogejose](https://openmc.discourse.group/u/rogejose)\
**Post date:** [March 3, 2025, 10:49pm UTC](https://openmc.discourse.group/t/creating-cross-section-files-at-959k-and-900k/5286/1 "2025-03-03T22:49:01Z")

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Hello, I am trying to create a custom cross section file at high temperatures (900K and 959K) for both thermal scattering data and incident neutron data for graphite, FLiBe, and U235. I’m not sure how to go about doing this or if what I was doing is correct? I believe I have code which makes the necessary .h5 files for uranium, but I don’t know how to do the same for thermal scattering with graphite at 900K:

Here is the uranium:

 ![image](https://global.discourse-cdn.com/free1/uploads/openmc/original/2X/2/2330054c1ffc2bccc22117c3d606b41104c77946.png)

Also when actually using these files in simulations, do you still need to define the temperature of the material? Or can I just register the cross section file “as is” and just run it?

Thank you!

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<div class="post-metadata">

**Author:** ![rogejose](https://avatars.discourse-cdn.com/v4/letter/r/7c8e57/32.png) [@rogejose](https://openmc.discourse.group/u/rogejose)\
**Post date:** [March 3, 2025, 10:49pm UTC](https://openmc.discourse.group/t/creating-cross-section-files-at-959k-and-900k/5286/2 "2025-03-03T22:49:38Z")

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The forum only allows me to upload one photo, here is the code snippet I have for graphite which uses ACE files:

 ![image](https://global.discourse-cdn.com/free1/uploads/openmc/original/2X/1/1cad4b57b3a495fcc5afa6f30916b285c7435844.png)

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**Author:** ![cjwiguna](https://yyz2.discourse-cdn.com/free1/user_avatar/openmc.discourse.group/cjwiguna/32/3728_2.png) [@cjwiguna](https://openmc.discourse.group/u/cjwiguna)\
**Post date:** [March 13, 2025, 4:02pm UTC](https://openmc.discourse.group/t/creating-cross-section-files-at-959k-and-900k/5286/3 "2025-03-13T16:02:44Z")

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Hello @rogejose, i think for the library from endf, we should also add the `temperatures` parameter in `from_njoy` method:

```auto
openmc.data.IncidentNeutron.from_njoy(endf_file_path, temperatures = [900, 959, other_temperatures...])

```

Also in the simulation, we should define the temperature in the material / cell (I prefer to put temperature in cell).