# Multi-group Reflected Slab

**URL:** <https://openmc.discourse.group/t/multi-group-reflected-slab/428>\
**Category:** User Support\
**Created:** [April 23, 2019, 1:44am UTC](https://openmc.discourse.group/t/multi-group-reflected-slab/428 "2019-04-23T01:44:04Z")\
**Posts on this page:** 2\
**Page:** 1

<div class="post-metadata">

**Author:** ![Hunter\_Belanger](https://avatars.discourse-cdn.com/v4/letter/h/13edae/32.png) [@Hunter\_Belanger](https://openmc.discourse.group/u/Hunter_Belanger)\
**Post date:** [April 23, 2019, 1:44am UTC](https://openmc.discourse.group/t/multi-group-reflected-slab/428/1 "2019-04-23T01:44:04Z")

</div>

Hello,

I am trying to learn OpenMC and figured I would start with a simple 1D reflected slab, with two energy groups. I followed the example notebook to define my MG cross section libraries, and then defined the geometry, but I get a k=91, while I know that k=1.00914, from MCNP. I am not sure what I am doing wrong however with this system. From plotting the geometry, it looks okay, so I am assuming my problem lies with the cross section definitions, but I am not sure. Any help or advice would be greatly appreciated!

Cordially,

Hunter

[Reflected\_Slab\_2\_Group.ipynb](https://openmc.discourse.group/uploads/short-url/zCVlj1rEWmkGqzx0AupLWDALJ0F.ipynb) (18.2 KB)

---

<div class="post-metadata">

**Author:** ![paulromano](https://yyz2.discourse-cdn.com/free1/user_avatar/openmc.discourse.group/paulromano/32/486_2.png) [@paulromano](https://openmc.discourse.group/u/paulromano)\
**Post date:** [April 24, 2019, 12:16pm UTC](https://openmc.discourse.group/t/multi-group-reflected-slab/428/2 "2019-04-24T12:16:23Z")

</div>

Hi Hunter,

The issue is that XSData.set\_nu\_fission expects the product of nu and the fission cross section, not just nu alone. When I change the values to be 2.5_0.00339 and 2.5_0.074 and rerun the notebook, I end up with a keff of 1.00906 +/- 0.00077.

Best regards,  
Paul