# Flux distribution and unit of reaction rate for the research reactor fuel assembly

**URL:** <https://openmc.discourse.group/t/flux-distribution-and-unit-of-reaction-rate-for-the-research-reactor-fuel-assembly/4408>\
**Category:** User Support\
**Created:** [June 18, 2024, 2:31pm UTC](https://openmc.discourse.group/t/flux-distribution-and-unit-of-reaction-rate-for-the-research-reactor-fuel-assembly/4408 "2024-06-18T14:31:43Z")\
**Posts on this page:** 2\
**Page:** 1

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**Author:** ![umarov](https://avatars.discourse-cdn.com/v4/letter/u/ecd19e/32.png) [@umarov](https://openmc.discourse.group/u/umarov)\
**Post date:** [June 18, 2024, 2:31pm UTC](https://openmc.discourse.group/t/flux-distribution-and-unit-of-reaction-rate-for-the-research-reactor-fuel-assembly/4408/1 "2024-06-18T14:31:43Z")

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Dear Users and Experts,

I have designed an IRT-4M fuel assembly for a research reactor and am seeking assistance in plotting 2D and 3D neutron flux distributions as in example ( [Jupyter Notebook Viewer (nbviewer.org)](https://nbviewer.org/github/openmc-dev/openmc-notebooks/blob/main/expansion-filters.ipynb)

 ![image](https://global.discourse-cdn.com/free1/uploads/openmc/original/2X/5/5eb7082c9c97a51909d77d27ca4546ef78061a68.png)  
). Despite multiple attempts, I have been unsuccessful. Could someone provide an example on my code to help me learn how to obtain and visualize flux distributions?

Additionally, I have plotted a fission reaction rate distribution but am uncertain about the units. I expected the reaction rate to be on the order of 10^4 10^6. How can I convert the reaction rate to real values?  
 ![image](https://global.discourse-cdn.com/free1/uploads/openmc/original/2X/2/2416b3cdc35a26946fbc8df4f4a0ded7a3247d3a.png)

Thank you for your assistance.

[IRT-4M.ipynb](https://openmc.discourse.group/uploads/short-url/susOKCtdNYMgvcVhFTW5B5zbtfg.ipynb) (136.8 KB)

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**Author:** ![wahidluthfi](https://yyz2.discourse-cdn.com/free1/user_avatar/openmc.discourse.group/wahidluthfi/32/3375_2.png) [@wahidluthfi](https://openmc.discourse.group/u/wahidluthfi)\
**Post date:** [June 18, 2024, 3:24pm UTC](https://openmc.discourse.group/t/flux-distribution-and-unit-of-reaction-rate-for-the-research-reactor-fuel-assembly/4408/2 "2024-06-18T15:24:44Z")

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I think you can read other discussions regarding tally normalization, Paul has been describing tally normalization properly in this discussion

> [@NOrmalizing Tally to get Flux value](https://openmc.discourse.group/t/normalizing-tally-to-get-flux-value/99/6):
>
> The tally is not normalized by volume, so the units are something like [neutrons-cm/source]. This can be understood quite easily: in a collision estimator tally, OpenMC accumulates 1/Sigma\_t which has units of cm, and similarly for a tracklength tally, OpenMC accumulates the tracklength in cm. When you divide by volume, you get [neutrons/cm^2-source]. The normalization factor Pnu/(Qk) gives you [J/secneutrons/fission/(J/fissionneutrons/source)] = [source/sec] so that when you multiply the flux b…