# Different Burnup Results **URL:** \ **Category:** User Support\ **Created:** [September 17, 2025, 10:24am UTC](https://openmc.discourse.group/t/different-burnup-results/6005 "2025-09-17T10:24:36Z")\ **Posts on this page:** 1\ **Showing post:** 4
**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:** [September 20, 2025, 4:00pm UTC](https://openmc.discourse.group/t/different-burnup-results/6005/4 "2025-09-20T16:00:16Z")
Hai hai, I haven’t tried the diff\_burnable\_mats feature of the openmc, but I have tried to manually create a specific material for each fuel assembly position, and it can change the trend of calculated keff under depletion calculation. You can try that on your SMART core model, either by using different material axially or radial for differentiating the same fuel assembly that is positioned at a non-symmetric position. Also, memory usage might be higher with more materials to deplete, so you can do some sensitivity analysis to get the proper number of different materials. Openmc also has various integrators that might interest you. I have tried some of it on a simple fuel assembly model, and the calculated nuclide density can be varied to some degree. > **[openmc.deplete – Depletion — OpenMC Documentation](https://docs.openmc.org/en/latest/pythonapi/deplete.html)** Regarding the normalization of openmc depletion calculation, this [topic](https://openmc.discourse.group/t/about-constant-power-normalization-during-eigenvalue-calculation-and-depletion-and-photo-fission-contribution-to-the-eigenvalue/4535/3) have been summarized the effects of slightly lower thermal power for each fission in openmc compared to other codes. It can be modified by changing the fission Q value at the chain file. I hope it could help you --- _[View the full topic](https://openmc.discourse.group/t/different-burnup-results/6005)._