# Depletion micro calculation mistake

**URL:** <https://openmc.discourse.group/t/depletion-micro-calculation-mistake/3145>\
**Category:** User Support\
**Created:** [June 25, 2023, 12:44pm UTC](https://openmc.discourse.group/t/depletion-micro-calculation-mistake/3145 "2023-06-25T12:44:27Z")\
**Posts on this page:** 3\
**Page:** 1

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**Author:** ![Lucky\_Kevin](https://avatars.discourse-cdn.com/v4/letter/l/ed655f/32.png) [@Lucky\_Kevin](https://openmc.discourse.group/u/Lucky_Kevin)\
**Post date:** [June 25, 2023, 12:44pm UTC](https://openmc.discourse.group/t/depletion-micro-calculation-mistake/3145/1 "2023-06-25T12:44:27Z")

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When I intend to tally macrosection of Xe135 in different depletion steps , the progress repoort " ERROR: Could not find the nuclide Xe135 specified in tally 1 in any material".  
The filter set in fuel cell which concludes U235 and U238. In addtion, tally file is added before integrator

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**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:** [June 26, 2023, 1:14pm UTC](https://openmc.discourse.group/t/depletion-micro-calculation-mistake/3145/2 "2023-06-26T13:14:49Z")

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@Lucky_Kevin welcome to the community! This is a limitation in the current version of OpenMC but has actually already been fixed in our developmental version and will be part of the next release.

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**Author:** ![Lucky\_Kevin](https://avatars.discourse-cdn.com/v4/letter/l/ed655f/32.png) [@Lucky\_Kevin](https://openmc.discourse.group/u/Lucky_Kevin)\
**Post date:** [June 29, 2023, 2:20am UTC](https://openmc.discourse.group/t/depletion-micro-calculation-mistake/3145/3 "2023-06-29T02:20:39Z")

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Okay ,thank you for replying me!