# Openmc v0.13 pincell example run error **URL:** \ **Category:** User Support\ **Created:** [January 28, 2023, 12:33pm UTC](https://openmc.discourse.group/t/openmc-v0-13-pincell-example-run-error/2661 "2023-01-28T12:33:13Z")\ **Posts on this page:** 2\ **Page:** 1
**Author:** ![emilmammadzada](https://avatars.discourse-cdn.com/v4/letter/e/c4cdca/32.png) [@emilmammadzada](https://openmc.discourse.group/u/emilmammadzada)\ **Post date:** [January 28, 2023, 12:33pm UTC](https://openmc.discourse.group/t/openmc-v0-13-pincell-example-run-error/2661/1 "2023-01-28T12:33:13Z")
## Reading settings XML file… Reading cross sections XML file… Reading materials XML file… ERROR: Could not find nuclide O18 in the nuclear data library. When I run Openmc v0.13 pincell example, I see this error output, how can I solve this problem? ---
**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:** [January 28, 2023, 4:00pm UTC](https://openmc.discourse.group/t/openmc-v0-13-pincell-example-run-error/2661/2 "2023-01-28T16:00:05Z")
This means that one of your materials contains O18 but your cross section library does not have data for O18. The most likely scenario is that you are using ENDF/B-VII.1, which lacked cross sections for O18. The easiest fix would be to download and use the ENDF/B-VIII.0 library [here](https://openmc.org/official-data-libraries/). Alternatively, you can make sure that at the time you are generating your materials.xml, you have set the `OPENMC_CROSS_SECTIONS` environment variable to the library you are using, which will ensure that whenever you call `material.add_element`, it will only use isotopes that are available in that library. Full information about cross section configuration can be found [here](https://docs.openmc.org/en/latest/usersguide/data.html).