# Missing nuclide when running pin cell depletion example **URL:** \ **Category:** User Support\ **Created:** [November 22, 2021, 11:57pm UTC](https://openmc.discourse.group/t/missing-nuclide-when-running-pin-cell-depletion-example/1558 "2021-11-22T23:57:16Z")\ **Posts on this page:** 5\ **Page:** 1
**Author:** ![Stepniac](https://avatars.discourse-cdn.com/v4/letter/s/df705f/32.png) [@Stepniac](https://openmc.discourse.group/u/Stepniac)\ **Post date:** [November 22, 2021, 11:57pm UTC](https://openmc.discourse.group/t/missing-nuclide-when-running-pin-cell-depletion-example/1558/1 "2021-11-22T23:57:16Z")
Hello. I’ve been running the pin cell depletion example provided with the OpenMC documentation. I’ve managed to run the depletion using the simple depletion chain. However I tried using the casl depletion chain with the pin cell example. OpenMC complained there is no Rh104 in the ENDF/B VII.1. I’ve confirmed that there is no Rh104 in the library on my computer. The library was downloaded from the OpenMC website. Has anyone encountered this issue? ---
**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:** [December 13, 2021, 1:16am UTC](https://openmc.discourse.group/t/missing-nuclide-when-running-pin-cell-depletion-example/1558/2 "2021-12-13T01:16:37Z")
Hi @Stepniac. I’m not sure why you would see such an error. Rh104 is indeed present in the depletion chain but it doesn’t have any associated neutron cross sections (only decay modes). For such cases, the depletion solver keeps track of their evolution, but it’s not supposed to try to insert them in the model for neutron transport purposes. I just tried running the pincell depletion example with the CASL depletion chain and didn’t run into any problems. ---
**Author:** ![Stepniac](https://avatars.discourse-cdn.com/v4/letter/s/df705f/32.png) [@Stepniac](https://openmc.discourse.group/u/Stepniac)\ **Post date:** [December 13, 2021, 6:27am UTC](https://openmc.discourse.group/t/missing-nuclide-when-running-pin-cell-depletion-example/1558/3 "2021-12-13T06:27:59Z")
Hi @paulromano I found out what was the problem. It appears that there was an inconsistency between the cross ---
**Author:** ![Stepniac](https://avatars.discourse-cdn.com/v4/letter/s/df705f/32.png) [@Stepniac](https://openmc.discourse.group/u/Stepniac)\ **Post date:** [December 13, 2021, 6:52am UTC](https://openmc.discourse.group/t/missing-nuclide-when-running-pin-cell-depletion-example/1558/4 "2021-12-13T06:52:36Z")
Hi @paulromano I finally found out what caused my problem. I usually use the JEFF3.3 library so I set up then environment variable `OPENMC_CROSS_SECTIONS` to reference the cross\_sections.xml for this library. However in the Python code I used the Materials attribute to reference the ENDF VII.1 cross\_sections.xml thinking that it will override the value set in the environment variable. Now in the JEFF cross\_sections.xml there is an entry ffor Rh104 that is not present in the cross\_sections.xml. It appears OpenMC uses the file referenced in `OPENMC_CROSS_SECTIONS` to determine the list of nuclides it intends to use for the depletion. The code then uses the ENDF cross\_sections.xml referenced in the Materials attribute. Since there is no Rh104.h5 in the ENDF library OpenMC raised an error. In summary the cross\_sections.xml needs to reference the library used in the depletion even though the user has used this library in the Python code. ---
**Author:** ![Stepniac](https://avatars.discourse-cdn.com/v4/letter/s/df705f/32.png) [@Stepniac](https://openmc.discourse.group/u/Stepniac)\ **Post date:** [December 13, 2021, 8:10pm UTC](https://openmc.discourse.group/t/missing-nuclide-when-running-pin-cell-depletion-example/1558/5 "2021-12-13T20:10:53Z")
What I mean is the environment variable need to reference the library to be used in the depletion.