# Problem with Pu239

**URL:** <https://openmc.discourse.group/t/problem-with-pu239/2462>\
**Category:** User Support\
**Created:** [November 9, 2022, 9:17pm UTC](https://openmc.discourse.group/t/problem-with-pu239/2462 "2022-11-09T21:17:39Z")\
**Posts on this page:** 3\
**Page:** 1

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**Author:** ![yang\_gs343](https://avatars.discourse-cdn.com/v4/letter/y/dbc845/32.png) [@yang\_gs343](https://openmc.discourse.group/u/yang_gs343)\
**Post date:** [November 9, 2022, 9:17pm UTC](https://openmc.discourse.group/t/problem-with-pu239/2462/1 "2022-11-09T21:17:39Z")

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Hi good people!

I tried to define some PuO2 material as follow:

material\_100 = mc.Material(name=“material\_100”)  
material\_100.add\_nuclide(‘H1’, 6.0070e-2)  
material\_100.add\_nuclide(‘O16’, 3.654e-2)  
material\_100.add\_nuclide(‘N14’,2.3699e-3)  
material\_100.add\_nuclide(‘Pu239’,2.7682e-4)  
material\_100.add\_nuclide(‘Pu240’,1.2214e-5)  
material\_100.add\_nuclide(‘Pu241’,8.3390e-7)  
material\_100.add\_nuclide(‘Pu242’,4.5800e-8)  
material\_100.set\_density(‘g/cm3’, 9.27)  
material\_200 = mc.Material(name=“material\_200”)  
material\_200.add\_nuclide(‘Cr50’,7.1866e-4)  
material\_200.add\_nuclide(‘Cr52’,1.3859e-2)  
material\_200.add\_nuclide(‘Cr53’,1.5715e-3)  
material\_200.add\_nuclide(‘Cr54’,3.9117e-4)  
material\_200.add\_nuclide(‘Fe54’,3.7005e-3)  
material\_200.add\_nuclide(‘Fe56’,5.8090e-2)  
material\_200.add\_nuclide(‘Fe57’,1.3415e-3)  
material\_200.add\_nuclide(‘Fe58’,1.7853e-4)  
material\_200.add\_nuclide(‘Ni78’,4.4318e-3)  
material\_200.add\_nuclide(‘Ni60’,1.7071e-3)  
material\_200.add\_nuclide(‘Ni61’,7.4207e-5)  
material\_200.add\_nuclide(‘Ni62’,2.3661e-4)  
material\_200.add\_nuclide(‘Ni64’,6.0256e-5)  
material\_200.set\_density(‘g/cm3’, 6.6)

The material card part runs well, but when I tried to run the plot part and running part, the error returns as: Could not find nuclide Pu239 in the nuclear data library.

I’m wondering where’s the problem?

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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:** [November 10, 2022, 4:27am UTC](https://openmc.discourse.group/t/problem-with-pu239/2462/2 "2022-11-10T04:27:40Z")

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@yang_gs343 welcome to the community! It sounds like you may not have set up a data library for use with OpenMC yet. Please see the instructions in our [user’s guide](https://docs.openmc.org/en/latest/usersguide/data.html) for configuring data sources. In particular, you will likely need to set the `OPENMC_CROSS_SECTIONS` environment variable.

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**Author:** ![yang\_gs343](https://avatars.discourse-cdn.com/v4/letter/y/dbc845/32.png) [@yang\_gs343](https://openmc.discourse.group/u/yang_gs343)\
**Post date:** [November 12, 2022, 12:31am UTC](https://openmc.discourse.group/t/problem-with-pu239/2462/3 "2022-11-12T00:31:08Z")

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Thank you! Solved after I set the configuration.