Problem with Pu239

User Support
1

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?

2

@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 for configuring data sources. In particular, you will likely need to set the OPENMC_CROSS_SECTIONS environment variable.

1 Like
3

Thank you! Solved after I set the configuration.