Nuclear data reading

User Support
1

Hi there! I’m trying to read the nuclear data of ENDFB VIII for some nuclides. Particularly, I am interested in knowing energy dependent

  1. Fission Cross-section
  2. Absorption Cross-section
    3.Transport cross-section
  3. Chi value
  4. Scattering Cross-section
    I tried using openmc.data.DataLibrary() API, and used the get_by_material() class. For example I used openmc.data.DataLibrary.get_by_material('U235', data_type='neutron') but I get an error telling me that
    TypeError: get_by_material() missing 1 required positional argument: 'name'
    My cross-section is globally declared in the .bashrc file, and also note that if I type openmc.data.DataLibrary.from_xm(), I get a valid output which is <openmc.data.library.DataLibrary at 0x7f291a5fed50>.
    Can anyone help me in this regard?
    Thank you!
1 Like
2

@fsabab you can look into MG-Cross_section example. Hope you’ll get your answer there.

1 Like
3

Thank you @Pranto. Got my answer.

4

I also encountered the same problem, but this link is invalid. Could you please give me a new link? Thank you.