About openmc.data.njoy.make_ace_thermal

User Support
1

Dear colleagues,

I am producing the ace files for thermal scattering S(a,b) at different temperatures. However, according to the openmc.data.njoy.make_ace_thermal function, I must input:

openmc.data.njoy.make_ace_thermal

openmc.data.njoy. make_ace_thermal (filename, filename_thermal, temperatures=None, ace=‘ace’, xsdir=‘xsdir’, error=0.001, **kwargs )[source]

Generate thermal scattering ACE file from ENDF files

Parameters: * filename (str) – Path to ENDF neutron sublibrary file

  • filename_thermal (str) – Path to ENDF thermal scattering sublibrary file

However, I ask myself, if I am producing, for instance U in UO2, what is the neutron sublibrary file I should use for producing the thermal scattering file. The nuclide U-235 or the U-238? How to produce thermal scattering for nuclides with many isotopes?

Juan

2

Juan,
openmc.data.njoy.make_ace_thermal() uses as backend NJOY, and in particular one of the steps of creating a thermal ACE file is to run THERMR. To run THERMR you need both a thermal ENDF-6 file and a neutron ENDF-6 file. But, at the end the neutron ENDF-6 is not used for anything, because the rest of the neutron data is dropped when you run ACER. TLDR: you can use U-235 or U-238 and you should get the same result.

The link between the thermal and neutron sublibraries is done in ACER, and it is controlled with the zaids= parameter of openmc.data.njoy.make_ace_thermal(). You should put all the nuclides present in the element.

I checked, and the material uuo2 is already associated with all the uranium nuclides. So you actually do not need to input the zaids yourself. You can check it in the header of the generated ACE file.

Best,
Ignacio

3

Thanks Ignacio,
I did not know about the existence of this dictionary in the njoy.py. Your help is super useful!
Have a nice day,
Juan