Hello everyone
I just work with OpenMC multi group mode
I computed all the cross sections in each group of energy for the use in the file mg_cross_sections.xml but I can’t calculate the neutron fission spectrum from a term OpenMC
Could anyone help me
Regards
Mohamed,
Consider this post to a previous thread about computing the fission matrix:
https://groups.google.com/d/msg/openmc-users/F6kCBG1BX6s/0qaL7a9GFwAJ
This details how to compute the fission spectrum using the fully documented “openmc.mgxs” module.
Best,
Will