Dear Prof.Paul Romano:
I finished to modify the ACE file as your said in last email “Question about how to use JEFF3.2 in openmc”.
It really works, and many thanks for your kindly help.
Nowadays, I want to use OPENMC to solve a external fixed source problem using the output neutron tally files from FLUKA.
I split a cylinder volume into several sub volumes and get its neutron energy spectrum and flux respectively.
How can I put these information into OPENMC?
Or what additional information I should have and put it into OPENMC?
Could you give me some advice or help me with a simple case?
Thank you for your help in advance.
Best Wishes
Yours
NiuOBA