Search found 2 matches

by iankolaja
Tue Jul 21, 2020 3:38 am
Forum: Gamma transport
Topic: Gamma Transport w/ Spent Fuel: Nuclide not found
Replies: 6
Views: 1306

Re: Gamma Transport w/ Spent Fuel: Nuclide not found

271581 nuclide corresponds with Co-158m, available at ENDF/B-VII and JEFF3.1.1 cross section data libraries (27358.ID). Off the top of my head: - did you use the same libraries for both runs (neutron transport and decay source)? and include all necessary nuclide data libraries? - which cross sectio...
by iankolaja
Tue Jul 07, 2020 3:19 am
Forum: Gamma transport
Topic: Gamma Transport w/ Spent Fuel: Nuclide not found
Replies: 6
Views: 1306

Gamma Transport w/ Spent Fuel: Nuclide not found

Hi! I am currently attempting to do some gamma transport calculations with a canister of spent fuel elements. I produced my spent fuel composition through burn-up calculations using a full core model. My goal is to combine this material definition with a gamma source definition in external source mo...