KCODE only normalized to power (with 1/keff) to obtain tally absolute value. I'm not sure if I use SDEF instead of KSRC in KCODE problem, do I normalize tally with source strength and the 1/keff factor? Or only the source strength?
Only the source strength, because for fixed source problem the MCNP tallies are normalized with per source partcle. Regards,
Be very cautious here: KCODE problems are not fixed source (i.e., normalized per source neutron) even if the initial source distribution is given by an SDEF code rather than KSRC. In KCODE, SDEF only defines the initial source distribution. After the first cycle, the fission source points generated in the previous cycle are used.
As for normalization of tallies, the tallies in KCODE are normalized PER FISSION NEUTRON. So, you must normalize to the value related to the reactor power, nu-bar, keff, and energy released per fission.
Depends on what you wish to calculate. Attached please find the following article on normalization of MCNP quantitites:
Article On normalization of fluxes and reaction rates in MCNP critic...
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