- #1
Kashif
- 4
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Greetings to all
need forum members help regarding MCNP5 college assignment.
The task is to calculate dose from a system comprising of a neutron source surrounded by natural uranium. the system is subcritical.
The problem is how to calculate dose due to fissions in the uranium because dose from neutron source can be calculated separately but i don't know how to incorporate dose due to fissions.
which tally and how i should use?
need forum members help regarding MCNP5 college assignment.
The task is to calculate dose from a system comprising of a neutron source surrounded by natural uranium. the system is subcritical.
The problem is how to calculate dose due to fissions in the uranium because dose from neutron source can be calculated separately but i don't know how to incorporate dose due to fissions.
which tally and how i should use?