Which Tally type should I use in MCNP6?

I have a problem where I want to model the dose in Gy of a gamma source on a surface as a function of distance. In the papers I have read several different tallys have been used which leaves me a little confused as to the appropriate tally. In the papers the *F4Mesh, F4, F6, and *F8 tallys were used.
 
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Can you tryout each one and decide from that?
 
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Hi,
the first question you need to answer is "What I want calculated ?" :
- dose,
- kerma,
- dose equivalent,
- equivalent dose
 

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