Which Tally type should I use in MCNP6?

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To model the dose in Gy from a gamma source on a surface as a function of distance in MCNP6, it's essential to determine the specific calculation needed, such as dose, kerma, dose equivalent, or equivalent dose. Various tally types, including *F4Mesh, F4, F6, and *F8, have been used in relevant literature, leading to confusion about the best choice. Users are encouraged to experiment with each tally type to see which one best meets their modeling needs. Understanding the desired output will guide the selection of the appropriate tally. Selecting the correct tally is crucial for accurate dose modeling in MCNP6.
khary23
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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?
 
Hi,
the first question you need to answer is "What I want calculated ?" :
- dose,
- kerma,
- dose equivalent,
- equivalent dose
 

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