Number of secondaries generated in a volume MCNP?

[Neo Tran]

Hello everyone,
I am having a problem with MCNP. My question is how to get number of secondaries in a certain volume. For example I have a neutron beam bombarded Pb target, and I want to count all of proton formed in the target. I considered tally F4, but the unit is 1/cm**2. Who can explain the tally to me? More over, help me solve this problem, please.
Thank you very much,Neo.

 

[quarkle]

For this, you will need to use the tally multiplier functionality with the F4 tally. F4 gives a flux averaged over the volume of the cell in units of particles/cm**2 per source particle. Using a tally multiplier card, you can get MCNP to multiply this flux with the atom density of the target nuclide and the microscopic cross section of the desired reaction (in this case, (n,p) reaction). If you multiply this value with the source term, you'll get the total number of protons formed in the target.
This is what your MCNP cards with look like (assuming a target cell 1):

F4:N 1
FM4 C M (103) (203)

Where C is the atom density of lead in your target, M is the corresponding material number in your input file for lead (assuming it is pure elemental lead)
103 is the reaction number for (n,p) reaction and 203 is a reaction number for total number of protons produced by all reactions. You can find a complete list of all reaction types and numbers in Appendix G of the MCNP5 manual Vol I
You can also find an example of the similar problem on Page 4-39 of the MCNP5 manual Vol II

 

[Neo Tran]

For this, you will need to use the tally multiplier functionality with the F4 tally. F4 gives a flux averaged over the volume of the cell in units of particles/cm**2 per source particle. Using a tally multiplier card, you can get MCNP to multiply this flux with the atom density of the target nuclide and the microscopic cross section of the desired reaction (in this case, (n,p) reaction). If you multiply this value with the source term, you'll get the total number of protons formed in the target.
This is what your MCNP cards with look like (assuming a target cell 1):

F4:N 1
FM4 C M (103) (203)

Where C is the atom density of lead in your target, M is the corresponding material number in your input file for lead (assuming it is pure elemental lead)
103 is the reaction number for (n,p) reaction and 203 is a reaction number for total number of protons produced by all reactions. You can find a complete list of all reaction types and numbers in Appendix G of the MCNP5 manual Vol I
You can also find an example of the similar problem on Page 4-39 of the MCNP5 manual Vol II

You was right. Thank you very much Quarkle. It is very useful for me in this time.