MT 107 is allowed for all types of incident particles, your tally should therefore also include (p,a) reactions. If you do not have incident neutrons then your tally should reflect the desired value. You should be able to verify by just adding a MT 1 tally for all neutron interactions.

See here for further details:

https://www.oecd-nea.org/dbdata/data/manual-endf/endf102_MT.pdf

Thank you very much.

Your link has been really useful, because In the mcnp manual It doesn`t specify that information.

MCNP6 Manual Section 3.2.5.7, Note 4:

"Any number of ENDF/B (MT) or special (R) reactions can be used in a multiplier set as long as they are present in the MCNP6 cross-section libraries."

This means that to calculate with the FM card the rate of the (p,a) reaction a) you need to have the library of cross-sections for protons for the material that you use at the FM card, and b) this library must include the (p,a) cross-section (identified by any number, as the Manual says, which will come next to your FM card).

However, to the best of my knowledge none of proton libraries that come with MCNP include (p,a) reactions.

The list of available libraries (now missing former Appendix G of the MCNP6 Manual that also lists the ENDF/B reaction types) can be found here:

https://laws.lanl.gov/vhosts/mcnp.lanl.gov/pdf_files/la-ur-13-21822v4.pdf