MCNPX normalizes all tallies to be per one starting particle. If you want to have them with respect to your reactor power, calculate the number of neutrons that is produced with a certain power, i.e. reactor power / energy deposited per fission * number of neutrons produced per fission, and multiply your result with that number. The latter two numbers can be calculated with MCNPX, too, or you can use a good estimate, depending on how exact you results need to be.

The best unit to present your result depends on what you want to tell your readers...

Thank you very much, so actually we need to calculate the total neutron number..since mcnp normalizes tally to one starting particle.

I hope that the following link can hep you

https://rsicc.ornl.gov/notebook/mcnp5/1230-mcnp_power_density_calc_port2006.pdf