Normalization of MCNP F4:P photon flux tally to experimental counts

[Marinaromany]

Hello everyone
I am using MCNP with a DT neutron source and I am interested in prompt gamma measurements. My detector is modeled as a simplified volume, and I am currently using an F4:P tally with energy bins to obtain the photon spectrum.

Since the F4 tally gives photon flux per source particle (ph/cm²/source), I would like to ask:

What is the correct way to convert or normalize an F4 photon flux tally to counts, so that it can be meaningfully compared with experimental spectra or reported as detector count

 

[Alex A]

Welcome to PF @Marinaromany,
It might be possible to do that with an F4 and a cross section multiplier but F8:p,e is a better way on the cell of the crystal, and the tally result is in total counts.

 

[DEvens]

Normalization of detectors is a thing.

What you need is some way of knowing the number of particles per second in your real system. You mention a DT neutron source and you are using 14.1 MeV neutrons. So I am expecting you have an accelerator firing D's into a T target. You would need some way to determine the number of neutrons per second this system is producing.

If you have a spec for your system, that is great! Use that number in the following.

If not, you have a problem. You need that number. How you proceed depends on what information you do have. For example, you might have some kind of detector at the facility. It might be calibrated to tell you neutrons per second at some location. That might tell you the neutrons per second at the source. Maybe *that* requires an MCNP calc.

Then you get your MCNP calc of flux-per-particle-started. And you multiply that by particles-started-per-second. And it gets you flux-per-second.

 

[Marinaromany]

Normalization of detectors is a thing.

What you need is some way of knowing the number of particles per second in your real system. You mention a DT neutron source and you are using 14.1 MeV neutrons. So I am expecting you have an accelerator firing D's into a T target. You would need some way to determine the number of neutrons per second this system is producing.

If you have a spec for your system, that is great! Use that number in the following.

If not, you have a problem. You need that number. How you proceed depends on what information you do have. For example, you might have some kind of detector at the facility. It might be calibrated to tell you neutrons per second at some location. That might tell you the neutrons per second at the source. Maybe *that* requires an MCNP calc.

Then you get your MCNP calc of flux-per-particle-started. And you multiply that by particles-started-per-second. And it gets you flux-per-second.

HeHello Thanks for your help. I have no of particle per second . Shall i multiply it with count getting from F4?