2-Neutron Correlation Pre-setup Calculations

Big Detector Solid Angle Calculations

Neutron Corr Home Page

MCNPX Simulation

14 MeV neutron source, emitted isotropically ([math]4\pi[/math])
Detector placed 1m away from source

face of the detector is 15.24cm x 76.2cm, and 3.6cm deep

The solid angle can be found from the number of particles hitting the detector as:

Results

Out of 1E9 neutrons generated, 8618287 neutrons hit the detector
- [math]\Delta \Omega = 0.108 Sr[/math]
  - if the detector is placed 70cm away from the source, [math]\Delta \Omega = 0.207 Sr[/math]
  - if the detector is placed 65cm away from the source, [math]\Delta \Omega = 0.236 Sr[/math]

As a test to verify our results
- We change the detector size to 2cm by 2cm and used 1E9 neutrons again
- 32061 neutrons struck the detector
- [math]\Delta \Omega = 0.0004 Sr[/math]

And, as a second test to verify our results
- We change the detector size to 1cm by 1cm and used 1E9 neutrons again
- 7965 neutrons struck the detector
- [math]\Delta \Omega = 0.0001 Sr[/math]

Now, what neutron singles rate into the detector should correspond to 1 fission per pulse?

If we have 1 fission per pulse and each fission emits on average 2.3 neutrons, we should expect 2.3 neutrons/pulse
The number of neutrons hitting the detector per pulse is found as [math]2.3*\frac{\Delta \Omega}{4\pi}[/math]
- @ 1 meter => 0.0198 neutrons hitting the detector per pulse
- @ 70 cm => 0.0379 neutrons hitting the detector per pulse
Taking into account the efficiency of the detector [math]\epsilon_0[/math], the number detected per pulse can be found as [math]2.3*\frac{\Delta \Omega}{4\pi}*\epsilon_0[/math]
- @ 1 meter from source => () neutrons detected per pulse
- @ 70 cm from source => () neutrons detected per pulse

Neutron and Photon Flux Through Lead

Using MCNPX, a simulation was done to determine how much of the neutrons and photons from the target will be blocked by various thickness of lead. With a monochromatic pencil beam of incident particles, the following results illustrate how much of the initial beam actually made it through the lead.

Neutron and Photon Flux Through Concrete

Detector Cross-Talk

Using MCNPX, I am trying to simulated the cross-talk we could expect between the big detectors. I used a fission neutron energy spectrum given in MCNPX, shown below.

These neutrons are incident upon a plastic scintillator and I will look at the neutrons and photons coming out of the scintillator's sides perpendicular to the incident beam. The first case I will look at is where the detectors are placed with only polyethylene between them, as illustrated below.