How to simulate a D-D neutron generator with MCNP6?

In summary, Hector is working on a way to simulate a neutron generator with MCNP6, but he needs to know the energy of neutrons in different directions. Kirk is going to calculate the energy of neutrons using classical energy conservation and then put a non-isotropic source in MCNP6.
  • #1
Hector_KIT
2
0
Hello everybody,

I am performing some experiments with a neutron generator. Specifically D-D reactions. I am trying to replicate the measurements with MCNP6, but I do not know how can I simulate the neutron generator with MCNP6 since the neutrons have a angular distribution. For MCNPX there was the MCUNED code which can represent the angular dependence, but it does not work with MCNP6. Another option may be to use DROSG-2000 to obtain the angular distribution and use it as source for MCNP6, but I am not able to obtain the probability of each energy.
Does anyone know any way to do it with MCNP6, or a complementary program similar to the above described?

Best regards
HSS
 
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  • #2
I'm not entirely sure about the syntax, but MCNP5 does have the option to explicitly model a source with an angular distribution of particles emitted. Is that what you're looking for?
 
  • #3
Hector_KIT said:
Hello everybody,

I am performing some experiments with a neutron generator. Specifically D-D reactions. I am trying to replicate the measurements with MCNP6, but I do not know how can I simulate the neutron generator with MCNP6 since the neutrons have a angular distribution. For MCNPX there was the MCUNED code which can represent the angular dependence, but it does not work with MCNP6. Another option may be to use DROSG-2000 to obtain the angular distribution and use it as source for MCNP6, but I am not able to obtain the probability of each energy.
Does anyone know any way to do it with MCNP6, or a complementary program similar to the above described?

Best regards
HSS
Hello Hector,

I'm curious if you have found a solution to this problem since it has been posted. I am personally working on a D-T generator project at Oregon State and am about to run into the problem of distribution.

Kirk
 
  • #4
Hi,
For my part I will calculate the energy of neutrons in different direction with classical energy conservation applied to the reaction (see for example chpter 4 of this book https://www.amazon.com/dp/3319486586/?tag=pfamazon01-20)
After that put a non istropic source in mcnp with dir :
SDEF DIR=d1 fdir=d2 vec= 0 0 1
SI1 H cos (teta1) cos (teta2) cos (teta3) ...
SP1 0 S1/S S2/S ...
DS2 s 10 20 30 ...
SI10 L energy_for_teta1
SI20 energy_for_teta2
SI30 energy_for_teta3
SP10 d 1
SP20 d 1
SP30 d 1

PSR
 

FAQ: How to simulate a D-D neutron generator with MCNP6?

What is MCNP6 and how does it simulate a D-D neutron generator?

MCNP6 is a Monte Carlo N-Particle transport code used for simulating the transport of particles, such as neutrons, through various materials and geometries. It uses a combination of random sampling and statistical analysis to simulate the behavior of particles in a system. To simulate a D-D neutron generator with MCNP6, the code uses the known physical properties of the materials and geometries involved to simulate the interactions between particles and their resulting behavior, such as scattering and absorption.

What is a D-D neutron generator and why is it important to simulate it?

A D-D neutron generator is a device that produces a high flux of neutrons through the fusion of deuterium gas. It is important to simulate a D-D neutron generator to understand its behavior and optimize its design for various applications, such as medical imaging or neutron activation analysis. Simulation allows for the study of the neutron flux and energy spectrum, as well as the effects of different materials and geometries on the neutron production and behavior.

How accurate is the simulation of a D-D neutron generator with MCNP6?

The accuracy of the simulation depends on the quality of the input data and assumptions made in the simulation. Generally, MCNP6 has been shown to produce results within a few percent of experimental data for most applications. However, more complex simulations may require more precise input data and adjustments to the simulation parameters to achieve accurate results.

Can MCNP6 simulate other types of neutron generators?

Yes, MCNP6 can simulate various types of neutron generators, including D-T, D-3He, and D-T fusion generators. It can also simulate other types of radiation sources, such as X-ray tubes and medical linear accelerators. However, the simulation parameters and input data may need to be adjusted for each specific type of generator.

Are there any limitations to simulating a D-D neutron generator with MCNP6?

While MCNP6 is a powerful simulation tool, it does have some limitations when simulating a D-D neutron generator. These include the complexity of the system and the accuracy of the input data, such as the cross-section data of materials used. Additionally, the simulation may not accurately capture certain physical phenomena, such as rare events or non-equilibrium conditions, without appropriate modifications to the simulation parameters.

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