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

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SUMMARY

The discussion focuses on simulating D-D neutron generator reactions using MCNP6. Users highlight the challenge of incorporating angular distributions of neutrons, noting that while MCNPX utilizes the MCUNED code for this purpose, it is incompatible with MCNP6. An alternative suggested is using DROSG-2000 to derive angular distributions, although users express difficulty in determining energy probabilities. Additionally, one participant mentions using classical energy conservation principles to calculate neutron energies in various directions for MCNP simulations.

PREREQUISITES
  • Familiarity with MCNP6 simulation software
  • Understanding of D-D neutron generation processes
  • Knowledge of angular distribution in particle physics
  • Basic principles of energy conservation in nuclear reactions
NEXT STEPS
  • Research the use of DROSG-2000 for angular distribution in MCNP6
  • Explore MCNP6 syntax for non-isotropic source definitions
  • Study classical energy conservation in neutron reactions
  • Investigate alternative simulation tools compatible with angular distributions
USEFUL FOR

Researchers and engineers working with neutron generators, nuclear physicists, and anyone involved in simulating neutron interactions using MCNP6.

Hector_KIT
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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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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?
 
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
 
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
 
Hello, I am designing an X-ray irradiator with MCNP simulation. But I am still in confusion, whether my X-ray housing will be a box or a cylinder. If the box (macrobody identifier of MCNP) is required, I am trying to match the dimension as that of the cylinder, i,e, the height will be that of the cylinder height, and the other two dimensions will be that of the radius of the cylinder (surface 52 and 53). Can anybody please help me define the surfaces? Below is my attached text file. Forgot...

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