Improving Efficiency in MCNP Neutron Simulations: Tips and Tricks for Beginners

In summary, the conversation is about using MCNP, specifically mcnpx 2.7, for neutron simulation. The speaker is struggling with obtaining low efficiency results in their simulation of a semi-opened detector with 16 1-inch He3 detectors made of Pu240. They mention having seen others' technical notes with higher efficiency results using a similar setup. The conversation also touches on the materials and models used in the simulation, as well as the speaker's inexperience with programming languages like C or C++. They receive a recommendation to use Excel to manipulate input and output files, and a helpful PDF resource for beginners.
  • #1
Estelassar
2
2
Hi, i am looking for some help on MCNP, more precisely mcnpx 2.7 for neutron simulation.

I created a model of semi opened detector with a various number of 1 inch He3 (here 16)
and i only obtain 6-7% efficiency.
the fact is, I've others technicals notes about same "types" of detector, for example 12 He3 1" disposed in square (wich is not the best positionning) where it's mentionned 11% of efficiency for approximatively the same sort of matrix.

In my case the matrix is composed of Pu240, below you will find my physicals parameters for the simulation.

Do you have some feedback about that ? can someone tell me if i am wrong somewhere to obtain an so low efficiency ?

I am in the blur...
Thx by advance
Qgd3Tsp.png


Code: 
mode n
c
Sdef cel=d1 X=d2 Y=d3 z=d4 erg=d5   
SI1 L 11
SP1 D  1
c
Si2 -14 14
SP2   0  1
c
SI3 -14 14
SP3   0  1
c
SI4   0  34.2
SP4   0   1
c
SP5 -3   0.79493   4.68927       $ 240Pu spectre de watt
c
c ----------------- Materiaux ----------------
c ACIER INOX
C --> Acier inoxydable 18-10 *************************************************
M1      6000.70c -0.001200  24050.70c -0.007513  24052.70c -0.150659
        24053.70c -0.017412  24054.70c -0.004416  26054.70c -0.040580
        26056.70c -0.660588  26057.70c -0.015529  26058.70c -0.002103
        28058.70c -0.067198  28060.70c -0.026776  28061.70c -0.001183
        28062.70c -0.003835  28064.70c -0.001008
c
c AIR ************************************************************************
M3   007014.70c   -0.80
     008016.70c   -0.20
c Helium 3 ****************************************************
M4   002003.70c   -1.0
c
M10   032073.70c -1.
c
c
C ------------------------------- polyethylene  high density (-0.95) -------------------------------
M25       1001.70c -0.143685   1002.70c -0.000033   6000.70c -0.856282
MT25      poly.01t  $ poly.70t
c
c ************************** Cadmium ******************************************
M27   48112.70c  -1.0
c
c --- Matrice source ---
c
c --- Hydrogénée 100% vinyle (densite = 0.4) ---
m50 &
   01001.70c 0.5 & $ H-Hydrogene
    06000.70c 0.33 & $ C-Carbone
    17035.70c 0.17  $ Cl-Chlore
c
c --- Mixte(densite = 0.8) ---
c m50 &
c 1001.70c -0.516    & $ H-Hydrogene
c   8016.70c -0.0645   & $ O-Oxygene
c   17035.70c -0.3655  & $ Cl-Chlore
c   25055.70c -0.00076 & $ Mn-Manganese
c   26056.70c -0.053   & $ Fe-Fer
c   29063.70c -0.0003    $ Cu-Cuivre
c --- Métallique (densite = 0.8 - 7.8) ---
c m50 &
c 06000.70c -0.00170 & $ C-Carbone
c 07014.70c -0.00012 & $ N-Azote
c 15031.70c -0.00035 & $ P-Phosphore
c 16032.70c -0.00035 & $ S-Soufre
c 25055.70c -0.01399 & $ Mn-Manganese
c 26056.70c -0.97801 & $ Fe-Fer
c 29063.70c -0.00548   $ Cu-Cuivre
c
c ******** TALLY neutron  ******
F4:N (340 341 342 343 344 345 346 347 348 349 350 351 352 353 354 355)
c
FM4   -1 4 103          $   -1 X 103  /// x= numéro mat hélium
SD4 1                   $ calcul volume tubes
c
ctme 5              $ TALLY Neutron
PRINT
 
Last edited:
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  • #2
Sorry i can't help you but I am a student in nuclear engineering. I want to use this program. I must learn about C or C++ or Mathlap? Thanks
 
  • #3
To use this program it's not necessary to know an programing language, it can help for sure (to automatize some tasks or grab data from the result files) but you can easily manipulate input and output files with some Excel.

If you begin, i recommend you this pdf http://www.nucleonica.net/wiki/images/6/6b/MCNPprimer.pdf
Good luck !
 
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Likes zhj2024 and Hamal_Arietis

FAQ: Improving Efficiency in MCNP Neutron Simulations: Tips and Tricks for Beginners

What is MCNP Neutron simulation?

MCNP Neutron simulation is a computer code used for simulating and analyzing the transport and interactions of neutrons in various materials and systems. It is widely used in nuclear engineering, medical physics, and other fields to model neutron behavior and radiation effects.

How does MCNP Neutron simulation work?

MCNP Neutron simulation uses Monte Carlo methods to model the random interactions of neutrons with atomic nuclei in a given system. It tracks the position, energy, and type of each neutron as it travels through the system, taking into account various physical processes such as scattering, absorption, and fission.

What are the applications of MCNP Neutron simulation?

MCNP Neutron simulation has a wide range of applications, including nuclear reactor design and safety analysis, radiation shielding design, nuclear waste management, and medical imaging and therapy. It is also used in basic research to study the fundamental behavior of neutrons and their interactions with matter.

What are the benefits of using MCNP Neutron simulation?

MCNP Neutron simulation offers many benefits, including the ability to model complex systems and materials, the flexibility to customize simulations for specific scenarios, and the accuracy to predict neutron behavior and radiation effects. It also allows for cost-effective and safe testing and optimization of designs without the need for physical experiments.

What are the limitations of MCNP Neutron simulation?

Like any simulation code, MCNP Neutron simulation has its limitations. It relies on input parameters and assumptions that may not accurately represent real-world conditions, and it can be computationally expensive and time-consuming. Additionally, it may not be suitable for modeling extreme conditions such as high-energy neutron interactions or very large systems.

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