# MCNP Neutron simulation

1. Sep 9, 2016

### Estelassar

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...

Code (Text):

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
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: Sep 9, 2016
2. Sep 10, 2016

### Hamal_Arietis

Sorry i cant help you but Im a student in nuclear engineering. I want to use this program. I must learn about C or C++ or Mathlap? Thanks

3. Sep 10, 2016

### Estelassar

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 !