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MCNP Neutron simulation

  1. Sep 9, 2016 #1
    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 (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
    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: Sep 9, 2016
  2. jcsd
  3. Sep 10, 2016 #2
    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
     
  4. Sep 10, 2016 #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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