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  1. Aug 26, 2015 #1
    Hi, I need assistance in performing statistical checks in MCNP5 i.e print table 160. I am not sure where the PRINT card should be placed and the format of it. I am using F4mesh tallies
  2. jcsd
  3. Aug 26, 2015 #2


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    Gold Member

    Hi her91. Welcome to the forum. Always glad to find another MCNP user.

    The PRINT command is in the user manual. For MCNP5, you should be able to find it on page 3-149. It goes in the data section. The default is to print everything, so if you just put the command


    on a line in the data section of your input, you will get all tables. Alternatively you can put the tables you want printed in a list behind the word PRINT. The integer for each kind of table is listed in the manual a few paragraphs after the description of the command.

    By default, without putting in a PRINT command, you get a statistical evaluation of tallies. Also, mesh tallies have an extra output file, in addition to the usual output file you get from running an MCNP input.
  4. Aug 27, 2015 #3
    Without using the print card, i get the following tables in the output file
    128 disabled

    If i add the print card, in the data section of the MCNP crashes.
  5. Aug 27, 2015 #4


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    Ok, debugging your input file remotely through vague hints is not going to work. If your file crashes, you have a syntax error. Round up the usual suspects.
    - Read that output file. Understand warnings and error messages.
    - Look for extra blank lines. Look for missing blank lines. Exactly one between cells and surfaces. Exactly one between surfaces and data. None inside data.
    - Look for lines too long. Remember the 72 character limit.
    - Look for non-accepted characters. Tabs are a real pain.

    If you post your input file I will have a look.
  6. Aug 28, 2015 #5
    Hello to everybody, i need help from MCNP users.
    How i can to plot color legend for tally mesh problems (i.e neutron flux distribution, energy deposition, and so on).
    Thank you!
  7. Oct 26, 2015 #6
    I am trying to use a 3D mesh tally over a lattice structure to find MeV/cm^2 using MCNPX v2.7. The MCNP visualizer shows both the mesh tally and the lattice structure in its correct location. I am running in mode P E with an energy cutoff for photon at 0.01 MeV and an ecut off of 0.3 MeV . No other tallys are in the input file. I generate file types of mdata and a mctal. Both of these files however have a matrix of all zeros. I have tried both type 1 and 3 meshes as well as varying the size resolution of the mesh tally.

    Input file excerpt:

    rmesh11:e PEDEP
    cora11 -25.55 72i 25.55
    corb11 -29.96 120i 29.4
    corc11 1 60i -41.7

    I was however able to use FMESH to determine MeV/cm^2 using MCNP6.1 using the same geometry and energy cut offs:

    *fmesh4:p GEOM=rec ORIGIN=-25.55 -29.95 -41.7
    IMESH=25.55 IINTS 73
    JMESH=29.4 JINTS 85
    KMESH=1 KINTS=61 out=ij

    Do you have any suggestions or trouble shooting ideas? I would prefer to run MCNPX to determine this information.
  8. Oct 26, 2015 #7
    You can plot within the mcnp plotter or outside with an external program such as matlab. I found that you can output mesh tallies as both mdata (binary) and mctal (ASCII) files using PRDMP J J -1.I have a matlab code that reads the data into a 3D matrix and then I am able to 3D color plot there.
  9. Oct 26, 2015 #8
    Hi everyone:
    Im trying to use the FMESH card of MCNP5 to obtain axial flux. I simulated an hexagonal reactor core having different enrichments of uranium (see uploaded file). When I run the program, I obtain a good value of keff (1.03941) but the meshtal file have all the values equal to zero. I only obtain values when the mesh is in the origin 0, 0, -50. I do the following in different positions:

    Fmesh14:n geom=cyl origin=0 0 -50
    imesh=8.77 iints=10
    jmesh=100 jints=10
    kmesh=1 kints=1
    emesh 1 10 out=ij

    Please, I need help.

    Attached Files:

  10. Oct 26, 2015 #9
    Are you using a TR card to move the mesh?
  11. Oct 26, 2015 #10
    No, I don´t use it. In a TRIGA reactor core simulation I used the same information to move the mesh (I only used "origin" to place the mesh in other location) and it result. In this case I don´t know whats is happening.
  12. Feb 11, 2016 #11
    I am Benali Abdel-Hai one user of MCNP code for simulation in different rescearche in medical physics, I need your help to give me some idea to simulate the DOSE RATE of photon beam and electron beam in external radiotherapie
  13. May 17, 2016 #12
    Hi experts,

    I am using MCNP5 for some calculations, I am pretty new here and I hope you all will help me to solve my problem.

    Well, my question is, how do I create a batch file to repeatedly run problems with a same condition but a little change in Energy bin.

    For example, say I have an Input like this

    Some inputs here....
    SDEF POS=0 0 0 PAR=1 ERG=D1
    SI1 H 0 4.13E-07 1.00E-02 5.00E-02 1.00E-01 2.00E-01 2.50E-01 3.00E-01
    4.00E-01 5.00E-01 6.00E-01 7.00E-01 8.00E-01 1.00E+00 1.20E+00
    1.40E+00 1.50E+00 1.60E+00 1.80E+00 2.00E+00 2.20E+00 2.30E+00
    2.40E+00 2.60E+00 2.80E+00 3.00E+00 3.40E+00 3.70E+00 4.20E+00
    4.60E+00 5.00E+00 5.50E+00 6.00E+00 6.50E+00 7.00E+00 7.50E+00
    8.00E+00 8.50E+00 9.00E+00 9.50E+00 1.00E+01 1.10E+01 1.20E+01
    1.30E+01 1.40E+01 1.60E+01
    SP1 D 0 4.44E-04 4.41E-03 8.60E-03 2.31E-02 1.35E-02 1.44E-02 3.08E-02
    3.26E-02 3.35E-02 3.41E-02 3.41E-02 6.72E-02 6.44E-02 6.09E-02
    2.89E-02 2.79E-02 5.25E-02 4.83E-02 4.40E-02 2.05E-02 1.95E-02
    3.63E-02 3.28E-02 2.95E-02 5.02E-02 3.09E-02 4.12E-02 2.53E-02
    1.99E-02 1.90E-02 1.40E-02 1.02E-02 7.49E-03 5.46E-03 3.98E-03
    2.89E-03 2.09E-03 1.51E-03 1.09E-03 1.35E-03 7.00E-04 3.61E-04
    1.85E-04 1.44E-04 3.80E-05
    ......some inputs here

    And now I want to create a batch file for automatically repeat this problem but with another SI card and other SP card. How should I do???

    I hope you all would kindly help me.
  14. Jun 1, 2016 #13
    Help with MCNP6 installation.
    I am currently getting the following error in my output.
    "bad trouble in subroutine ffetch of xact

    cannot find xs library file specified in xsdir "

    I have installed MCNP6 before and had the same error. I cant remember how i fixed it.
  15. Jun 13, 2016 #14
    Perhaps you forgot to install the MCNP Data? It has a separate installation exe. If not, just reinstall the MCNP Data portion again, there might be a path setting error.
  16. Sep 26, 2016 #15
    Hi Experts,
    I am wondering about the problem that I encountered in my research. Suppose I have a neutron surface source with 44 energy bins (from 4.13E-7 to 16.0 MeV). Now I want to rebin this source to 30 Energy bins (also from 4.13E-7 to 16.0 MeV). I intend to use MCNP5 to do this. But I don't know how to do and which tally should I use?

    I hope experts from HPS can help me in solving this problem.
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