What is Mcnpx: Definition and 35 Discussions

Monte Carlo N-Particle Transport (MCNP) is a general-purpose, continuous-energy, generalized-geometry, time-dependent, Monte Carlo radiation transport code designed to track many particle types over broad ranges of energies and is developed by Los Alamos National Laboratory. Specific areas of application include, but are not limited to, radiation protection and dosimetry, radiation shielding, radiography, medical physics, nuclear criticality safety, detector design and analysis, nuclear oil well logging, accelerator target design, fission and fusion reactor design, decontamination and decommissioning. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and fourth-degree elliptical tori.
Point-wise cross section data are typically used, although group-wise data also are available. For neutrons, all reactions given in a particular cross-section evaluation (such as ENDF/B-VI) are accounted for. Thermal neutrons are described by both the free gas and S(α,β) models. For photons, the code accounts for incoherent and coherent scattering, the possibility of fluorescent emission after photoelectric absorption, absorption in pair production with local emission of annihilation radiation, and bremsstrahlung. A continuous-slowing-down model is used for electron transport that includes positrons, k x-rays, and bremsstrahlung but does not include external or self-induced fields.
Important standard features that make MCNP very versatile and easy to use include a powerful general source, criticality source, and surface source; both geometry and output tally plotters; a rich collection of variance reduction techniques; a flexible tally structure; and an extensive collection of cross-section data.
MCNP contains numerous flexible tallies: surface current & flux, volume flux (track length), point or ring detectors, particle heating, fission heating, pulse height tally for energy or charge deposition, mesh tallies, and radiography tallies.
The key value MCNP provides is a predictive capability that can replace expensive or impossible-to-perform experiments. It is often used to design large-scale measurements providing a significant time and cost savings to the community. LANL's latest version of the MCNP code, version 6.2, represents one piece of a set of synergistic capabilities each developed at LANL; it includes evaluated nuclear data (ENDF) and the data processing code, NJOY. The international user community’s high confidence in MCNP’s predictive capabilities are based on its performance with verification and validation test suites, comparisons to its predecessor codes, automated testing, underlying high quality nuclear and atomic databases and significant testing by its users.

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  1. I

    MCNPX Transformation

    Hello, I doing my thesis, I'm beginner in MCNPX. I want to ask about transformation in MCNPX, I want to transform this head phantom in inside and outside field ( radiation using LINAC from face to ear) in angle 90° and 270 °. Can anyone help me to solve this problem with transformation code?
  2. Elizabeth Vega

    MCNPX Fatal error: "electron importances are zero"

    When I run the application, I get an error message. Fatal error electron importances are zero, I can't find what's wrong in the code.
  3. jj8bmk

    What is a unit of time listed in MCNP6 PTRAC output file?

    Hi everyone, I've been trying to analyze PTRAC output file from MCNP6 here we can see the location , cell, particle, time, and so on... My question is, I have trouble finding the unit of time listed in PTRAC, (ex, 0.30113E-02), which is hard to find in MCNP manual My intuition is that the...
  4. G

    Calculation of activation with mesh tally in MCNPX 2.6

    Hi, All! Can anyone help me to construct a card for calculation of a spatial activation? Reaction is 27Al(n,p). I used a tmesh card for calculation of neutron flux as: tmesh rmesh21:n flux cora21 190 9i 210 corb21 -10 9i 10 corc21 -10 9i 10 ergsh21 0 20 endmd Somewhere in a problem I...
  5. emilmammadzada

    I get this error when I try to run the code MCNPX

    I get this error"bad trouble in imcn in routine pass1 unexpected eof in file depletion.inp" when I try to run the code MCNPX my input file c Depletion pincell input file for MCNPX c Define cells c Cell 1: Fuel 1 0 -1.0 -4 -5 -6 c Cell 2: Cladding 2 0 -2.0 4 -7 c Cell 3: Moderator 3 0 -3.0...
  6. W

    Skyshine vs Direct Dose in MCNP5

    Hello - what is an accepted definition of the skyshinne dose in MCNP and how would you calculate this? If you have a source and a shield a few meters away between the dose point, the contribution that goes around the shield would be skyshine....but..what if you have a big source region (e.g., a...
  7. emilmammadzada

    Solve MCNPX Vol Card Problem: Adding/Calculating Vol for Each Cell

    Cell vol card fatal error appears in this input file. How do I add a vol card to this input file. How can I calculate the vol for each cell. Is there a method?
  8. emilmammadzada

    Fixing MCNPX Fatal Error: Too Many Numbers First Entry

    When I run the application, I get an error message. This message: Fatal error too many numbers first entry. What could be the reason?
  9. T

    Problem with Moritz (3D viewer of MCNPX)

    Hello everyone! We get Problem to run Moritz on a given PC when its works on others PCs What could be the solution in fact , we can see the exe running in the taskmanager but not GUI is opened.. Thanks in advance for help Thibaut
  10. A

    [MCNPX] How to run an entire folder?

    I know that "mcnpx n=filename" I run the file, but how do I run the entire folder with the entries inside?
  11. nuclearsneke

    MCNP5 tallies conversion and MCNPX

    Hi there I want to convert the flux (F4:N tally) from mcnp units to cm-2s-1 units. How to do that? Also I have some bug in MCNPX: while running the file, I get an error like " >bad trouble in imcn in routine xin >Cannot find bertin " How to solve it? Database for MCNP5-MCNPX got installed already.
  12. A

    The length of the line in the MCNP cell card MCNP

    Homework Statement:: I go back to the line to finish the previous line in MCNP cell card but I had the error message shown in the photo. Please make a solution to my problem Relevant Equations:: c ********************* BLOCK 1: cartes des cellules **************** 1 2 -1.184 -40 #3 #19 #18...
  13. nuclearsneke

    MCNP5 or MCNPX, parrallel beam

    Hi, I am interested in simulation of parallel beams for neutrons and photons (separately of course). Any ideas on how to simulate them in MCNP5 or MCNPX?
  14. A

    Model UO2(5%)+Th+U233 Fuel in MCNPX for SCWR

    Hey there, I'm working on an MCNPX modelling for SCWR using different clads and fuels, the first fuel was UO2(5%) and I have calculated the number density correctly since there was only one vector U. But now I don't know how top deal with the Th+U233 due to the existence of Thorium. Can anyone...
  15. tmanici

    MCNPX , problem of detecting the photons in lattices

    Hello everyone! I hope you all doing well :) I am having a trouble with detection the radiation in lattices. i am adding the input and the result file here for makes everything clear, If someone can help me i would really be appreciate! thank you! ☺
  16. L

    How to use mesh tally in MCNPX to calculate dose?

    Hello,guys, I wonder how to use mesh tally to calculate dose.I set a cylinder,and set the material of the cylinder.Then I want to divide a cylinder into smaller cylinders in a direction perpendicular to the z-axis.And I need to record the flux of each mesh,use dedf card to get dose. I have tried...
  17. Ericdjs

    Could anyone help me with my problem, my MCNPX simulation is not working

    My MCNPX is not working when I start the calculation. It says "bad trouble in imcn in routine xin cross-section file bertin does not exist."
  18. A

    Troubleshooting Geometry Cutting with Vised X_225 and MCNPX 2.7

    Could someone tell me why this happens when I cut geometry? The program that i used is Vised X_225 and my mcnpx version is 2.7 Sorry for my posts, I'm really in trouble.
  19. A

    MCNPX - problem in cross-section

    Why does mcnpx not recognize the shell when I crop the cell in half? I put on a lead shield. I put everything (covering everything) and it worked. I cut half and the shield stop of work, but the cell is there. 10 2 -0.9500 (-1 2 -3) #20 imp:p=1 VOL=149.2256511 $ espessura / thickness...
  20. A

    MCNPX - Question in SDEF card about AXS and EXT

    My code version is 2.7 I have a disk source of R=0.3 cm, 60 cm above in z axis. I want set limits for the x and y axis, but, I can only put one command "axs" and "ext". How can i define two limits with one command? my code it is like this SDEF pos=0 0 60 rad=d1 axs=1 0 0 ext=d2 PAR=2 ERG=0.018...
  21. A

    MCNPX - How calculate Kerma (kinetic energy released per unit mass) in Air?

    Hi, my name is alexander, i am student from Institute of radioprotection and dosimetry (IRD). My project is calculate MGD (mean glandular dose) from womans with augmented breast. i am having dificulties to calculate Kerma in air with mcnpx. I drew a block of air above the breast, i am using the...
  22. A

    MCNPX Mesh Tally Problem: Offset of proton flux at surfaces

    Dear all, we are using MCNPX for the simulation of proton beam interactions at our proton therapy facility. But now we found a very strange behavior within a RMESH1:h flux tally: In an even very simple geometry we see a constriction or rather offset of the Proton flux at surfaces (surface 1118...
  23. P

    Rotational Symmetry in MCNPX core design

    Designing a PWR core in MCNPX for burnup using 4 folds rotational symmetry to reduce computational time of the core, taking reflective boundary conditions on rotational symmetry planes. should the power be reduced to 1/4th of original power (3000 MWth) in burnup card or does the reflective...
  24. K

    Is there a maximum limit for particle histories in MCNPX input files?

    Hi, I would like to know if there is any maximum limit for number of particle histories can be used in mcnpx input file. Thank you. Regards,
  25. M

    Solve MCNP Error: Low Sampling Efficiency with Lattice Source

    Hi, I have a sphere that it contains many sub-spheres. I want to define these small spheres as volumetric source. But when I run MCNP, it doesn't work. MCNP error: the sampeling effeiciency is too low Maybe someone can help me. 100 0 10 200 1 -1 -10 fill=2 300 1 -1 -2 3 -4 5 -6 7...
  26. H

    Error: no plot because it would be empty, using mcnpx

    Hello everyone! I would like to ask you if anyone can help me solving my Problem with mcnpx. I created a file with a watertank, a nozzle and a slit blend on the surface. I also inserted a protonbeam and would like to plot the dose and fluence for examplte with Moritz. But when I execute the...
  27. H

    MCNPX do not finish a simulation possible bug in the code

    I am testing the MCNPX plugging MCUNED to make calculations with neutron generators. After the compilation many examples to test the installation are provided. But one of them (I attached the code below) starts but it never finish. Just keeps in the first rendezvous. I first though in a problem...
  28. A

    Help Needed for Modeling Hybrid Reactor Geometry in MCNPX

    Hello, i am Ali doing PhD studies i am working on Activation Analysis of Hybrid reactor but i just stat studying this geometry but i am not able to understand this geometry for MCNPX modeling, so i need some help in modelling this geometry in MCNPX. please guide me in this regard
  29. Mr. mir

    Mcnpx fatal error in continue run

    I got and fatal error like below. " fatal error. continue run not yet consistent with histp writing. " Couldn't I use continue run when I make histp file? please, help me... calculation time is too long, so I want to use continue run.
  30. Rosenti 87

    MCNP5 vs MCNPX: What's the Difference?

    Hello everyone, I need help. Anyone can explain what is the basic differences between MCNP5 and MCNPX? I appreciate every suggestion for help me. Thanks. Rose
  31. K

    MCNPX dose result is too small

    I want to calculate absorbed dose by MCNP and phantom. describe x-ray tube and set radiation filed is 30cm x 30cm I use SRS-78, describe sorce card. 100kev 20mAs 14degree target is tungsten and film is Al 3mm. I use *f6 tally so I convert jerk/g ->Mev/g use 1Mev = 1.60E-22jerk and then...
  32. E

    MCNP: How to specify a small source in a large lattice

    I am working on an input file in MCNPX/6 that uses a CT scan lattice geometry. I want to specify a small source in a large universe (lung). Right now I have a source uniformly distributed through the universe. The existing documentation is vague on this topic. Is it possible to contain the...
  33. N

    Solving "Fatal Error" in MCNPX Input Code

    I am running a simple MCNPX input code and am getting a fatal error that says: "Fatal Error: no m card for material no. X". I thought it was something with my data card or my cell cards but can not figure out what the problem is.
  34. B

    Where Can I Download MCNPX or MCNP5C for Free?

    hi please help me to download free MCNPX Or MCNP5C ! Where can i download those soft? Or MCNPX visual basic? please help me! tanx
  35. C

    Can MCNPX Perform Time Dependent Calculations?

    Hi there any body know that mcnpx has abality to do time dependent calculations? please lead me quickly, I need it. regards