What is Mcnp: Definition and 187 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. S

    A MCNP for the ionization chamber

    I need help to develop an mcnp geometry for input and output of Percent depth dose.
  2. emilmammadzada

    Design error in reactor nuclear simulation in serpent code

    Dear experts, I would like to get help from you on something. I want to design a vver -1200 core in the serpent nuclear code, but I am getting errors in the geometry I defined below. I would like to get help from you on how to fix these errors.I seem to be getting errors in cell definitions and...
  3. Hamidul

    MCNP code for Neutron Spectroscopy

    Hello everyone , in my mcnp coding for finding neutron spectroscopy I used F2 tally across a surface. Is it correct or I should use f4 tally? Morever I need to transform the flux data into neutron fluence. How can I do that. Here I uploaded my code. Though my data from codes is way more...
  4. Hamidul

    Help with neutron spectroscopy experiments in MCNP code

    Hello everyone, currently I am doing a neutron spectroscopy experiments. I am doing it with the MCNP code. I designed my Geometry there, but facing problems in data cards, is there anyone who can help me in this sector?
  5. Anisur Rahman

    MCNP: Can I input the X-Ray tube voltage in MCNP source specification?

    Can I input the X-Ray tube voltage in MCNP source specification?
  6. Anisur Rahman

    Question about Source Probability in MCNP

    Here, SP stands for source probability. But probability needs to be normalized. Here values in SP3, SP4 are larger than 1, It means that SP is not ordinary probability here. But what actually SP represent here?
  7. Anisur Rahman

    What is the physical significance of WGT and SBn in MCNP source specifications?

    What is meant by SP1,SI1 and SB here? I actually can't get the physical significance. And What is the physical significance of WGT here? Sorry for my this kind of questions. I am novice in MCNP.
  8. A

    How to check the isotropy of the source in MCNP?

    Hello everyone! I need to make sure that my source is isotropic. How can I check that? I have point source pos -11 0 0 erg=d1 with Maxwellian spectrum of energy and some surfaces through which neutron flux passes.
  9. A

    Meaning of "Average" Flux Tallies in MCNP

    Hello, I've been working with MCNP on and off for a few years now, but just recently realized that I don't entirely understand how tallies are actually calculated in MCNP, and what they signify. Taking the example of the F2 tally, the user manual (Section states that F2 is the "flux...
  10. A

    Isotropic distribution for a surface source MCNP

    Hello everyone! I have to use isotropic distribution for my mcnp program. But I didn't find how can I create that one.
  11. A

    Please help with this MCNP program -- in the output response I get too small a value

    Hello everyone! I have some troubles with my MCNP programm: I have a source, a moderator and a tally. The source is surface, the moderator is water (but I need to calculate for vacuum as well). Only neutrons are used in this task. The neutron flux is unidirectional. I take 1e6 the number of...
  12. 1

    MCNP6.2 - SSW and surfaces defining universe boundaries

    Hi everyone, I am using SSW card. Although the manual is very clear about the fact that the cells used in SSW card have to belong to the lowest level, the manual is not that clear the surfaces. Is SSW able to track particles crossing surfaces defining higher level universes? Let's assume this...
  13. E

    Revisiting MCNP: Refreshing Skills for a New Job in Nuclear Engineering

    Hello, After some time away I've gotten back into MCNP. I've been in the field of Nuclear Engineering for over ten years, but I recently changed jobs and need to use MCNP. I'm trying to get my skills back up, since I haven't been using it as much in my old job. Looking forward to some great...
  14. E

    Efficient MCNP Lattice Source Help: Defining Universes and Tallies in Cell File

    This is what I hate about MCNP, not a lot of documentation. How do I define all of a universe as a source and a tally? I have a lattice like the below code. How do I get this code to work with tallies for positions 1,2, and 3 in the lattice; and a source for the 2's. I get the error "sampling...
  15. S

    MCNP: Ctme Card Impacts on Results

    Ctme card is used to limit the running time of the mcnp input file, does it affect the telly result or not ?
  16. S

    MCNP: How to display particles outside the source?

    Particle display in visual editor of mcnp input file only shows particles inside source, I am interested to see particle path towards tally region. Plz guide thanks
  17. S

    Run MCNP 5 input file of certain geometry for flux calculation

    Can any one please explain if I want to run mcnp 5 input file of certain geometry for flux calculation on different surfaces. So far as I know If I increase the NPS (number of particles) it wll give more accurate result but when I increase NPS from 10e9, input file do not run and close within a...
  18. chengmo

    How to use mdata file data output by mesh card in mcnp software?

    How to use mdata file data output by mesh card in mcnp software? I converted the mdata data using gridconv.exe, but I don't know how to use the data to identify the section graph for the xyz axis? Does anyone know how it's arranged?
  19. K

    MCNP Problem - Bad character in column 2

    hello , I am getting an error that reads as follows: 1 PX 12.5 bad trouble in imcn in routine ckchar bad character, probably a control character, in column 2Cell cards 1 0 1 -2 -7 8 10 -9 imp: n=1 2 0 2 -3 -7 8 10 -9 imp: n=0 3 0 3 -4 -7 8 10 -9 imp: n=1 4 0 4 -5 -7 8 10 -9 imp: n=0 5 0 5 -6 -7...
  20. chengmo

    MCNP: Measure Dosage in Concrete with tmesh Cards

    When using tmesh cards to measure dosage in concrete, can mesh3 cards be used? My friend said it might be a mesh1 card but I can't use it. Can someone guide me? I want to look at examples
  21. 1

    MCNP6.2 - ENDF/B reaction numbers for tally multiplier

    Hi everyone, I am trying to evaluate the spectral index of an nonelastic (n,n') reaction. For that I want to set up a tally multiplier on a cell (let's call it cell 10). The reaction is present in the ENDF/B library as MT=4 but I have not seen it in the table of the special reaction numbers...
  22. R

    MCNP terminology question -- Effective Full Power Days (EFPDs)

    for MCNP users, i would like to ask about terminologies: if i depleted a fuel assembly under constant power, is the number of days in the out can be used as Effective Full Power Days (EFPDs), or this term has another specific meaning?
  23. chengmo

    Can MCNP solve the geometric coincidence issue with a semi-cylinder and cuboid?

    How do we solve the geometric coincidence problem? I need a semi-cylinder that fits into the cuboid but if I use the cuboid and the cylinder directly it's geometrically problematic
  24. chengmo

    MCNP macro definition modeling

    I use macro definition to model the results can be viewed in vised but vised does not display all the big guy know? Is there something wrong with my modeling?
  25. chengmo

    MCNP macro definition modeling

    C 1 1 -19.35 -7 1 -2 2 1 -19.35 -8 3 -4 3 1 -19.35 -5:-6 4 0 5 6 #1 #2 C 1 RCC 0 -10 0 0 -10 10 2 2 RCC 0 -10 0 0 -10 10 5 3 RCC 0 10 0 0 10 10 2 4 RCC 0 10 0 0 10 10 5 5 RPP 2 5 -10 10 0 10 6 RPP -3 0 -15 15 0 10 7 RPP 0 5 -15 -10 0 10 8 RPP 0 5 10 15 0 10 C M1 074184 1
  26. A

    Help debugging MCNP code - particle lost and zero latice element found

    I keep getting particle lost error even though there were no hole in the lattice. Can someone identify any mistake in my code?
  27. jianggong

    14MeV neutrons in MCNP interact with carbon without producing alpha

    14MeV neutrons in MCNP interact with carbon without producing alpha particles and protons, yes Questions about my cross section data? I've tried ENDFB8/B7.1, JEFF3.3, JENDL5, CENDL3.2 without any results, but if you use phys:n model, it looks like alpha particles will be produced, but it doesn't...
  28. 1

    Is it possible to use TR or TRCL to translate an F4 or FMESH4 tally in MCNP6.2?

    Hi everyone, I'd like to know if it is possible to use TR or TRCL to translate an F4 or FMESH4 tally. Let me better explain: I have a cell, centered at 0,0,0 and for this cell I set up a segmentation tally along z-axis with a series of planes orthogonal to the z-axis. Moreover, I have also set...
  29. Islam Nabil

    I need an MCNP simulated APR1400 input file

    need a MCNP simulated APR1400 input file? which consists of lattice 16x16 inside a lattice 17x17 ??
  30. A

    MCNP FMESH for Plotting power distribution

    Hello I'm trying to use FMESH command to get power distribution of this core geometry. I want to use xyz coordinate in a 1/12 slice of a core so I could use the output of the MCNP sim for a CFD input How should I approach this? Thank you
  31. Islam Nabil

    MCNP lattice of the fuel assembly input file?

    There is an input file for a simple 16 x 16 lattice fuel assembly. I have a message blocking the run of the code; "bad trouble in subroutine newcel of mcrun source particle no 1 random number 6647299061401 zero lattice element hit." What is wrong?
  32. 1

    MCNP Surface Tallies: F1 & F2 on Infinite Cylinders & Planes

    Hi, I have a question concerning surface tallies like F1 and F2. You have to provide a surface for them. Since, surfaces are defined as infinite (infinitely long cylinders, infinitely extended planes) how can you write the surface tally of a cell? What are the actual tally surfaces for F1 anf...
  33. 1

    What is the difference between FMESH and MESH in MCNP?

    Hi everyone, I am struggling to understand the difference between FMESH and MESH. FMESH is used to create a mesh superimposed with the geometry but...what does MESH? Is it only involved in weight-window generation and not needed for mesh tallies? Thanks in advance for the clarification.
  34. 1

    MCNP - z-coordinates of cylindrical mesh >0?

    Hi everyone, I am using MCNP6.2 and trying to set up a cylindrical coordinate in a reactor channel. The origin as the midplane of the channel. In my attempt of setting up a cylindrical FMESH with the origin on the z-axis at the bottom of the channel (so z<0) I got this fatal error message...
  35. 1

    MCNP - Tallies definition with "<"

    Hi everyone, In MCNP manual there are often examples of Listing containing examples of tallies which have, in the definition of the cells/surfaces of the tally itself, the "<" symbol. I could not find in the document any reference to the use of logical expression in the definition of tallies...
  36. 1

    If MCNP tallies are normalized, shouldn't they be <1?

    Hi everyone, I'm really new to MCNP here and I'm "playing" around trying to understand what is going on. I think I am having problems understanding what, in a criticality calculation, the MCNP tallies are normalized to consequently, how comes they can be >1. I was thinking that, in a...
  37. 1

    MCNP6.2 - How to plot normalized tallies with MCNP6.2?

    Hi everyone, I'm really new to MCNP here and I'm "playing" around trying to understand what is going on. I'd like to plot my tallies (F2, F4 and F6). Is there any tool (e.g. python or matlab package) you recommend? I know that the internal plot MCPLOT is available but I'm wondering how you...
  38. 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...
  39. W

    Where Are the Other Detectors for F5 Tally in MCNP?

    The MCNP manual states that you can have multiple detectors for a single F5 tally. Say you have f15:n x1 y1 z1 r x2 y2 z2 r.....Thing is, my output file only lists the tally result for the first f5 detector (x1,y1,z1). Where are other detectors for this tally? Is there a reason code developers...
  40. A

    MNCP deleting input surfaces

    Hello Tried to model gas cooled reactor with hexagonal fuel elements. MCNP keep deleting surfaces (If you could, run my input and check the .txto file) so the simulations won't run Any advice?
  41. A

    Troubleshooting MCNP k_eff for Space Reactor Core: Tips and Tricks"

    Hello! I tried modeling a space reactor core with MCNP. I'm pretty sure the geometry and material properties are correct. Got k_eff of 1.4, much higher than 1.003 from the reference. Could anyone spot the mistake in my code? I couldn't figure out anything else
  42. Luizpo

    PTRAC File - MCNP - Multi-core computing

    Homework Statement:: PTRAC File - MCNP - Multi-core computing Relevant Equations:: No equations My name is Luiz. I am a postdoc at the institute of energy and nuclear research in São Paulo-Brazil. Our group models a cold neutron source (CNS) for the Brazilian multipurpose reactor project...
  43. 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?
  44. D

    How can MCNP code be used to calculate photoneutron doses in radiotherapy?

    I worked monte carlo simulation for dose calculations in field radiotherapy (external and brachytherapy) using EGSnrc and MCNP codes.
  45. 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?
  46. lee6853

    [MCNP] Lost too much Keff with Burnup card

    Hi there! Me again. I am doing my research about converting HEU research reactor to LEU. I made FA and core finally and started using the burnup card to check changing of Keff and fission products. Well, the thing was only with one-month burnup my Keff was decreased drastically from 1.118 to...
  47. lee6853

    MCNP and simple nuclear physics

    Hi guys! I'm a master's student majoring in nuclear engineering in graduate school. I have a few questions while doing research, so I'm writing this here. My research is simple. We conduct neutron analysis to convert a research reactor using highly enriched uranium into a low enriched uranium...
  48. A

    MCNP6 with mpi failed with signal 11 (Segmentation fault)

    I use Python scripts to run mcnp.mpi like And I encountered this bug report The scipts has run normally for a few hours. I extracted the inp file and it can be run normally. I searched on Internet and found it seems to be the problem related to memory, but i checked the log, there's still...
  49. 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?