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.
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...
Hi there,
I´m just finishing an input file for MCNP5 and I can´t find a value of density for UO2 enriched to 3,25% - 3,6%.
Does anyone know it or know where I can find it?
Thanks in advance!
(P.D.: wikipedia is not my friend... )
I had some problems finding out an equivalent way to describe a particular gama source.We can get the original way describing the particular gama source in this file: "1.txt", as well as its model in this file:"model.jpg",with this describing way,we can get the distribution of gama flux like...
Hello,
I am a graduate student attempting to run evaluate the depletion of a ceramic film attached to the moderator-side of the fuel clad. I am having some issues with my MCNP syntax/code and I was wondering if one could assist.
My input file is attached. I am not looking for someone to...
Hi,
I have written a source subroutine and I am trying to link it with mcnp6. When I run my input file it says you need a source subroutine. My input file and the source subroutine(written in fortran) are in the same directory. Where do I have to keep the subroutine to link with mcnp? Any help...
Hello to everybody,
I need some explanation on how to use SDEF variables to define correctly a planar rectangular source. Let say this source is emitting in all direction but I am interested in a side of the source surface where a point detector is located. I used in my example VEC (VEC =001)...
Assume no boron and all control rods out, so the core is super-critical. if KCODE mode is used, and F4 card is for tally the neutron for each assamblies. Can the results represent the power distribution of the core, whether the multiply-factor can affect.
the results i get (power...
mcnp error "the new source has overrun the old source"
I am a beginner with MCNP. This error really confused me. Who knows what this error means? Thank you.
Hello,everyone.Lastly,I use MCNP to claculate the dose distribution of CT phantom,but when I run my code,I get a message that says
' bad trouble in subroutine newcel of mcrun
source particle no. 18859
starting random...
fatal error. *** cross-section tables are missing from xsdir
10001.70m
10002.70m
10003.70m
...
bad trouble in subroutine ixsdir of imcn
cannot continue with missing cross-section table(s).
is the message I am getting. I have created a link using
ln -s...
I have a question about diffrence between ways of defining a certain plate like (3^1/2)x+y+17.3=0 to MCNP code .for example I know we can define it to MCNP with 15 p 3 2 0 6 but I couldn't understand this method 15 10 0 5 8.66 0 10 0 1
Hi everyone,
I’m a student and in my master degree, I’m working in a project about a bunker for industrial radiography of steel, and I need find some input of mcnp about this matter.
Somebody can help me?
Thanks a lot
The only thing I know how on the basis of nuclear reactor design is how to run kcode in MCNP and see if my theoretical reactor is critical. How would I be able to calculate how the fuel is burning in my reactor over some period of time, and change my core composition accordingly?
hello, I am new to MCNP, could somebody tell me how to use imp:n, what is imp:n=0 means, if neutron importance is 0 in one cell, why the F4 tally is 0 in this cell? how about imp:n=1 or some large number?
Thanks for all.
I am a beginner user of MCNP and is still learning how to use it. When I run my program, I get a message that says 'bad trouble in subroutine main of mcnp'. What exactly does that mean and how can i correct it.
Thanks.
I've been having difficulty in successfully compiling the MCNP executable with the MPI/MPICH libraries on Scientific Linux 6.x (whatever the latest version is). I'm currently using MCNP5, and the latest distribution DVD. The compiler is GCC, and the MPI is MPICH 1.4.x (the version immediately...
Hi
I have problems in mcnp input of voxels. I define a voxel arrays and their universes,for example universe1 is defined as:
2001 43 -1.030 -70 u=1 vol=6.795352
but in runing I have below error:
fatal error.the surface type is not recognized: -1.03
While 1.03 is the density of...
I have just started to use MCNP code for nuclear reactor modelling. I would like to practice using a visual editor for input file. Can anyone tell me how can I get a free version of visual editor or other similar editor compatible with MCNP?
I'm curious if anyone has any opinions in the results from MCNP vs Geant4 for shielding design.
I like that Geant4 uses a more modern syntax, but iv'e also heard that MCNP gives better results.
Hello all. I'm am a first time poster but a long time visitor. I am having a little trouble that I was hoping someone far wiser and more knowledgeable than myself might be able to help with.
I've been using MCNP to investigate criticality in a simple geometry consisting of a central natural...
hi there
I wnat to calculate the group constants for a FA(fuel assembly) using MCNP (similar to lattice calculations).
How can I do it? Please lead me.
Thanks alot!
Hello all, I'm a senior Nuclear Engineering student.
This semester I've been working with DANTSYS and MCNP (more like failing to learn properly). I was wondering if anyone had some experience with either program and would be willing to allow me to ask them the occasional question. I think...
hello,
i am using mcnp for a reactor kinetic study.the only problem is, to establish the model i need to calculate the temperature coefficients, and in order to do that i need to calculate k in different temperatures and therefor need a mcnp library that contains cross-sections for a multitude...
Hey,
I'm modeling the criticality of a core for a university project using MCNP4C. I've run the core on the appropriate KCODE parameters, specifying the source using the KSRC card, and I was looking to cut down on the computing time using the SRCTP files.
My problem is that I can't find...
Hi there
I have a question about FMn card in mcnp code.
there is a parameter (is named C) in front of FMn card, I can not understand to calculate the value of that parameter. please help me.
thanks alot.
Hi there,
I want to know that how can involve the temperature in mcnp code.
for example; the library endf7 for mcnp has five certain temperature:300 kelvin, 600 kelvin, 900 kelvin, 1200 kelvin & 1500 kelvin. if I want to calculate flux distribution in 330 kelvin for a research reactor, how can...
I am running (a very basic) simulation of the proposed LIFE concept reactor at LLNL as part of my MSc thesis. What I hope to achieve is to calculate the fission energy gain from a fissile blanket surrounding a source of fusion neutrons (ie D-T pellet blasted by lasers)
The problem summary...
Im having some trouble interpreting the MCNP output file, more specificaly table 140 that describes reaction rates.
The problem of course is that the acctual reaction rates isn't written but rather things like total colissions, collions*weight, weight lost to capture etc.
How do I convert...
Does anyone know if there are templates for things such as the ICRP 23 reference man (70kg)? I am trying to create one for a little side project I am doing (calculating the photon dose my friend gets from sleeping next to his cat that has been treated with I-131). I am trying to create the...
I have to write some input for MCNP for one of my classes. The graduate student had given us a lecture and some examples, but I'm trying to find some more detailed documentation, but even anything basic would be nice.
The one I did run across, from LANL, I needed authorization to view, why is...
" 'New Mexico State University Radiation Safety Manual', is a program prepared by Katrina D. Doolittle with assistance from Trina F. Witter in partial fulfillment of the requirements for a Broad Scope Type AB Radioactive Materials License issued to New Mexico State University.
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