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 would like to ask about this error " bad trouble in imcn in routine dynamic allocate"
this came up when I described a hexagonal fuel assembly with lat=2 fill=-17:17 -17:17 0:0 but it gave the error above. is there a maximum for this?
Hi!
First of all, thank you for your time.
I am simulating a nuclear engine for space applications. I want to know the fission rate of the engine but i dont know how. I am using xming to plot the fmesh 4 and the tally is:
fmesh4:n geom=xyz origin= -50. -50. -50.
imesh= 50...
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?
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...
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...
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...
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...
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?
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
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.
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...
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?
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...
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!
☺
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...
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.
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...
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...
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...
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...
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...
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...
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...
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...
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
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.
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
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...
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...
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.