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.
Hi all,
I am working on some criticality problems using MCNP4C. A key aspect of what I am attempting to do is to use Python to automate the creation and running of MCNP input files.
One of the issues so far is that after each command entered into the Command Prompt window, the MCNP...
During a reactor assembly calculation, I need to determine axial and radial flux distribution over the surface. When I use F2 and F4 tally I get some value with unit 1/cm**2
What does the value means, neutron flux is supposed to be in the 10^14 range but output values are 10^2 range.
Can anyone...
Hello everybody. I would like to ask a question;
if I change the dimensions and densities and the material number identifier ZAID to a specific temperature.
Does MCNP change automatically the volume of the cells? or I have to change the volume of each cell manually and indicate it in the cell...
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 there,
I have a very simple question about MCNP (6.2 to be precise) ... maybe someone here might enlighten me ...
Based on the (more than simple) MCNP input file below, which describes a sphere with R=200cm, filled with air and a point source in the center. There's a single ring tally on...
I need help to construct this source on mcnpx. I tried a lot of thing, but nothing worked. The source is a isotropic point source, but the beam is rectangular with dimensions of the same of scoring plane. The source is the standard mammography beam. Please, help.
My parameters in use:
SDEF...
Hello, I am a student who started studying MCNP. I'm not used to writing in English, so I'd appreciate it if you could understand even if there were grammatical errors in my thread.
I want to check the energy of gamma rays from neutrons reacting with matter. So, I wrote this in the content of...
Hi, Is there any tutorial or pdfs that can help me with the MCNP output data?
I'm working on the criticality of the SCWR, and I designed the fuel assembly and run it on the MCNP, but I have no idea about data extraction.
I wrote a program to determine the critical mass of uranium oxide with an enrichment of 10%. I got a keff approximately equal to 1 with the selected volume and density (attached a file). Is it possible to somehow run the program without writing the initial density and volume into the conditions...
How do I express the temperature in the cell cards of the MCNP?, Say the temperature of the fuel is 500K, how do I write it as the following PWR example?
Hello everyone,
trying a run I've got the following error message, and thus no output file since the run stopped by itsefl.
Please, has anyone encountered this error and would happen to know how to correct the input file ? (any hint help)
Thanks to all of you
The MCNP6.2 manual (page 3-37) says: "There are two nj values that can be used in the lattice array that have special meanings. A zero in the level-zero (real world) lattice means that the lattice element does not exist, making it possible, in effect, to specify a non-rectangular array."
How...
hi dears
I am trying to plot my mesh tally with no success.
Try it on vised (no tally no plot) and on promt (see photo)...
if anyone can give an insight is highly appreciated.
tks
Carla
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!
☺
I am trying to simulate fission product ejection from thin fissile films in gas filled detectors (fission chamber). Does MCNP 6.2 produce recoil fission products that will be transported through the system?
I have enabled "heavy ion physics" (#), tried options 3 and 5 for NCIA on the neutron...
Hi everybody,
I am trying to score fission spectrum in MCNP for a kcode calculation. I would like to check at which energy neutrons produced by fission are generated. I have no idea how to perform since tallies are usually volume or material dependent and I just want to build a spectrum...
Hi,
I have a problem with MCNP. The problem is : the input file can run correctly on WindowsXP system,but when I run it on a more powerful computer with win10 system , the MCNP showed "chg_mem error". How to solve this problem? can MCNP only operate on winXP or win2000 system?
I was reading papers on neutron flux traps. Some people in the University of Tehran used MCNP to determine effectiveness of different neutron moderators. Context out of the way, my question is regarding the situation. How did researchers paid by an organization like the University of Tehran get...
I am running some depletion calculations and have noticed odd behavior when keff is much greater or lower than one in the number of absorption's taking place on the outside the reactor. MCNP seems to be deleting the neutrons after it has reached the number of fissions required for the power...
Hello,
I am getting the following error
Unexpected error in file
I have gone through the code and noting seems to be amiss. Does anyone have any ideas?
Hallo everybody,
I am using a mesh-based-weight window generator. It is clear to me that the coarse mesh must cover the full geometry; my doubt is about the fine mesh. Is it possible to define the fine mesh only in a part of a geometry, for example telling to MCNP to make 5 bins between y1 and...
Hello All,
I have yet another MCNP question. I received the following error "geometry error: no intersection found mcnp" when trying to run a a simulation. I looked at the output and according to it I have an infinite volume in cells 14 and 500. I plotted the geometry and don't see how its...
Hello,
I am running a MCNP calculation where there is a source in the center of a building. I have put a fmesh tally parallel to the ground so that I can see how the particles disperse throughout the building and outside the building. However, I only really care about the results from the fmesh...
I need some help defining a tally volume. I want a volume bounded by three surfaces, but when I do an initial plot the volume is in red dashed lines. I know that each cell needs to be uniquely defined, but I am not seeing how my volume is not unique. The cell in question is cell 200 in the code...
I have been encountering "zero lattice element hit" when I try to run it in MCNP5. The input file is provided below. The square lattice runs fine, but the hexagonal does not. Hope someone can help me here. Thank you.
MCNP (version 6 at least, possibly other versions) produces a summary table. An example is attached. The example shows the neutron portion of the summary table in a combined neutron/photon case. There is an SDEF card that produces photons, and the weights on that SDEF card give the relative...
Hello,
I am working through the MCNP manual and am experiencing the following error as well as warning when trying to run the sample problem from the manual.
The fatal error I get is
"fatal error. surface 0 not found for cell 1050."
I have upload the output file with the errors in case I...
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.
Hello for all of you,
I am having a problem with the visual editor of the MCNP.
When I use cards like this cell card:
(( 10 2 -1.04 -2600 imp:p=1 vol=176 ))
in the -visual editor- The VISED automatically rewrite the code after updating it, and eliminate the importance and the volume...
Hi,
I did an MCNP simulation to see the neutron energy deposition in water. I used 14-MeV neutrons, and big enough water body to make sure all the neutrons stay in the water and give all their energies to water. I used F6 tally. However, I got energy deposition of around 10.3 MeV, not 14 MEV...
I don't have anyone in my university who can help me with MCNP. I'm trying to write an input deck to calculate fluence in a zinc sulphide scintillator. The alpha particles are being emitted from a solution containing plutonium, plutonium nitrate and plutonium nitride.
There are lots of things I...
Dear colleagues,
I'm trying to make calculation of flux in ex-core detector. I have to evaluate neutron flux from the part of fuel assembly in full-scale model of PWR.
I don't know very well MCNP Code, so I decided to use F5 card with option FT5 ICD and FU5 671 number of cells, where the part...
For example, I don't understand slide 4, line 4 of the link below.
I know that the first three numbers are the cell number, material number and density of water. But what do the next 6 numbers mean (2, -1, 4, -3, 5, -6)?
Similarly on line 8, what do -0.001293 (101, -102, -100) mean?
Thank you...
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...
Hi there,
I would like some help understanding the attached MCNP output file.
The file tells me that the mean alpha energy is 7.1931E-04 after a million simulations.
I have two questions:
Does the file tell me anywhere what the error in the mean value is (+/-)? Or can I simply work this out...
Hi,
I have been doing some simulation with MCNP. They take a long time. I think there is a way to ignore some particle histories, which reduces the time for simulations. In my model, there is a point neutron source, an object, and a detector. Most of the neutrons do not reach the detector. So...
I am a masters student in the UK. For my project I have to monitor the efficiency of zinc sulphide detectors for monitoring alpha particles in liquid solutions. I need to model things like the proximity of the detector to the liquid, the size of the detector and the thickness of the light-tight...
Hello my friends,
I would like to ask a question about MCNP.
I want to score energy fluence on a rectangular radiograph planar grid. Do you know which tally should I use?
I use FIR tally for particle flux.
Hope you know an answer to my question :)
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. I need some help with the use of tally card in MCNP. I have been trying to use the f1, f4 and f2 tally to calculate surface current, average flux on a cell and avergage flux on a surface respectively, my question is: It's possible use those kind of tallies with macrobodies and surfaces...