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, 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...
Hi,
I am trying to model the distribution of the light emission from a material when excited with neutrons in MCNP. I have been searching literature and found not many things. Could anyone provide me with sources from which I can get info?
Thank you in advance.
Hello, it's a privilege to enter in this forum. My name is Geovanny, I'm from Mexico and I'm a student from the Autonomous Yucatan University. I'm studying in my last semester of Physical Engineering and i have some doubts about de MCNP software.
One of those is the next message that VISEDX...
Hello, every body, I'm a new gust in this forum , and I have a question to the users of the simulation VISED version of MCNP software.
Can we change some parameters in the materials part of the input file then reread the tally again to plot two different curves together on the same graph, or...
Dear all,
I would like to simulate X-ray tube and check the dose rate in the room.
My problem is when I simulate 80kV electrons bombard to the tungsten target, there is just a few photons coming out from the tube window. And the dose rate at 1m from the tube is nearly zero which is impossible...
Greeting,
I am trying to figure out how can I include the activity of a Co-60 source in MCNP code.
I have the following problem Co-60 source in a cylinder surrounded by concrete. I just need to know how to include the source activity and whether it should be in Bq or Ci.
Thank you in advance
Sorry for two questions in a row, this one we have been stumped on for the entire day
We're getting "fatal error. detector no. 1 of tally 5 is not in any cell.", What could be causing this?
In the manual and all examples we've seen, nobody has parameters specifying cell location for the...
I'm modelling a scenario for the research that I'm working on, and I got the cells and surfaces all mapped out for the environment finally, but now I'm totally stuck on creating a source.
I'd like a point source of Californium-252, but after hours of looking, I don't see any out of the 1000...
Hey all,
I was wondering if anyone had any good tips on debugging mcnp geometry? I'm an intermediate user working on better understanding the program. Does anyone have any tips or tricks that go beyond simply reading the manual?
Hello,
I use free VisEd and I want to plot the collision, the particle's transport and the particle in tally but I have an error. For the source the plot is ok. How can I solve this problem freely?
Someone can help me?
My input :C Cellules30 50 -1 (-1 3 19 ):(-2 3 ):(-4 3 -6 5 -7 1 2 ) imp:p=1...
I've studied radiography recently and find a function of radiography tally in MCNP6.
After i run input file, the result of the radiography represent the color map that consists of red, green and blue etc... but i want to see the gray color.
How to change the color map??
I'm trying to learn MCNP and need some training manuals to get it completely.
please introduce the best ones.
and tell me what are differences between MCNP4X,MCNPX,MCNP5 and MCNP6
Hey all,
I am a nuclear engineering student at OSU. Just stumbled across Physics Forums while attempting to make sense of MCNP (Monte Carlo N-Particle code) .
The majority of my interests in my studies involves data manipulation and dosimetry.
Kirk
Hi, I'm new in this forum and I have a question about MCNPX.
I would like to determine the position of each interaction between photons and matter (specially by Compton scattering.) And I would like to know how much energy the incident photon gave to the electron.
Is it possible to do that with...
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...
I'm new to mcnp and trying to perform this calculation, if anyone can provide some feedback to see if I'm even going in the right direction that would be much appreciated. The geometry is correct with only 1 transverse, the issue I am having is making sure my data block is correct and how I am...
I have to calculate the gammas produce due to oxygen activation in the coolant of PWR reactor core. the tally i used is F4 and FM.. but the problem is in core their are different cells within a cell. so how to specify the required cell . i have selected the whole core as a cell and then specify...
hi everyone, i have a mcnp error, it don't finish, it only run a half! i hope someone can help me! thank you!
and my code
JCO
c cell cards
1 1 -0.00117 -100 300
2 2 -1.52 (-100:-101) -300 #1 $solution
3 4 -7.93 (-110:-111)#1#2 $clad
4 3 -1 (-200:-201)#3#1#2 $water
5 4 -7.93...
Hello everyone !
I wanted to ask you if anyone ever encountered problems with proton transport using MCNP. In particular, the production of electrons when protons interact with matter. I have been trying to figure it out modifying the different physics cards (phys, lca, ...) with no result...
Hello everyone,
I am having a problem with MCNP. My question is how to get number of secondaries in a certain volume. For example I have a neutron beam bombarded Pb target, and I want to count all of proton formed in the target. I considered tally F4, but the unit is 1/cm**2. Who can explain the...
Hello everybody,
I am performing some experiments with a neutron generator. Specifically D-D reactions. I am trying to replicate the measurements with MCNP6, but I do not know how can I simulate the neutron generator with MCNP6 since the neutrons have a angular distribution. For MCNPX there was...
I need help with creating repeated structures.
I have a hexagonal aluminium container, inside of which i need to put cylindrical rods. I defined a rod and its cladding and air surrounding it, as universe=5 then i defined another hexagonal surface with 0 material but fill=5 and i define this...
Dear all,
This is my first post in this forum.
I would like to know how to obtain the result data of an MCNPX or MCNP6 tally for each simulated history, before the data of different histories is averaged and normalized by the total number of simulated source particles (nps).
I'm calculating...
Hello,
I've heard that MCNP outputs delayed neutron fractions (beta) and neutron generation times (Lambda).
Any ideas as to where in the output file it writes these? I've coppied and pasted the Godiva reactor code and run it, but I can't seem to find it in there.
Best Regards,
-TP
I am running MCNP. It gives a error massage : fatal error description of cell 27 uses 1098 words. 1000 words max.
I am not able to resolve this. If anybody can explain me about this error and how to correct it. regards
Hello! I am really glad to find this forum!
I was wondering if anyone has a MATLAB code to parse the MCNP output file MCTAL? I did some Lattice simulation but it seemed hard to read the MCTAL file directly..
Best regards
Hi ,
This is my first post in this forum, I am new and happy to be in this forum :)
My question is, during the calculation of neutron and photon of a single-point reactor core, does MCNP5 taking into account the gamma decay? because during fission process, fission product can emit gamma. Does...
In PWR, fast neutron produced from fission in fuel has been moderated into thermal neutron by the a series of collisiion with coolant,i.e. H2O. So the multi-group neutron flux in coolant and fuel pin has much diffenrce, e.g. the relative higher fast neutron in fuel pin and relative higher...
Hi
I am trying to run a fast reactor model but I cannot figure out what is happening because I get the error: run terminated because 10 particles got lost. I am using the sdef card (with and without parameters) and the model is not running. Hope you can help me to fix any mistake I did. The...
Hi everyone
Does anyone help about unit of below MCNP line for neutron? Is it "neutron*barn/cm^2" right ?
fm14 (1 108 102)
And also what's the unit difference between "fm14 (1 108 -6 -8)" and "fm14 (1 108 -6 )" ?
Many thanks
Hi, i am looking for some help on MCNP, more precisely mcnpx 2.7 for neutron simulation.
I created a model of semi opened detector with a various number of 1 inch He3 (here 16)
and i only obtain 6-7% efficiency.
the fact is, I've others technicals notes about same "types" of detector, for...
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...
HI, i´m trying to build a lattice in mcnp. Actually it works but there are still lots of red lines in the plot.
Maybe someone can help me.
The Lattice code looks like that :
lattice
1 0 2 $ outer space
2 3 -0.00126 -2 #4 #39 #41 #42 #43 #44 #45 #46 #47 $ universe
4 2 -8.4 -1 $ ball to enable...
Hi,
I'm working on a MCNP simulation where I have to use F6 tallies. According to the manual: "In the F6 and +F6 tallies, material density is available for the chosen cells, and normalization is MeV/gm/source-particle."
To which source-particles is this value normalized: the source-particles...
Hello there,
I am using mcnp6 to simulate a tokamak. I am interested in the energy deposition in the blanket and I am using a fmesh4 and the tally multiplier fm4 as follow:
fmesh4:n ORIGIN=0 -24.2 -50 OUT=CF
imesh=35.2 iints=352
jmesh=24.2 jints=484
kmesh=50...
I'm currently working on a project using mcnp to model a HPGe detector with a Co-60 source. I have defined my cells and got my geometry spot on but the project requires me to move the source around about 50 different positions relative to the detector. I was wondering if there was any way to...
Hello
I am a lower-intermediate user of MCNP and I do not know how to obtain the diffusion coefficient (or maybe the angle of scattering) using tallies. I also have read a paper (Multigroup Scattering Matrix Generation Method using Weight-to-Flux Ratio Based on a Continous Energy Monte Carlo...
Dear all,
I have a problem related to SSR card.
I created a WSSA file (>1GB) from the input file by SSW card. I change the file name to RSSA in order to use for SSR card.
When I run SSR card, I got the the error: bad trouble in subroutine issrc of imcn unexpected eof on file rssa
How to fix...
I am currently using KCode and modeling a fuel array (Lattice), reflector array (Lattice), and a Stainless steel reflector (Cell filled with Stainless steel, simple geometry). The cells are situated right next to one another, such as 3 squares of similar sizes which make a rectangle. All...
i have a simulation problem about a fuel assembly, after running this warring pop up:- " non-lattice cell in lattice universe "; and visual editor crash with one warning message
" warning. 2 surfaces were deleted for being the same as others."
So, what may be the problem with the input?
Hi, I need assistance in performing statistical checks in MCNP5 i.e print table 160. I am not sure where the PRINT card should be placed and the format of it. I am using F4mesh tallies
Homework Statement
the radius from the symmetry to center of the plasma is about 6.2 metres and the minor radius is 2 metres
Homework Equations
Can you guys help me to make the plasma geometry for MCNP?
The Attempt at a Solution
the softcode of plasma geometry