Mcnp Definition and 207 Threads

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...

According to the manual FMESH only supports Cartesian or cylindrical geometry has there been any update to this or is that it?

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 am trying to build a honeycomb structure in MCNP. The code is given below. Hexagonal lattice 1 0 -1 -19 29 fill=1 imp:n=1 2 1 -2.26 -10 11 -12 13 -14 15 u=1 lat=2 imp:n=1 3 0 1:19:-29 imp:n=0 1 c/z 50 50 200 19 pz 31 29 pz -31 10 px 1 11 px -1 12 p 1...

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...

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

hello i want to know if MCNP version 5 can define 8 sources by sdf card i need a help please

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...

neutronics and burnup data step duration time power keff flux ave. nu ave. q burnup source (days) (days) (MW) (GWd/MTU) (nts/sec) 0...

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??

Hi! Are there any experts out there using MCNP on a Mac? I could do with a little help setting up, I am new to it. Here's hoping! Shahed.

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

Hi, I would like to know if there is any maximum limit for number of particle histories can be used in mcnpx input file. Thank you. Regards,

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...

I am now in the process to define a cubic volumetric source in mcnp code, does anyone have a clear explanation for that.

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