Mcnp6 Definition and 51 Threads

Hello everyone, I am currently working on a burnup calculation for a fuel assembly with repeated geometric structures using MCNP6. I have defined two materials (Material 1 and Material 2) which are actually the same material but located in different positions. However, after running the...

Hi everyone, I'm a complete beginner with MCNP and trying to learn how to perform burnup calculations. Right now, I'm feeling a bit lost and not sure where to start. I found the OECD-NEA Burnup Credit Calculational Criticality Benchmark (Phase I-B) and was wondering if anyone has worked through...

I was tasked with designing a shield for a Co-60 source, I decided to use ambient dose equivalent, tally 5. I got my results, followed "Radiation Problems: From Analytical to Monte-Carlo Solutions" example and arrived at a dose of X. I performed some shielding equations and arrived on a required...

I am facing a problem after running my input file through MCNP 5 it give me these comment👇👇 Can any one help me to resolve the error? mode n p comment. photonuclear physics may be needed (phys:p). comment. 9 surfaces...

Hello Dear Experts. I need help with the calculation of total thermal power in nuclear reactors. Can someone explain the general mathematical approach or provide useful formulas for this calculation? Thanks!

Hey all. I am creating an MCNP input deck but my geometry is appearing red in the vised plot and I am receiving errors relating to the geometry and material card sections. Could anyone potentially help me figure this issue out?

Hello everyone, I came across the following atom/b-cm values for a specific nuclear fuel composition: U-234.81c 1.1369e-06 U-235.81c 3.0421e-05 U-236.81c 2.4896e-06 U-238.81c 0.019613 Np-237.81c 4.6686e-05 Pu-236.81c 4.97e-10 Pu-238.81c 0.00011695 Pu-239.81c 0.0022076 Pu-240.81c 0.0013244...

Hello everyone, I'm working on a PWR fuel pin depletion simulation in MCNPX, but I'm encountering several warnings and an error that stops my simulation. Here’s my input setup: [c *** PWR pincell *** c c --- cell cards --- 1 1 -10.4 -1 imp:n=1 vol=192.29 $ fuel 2 2 -6.55 1 -2 imp:n=1...

I have encpuntered this error with the gamma spectra "entries are not monotonically increasing". Despite attempting the following solutions, the issue remains unresolved: Rearranging Energies in ascending order. Removed any duplicate energy values. What may be causing this error? and how can I...

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...

In my intro class, I am trying to design for radiation sources. Currently, I am trying to plot the flux from neutron and photon source locations progressively moving further and further away, however the code is not running and it says "geometry error: no cell found run terminated because 10...

I'm dealing with a specific situation: I'm analyzing Linear Energy Transfer (LET) in a cylindrical sample. According to the definition of LET in the manual, we have: “The linear energy transfer (LET) special tally option allows track length tallies to record flux as a function of stopping power...

Hello, I am a very new MCNP user and have been doing my best to learn on my own. I am struggling to get this problem I am working on to be even somewhat correct. I am trying to determine the dose rate for a person outside a lead-lined room(box) with a 20 TBq Yb-177 source inside. I'm using an...

15- mode a 16- m1 1000. -0.111894 $MAT1 17- 8000. -0.888106...

When I used PAR=-CR of mcnp6 to describe cosmic particles, there was an error:"Expire parameter is too many cases of erg > emax,bad trouble in subroutine startp of mcrun,"Any idea on how to resolve this problem?

the code file is attached

I would like to analyse mdata of tmesh in MCNP6 by python, but I am struggling how to convert mdata (unformatted binary) to any text file in python. I used GRIDCONV but it's not suitable for automations. I also used scipy.io.Fortrunfile but could not convert the file because I was not sure...

Hello, I'm new to MCNP deck-building, and I'm trying to acquire an X-ray energy spectrum using MCNP6, on Windows 10 environment. I'm running MCNP6 via MCNPX Visual Editor Version X_24E, and the deck is input using built-in "Input File" tab in MCNPX Visual Editor. My input deck is given below...

Hello everyone! I have to use isotropic distribution for my mcnp program. But I didn't find how can I create that one.

Hello everyone! I have some troubles with my MCNP programm: I have a source, a moderator and a tally. The source is surface, the moderator is water (but I need to calculate for vacuum as well). Only neutrons are used in this task. The neutron flux is unidirectional. I take 1e6 the number of...

For a novice research problem, I am approximating a system as a spherical reactor of homogenized natural uranium and heavy water, reflected by infinite graphite. I was attempting to find the critical mass and dimensions for it (very similarly to Lamarsh 3e Ex.6.7-8). To do so, I need to...

Hi everyone. I am struggling understanding how to combine more than one transformations, especially rotations. This stems mainly form the fact that it's unclear to me what reference frame is used to define the transformations angle if two consecutive transformations are applied. If I have a...

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...

Hello, I'm testing about burnable poison depletion with mcnp6. I'm using m1 92233.82c -7.91619E-01 90232.82c -8.75802E-02 8016.82c -1.20800E-01 for fuel and m12 64152.82c -1.39023E-03 $ gd-152 64154.82c -1.53529E-02 $ gd-154...

Hi everyone, I'm really new to MCNP here and I'm "playing" around trying to understand what is going on. I think I am having problems understanding what, in a criticality calculation, the MCNP tallies are normalized to consequently, how comes they can be >1. I was thinking that, in a...

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...

The MCNP manual states that you can have multiple detectors for a single F5 tally. Say you have f15:n x1 y1 z1 r x2 y2 z2 r.....Thing is, my output file only lists the tally result for the first f5 detector (x1,y1,z1). Where are other detectors for this tally? Is there a reason code developers...

Homework Statement:: PTRAC File - MCNP - Multi-core computing Relevant Equations:: No equations My name is Luiz. I am a postdoc at the institute of energy and nuclear research in São Paulo-Brazil. Our group models a cold neutron source (CNS) for the Brazilian multipurpose reactor project...

I use Python scripts to run mcnp.mpi like And I encountered this bug report The scipts has run normally for a few hours. I extracted the inp file and it can be run normally. I searched on Internet and found it seems to be the problem related to memory, but i checked the log, there's still...

Hi everyone, I'm making a fuel assembly model and I would like to have a cylindrical gamma source on each fuel rod to measure the decay heat of a fuel assembly but I'm struggling to define the source since it is in a repeated structure. This is the geometry of my model: Level 2 : cells 1, 2 3...

Hi! so its me again. After previous problem already solved now i ran into another problem. now about BURN card. its said need model? what kind of model? I tried using .84c for all my material but doesn't seems to work! heres my code if anyone wants to try :-( c TWR-P...

Hi! so i kinda stuck when i tried to run my code in MCNP6 because the output keep showing me "bad trouble in subroutine source of mcrun you need a source subroutine." While I am sure i already put my KCODE and KSRC in my code (on the picture below). Could anyone help...

MCNP6 gives me a "10 particles got lost" error when I try to run the attached input file modeling a 3x3 fuel lattice surrounded in coolant. As I understand it, this error is usually related to the geometry/surface definitions of each component, but I'm unsure of what the source of the error is...

uranium dioxide with 4.2% enrichment c Cell Cards 101 2 -0.0003922 -7 -5 6 imp:n=1 vol= 0.26195 tmp= 1.0109E-7 201 1 -10.55 7 -8 -5 6 imp:n=1 vol= 8.84672 tmp= 1.0109E-7 301 2 -0.001598 8 -9 -5 6 imp:n=1 vol= 0.288775 tmp= 1.0109E-7 401 3...

Hello everyone, I do need your help in this matter, please kindly help me solve this problem. I use MCNP5 and i want to use Tally E with Fmesh. I use Tally E and Fmesh this way with MCNP5. F4:P,E 6 E4 0 200i 2 FMESH4:P GEOM=REC ORIGIN=-550 -550 -1 IMESH=-50 IINTS=5 JMESH=550...

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...

Hello all, Starting yesterday I have been intermittently getting a " forrtl: severe <157> : program exception -access violation" and my runs stopping . Does anyone have any ideas as to what could be causing this ? I have attached a screen shot of the entire error message thanks

I am having an issue with MCNP6. When I run a simulation an output file is created, but not a mctal file. This behavior started after getting the following error forrtl: severe <157> : program exception -access violation I restarted and didn't get the error, but now not the mctal file and I...

I am trying yo find the flux in a cell which is bounded by two concentric spheres and a cone. When I run the code I get a warning that no cross section tables are called for in this problem and a tally result of zero. The way I defined the surfaces and cells is below if anyone sees where I...

I am modeling a cylindrical source in MCNP6 and would like to use the FMESH tally in cylindrical coordinates. I am looking for the dose to water from the source as a function of radial distance as well as polar angle running from 0 to 180 degrees in the YZ plane not around Z. Is there a way to...

I am using the TMESH tally to try and get the flux of a 1 MeV point source through a spherical MESH a distance of 1cm away from the origin and covering all polar and azimuthal angles. When I try to plot using the RUNTPE function I get the error "no tally no plot". I also do not see tally...

I am attempting to build a sodium iodide detector on MCNP. I am using a disc source and I have been trying to define it in terms of a cell that I placed underneath the detector. When I run it, I get 0's for my counts. This indicates to me that my source isn't hitting my detector. I keep...

I have a problem where I want to model the dose in Gy of a gamma source on a surface as a function of distance. In the papers I have read several different tallys have been used which leaves me a little confused as to the appropriate tally. In the papers the *F4Mesh, F4, F6, and *F8 tallys were...

I am working on a problem determining dose rate using MCNP6. I am following two papers that did the same type of simulation and in them they multiply the tally results by the photon yield also called the photon intensity to detrime the dose rate. My question is where does one find this value ...

Hello, I am working through the MCNP6 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. 1 entries not equal to number of cells = 4." From the IMP card. My entry for...

Hi, I'm simulating in MCNP6 the reaction of proton beam on targets but in the simulation the 96% of the proton are lost for energy cutoff. I don't understand why happen that. I use a tally4 to obtain the reaction in the target, but I suppose the results are wrong because the lost of protons (if...

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 everyone here, I do need your help in this matter, please kindly help me solve this problem. I am new to this forum and now am seeking for help. I am new to MCNP code, and I run MCNP6 code using BURN card and it give me fatal error say "Model required. Cannot use memory reduction...

I was running the MCNP input file. after simulation it shows that 10 particle are lost,( using MCNP6) I do not know where is a problem. How can i get to the solution?

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...