Mcnp Definition and 207 Threads

I would like some help to configure the SDEF card for a lattice cell. I am configuring it but I am getting the error "fatal error. Cell 889 in SDEF CEL path not at lev=0" I have tried several configurations but I am not having any success, please help me. I want to set cell 132 as my source...

Hello everyone, I am facing difficulties while trying to calculate the dose in the tibia due to brachytherapy in an MCNP simulation. We are working with the radionuclide Ho-166, and therefore, we need to account for both photon and electron contributions to the deposited dose. Initially, I...

I'm trying to rotate an RPP 45 degrees around the y axis (BUT NOT THE ORIGIN Y AXIS, rather the y axis at x=a, z=b). Is there a way to do this in MCNP? I've tried every single possible combination of angles and inputs to no avail. Again, I have an RPP that is not centered at the origin and I...

Hi all, I'm working on building a fuel compact using hexagonal lattice cells, but I'm running into trouble with the RHP/HEX definition. The deck is error free, but for one reason or another the lattice cell isn't replicating along the Z axis.. only the X and Y, or at least this is what I'm...

Hello, I'm working on BURN UP analysis on a NuScale PWR with variations on U235 enrichment and the number of Gd2O3 burnable poisons rods used. I use KCODE and KSRC card as attached below. I'm having trouble with the error message on the output file. At first, I fixed that by increasing the...

I'm looking for someone to help troubleshoot my MCNP input deck. I'm getting a geometry related error most likely due to some sort of surface overlap. Haven't been able to identify the issue myself, so I'd appreciate a secondary check. Thanks!

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'm wondering how exactly MCNP calculates an f6 tally? I'm trying to compare a theoretical result with an MCNP f6 tally (MeV/g). I have an initial energy spectrum and a thin layer of lead that attenuates the x-rays. Using the attenuation coefficient at each energy (bin width of 0.5 kev from...

Hello All, I'm trying to simulate in MCNP the energy response of a PIN diode. To do this, I have modelled a "slab" of silicon in an epoxy case at 2cm away from the source and with the F8 tally set to 25keV bin increments to 1MeV, I do as follows: Set the source energy to 33keV Run the...

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...

Hello everyone. I am simulating a Cesium-137 source with an energy of 0.662 MeV and an activity of 225 mCi. When I use the "T: tally time bins" card, for example: F24:P 1 E24 20 T24 0 1000 25I 3600 196I 200600 I understand that I am asking the program to give me the average flux in this cell...

Hello, I've been running into some frustrating issues with my MCNP deck. Photons are getting lost which is terminating the run file prematurely. When consulting the output file there seems to be some sort of geometry issue, but there are no fatal errors that I can see so I'm lost on how to...

Hi everyone. is it possible to simulate low energy photon in wavelength range (300 nm to 1000 nm) in MCNP. If not possible in mcnp please suggest any other code to simulate it. thanks

Hello, I'm getting an odd fatal error that seems to be triggered due to my material card. Below is the material card for my input deck and the associated error. Appreciate any help that can be given. M1 6012.80c -0.000124 7014.80c -0.755267 8016.80c -0.231781 18040.80c -0.012827 $...

wanted to learn the MCNP for my research, but need some help, don't know nothing about that. There some step by step tutorial in youtube or website focus on that. Wanna make one analysis in one HTGR reactor in the case

Hi everyone, I'm trying to compare 3 different fuels and MCNP and I want to recover the burnup of each. When I do that however, I get identical numbers for burnup, which doesn't make sense to me, as they have different materials (LEU vs LEU+ vs a thorium-based fuel). Does anyone know what...

Hello everyone, I am currently working on a Monte Carlo N-Particle (MCNP) simulation and have encountered several issues that I hope someone here can help me resolve. My project involves simulating a brachytherapy treatment room, and I am struggling with defining the cells, surfaces, and...

Hi everyone, I'm a newbie to MCNP, I'm trying to calculate burnup for this pellet I include here in a PWR in an infinitely repeated geometry, but it seems to be failing for some reason. I get the error message: ctm = 0.00 nrn = 0 dump 1 on file runtpp.h5 nps =...

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...

I need help to develop an mcnp geometry for input and output of Percent depth dose.

Dear experts, I would like to get help from you on something. I want to design a vver -1200 core in the serpent nuclear code, but I am getting errors in the geometry I defined below. I would like to get help from you on how to fix these errors.I seem to be getting errors in cell definitions and...

Hello everyone , in my mcnp coding for finding neutron spectroscopy I used F2 tally across a surface. Is it correct or I should use f4 tally? Morever I need to transform the flux data into neutron fluence. How can I do that. Here I uploaded my code. Though my data from codes is way more...

Hello everyone, currently I am doing a neutron spectroscopy experiments. I am doing it with the MCNP code. I designed my Geometry there, but facing problems in data cards, is there anyone who can help me in this sector?

Can I input the X-Ray tube voltage in MCNP source specification?

Here, SP stands for source probability. But probability needs to be normalized. Here values in SP3, SP4 are larger than 1, It means that SP is not ordinary probability here. But what actually SP represent here?

What is meant by SP1,SI1 and SB here? I actually can't get the physical significance. And What is the physical significance of WGT here? Sorry for my this kind of questions. I am novice in MCNP.

Hello everyone! I need to make sure that my source is isotropic. How can I check that? I have point source pos -11 0 0 erg=d1 with Maxwellian spectrum of energy and some surfaces through which neutron flux passes.

Hello, I've been working with MCNP on and off for a few years now, but just recently realized that I don't entirely understand how tallies are actually calculated in MCNP, and what they signify. Taking the example of the F2 tally, the user manual (Section 3.3.5.1) states that F2 is the "flux...

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...

Hi everyone, I am using SSW card. Although the manual is very clear about the fact that the cells used in SSW card have to belong to the lowest level, the manual is not that clear the surfaces. Is SSW able to track particles crossing surfaces defining higher level universes? Let's assume this...

Hello, After some time away I've gotten back into MCNP. I've been in the field of Nuclear Engineering for over ten years, but I recently changed jobs and need to use MCNP. I'm trying to get my skills back up, since I haven't been using it as much in my old job. Looking forward to some great...

This is what I hate about MCNP, not a lot of documentation. How do I define all of a universe as a source and a tally? I have a lattice like the below code. How do I get this code to work with tallies for positions 1,2, and 3 in the lattice; and a source for the 2's. I get the error "sampling...

Ctme card is used to limit the running time of the mcnp input file, does it affect the telly result or not ?

Particle display in visual editor of mcnp input file only shows particles inside source, I am interested to see particle path towards tally region. Plz guide thanks

Can any one please explain if I want to run mcnp 5 input file of certain geometry for flux calculation on different surfaces. So far as I know If I increase the NPS (number of particles) it wll give more accurate result but when I increase NPS from 10e9, input file do not run and close within a...

How to use mdata file data output by mesh card in mcnp software? I converted the mdata data using gridconv.exe, but I don't know how to use the data to identify the section graph for the xyz axis? Does anyone know how it's arranged?

hello , I am getting an error that reads as follows: 1 PX 12.5 bad trouble in imcn in routine ckchar bad character, probably a control character, in column 2Cell cards 1 0 1 -2 -7 8 10 -9 imp: n=1 2 0 2 -3 -7 8 10 -9 imp: n=0 3 0 3 -4 -7 8 10 -9 imp: n=1 4 0 4 -5 -7 8 10 -9 imp: n=0 5 0 5 -6 -7...

When using tmesh cards to measure dosage in concrete, can mesh3 cards be used? My friend said it might be a mesh1 card but I can't use it. Can someone guide me? I want to look at examples

Hi everyone, I am trying to evaluate the spectral index of an nonelastic (n,n') reaction. For that I want to set up a tally multiplier on a cell (let's call it cell 10). The reaction is present in the ENDF/B library as MT=4 but I have not seen it in the table of the special reaction numbers...

for MCNP users, i would like to ask about terminologies: if i depleted a fuel assembly under constant power, is the number of days in the out can be used as Effective Full Power Days (EFPDs), or this term has another specific meaning?

How do we solve the geometric coincidence problem? I need a semi-cylinder that fits into the cuboid but if I use the cuboid and the cylinder directly it's geometrically problematic

I use macro definition to model the results can be viewed in vised but vised does not display all the big guy know? Is there something wrong with my modeling?

C 1 1 -19.35 -7 1 -2 2 1 -19.35 -8 3 -4 3 1 -19.35 -5:-6 4 0 5 6 #1 #2 C 1 RCC 0 -10 0 0 -10 10 2 2 RCC 0 -10 0 0 -10 10 5 3 RCC 0 10 0 0 10 10 2 4 RCC 0 10 0 0 10 10 5 5 RPP 2 5 -10 10 0 10 6 RPP -3 0 -15 15 0 10 7 RPP 0 5 -15 -10 0 10 8 RPP 0 5 10 15 0 10 C M1 074184 1

I keep getting particle lost error even though there were no hole in the lattice. Can someone identify any mistake in my code?

14MeV neutrons in MCNP interact with carbon without producing alpha particles and protons, yes Questions about my cross section data? I've tried ENDFB8/B7.1, JEFF3.3, JENDL5, CENDL3.2 without any results, but if you use phys:n model, it looks like alpha particles will be produced, but it doesn't...

Hi everyone, I'd like to know if it is possible to use TR or TRCL to translate an F4 or FMESH4 tally. Let me better explain: I have a cell, centered at 0,0,0 and for this cell I set up a segmentation tally along z-axis with a series of planes orthogonal to the z-axis. Moreover, I have also set...

need a MCNP simulated APR1400 input file? which consists of lattice 16x16 inside a lattice 17x17 ??

Hello I'm trying to use FMESH command to get power distribution of this core geometry. I want to use xyz coordinate in a 1/12 slice of a core so I could use the output of the MCNP sim for a CFD input How should I approach this? Thank you