Mcnp Definition and 204 Threads

Hi all, I'm modeling an HPGe detector and want to determine the amount of downscatter that contributes to a Ba-133 spectra. I'm using a PTRAC card to filter scattering events that occur in my Ta4C3 shield that contribute to tally 1 (surface fluence across the front face of the detector). I want...

Hi all, I'm attempting to simulate a very specific setup in MCNP. I want to know the fraction of particles contributing to a surface tally that previously interacted (scattered) with a specified cell. I'm currently doing this via the PTRAC card as well as FILTER. Currently, my PTRAC is...

Hello, I'm attempting to model a 300x300x300 cm room in MCNP with a doorway and walk-in section, but I'm struggling with some of the cell definitions, particularly in the Z plane. I've attached the input deck below. It is fairly short, so it's probably going to be a relatively quick fix. Any...

Hi all, I'm working with an MCNP deck with 4 embedded universes and am having trouble with a srctp error when attempting to run the deck. I had the simplified version with the larger two universes running fine, but when I try to add in the third and fourth layers I experience this error. Would...

Dear all, I want to simulate an X-ray tube and check the dose rate in the room. My problem is in the data card. I don't understand how to write sdef, tally, and others. I simulate 85kV with 25 mAs and the source position is at 0 0 0. Regards, anggi

c *************** BLOCK 2: SURFACE CARDS ************** 10 PZ 100 110 SO 110

Hi folks, I'm attempting to define an SDEF cone source but am getting tripped up in the SI/SP/SB distributions. I need all particles to be generated at angles within the cone, with all angles in the cone having equal probability. I feel like this should be relatively simple to define since the...

The mcnp6 code I wrote contains a "zero lattice element hit" error, but I am unable to identify the problem. My code has been uploaded, and I kindly ask someone with the ability to assist me. Thank you

I'm trying to resolve two separate fatal errors in my MCNP deck. One is claiming that I'm mixing atom and weight fractions within a single material card, which I'm clearly not if you take a peek at the material definitions. The other is saying that I have an odd number of entries, indicating...

I've got a small geometry related error in my MCNP input deck, corresponding to cell #14 (the outer edge of my detector model). This should be a quick fix, but I'm running into issues defining that particular union of surfaces. Any assistance is appreciated.

Dear Experts I would like to create a small modular reactor-like geometry for Serpent code (example: ABV reactor). How can I add a control rod to the center like in the picture below? Which code scripts should I add in the geometry part. And how can I do this in Hexagonal laticce?

Hello everyone, I am trying to calculate the atomic density (in atom/barn-cm) for borated water (750 ppm) at 300 K in the active core of a reactor. Could anyone guide me on the correct approach or provide a reference formula for this calculation? I came across the following atomic densities (in...

I'm working on a very basic setup to activate cobalt 59 with neutrons emitted from californium 252. MCNP is giving me the following warnings: warning. photonuclear physics may be needed (phys:p)...

Hi, everyone. I am using MCNP to simulate an incident electron beam hitting a tungsten target and obtaining the bremsstrahlung spectra in natural element samples placed behind the initial setup. I want to use the Subroutine Source option in MCNP to record all the directions, weights, energies...

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