Mcnp Definition and 204 Threads

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

There is an input file for a simple 16 x 16 lattice fuel assembly. I have a message blocking the run of the code; "bad trouble in subroutine newcel of mcrun source particle no 1 random number 6647299061401 zero lattice element hit." What is wrong?

Hi, I have a question concerning surface tallies like F1 and F2. You have to provide a surface for them. Since, surfaces are defined as infinite (infinitely long cylinders, infinitely extended planes) how can you write the surface tally of a cell? What are the actual tally surfaces for F1 anf...

Hi everyone, I am struggling to understand the difference between FMESH and MESH. FMESH is used to create a mesh superimposed with the geometry but...what does MESH? Is it only involved in weight-window generation and not needed for mesh tallies? Thanks in advance for the clarification.

Hi everyone, I am using MCNP6.2 and trying to set up a cylindrical coordinate in a reactor channel. The origin as the midplane of the channel. In my attempt of setting up a cylindrical FMESH with the origin on the z-axis at the bottom of the channel (so z<0) I got this fatal error message...

Hi everyone, In MCNP manual there are often examples of Listing containing examples of tallies which have, in the definition of the cells/surfaces of the tally itself, the "<" symbol. I could not find in the document any reference to the use of logical expression in the definition of tallies...

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...

Hi everyone, I'm really new to MCNP here and I'm "playing" around trying to understand what is going on. I'd like to plot my tallies (F2, F4 and F6). Is there any tool (e.g. python or matlab package) you recommend? I know that the internal plot MCPLOT is available but I'm wondering how you...

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...

Hello Tried to model gas cooled reactor with hexagonal fuel elements. MCNP keep deleting surfaces (If you could, run my input and check the .txto file) so the simulations won't run Any advice?

Hello! I tried modeling a space reactor core with MCNP. I'm pretty sure the geometry and material properties are correct. Got k_eff of 1.4, much higher than 1.003 from the reference. Could anyone spot the mistake in my code? I couldn't figure out anything else

mcnp5

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 worked monte carlo simulation for dose calculations in field radiotherapy (external and brachytherapy) using EGSnrc and MCNP codes.

When I run the application, I get an error message. This message: Fatal error too many numbers first entry. What could be the reason?

Hi there! Me again. I am doing my research about converting HEU research reactor to LEU. I made FA and core finally and started using the burnup card to check changing of Keff and fission products. Well, the thing was only with one-month burnup my Keff was decreased drastically from 1.118 to...

Hi guys! I'm a master's student majoring in nuclear engineering in graduate school. I have a few questions while doing research, so I'm writing this here. My research is simple. We conduct neutron analysis to convert a research reactor using highly enriched uranium into a low enriched uranium...

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...

I know that "mcnpx n=filename" I run the file, but how do I run the entire folder with the entries inside?

Hi all, I am working on some criticality problems using MCNP4C. A key aspect of what I am attempting to do is to use Python to automate the creation and running of MCNP input files. One of the issues so far is that after each command entered into the Command Prompt window, the MCNP...

During a reactor assembly calculation, I need to determine axial and radial flux distribution over the surface. When I use F2 and F4 tally I get some value with unit 1/cm**2 What does the value means, neutron flux is supposed to be in the 10^14 range but output values are 10^2 range. Can anyone...

Hello everybody. I would like to ask a question; if I change the dimensions and densities and the material number identifier ZAID to a specific temperature. Does MCNP change automatically the volume of the cells? or I have to change the volume of each cell manually and indicate it in the cell...

Hi there I want to convert the flux (F4:N tally) from mcnp units to cm-2s-1 units. How to do that? Also I have some bug in MCNPX: while running the file, I get an error like " >bad trouble in imcn in routine xin >Cannot find bertin " How to solve it? Database for MCNP5-MCNPX got installed already.

Homework Statement:: I go back to the line to finish the previous line in MCNP cell card but I had the error message shown in the photo. Please make a solution to my problem Relevant Equations:: c ********************* BLOCK 1: cartes des cellules **************** 1 2 -1.184 -40 #3 #19 #18...

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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...

Hi there, I have a very simple question about MCNP (6.2 to be precise) ... maybe someone here might enlighten me ... Based on the (more than simple) MCNP input file below, which describes a sphere with R=200cm, filled with air and a point source in the center. There's a single ring tally on...

What are the most important fission products should I include/care about that comes out from the MCNP output?

I need help to construct this source on mcnpx. I tried a lot of thing, but nothing worked. The source is a isotropic point source, but the beam is rectangular with dimensions of the same of scoring plane. The source is the standard mammography beam. Please, help. My parameters in use: SDEF...

Hello, I am a student who started studying MCNP. I'm not used to writing in English, so I'd appreciate it if you could understand even if there were grammatical errors in my thread. I want to check the energy of gamma rays from neutrons reacting with matter. So, I wrote this in the content of...

Hi, Is there any tutorial or pdfs that can help me with the MCNP output data? I'm working on the criticality of the SCWR, and I designed the fuel assembly and run it on the MCNP, but I have no idea about data extraction.

I wrote a program to determine the critical mass of uranium oxide with an enrichment of 10%. I got a keff approximately equal to 1 with the selected volume and density (attached a file). Is it possible to somehow run the program without writing the initial density and volume into the conditions...

What does the number 0.62 refer to? Does it the water density at high temperature?

How do I express the temperature in the cell cards of the MCNP?, Say the temperature of the fuel is 500K, how do I write it as the following PWR example?

Hello everyone, trying a run I've got the following error message, and thus no output file since the run stopped by itsefl. Please, has anyone encountered this error and would happen to know how to correct the input file ? (any hint help) Thanks to all of you