Mcnp Definition and 207 Threads

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...

Cell vol card fatal error appears in this input file. How do I add a vol card to this input file. How can I calculate the vol for each cell. Is there a method?

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

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...

hi dears I am trying to plot my mesh tally with no success. Try it on vised (no tally no plot) and on promt (see photo)... if anyone can give an insight is highly appreciated. tks Carla

Hello everyone! I hope you all doing well :) I am having a trouble with detection the radiation in lattices. i am adding the input and the result file here for makes everything clear, If someone can help me i would really be appreciate! thank you! ☺

I am trying to simulate fission product ejection from thin fissile films in gas filled detectors (fission chamber). Does MCNP 6.2 produce recoil fission products that will be transported through the system? I have enabled "heavy ion physics" (#), tried options 3 and 5 for NCIA on the neutron...

Hi everybody, I am trying to score fission spectrum in MCNP for a kcode calculation. I would like to check at which energy neutrons produced by fission are generated. I have no idea how to perform since tallies are usually volume or material dependent and I just want to build a spectrum...

Hi, I have a problem with MCNP. The problem is : the input file can run correctly on WindowsXP system,but when I run it on a more powerful computer with win10 system , the MCNP showed "chg_mem error". How to solve this problem? can MCNP only operate on winXP or win2000 system?

I was reading papers on neutron flux traps. Some people in the University of Tehran used MCNP to determine effectiveness of different neutron moderators. Context out of the way, my question is regarding the situation. How did researchers paid by an organization like the University of Tehran get...

I am running some depletion calculations and have noticed odd behavior when keff is much greater or lower than one in the number of absorption's taking place on the outside the reactor. MCNP seems to be deleting the neutrons after it has reached the number of fissions required for the power...

Hello, I am getting the following error Unexpected error in file I have gone through the code and noting seems to be amiss. Does anyone have any ideas?

Hallo everybody, I am using a mesh-based-weight window generator. It is clear to me that the coarse mesh must cover the full geometry; my doubt is about the fine mesh. Is it possible to define the fine mesh only in a part of a geometry, for example telling to MCNP to make 5 bins between y1 and...

Hello All, I have yet another MCNP question. I received the following error "geometry error: no intersection found mcnp" when trying to run a a simulation. I looked at the output and according to it I have an infinite volume in cells 14 and 500. I plotted the geometry and don't see how its...