Mcnp Definition and 207 Threads

Hello! I am really glad to find this forum! I was wondering if anyone has a MATLAB code to parse the MCNP output file MCTAL? I did some Lattice simulation but it seemed hard to read the MCTAL file directly.. Best regards

Hi , This is my first post in this forum, I am new and happy to be in this forum :) My question is, during the calculation of neutron and photon of a single-point reactor core, does MCNP5 taking into account the gamma decay? because during fission process, fission product can emit gamma. Does...

In PWR, fast neutron produced from fission in fuel has been moderated into thermal neutron by the a series of collisiion with coolant,i.e. H2O. So the multi-group neutron flux in coolant and fuel pin has much diffenrce, e.g. the relative higher fast neutron in fuel pin and relative higher...

Hi I am trying to run a fast reactor model but I cannot figure out what is happening because I get the error: run terminated because 10 particles got lost. I am using the sdef card (with and without parameters) and the model is not running. Hope you can help me to fix any mistake I did. The...

Hi everyone Does anyone help about unit of below MCNP line for neutron? Is it "neutron*barn/cm^2" right ? fm14 (1 108 102) And also what's the unit difference between "fm14 (1 108 -6 -8)" and "fm14 (1 108 -6 )" ? Many thanks

Hi, i am looking for some help on MCNP, more precisely mcnpx 2.7 for neutron simulation. I created a model of semi opened detector with a various number of 1 inch He3 (here 16) and i only obtain 6-7% efficiency. the fact is, I've others technicals notes about same "types" of detector, for...

I am testing the MCNPX plugging MCUNED to make calculations with neutron generators. After the compilation many examples to test the installation are provided. But one of them (I attached the code below) starts but it never finish. Just keeps in the first rendezvous. I first though in a problem...

HI, i´m trying to build a lattice in mcnp. Actually it works but there are still lots of red lines in the plot. Maybe someone can help me. The Lattice code looks like that : lattice 1 0 2 $ outer space 2 3 -0.00126 -2 #4 #39 #41 #42 #43 #44 #45 #46 #47 $ universe 4 2 -8.4 -1 $ ball to enable...

Hi, I'm working on a MCNP simulation where I have to use F6 tallies. According to the manual: "In the F6 and +F6 tallies, material density is available for the chosen cells, and normalization is MeV/gm/source-particle." To which source-particles is this value normalized: the source-particles...

Hello there, I am using mcnp6 to simulate a tokamak. I am interested in the energy deposition in the blanket and I am using a fmesh4 and the tally multiplier fm4 as follow: fmesh4:n ORIGIN=0 -24.2 -50 OUT=CF imesh=35.2 iints=352 jmesh=24.2 jints=484 kmesh=50...

I'm currently working on a project using mcnp to model a HPGe detector with a Co-60 source. I have defined my cells and got my geometry spot on but the project requires me to move the source around about 50 different positions relative to the detector. I was wondering if there was any way to...

I don't understand about variance reduction technique. Some one explain PHYS:N, PHYS:P and cutoff. when do I use them? Thank you!

Hello I am a lower-intermediate user of MCNP and I do not know how to obtain the diffusion coefficient (or maybe the angle of scattering) using tallies. I also have read a paper (Multigroup Scattering Matrix Generation Method using Weight-to-Flux Ratio Based on a Continous Energy Monte Carlo...

Dear all, I have a problem related to SSR card. I created a WSSA file (>1GB) from the input file by SSW card. I change the file name to RSSA in order to use for SSR card. When I run SSR card, I got the the error: bad trouble in subroutine issrc of imcn unexpected eof on file rssa How to fix...

I am currently using KCode and modeling a fuel array (Lattice), reflector array (Lattice), and a Stainless steel reflector (Cell filled with Stainless steel, simple geometry). The cells are situated right next to one another, such as 3 squares of similar sizes which make a rectangle. All...

i have a simulation problem about a fuel assembly, after running this warring pop up:- " non-lattice cell in lattice universe "; and visual editor crash with one warning message " warning. 2 surfaces were deleted for being the same as others." So, what may be the problem with the input?

Hi, I need assistance in performing statistical checks in MCNP5 i.e print table 160. I am not sure where the PRINT card should be placed and the format of it. I am using F4mesh tallies

Homework Statement the radius from the symmetry to center of the plasma is about 6.2 metres and the minor radius is 2 metres Homework Equations Can you guys help me to make the plasma geometry for MCNP? The Attempt at a Solution the softcode of plasma geometry

I just started learning MCNP for research reactor analysis. Can anybody advise where and how I could easily get material to read and practice

How to use HEX card in MCNP input, i couldn't figure out its definition?

I am working on an input file in MCNPX/6 that uses a CT scan lattice geometry. I want to specify a small source in a large universe (lung). Right now I have a source uniformly distributed through the universe. The existing documentation is vague on this topic. Is it possible to contain the...

Hi there, I´m just finishing an input file for MCNP5 and I can´t find a value of density for UO2 enriched to 3,25% - 3,6%. Does anyone know it or know where I can find it? Thanks in advance! (P.D.: wikipedia is not my friend... )

I had some problems finding out an equivalent way to describe a particular gama source.We can get the original way describing the particular gama source in this file: "1.txt", as well as its model in this file:"model.jpg",with this describing way,we can get the distribution of gama flux like...

Hello, I am a graduate student attempting to run evaluate the depletion of a ceramic film attached to the moderator-side of the fuel clad. I am having some issues with my MCNP syntax/code and I was wondering if one could assist. My input file is attached. I am not looking for someone to...

Hi, I have written a source subroutine and I am trying to link it with mcnp6. When I run my input file it says you need a source subroutine. My input file and the source subroutine(written in fortran) are in the same directory. Where do I have to keep the subroutine to link with mcnp? Any help...

I want to know the best tally for MCNP Flux (avg and for each assembly) Please help

Hello to everybody, I need some explanation on how to use SDEF variables to define correctly a planar rectangular source. Let say this source is emitting in all direction but I am interested in a side of the source surface where a point detector is located. I used in my example VEC (VEC =001)...

Assume no boron and all control rods out, so the core is super-critical. if KCODE mode is used, and F4 card is for tally the neutron for each assamblies. Can the results represent the power distribution of the core, whether the multiply-factor can affect. the results i get (power...

Hello Every body, I hope all is well. I have a problem with MCNP code for 1/8 AP1000, so please can anybody help me ? Regrads

mcnp error "the new source has overrun the old source" I am a beginner with MCNP. This error really confused me. Who knows what this error means? Thank you.

Hello,everyone.Lastly,I use MCNP to claculate the dose distribution of CT phantom,but when I run my code,I get a message that says ' bad trouble in subroutine newcel of mcrun source particle no. 18859 starting random...

fatal error. *** cross-section tables are missing from xsdir 10001.70m 10002.70m 10003.70m ... bad trouble in subroutine ixsdir of imcn cannot continue with missing cross-section table(s). is the message I am getting. I have created a link using ln -s...

I have a question about diffrence between ways of defining a certain plate like (3^1/2)x+y+17.3=0 to MCNP code .for example I know we can define it to MCNP with 15 p 3 2 0 6 but I couldn't understand this method 15 10 0 5 8.66 0 10 0 1

How to learn MCNP basically?

Hi everyone, I’m a student and in my master degree, I’m working in a project about a bunker for industrial radiography of steel, and I need find some input of mcnp about this matter. Somebody can help me? Thanks a lot

The only thing I know how on the basis of nuclear reactor design is how to run kcode in MCNP and see if my theoretical reactor is critical. How would I be able to calculate how the fuel is burning in my reactor over some period of time, and change my core composition accordingly?

hello, I am new to MCNP, could somebody tell me how to use imp:n, what is imp:n=0 means, if neutron importance is 0 in one cell, why the F4 tally is 0 in this cell? how about imp:n=1 or some large number? Thanks for all.

I am a beginner user of MCNP and is still learning how to use it. When I run my program, I get a message that says 'bad trouble in subroutine main of mcnp'. What exactly does that mean and how can i correct it. Thanks.

I've been having difficulty in successfully compiling the MCNP executable with the MPI/MPICH libraries on Scientific Linux 6.x (whatever the latest version is). I'm currently using MCNP5, and the latest distribution DVD. The compiler is GCC, and the MPI is MPICH 1.4.x (the version immediately...

Hi I have problems in mcnp input of voxels. I define a voxel arrays and their universes,for example universe1 is defined as: 2001 43 -1.030 -70 u=1 vol=6.795352 but in runing I have below error: fatal error.the surface type is not recognized: -1.03 While 1.03 is the density of...

I have just started to use MCNP code for nuclear reactor modelling. I would like to practice using a visual editor for input file. Can anyone tell me how can I get a free version of visual editor or other similar editor compatible with MCNP?

I'm curious if anyone has any opinions in the results from MCNP vs Geant4 for shielding design. I like that Geant4 uses a more modern syntax, but iv'e also heard that MCNP gives better results.

Hello all. I'm am a first time poster but a long time visitor. I am having a little trouble that I was hoping someone far wiser and more knowledgeable than myself might be able to help with. I've been using MCNP to investigate criticality in a simple geometry consisting of a central natural...

hi there I wnat to calculate the group constants for a FA(fuel assembly) using MCNP (similar to lattice calculations). How can I do it? Please lead me. Thanks alot!

Hello all, I'm a senior Nuclear Engineering student. This semester I've been working with DANTSYS and MCNP (more like failing to learn properly). I was wondering if anyone had some experience with either program and would be willing to allow me to ask them the occasional question. I think...

hello, i am using mcnp for a reactor kinetic study.the only problem is, to establish the model i need to calculate the temperature coefficients, and in order to do that i need to calculate k in different temperatures and therefor need a mcnp library that contains cross-sections for a multitude...

Hey, I'm modeling the criticality of a core for a university project using MCNP4C. I've run the core on the appropriate KCODE parameters, specifying the source using the KSRC card, and I was looking to cut down on the computing time using the SRCTP files. My problem is that I can't find...

Hi there I have a question about FMn card in mcnp code. there is a parameter (is named C) in front of FMn card, I can not understand to calculate the value of that parameter. please help me. thanks alot.

Hi there any body know that mcnpx has abality to do time dependent calculations? please lead me quickly, I need it. regards

Hi there, I want to know that how can involve the temperature in mcnp code. for example; the library endf7 for mcnp has five certain temperature:300 kelvin, 600 kelvin, 900 kelvin, 1200 kelvin & 1500 kelvin. if I want to calculate flux distribution in 330 kelvin for a research reactor, how can...