Mcnp5 Definition and 43 Threads

I am facing a problem after running my input file through MCNP 5 it give me these comment👇👇 Can any one help me to resolve the error? mode n p comment. photonuclear physics may be needed (phys:p). comment. 9 surfaces...

This is my mcnp code: 1 4 -0.7119 1 -2 3 -4 5 -6 fill=1 imp:n=1 26 4 -0.7119 11 -12 13 -14 5 -6 u=12 lat=1 fill=13 imp:n=1 20 4 -0.7119 46 -47 -45 #1 fill=12 imp:n=1 19 0 #20 imp:n=0 3 4 -0.7119 11 -12 13 -14 u=1 lat=1 imp:n=1 fill=-4:4 -4:4 0:0 12 2r 8 9 8 12...

In a study I did in MCNP, I want to create a sliced geometry in a torus structure. But MCNP cannot calculate the volume of the geometry I created. What should I do about this issue? 2 0 -50 -12 13 3 0 #2 12 pz 34.4985 13 pz -34.4985 50...

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 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...

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...

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.

Can anyone tell me how I can solve the problem of non-verification of statistical tests done by MCNP5 (relative error, VOV, figure of Merite, slope). I tried to increase the number of particles generated in order to hope to verify the tests but it did not work.

I have a problem with my MCNP5 cross-section library. Se and Te elements with atomic numbers; 34000 and 52000, respectively, have no neutron cross-section stored, either neutron contentious or neutron other, in the library, and so I can't run any input file that has any material of them because...

I am working on the optimization of the thickness of filters in order to reduce the continuous Bremsstrahlung spectrum emitted by phosphorus 32 measured by the Ge(HP) detector using some MCNP5 code but unfortunately I couldn't find the exact form of declaration of bremsstrahlung in the MCNP...

I am working on optimizing the thickness of the filters in order to reduce the continuous Bremsstrahlung spectrum emitted by phosphorus 32 with a maximum energy of 1.8 MeV measured by the Ge(HP) detector using MCNP5 code but unfortunately I don't haven't found the exact form of declaration of...

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...

mcnp5

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

MCNP Output: energy stopping power range radiation beta**2 density rad/col drange dyield n collision radiation total yield corr mev mev cm2/g mev cm2/g...

Is it possible to transport only photon , electron and neutron in mcnp5?

How can I make a sample code or a simulation about mcnp5 charged-particle energy-loss straggling?How can I find sample examples?

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.

Hi, I am interested in simulation of parallel beams for neutrons and photons (separately of course). Any ideas on how to simulate them in MCNP5 or MCNPX?

hello guys. I would like to ask you if anyone knows how to upgrade the library at mcnp5. I downloaded the last version for mcnp5 but I don't know how to change it. please help. Thank you.

Greetings to all need forum members help regarding MCNP5 college assignment. The task is to calculate dose from a system comprising of a neutron source surrounded by natural uranium. the system is subcritical. The problem is how to calculate dose due to fissions in the uranium because dose from...

So I am getting a fatal error on a tally that contains a universe definition in it. f12:n (10<(u=1)) sd12 1.0 e12 0.1 20.0 the error specifically says "fatal error. invalid universe format in f card bin 1 tally 2." However, this input is able to run on MCNP5 without experiencing...

I am a graduate nuclear engineer(no Master yet), and I ask if nuclear battery(radioisotope thermoelectric generator, that use plutonium-238 as heat source) can be simulate by MCNP5. And is their any card in MCNP5 treat or deal with thermoelectric converions. Thanks

I'm modelling a scenario for the research that I'm working on, and I got the cells and surfaces all mapped out for the environment finally, but now I'm totally stuck on creating a source. I'd like a point source of Californium-252, but after hours of looking, I don't see any out of the 1000...

Hey all, I was wondering if anyone had any good tips on debugging mcnp geometry? I'm an intermediate user working on better understanding the program. Does anyone have any tips or tricks that go beyond simply reading the manual?

Hello everyone, I do need your help in this matter, please kindly help me solve this problem I use MCNP5 and i run a simple example using Fmesh4 and Fm, De / Df but I get an error: " fatal error. no tally associated with response function -1" " fatal error. no tally associated with fm card -1...

Hello, I use free VisEd and I want to plot the collision, the particle's transport and the particle in tally but I have an error. For the source the plot is ok. How can I solve this problem freely? Someone can help me? My input :C Cellules30 50 -1 (-1 3 19 ):(-2 3 ):(-4 3 -6 5 -7 1 2 ) imp:p=1...

Hello, can you help me? In my model, I have to simulate source distribution in two volumetric cells (30 and 31). For example, with a photon source of 356 keV, I have to simulate the distribution according 0.2 (20% in source 30: thyroid without nodule) and 0.8 (80% in source 31: nodule). I use...

Hi, I'm working on a project of coupled simulation involving MCNP and Fluent, and having trouble with how to apply the pseudo-material method in cross section handling. Is there any related material or tutorial? Or, does anyone know the procedure of applying the method and what kind of files the...

I recently installed MCNP5 on rocks cluster running centOS 6.5, with mpich2-1.4.1p. While running simulations, MNCP5 is stuck in loop in 2 of the simulations. When I tried to perform one of the stuck simulations on windows with mpich-1.2.4, it didn't stuck. Can anyone guide me why MNCP5 is...

Hello, Is there anyone who has any experience of installing (compiling) MCNP5 on Scientific Linux or CentOS? Previously I had used MCNP5 on Windows. But Recently I thought of moving to Linux based cluster. But I am having difficulty in compiling MCNP5 on Linux. The environment that I am using...

After running MCNP5, the output run terminated because 10 particles got lost how i can fix this problem ?

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

Hello everyone, I need some continuous-energy neutron cross sections libraries in ACE format for MCNP5 (endf66a, endf66b, endf66c, misc5xs) as well as the continuous-energy photon tables (mcplib, mcplib02, mcplib03, mcplib04). Here: http://www.oecd-nea.org/janisweb/search/endf , we have...

Hello everyone, I need help. Anyone can explain what is the basic differences between MCNP5 and MCNPX? I appreciate every suggestion for help me. Thanks. Rose

Hi for everyone, anyone knows whether it is possible to execute burnup process in MCNP5? I want to get K-inf as fuel depletion goes on. Thanks

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 model the core using MCNP 5, and I want to calculate the PPF for each assembly, and I want a way to do that.

Greetings to all I had some problems using MCNP5 for gamma shielding calculation. The original input file is from a document by George E. Chabot, Jr., PhD, CHP Shielding of Gamma Radiation. The description of the problems are following: 1) Three different shielding materials, namely...

HIya, I am not actually sure why this is happening so was wondering if anyone can help. im running in photon mode and I am trying to generate a weight window. initially i think my splitting mesh was over done a little bit (ok too much say about 15 fine intervals in a 10cm region) and i...

hello.im having project to find neutron flux from cf-252 using mcnp.but I am stuck at the output that i dono have to convert it to become flux.help please.

Hello fellow nuclear engineers and physicists, Does anyone know how to just get the scattered dose from a surface like concrete using MCNP5? I'm doing a project for a shielding class and I need to compare scattered dose and direct (uncollided) dose for a point detector. The photons are coming...