What is Mcnp5: Definition and 41 Discussions

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  1. A

    How to check the isotropy of the source 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.
  2. V

    Why Do MCNP5 Statistical Tests Fail Despite Increasing Particle Numbers?

    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.
  3. Islam Nabil

    I have a problem on my MCNP5 cross-section library

    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...
  4. A

    Optimizing Filter Thickness in MCNP5 for Phosphorus 32 Bremsstrahlung Reduction

    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...
  5. A

    Optimize Filter Thickness with MCNP5 for Phosphorus 32 Bremsstrahlung

    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...
  6. W

    Skyshine vs Direct Dose in MCNP5

    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...
  7. W

    Where Are the Other Detectors for F5 Tally in MCNP?

    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...
  8. emilmammadzada

    Fixing MCNPX Fatal Error: Too Many Numbers First Entry

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

    Why do I always get 0 values in the radiation part? (MCNP5)

    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...
  10. emilmammadzada

    MCNP5 stopping power for input file

    Where can I find stopping power input file examples for mcnp5?
  11. emilmammadzada

    Particles transporting in MCNP5

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

    How to Simulate Charged-Particle Energy-Loss Straggling 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?
  13. nuclearsneke

    MCNP5 tallies conversion and MCNPX

    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.
  14. nuclearsneke

    MCNP5 or MCNPX, parrallel beam

    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?
  15. L

    Understanding IMP:N in MCNP5 for Beginners

    I knew MCNP at 3 month ago but I do not know Why followed IMP:N (or IMP:P) IMP:N 1 2 2M or IMP:N 1 3 2M. I do not understand Why 2 (or 3) behind 1. Please, Someone indicate an answer physic, thank
  16. bozinion

    How Do You Upgrade the MCNP5 Library?

    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.
  17. Kashif

    MCNP5, dose calculation, subcritical system

    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...
  18. W

    MCNP5 vs MCNP6.2 syntax difference

    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...
  19. Al-mutawakel

    Could a nuclear battery (RTG) be simulated by MCNP5?

    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
  20. L

    MCNP source definition question

    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...
  21. W

    A problem about MCNP5 with mpi

    when I set nps=100000000 and 35 threads , there is a problem that cpu are always 100% but the process of mcnp stays at 20000000 ; But if I set nps=50000000 and 35 threads , there will be normal. Anyone knows why?
  22. Kirk Truax

    Any tips for debugging MCNP geometry errors?

    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?
  23. K

    [MCNP5] Error when use Fm, De/Dm with Fmesh

    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...
  24. D

    MCNP VisEd Plot tracks :Error->File xsdir not available

    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...
  25. D

    MCNP5 error: Source distribution in 2 volumetric cells (help please)

    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...
  26. Kashif

    MCNP5: Pseudolevels Explained for O16(n,p)N16 Reaction

    Hi ! can anyone explain what are "pseudolevels" in MCNP5 . ? i want to calculate total number of protrons produce due to O16(n,p)N16 reacctions in core. whether i should use 103 or 203 as reaction number in FM tally ?
  27. G

    Guidance or tutorial of applying pseudo-material method

    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...
  28. J

    Why Does MCNP5 Behave Differently Across Operating Systems?

    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...
  29. J

    MCNP5 Installation Help for Scientific Linux & CentOS | Rocks Cluster Guide

    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...
  30. R

    Solving the 10 Particle Loss Problem After MCNP5 Run

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

    Performing statistical checks in MCNP5

    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
  32. AlbeFerS

    Nuclear data libraries in ACE format for MCNP5

    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...
  33. Rosenti 87

    MCNP5 vs MCNPX: What's the Difference?

    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
  34. D

    Anyone knows how to execute burnup process in MCNP5?

    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
  35. AlbeFerS

    Finding UO2 Enrichment Value for MCNP5 Input File

    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... )
  36. zaidtaher

    Power Peaking Factor usnig MCNP5

    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.
  37. D

    Shielding calculation with MCNP5

    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...
  38. G

    Why Does Changing the Splitting Mesh in MCNP5 Affect Histories and File Outputs?

    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...
  39. Y

    Help with MCNP5 Output Analysis

    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.
  40. D

    How to get the scattered dose from concrete using MCNP5?

    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...