Nuclear Engineering Forum

Post thread
Join other experts in discussion on nuclear engineering. Topics on design and construction and operation of nuclear reactors. Radiation Protections, Radioactive Waste, Design, Reactors as well!
Why is My MCNP Program Outputting Too Small a Value?
  • angfells
  • · Replies 2 ·
Replies
2
Best Dosimeter for 10 ns Pulse Dose Measurement
  • Salman Khan
  • · Replies 9 ·
Replies
9
Reactor refrigeration without external power supply
  • Gruxg
  • · Replies 5 ·
Replies
5
Is there anyone here who uses the Mesh tally in MCNPX?
  • QEROURI ASSIYA
  • · Replies 1 ·
Replies
1
Creating FMESH Tally Over Region - Troubleshooting
  • nrat320
  • · Replies 1 ·
Replies
1
Thermal neutron detection using MCNP
  • Islam Nabil
  • · Replies 3 ·
Replies
3
Product Cross Section For Zn66 in Geant4
  • emilmammadzada
  • · Replies 1 ·
Replies
1
Dose recording on Pocket dosimeter -- Is it accurate for a person's whole body?
  • Salman Khan
  • · Replies 7 ·
Replies
7
Why am I seeing a discrepancy in photon energy when doping ZnS:Ag in MCNP 6.1?
  • nrat320
  • · Replies 6 ·
Replies
6
Codes to calculate diffusion parameters of homogeneous reactor
  • DeltaMed910
  • · Replies 3 ·
Replies
3
MCNP6.2 - SSW and surfaces defining universe boundaries
  • 19matthew89
  • · Replies 1 ·
Replies
1
Plot Mesh Tally with MCNP - Help request from Carla
  • carlepy
  • · Replies 3 ·
Replies
3
MCNP6.2 - Combination of transformations
  • 19matthew89
  • · Replies 1 ·
Replies
1
Efficient MCNP Lattice Source Help: Defining Universes and Tallies in Cell File
  • ethnscot
  • · Replies 2 ·
Replies
2
MCNP: How to display particles outside the source?
  • Salman Khan
  • · Replies 1 ·
Replies
1
MCNP Tally f8 unreliable since neutron transport is nonanalog
  • dando
  • · Replies 3 ·
Replies
3
Is quartarad radex 1503+ or GQ GMC-300E a good geiger counter?
  • Alberto Spain
  • · Replies 1 ·
Replies
1
MCNP: Ctme Card Impacts on Results
  • Salman Khan
  • · Replies 1 ·
Replies
1
Run MCNP 5 input file of certain geometry for flux calculation
  • Salman Khan
  • · Replies 8 ·
Replies
8
How to use mdata file data output by mesh card in mcnp software?
  • chengmo
  • · Replies 1 ·
Replies
1
What is a unit of time listed in MCNP6 PTRAC output file?
  • jj8bmk
  • · Replies 1 ·
Replies
1
MCNP: Measure Dosage in Concrete with tmesh Cards
  • chengmo
  • · Replies 1 ·
Replies
1
MCNP6.2 - ENDF/B reaction numbers for tally multiplier
  • 19matthew89
  • · Replies 10 ·
Replies
10
I use tmesh card to get mdata file, what tool should I use to draw？
  • chengmo
  • · Replies 1 ·
Replies
1
Basic FMESH question: "Jmesh keyword missing error"
  • AlexFi
  • · Replies 4 ·
Replies
4
Fluka user routines -- Questions about Input and Output files
  • emilmammadzada
  • · Replies 1 ·
Replies
1
MCNP6 BURN -- Testing about burnable poison depletion
  • goyu96
  • · Replies 1 ·
Replies
1
MCNP terminology question -- Effective Full Power Days (EFPDs)
  • Rofida
  • · Replies 1 ·
Replies
1
SSW of a translated cell (via TRCL)?
  • 19matthew89
  • · Replies 1 ·
Replies
1
How Does MCNP Macro Definition Modeling Work?
  • chengmo
  • · Replies 1 ·
Replies
1
Can MCNP solve the geometric coincidence issue with a semi-cylinder and cuboid?
  • chengmo
  • · Replies 6 ·
Replies
6
Why is my SSW/SSR code not producing readable output from a KCODE run?
  • 19matthew89
  • · Replies 3 ·
Replies
3
Help debugging MCNP code - particle lost and zero latice element found
  • AlexFi
  • · Replies 3 ·
Replies
3
Why Do MCNP5 Statistical Tests Fail Despite Increasing Particle Numbers?
  • Van525
  • · Replies 6 ·
Replies
6
Is There an Issue with My MCNP Macro Definition Modeling?
  • chengmo
  • · Replies 1 ·
Replies
1
I have a problem on my MCNP5 cross-section library
  • Islam Nabil
  • · Replies 1 ·
Replies
1
Absorbed dose: organ dose manual calculation
  • emilmammadzada
  • · Replies 1 ·
Replies
1
MCNP6.2 - SDEF on a RCC surface
  • 19matthew89
  • · Replies 1 ·
Replies
1
Need a 3d scatter plot with connected lines
  • Islam Nabil
  • · Replies 1 ·
Replies
1
14MeV neutrons in MCNP interact with carbon without producing alpha
  • jianggong
  • · Replies 5 ·
Replies
5
Radionuclides used in nuclear medicine
  • emilmammadzada
  • · Replies 3 ·
Replies
3
I need an MCNP simulated APR1400 input file
  • Islam Nabil
  • · Replies 10 ·
Replies
10
How can I find the stopping power for a single energy value in FLUKA?
  • emilmammadzada
  • · Replies 1 ·
Replies
1
MCNP FMESH for Plotting power distribution
  • AlexFi
  • · Replies 15 ·
Replies
15
MCNP6.2 - Source particle started in a cell of zero importance
  • 19matthew89
  • · Replies 7 ·
Replies
7
Jules Horowitz Reactor (JHR) Material Test Reactor
  • Astronuc
  • · Replies 0 ·
Replies
0
What is the current status of Vogtle Unit 3?
  • gmax137
  • · Replies 1 ·
Replies
1
MCNP6.2 - SDEF position related keywords
  • 19matthew89
  • · Replies 3 ·
Replies
3
MCNP lattice of the fuel assembly input file?
  • Islam Nabil
  • · Replies 10 ·
Replies
10
Calculation of activation with mesh tally in MCNPX 2.6
  • Gogy
  • · Replies 5 ·
Replies
5
I get this error when I try to run the code MCNPX
  • emilmammadzada
  • · Replies 5 ·
Replies
5
MCNP Surface Tallies: F1 & F2 on Infinite Cylinders & Planes
  • 19matthew89
  • · Replies 5 ·
Replies
5
MCNP6.2 - Are results of FMESH tallies already divided by cell volume?
  • 19matthew89
  • · Replies 2 ·
Replies
2
Hysplit.v5.2.3_UbuntuOS20.04.4LTS_public compiling erorr
  • emilmammadzada
  • · Replies 7 ·
Replies
7
MCNP6.2: F6:N and F6:P tallies vs +F6 tallies?
  • 19matthew89
  • · Replies 3 ·
Replies
3
Optimizing Filter Thickness in MCNP5 for Phosphorus 32 Bremsstrahlung Reduction
  • Addali sabah
  • · Replies 6 ·
Replies
6
Does Jimmy Carter's long life prove some exposure to radiation is OK?
  • swampwiz
  • · Replies 1 ·
Replies
1
Why Does MCNP Delete Surfaces in Hexagonal Fuel Element Simulations?
  • AlexFi
  • · Replies 5 ·
Replies
5
If MCNP tallies are normalized, shouldn't they be <1?
  • 19matthew89
  • · Replies 5 ·
Replies
5
MCNP - z-coordinates of cylindrical mesh >0?
  • 19matthew89
  • · Replies 2 ·
Replies
2
FRENDY nuclear code -- Where to download it? (for ubuntu 22.04)
  • emilmammadzada
  • · Replies 6 ·
Replies
6
What is the difference between FMESH and MESH in MCNP?
  • 19matthew89
  • · Replies 1 ·
Replies
1
MCNP - Tallies definition with "<"
  • 19matthew89
  • · Replies 2 ·
Replies
2
Simulate3: Modeling 3-Batch Equilibrium Cycle for PWR
  • tonidous
  • · Replies 4 ·
Replies
4
Major Nuclear Fusion Milestone: Ignition Achieved
  • phyzguy
  • · Replies 34 ·
2
Replies
34
Optimize Filter Thickness with MCNP5 for Phosphorus 32 Bremsstrahlung
  • Addali sabah
  • · Replies 6 ·
Replies
6
Skyshine vs Direct Dose in MCNP5
  • Will_007
  • · Replies 4 ·
Replies
4
FRENSIE installing and compiling error
  • emilmammadzada
  • · Replies 6 ·
Replies
6
Running the tests njoy16 in ubuntu terminal
  • emilmammadzada
  • · Replies 15 ·
Replies
15
NuScale Modular Reactor gets NRC certification
  • gleem
  • · Replies 1 ·
Replies
1
Stopping power and projected range for FLUKA code
  • emilmammadzada
  • · Replies 1 ·
Replies
1
Where Are the Other Detectors for F5 Tally in MCNP?
  • Will_007
  • · Replies 4 ·
Replies
4
Modeling and Simulation in Nuclear Energy
  • Astronuc
  • · Replies 1 ·
Replies
1
Troubleshooting MCNP k_eff for Space Reactor Core: Tips and Tricks"
  • AlexFi
  • · Replies 11 ·
Replies
11
Why Don't GM Counters Detect Alpha Particles?
  • Vilius
  • · Replies 7 ·
Replies
7
How to Fix OpenMc Installing Error in Terminal?
  • emilmammadzada
  • · Replies 2 ·
Replies
2
Big announcement about fusion energy coming soon (Dec-2022)
  • swampwiz
  • · Replies 129 ·
5
Replies
129
Did Installing ENDFVII Fix the MCNP5 Error in Routine IMCN?
  • Limitzone
  • · Replies 6 ·
Replies
6
Searching for a single and complete resource about nuclear fuels
  • mark_bose
  • · Replies 8 ·
Replies
8
PTRAC File - MCNP - Multi-core computing
  • Luizpo
  • · Replies 1 ·
Replies
1
How do I install FLUKA on Windows 10 using WSL and Ubuntu 20.04?
  • emilmammadzada
  • · Replies 3 ·
Replies
3
Particle transporting (FLIBE) -Geant4
  • emilmammadzada
  • · Replies 1 ·
Replies
1
Fixing MCNPX Fatal Error: Too Many Numbers First Entry
  • emilmammadzada
  • · Replies 19 ·
Replies
19
MCNP6.2 BURN Problem Uranium Dioxide 4.2% Enrichment
  • mhovi
  • · Replies 4 ·
Replies
4
Hybrid/variable spectrum reactor
  • armoredchestnut
  • · Replies 4 ·
Replies
4
Why do I always get 0 values in the radiation part? (MCNP5)
  • emilmammadzada
  • · Replies 58 ·
2
Replies
58
Particles transporting in MCNP5
  • emilmammadzada
  • · Replies 11 ·
Replies
11
Cylindrical and Symmetrical Nuclear Fusion Reactor
  • MartinG
  • · Replies 19 ·
Replies
19
How to Simulate Charged-Particle Energy-Loss Straggling in MCNP5?
  • emilmammadzada
  • · Replies 6 ·
Replies
6
Nuclear fusion and anti-nuclear technology
  • thewowsignal
  • · Replies 8 ·
Replies
8
Fusion catalized from graphite?
  • CCatalyst
  • · Replies 16 ·
Replies
16
Error: no plot because it would be empty, using mcnpx
  • hu94mcnpx
  • · Replies 3 ·
Replies
3
Fusion tritium problem solution?
  • artis
  • · Replies 9 ·
Replies
9
MCNP6 with mpi failed with signal 11 (Segmentation fault)
  • Albert ZHANG
  • · Replies 3 ·
Replies
3
[MCNP] Lost too much Keff with Burnup card
  • lee6853
  • · Replies 5 ·
Replies
5
Would wearing sunscreen help during a nuclear disaster/attack?
  • MELJA
  • · Replies 16 ·
Replies
16
Small yield of tactical nuclear weapons
  • Thecla
  • · Replies 2 ·
Replies
2
Does radon remain inside solid radium?
  • Philip Koeck
  • · Replies 9 ·
Replies
9
Nuclear Energy reading material
  • DTMsurf
  • · Replies 9 ·
Replies
9
New approach to stabilize fusion plasmas
  • Tom.G
  • · Replies 4 ·
Replies
4