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The Nuclear Power Thread
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MCNP and Nuclear Resources
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Fukushima Fukushima Daiichi nuclear disaster discussions
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MCNP6.2 - ENDF/B reaction numbers for tally multiplier
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10
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115
Fukushima Japan Earthquake: nuclear plants Fukushima part 2
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MCNP Problem - Bad character in column 2
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6
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94
How to use mdata file data output by mesh card in mcnp software?
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42
MCNP -- When using tmesh cards to measure dosage in concrete, can mesh3 cards be used?
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0
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49
I use tmesh card to get mdata file, what tool should I use to draw？
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1
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59
Basic FMESH question: "Jmesh keyword missing error"
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4
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360
Are spent nuclear fuel rods radioactive?
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20
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MCNP6 BURN -- Testing about burnable poison depletion
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96
Fluka user routines -- Questions about Input and Output files
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118
MCNP terminology question -- Effective Full Power Days (EFPDs)
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1
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128
Particles produced and energies in the target Geant4
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6
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251
MCNP geometric coincidence
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6
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215
SSW of a translated cell (via TRCL)?
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0
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191
SSW/SSR with a KCODE run
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3
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251
Help debugging MCNP code - particle lost and zero latice element found
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3
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230
Statistical checks MCNP5
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6
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210
MCNP macro definition modeling
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167
MCNP macro definition modeling
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146
I have a problem on my MCNP5 cross-section library
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1
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216
Absorbed dose: organ dose manual calculation
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1
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226
MCNP6.2 - SDEF on a RCC surface
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249
Need a 3d scatter plot with connected lines
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247
14MeV neutrons in MCNP interact with carbon without producing alpha
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5
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250
MCNP6.2 - translate a Tally
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2
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285
Radionuclides used in nuclear medicine
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3
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249
I need an MCNP simulated APR1400 input file
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10
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321
How can I find the stopping power for a single energy value in FLUKA?
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2
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350
MCNP FMESH for Plotting power distribution
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15
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620
MCNP6.2 - Source particle started in a cell of zero importance
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7
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398
Jules Horowitz Reactor (JHR) Material Test Reactor
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1
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254
Absorbed dose calculating in Fluka
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1
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291
Vogtle 3 Initial Criticality
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1
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315
Depth dose calculation in Geant4
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1
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275
MCNP6.2 - SDEF position related keywords
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3
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429
MCNP lattice of the fuel assembly input file?
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10
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409
Calculation of activation with mesh tally in MCNPX 2.6
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5
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421
I get this error when I try to run the code MCNPX
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5
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393
MCNP Surface tallies
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5
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496
MCNP6.2 - Are results of FMESH tallies already divided by cell volume?
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2
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341
How can I calculate the stopping power for aluminum in Fluka?
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1
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381
TRIPOLI-4 version 9S How to installing on Unix system for Ubuntu22.04
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1
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327
hysplit.v5.2.3_UbuntuOS20.04.4LTS_public compiling erorr
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7
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409
MCNP6.2: F6:N and F6:P tallies vs +F6 tallies?
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3
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413
MCNP5 declaration
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6
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414
Does Jimmy Carter's long life prove some exposure to radiation is OK?
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1
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380
MNCP deleting input surfaces
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5
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571
If MCNP tallies are normalized, shouldn't they be <1?
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5
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470
MCNP - z-coordinates of cylindrical mesh >0?
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2
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433
FRENDY nuclear code -- Where to download it? (for ubuntu 22.04)
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6
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479
MCNP - FMESH vs. MESH cards
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1
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455
MCNP - Tallies definition with "<"
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2
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462
MCNP6.2 - How to plot normalized tallies with MCNP6.2?
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2
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462
Casmo4e/simulate
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4
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496
Major Nuclear Fusion Milestone: Ignition Achieved
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34
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MCNP5 declaration
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6
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468
Skyshine vs Direct Dose in MCNP5
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4
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488
FRENSIE installing and compiling error
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6
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449
Running the tests njoy16 in ubuntu terminal
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15
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554
NuScale Modular Reactor gets NRC certification
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1
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557
Stopping power and projected range for FLUKA code
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1
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525
F5 Tally Multiple Detectors
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4
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564
Modeling and Simulation in Nuclear Energy
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529
MCNP k_eff question
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11
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739
GM counter and alpha particles
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7
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638
OpenMc installing error
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2
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593
Big announcement about fusion energy coming soon (Dec-2022)
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129
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MCNP5: error in routine imcn
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6
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707
Searching for a single and complete resource about nuclear fuels
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8
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776
PTRAC File - MCNP - Multi-core computing
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676
Fluka installation problem
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652
Particle transporting (FLIBE) -Geant4
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598
MCNPX Fatal Error
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19
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MCNPX Vol card Problem
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1
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600
MCNP6.2 BURN Problem
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Hybrid/variable spectrum reactor
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703
Why do I always get 0 values in the radiation part? (MCNP5)
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58
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Particles transporting in MCNP5
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MCNP5 stopping power for input file
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Cylindrical and Symmetrical Nuclear Fusion Reactor
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Nuclear Engineering MCNP5
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Nuclear fusion and anti-nuclear technology
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Problem with Moritz (3D viewer of MCNPX)
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744
Fusion catalized from graphite?
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Error: no plot because it would be empty, using mcnpx
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Fusion tritium problem solution?
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999
MCNP6 with mpi failed with signal 11 (Segmentation fault)
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[MCNP] Lost too much Keff with Burnup card
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Would wearing sunscreen help during a nuclear disaster/attack?
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Small yield of tactical nuclear weapons
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Does radon remain inside solid radium?
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Nuclear Energy reading material
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New approach to stabilize fusion plasmas
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MCNP and simple nuclear physics
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Halite / Centurion and fission - fusion power plants
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Exceeding the saturation temperature of coolant on the clad for PWR
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Nuclear Fusion Breakthrough....What Does This Mean?
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Zaporizhya Nuclear Power Plant Issues
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Where to find detailed control panel pictures of nuclear plants?
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[MCNPX] How to run an entire folder?
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4
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1K

Nuclear Engineering

Welcome to the Nuclear Engineering forum!

Join other experts in discussion on nuclear engineering. Nuclear engineering deals primarily with the designing and upkeep of nuclear power systems (ie. nuclear reactors).

Topics may include nuclear physics and those concerned with the design and construction and operation of nuclear reactors.

Radiation Protections, Radioactive Waste, Design, Reactors as well!

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