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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!
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The Nuclear Power Thread
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Looking for MCNP tutorials for a beginner
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Understanding Radiation and Radiological Health: A Comprehensive Guide
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Creating the geometry using MNCPX VisEd
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MCNP Code Error -- Need to find the cause of the error
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Separate the inelastic and capture reactions in a tally
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MCNP6 BURN Card Issue: Zero Burnup for Material in Fuel Assembly
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Interesting Documentary on the Dounreay Breeder Reactor
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Beginner Seeking Help: MCNP6 Burnup Example (OECD-NEA Benchmark)
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Fusion research idea involving more pressure
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eVinci MCNP input file
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What Could Be Causing Unexpected Results in the LET Analysis with MCNP6.2?
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Help Needed: Discrepancy in 2D Monte Carlo Neutron Transport Simulation
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Fukushima Japan Earthquake: nuclear plants Fukushima part 2
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How to apply Nemtab cross section library into PARCS?
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Is nuclear fusion without radioactive waste possible?
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How effective is nuclear propulsion?
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488
Mesh Tally in MCNPX for Dose Calculation / How to Add & Visualize?
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Shielded and unshielded dose from a Co-60 source are the same?
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MCNP: Is this a valid way to define "two sources with different sizes"?
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Atomic displacements, replacements, and radiation damage of structural materials
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MCNPX output proton energy is zero?
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MCNP6.2 - How to plot normalized tallies with MCNP6.2?
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Geometry error in newcel
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How to deal with Strong Anisotropic Scattering in multigroup Monte Carlo
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Help to run input file (MCNPX)
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Simulation of Pu-238 Decay
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Anode in center of a magnetic confinement fusion device?
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MCNP PTRAC filters
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China plans world’s first fusion-fission power plant
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MCNP PTRAC card help
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Reflector flux contribution
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The MCNP code never ends, what went wrong?
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Need help with this MCNPX Error
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How to convert MCNP-generated data
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I need help please for Simulation of Xenon Oscillations in a Nuclear Reactor
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MCNP Geometry Error
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MCNP6 Reproduction of 40 MeV D - Li Target Experiment
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I want to see the peak for an explosive detector simulation in MCNP
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Calculating Electron Yield by Photon Saturation in MCNP
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How to use this MCNP physics model?
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Error: "MCNP bank is full. Bank backup file is being created."
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MCNP Deck Error: cannot create srctp
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How to draw a complex geometry in MCNP?
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12 yr old Makes a Nuclear Fusion Reactor
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MCNP: Integral flux crossing the spherical surface of a spherical cap
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MCNP cone source definition
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Zero lattice element hit in MCNP6 (but I cannot find the wrong area)
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Constellation Energy buying Calpine
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Assistance Needed with OpenMC to Calculate Air Kerma and Mean Glandular Dose for Breast Voxel Phantom
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MCNP Data Card Errors
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Chernobyl New Safe Confinement structure struck by drone… what now?
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Technical Discussion on TMESH Tally in MCNP 6.2 and 6.3
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MCNP geometry error
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Best Point Gamma Source in MCNP simulation?
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How to Calculate Total Thermal Power in Nuclear Reactors?
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Subroutine Sources option in MCNP
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Addition of Central Control Rod in Small Modular Reactor Geometry Using Serpent
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How is Uranium refined so it can be used in nuclear reactors?
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How to Calculate Atomic Density for Borated Water (750 ppm) at 300 K?
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Problem with photonuclear physics phys:p in MCNP
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Questions about beta radiation shielding
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MCNP Barrel Geometry Help
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Literature on VVER-1200
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MCNP Geometry Error when I try to create a sliced geometry in a torus structure
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Fukushima Fukushima -- possible alternate scenario?
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Help with error in MCNP6: "Geometry error: no intersection found"
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MCNP6 Error: "Geometry error: no cell found"
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MCNP RHP/HEX Geometry Clarifications
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"10 particles lost" warning in MCNPX
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Help with F4 Fm4 dose calculation in MCNP simulation
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MCNPX ver 2.7 error: ''bad problem in xact in routine tpefil chg_mem error''
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How do I create an SDEF card for lattice geometry?
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How to Calculate Atom/B-cm Values for Nuclear Fuel Composition?
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How to fix this error in MCNPX 2.7?
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LCS error: bert.3 bert.2 when running BURN card MCNP
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MCNP TR Transform Card Question
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Help debugging a geometry-related error in my MCNP input deck
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MCNPX pwr pin depletion input file running error
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MCNP6: Cell Volume Not Calculated - Assymetrical
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Sabine on Small Nuclear Reactors
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Question about multiple runtimes on MPI for MCNP6
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Collecting MCU (Monte Carlo Universal) Input files
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Catalyzed fusion… hype, or reality?
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Gamma electron coupled simulation
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MCNPX memory error: "bad trouble in imcn in routine dynamic allocation"
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How to Calculate the Source Activity from Dose Rate in MCBEND
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Energy Spectrum Attenuation
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How does MNCP calculate an F6 Tally?
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MCNP Modelling Detector Energy Response Vs Empirical Results
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Nuclear Power Gets a Big Boost in the US
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Analyzing Pulse Height vs. Time Response of a Scintillator Detector in Front of a Pulsed X-Ray Source
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Why Are MCNP Photons Getting Lost?
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radioactive iodine - risk of exposure
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Low energy Photon Simulation in MCNP
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Approval from China’s State Council to add four AP1000 plants
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Another SMR on the Horizon
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NuScale, UAMPS terminate small modular reactor project in Idaho
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Fatal error: "entries are not monotonically increasing"
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Question about counting rate in a detector
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Compton scattering in MCNP
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Strange MCNP Fatal Error due to my material card?
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Dose rate measurement through MCNP
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