Nuclear Engineering Forum

Nuclear reactor design with refuel period of more than 20 years

MCNP5 LINAC Isocenter Dose Calculation Produces No Tally Results

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Realta Fusion claims electrical generation from plasma!

Can I use MCNP on ARM devices?

Simulating an RBMK-1500 like the one in Ignalina

How close are we to getting micro reactors?

Calculating Electron Yield by Photon Saturation in MCNP

Regarding Weight Window Generation Iteration in MCNP

What Geiger Counter is best for this?

How to draw a complex geometry in MCNP?

MCNP PTRAC card outputting gibberish

Geiger Counter Question

Configuration of JEFF-3.3 Nuclear Data Library in MCNP6

Particles lost in MCNP running

How effective is nuclear propulsion?

Penning ion source test

How to apply Nemtab cross section library into PARCS?

OKLO's 15 MW liquid metal-cooled fast reactor for OpenAI power

Normalization of MCNP F4:P photon flux tally to experimental counts

Nuclear Engineering and Propulsion Journals

Geometry modeling of an X-ray Irradiator

MCNP IMP:n=0 when IMP:p,e=1

MCNP vs Serpent pin-cell burnup discrepancy in keff evolution

MCNP always saying my codes missed a cell

Creating the geometry using MNCPX VisEd

MCNP Code Error -- Need to find the cause of the error

Separate the inelastic and capture reactions in a tally

MCNP6 BURN Card Issue: Zero Burnup for Material in Fuel Assembly

Interesting Documentary on the Dounreay Breeder Reactor

MCNP6 burnup calculation with OECD-NEA benchmark

Fusion research idea involving more pressure

eVinci MCNP input file

What Could Be Causing Unexpected Results in the LET Analysis with MCNP6.2?

Debugging 2D Monte Carlo neutron transport code with fuel pin geometry

Is nuclear fusion without radioactive waste possible?

Shielded and unshielded dose from a Co-60 source are the same?

MCNP: Is this a valid way to define "two sources with different sizes"?

Atomic displacements, replacements, and radiation damage of structural materials

MCNPX output proton energy is zero?

MCNP6.2 - How to plot normalized tallies with MCNP6.2?

Geometry error in newcel

MCNPX internal radiotherapy simulation input file validation

Simulation of Pu-238 Decay

Anode in center of a magnetic confinement fusion device?

MCNP PTRAC filters

China plans world’s first fusion-fission power plant

Increasing neutron flux in reflector cells above 5e8

The MCNP code never ends, what went wrong?

MCNPX burnup calculation with lattice geometry

How to convert MCNP-generated data

Xenon oscillation simulation with diffusion model and Newton-Raphson solver

MCNP cell definitions for 300x300x300 cm room with doorway

MCNP6 Reproduction of 40 MeV D - Li Target Experiment

I want to see the peak for an explosive detector simulation in MCNP

How to use this MCNP physics model?

MCNP Deck Error: cannot create srctp

12 yr old Makes a Nuclear Fusion Reactor

MCNP: Integral flux crossing the spherical surface of a spherical cap

MCNP cone source definition

Zero lattice element hit in MCNP6 (but I cannot find the wrong area)

Constellation Energy buying Calpine

Assistance Needed with OpenMC to Calculate Air Kerma and Mean Glandular Dose for Breast Voxel Phantom

MCNP Data Card Errors

Drone strike damage to Chernobyl New Safe Confinement structure

Technical Discussion on TMESH Tally in MCNP 6.2 and 6.3

MCNP cell 14 geometry error in detector model

Best Point Gamma Source in MCNP simulation?

How to Calculate Total Thermal Power in Nuclear Reactors?

Subroutine Sources option in MCNP

How is Uranium refined so it can be used in nuclear reactors?

How to Calculate Atomic Density for Borated Water (750 ppm) at 300 K?

Problem with photonuclear physics phys:p in MCNP

Questions about beta radiation shielding

MCNP geometry appearing red in vised plot

Literature on VVER-1200

Cooling systems in thermal power stations

MCNP Geometry Error when I try to create a sliced geometry in a torus structure

MCNP6 geometry error: no intersection found with lost particles

MCNP6 Error: "Geometry error: no cell found"

Hexagonal lattice cells not replicating in Z axis in MCNP

"10 particles lost" warning in MCNPX

MCNP dose calculation in tibia with Ho-166 brachytherapy

MCNPX ver 2.7 error: ''bad problem in xact in routine tpefil chg_mem error''

How do I create an SDEF card for lattice geometry?

How to Calculate Atom/B-cm Values for Nuclear Fuel Composition?

How to fix this error in MCNPX 2.7?

LCS error: bert.3 bert.2 when running BURN card MCNP

MCNP TR Transform Card Question

Debugging surface overlap error in MCNP input deck

MCNPX pwr pin depletion input file running error

MCNP6: Cell Volume Not Calculated - Assymetrical

Sabine on Small Nuclear Reactors

Why does MCNP6 MPI generate different runtpe files and results per node?