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Help, try to make one analysis in the RITM-200 reactor series, but can't find some informartion... any tips?
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MCNP: Help constructing the geometry for this pinpoint chamber please
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Design error in reactor nuclear simulation in serpent code
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Why Are Burnup Numbers Identical for Different Fuels in MCNP?
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Help Needed with MCNP Simulation for Brachytherapy Treatment Room
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Why is My MCNP Burnup Calculation Failing?
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Computation for nuclear reactor systems
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How to Determine Fission Rate in MCNP6?
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PIG (Penning Ionization Gauge) Ion source
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Could Flattened End Magnets Improve Magnetic Mirror Traps?
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MCNPX Transformation
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Endothermic and Threshold reactions - Are they Equivalent?
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Question regarding fusion reaction containment and procedure
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I need help in solving the problem of the code written with MCNPX 2.6
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Vogtle 4 initial criticality
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MCNP: Declaring two sources in two cells
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MCNP6: Energy and angular distribution of neutrons after moderation
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Plotting more than one tally together in VISED MCNP version
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Learning VisEd for MCNP
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MCNP Summary Table: Explanation and Use
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Fukushima Fukushima Daiichi nuclear disaster discussions
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Radioisotope Decay Simulation in MCNP6
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To simulate the interaction of gamma photons with only oxygen (O) atom
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Are spent nuclear fuel rods radioactive?
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Can Any Country Achieve Net Zero Without Nuclear?
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New Salt Cooled Nuclear Reactor Approved by NRC
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How can I fix the illegal data signal error in my FMESH tally code in MCNP6?
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Have pulsed fission reactors got any potential in nuclear energy?
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MCNP6 gamma dose tally problems
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Reproducing (almost) Oklo reactor conditions in the lab
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Confirmation of Laser Fusion Break Even Point
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MCNP: Can I input the X-Ray tube voltage in MCNP source specification?
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Can Small Modular Reactors Revitalize Nuclear-Powered Surface Combatants?
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Is it possible to use TR or TRCL to translate an F4 or FMESH4 tally in MCNP6.2?
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How is Alpha Particle Energy Defined in MCNP6?
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Solving Fatal Error in MCNP6 Particle Alpha Mode
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Describing cosmic particles in mcnp6
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Does the hydrogen bomb emit beta radiation?
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What is the practical size of a neutron research reactor?
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Atomic density of Oxygen in a composite
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What Is the Difference Between Angular Neutron Flux and Neutron Current Vector?
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Elevated Geiger Counter Radiation Reading for Envelope?
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MCNP code for Neutron Spectroscopy
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Fukushima Candu Reactor compared k_eff values with Serpent
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Why can't we simply use an atomic bomb to get the fusion working?
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Need help ordering parts for a nuclear fusion reactor
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Calculating Microdosimetry in MCNP
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How to Declare a Spherical Neutron Source for Thermal Neutron Flux Calculation?
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Sources for nuclear fuel pellet fabrication methods?
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Error running example ICRP145_HumanPhantoms in Geant4
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Westinghouse Very Small Modular Reactor Progress
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Help with neutron spectroscopy experiments in MCNP code
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Radioistope Power Systems (RPS) and Pu-238
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MCNP Problem - Bad character in column 2
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Best software to simulate the radiation emitted by uranium powder inside a pipe?
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Double differential cross section in Fluka
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How Can I Simulate a Narrow Spectrum X-Ray Beam Using MCNP?
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Problem with Moritz (3D viewer of MCNPX)
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FLUKA: Why Does Neutron Flux Increase When Passing Through a Moderator?
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How to Convert TMESH Data to Text Files in MCNP6 Using Python"
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Background Analysis for MCBEND Simulation
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Question about Source Probability in MCNP
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MCNPX Fatal error: "electron importances are zero"
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What is the physical significance of WGT and SBn in MCNP source specifications?
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MCNP -- What is meant by 2M and 20R?
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How to check the isotropy of the source in MCNP?
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Need Recommendations for Nuclear Reaction Modelling Software
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Particles produced and energies in the target Geant4
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Positioning of Capacitance in Ionization Chamber's Equivalent Circuit
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Gaussian Energy Broadening tally
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How far do fission products travel?
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MCNP6: error -- "bad trouble in imcn" is usually tangled with spacing error?
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Understanding MCNP Tally F5 Output: Tips for Beginners
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What actually happens to the U-238 part of fuel in a nuclear reactor?
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Real Experiment-MCBEND Simulation
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Experiment-Simulation comparing count values in MCNP
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Meaning of "Average" Flux Tallies in MCNP
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Isotropic distribution for a surface source MCNP
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Chernobyl Chernobyl tourism visit caused contamination?
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Dose calculation for oscilloscope
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Gamma Ray Shielding -- How much is enough?
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Why is My MCNP Program Outputting Too Small a Value?
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Best Dosimeter for 10 ns Pulse Dose Measurement
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Reactor refrigeration without external power supply
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Is there anyone here who uses the Mesh tally in MCNPX?
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Creating FMESH Tally Over Region - Troubleshooting
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Thermal neutron detection using MCNP
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Product Cross Section For Zn66 in Geant4
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Dose recording on Pocket dosimeter -- Is it accurate for a person's whole body?
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Why am I seeing a discrepancy in photon energy when doping ZnS:Ag in MCNP 6.1?
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Codes to calculate diffusion parameters of homogeneous reactor
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MCNP6.2 - SSW and surfaces defining universe boundaries
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Plot Mesh Tally with MCNP - Help request from Carla
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MCNP6.2 - Combination of transformations
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Efficient MCNP Lattice Source Help: Defining Universes and Tallies in Cell File
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MCNP: How to display particles outside the source?
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MCNP Tally f8 unreliable since neutron transport is nonanalog
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Is quartarad radex 1503+ or GQ GMC-300E a good geiger counter?
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MCNP: Ctme Card Impacts on Results
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Run MCNP 5 input file of certain geometry for flux calculation
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