Keywords: MCNP6; DXTRAN; Angular biasing; Convex hull;
مقالات ISI (ترجمه نشده)
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Keywords: Recycled concrete aggregate; Gamma radiation shielding; Linear attenuation coefficient; Transportable container; Radioactive waste; Surface dose rate; MCNP6;
Keywords: Radiation field; Deep geological repository; High-level nuclear waste; Low-pH concrete; Monte Carlo; MCNP6;
Keywords: DT neutron generator; PGNAA; CEARCPG; MCNP6; Crude oil; Chlorine;
Keywords: Occupational radiation exposure; Effective dose; ICRP/ICRU reference adult voxel phantom; Deep geological disposal facility; High-level nuclear waste; MCNP6;
Keywords: Epithermal spectrum reactor; Thorium; NEWT; MCNP6; Core analysis; Void reactivity; Doppler coefficient;
Keywords: Reactivity; Nuclear data; IFP; IPEN; SPERT; Tripoli-4®; MCNP6;
Keywords: Space radiation; Radiation shielding; Radiation transport; HZETRN; Geant4; FLUKA; PHITS; MCNP6;
Keywords: Neutron imaging; Gamma-ray imaging; MCNP6; Geant4; Cs2LiYCl6:Ce (CLYC); 95% 6Li enriched Cs2LiYCl6:Ce (CLYC-6); Natural Li-glass (GS10); EJ-309 liquid scintillator;
Keywords: Albedo neutron dosimetry; Field calibration; Monte Carlo method; MCNP6; Deep geological repository; High-level nuclear waste;
Keywords: Radionuclide; On-site inspection; 37Ar; MCNP6; Cosmic ray neutrons;
Investigation of uncertainty caused by random arrangement of coated fuel particles in HTTR criticality calculations
Keywords: HTTR; MCNP6; Random arrangement; Coated fuel particle; Criticality calculation;
Sensitivity evaluations of design parameters for various thorium based epithermal reactor (TBER) concepts
Keywords: Epithermal reactor; Thorium; MCNP6; Coolant impurity; Boron content;
SUMCOR: Cascade summing correction for volumetric sources applying MCNP6
Keywords: Cascade summing; Monte Carlo; Gamma-ray emission; HPGe; MCNP6;
Kinetics calculation of fast periodic pulsed reactors using MCNP6
Keywords: Fast periodic pulsed reactors; Prompt supercritical state; MCNP6; Russian roulette;
Neutronic design of homogeneous thorium/uranium fuel for 24 month fuel cycles in the European pressurized reactor using MCNP6
Keywords: EPR; Homogeneous thorium/uranium; 24 month fuel cycles; MCNP6; Thorium fuel;
New perturbation and sensitivity capabilities in Tripoli-4®
Keywords: Perturbation theory; Sensitivity analysis; IFP; Differential operator sampling; Correlated sampling; Monte Carlo; Tripoli-4®; MCNP6; COMAC; SCALE;
Validation of UNIST Monte Carlo code MCS for criticality safety analysis of PWR spent fuel pool and storage cask
Keywords: PWR spent fuel; Criticality safety analysis; MCS; Validation; ICSBEP; MCNP6;
The effects of nuclear data library processing on Geant4 and MCNP simulations of the thermal neutron scattering law
Keywords: Thermal neutron scattering; Scattering law; Double-differential cross section; Geant4; MCNP6; NJOY;
Comparison of the thermal neutron scattering treatment in MCNP6 and GEANT4 codes
Keywords: Benchmarking; Monte Carlo simulation; Thermal scattering law; MCNP6; GEANT4; CANS;
Review of heavy charged particle transport in MCNP6.2
Keywords: MCNP6; Hadrons; Charged particles; Ions; Monte Carlo; Radiation transport;
Study by Monte Carlo methods of an explosives detection system made up with a D-D neutron generator and NaI(Tl) gamma detectors
Keywords: Explosives detection; MCNP6; RDX; Ammonium nitrate; Neutron generator; Deuterium-deuterium reaction;
MCNP6 unstructured mesh application to estimate the photoneutron distribution and induced activity inside a linac bunker
Keywords: MCNP6; Unstructured mesh; Photoneutron reactions; Bunker; Induced activity; LinAc;
Background-Source Cosmic-Photon Elevation Scaling and Cosmic-Neutron/Photon Date Scaling in MCNP6
Keywords: MCNP6; MCNP 6.2.0; Cosmic-Source; Background-Source; Elevation Scaling; Date Scaling; radiation transport;
Studies on the liquid fluoride thorium reactor: Comparative neutronics analysis of MCNP6 code with SRAC95 reactor analysis code based on FUJI-U3-(0)
Keywords: MCNP6; Molten salt reactor; MSR; FUJI-U3;
Evaluation of the new electron-transport algorithm in MCNP6.1 for the simulation of dose point kernel in water
Keywords: MCNP6; Electrons transport algorithm; Condensed history method; Single event algorithm; Dose Point Kernel (DPK);
Numerical investigation of the random arrangement effect of coated fuel particles on the criticality of HTTR fuel compact using MCNP6
Keywords: HTTR; Coated fuel particle; Stochastic geometry; Realized random packing; MCNP6;
An energy-dependent numerical model for the condensation probability, ?j
Keywords: Condensation probability; Nuclear spallation reactions; Monte Carlo; Transport codes; MCNP6; Cascade exciton model (CEM); Fragment spectra;
Performance of 10B+ZnS(Ag) neutron detectors in RPM for the detection of special nuclear materials
Keywords: Special nuclear materials; Neutron; Radiation portal monitors; MCNP6;
Fluence to local skin absorbed dose and dose equivalent conversion coefficients for monoenergetic positrons using Monte-Carlo code MCNP6
Keywords: Skin dose; MCNP6; Local-Skin equivalent Dose; Positrons
Well GeHP detector calibration for environmental measurements using reference materials
Keywords: Well-type detector; Full energy peak efficiency; True coincidence summing; MCNP6; Efficiency correction;
Positron-based attenuation correction for Positron Emission Tomography data using MCNP6 code
Keywords: PET; MCNP6; Image reconstruction; Attenuation correction
Lumped pseudo fission products during burnup step in MCNP5-ORIGEN coupling system
Keywords: MCNP5; ORIGEN; MCNP6; Fission products; UZrH1.6 fuel; End of cycle
Verification of MCNP6 model of the Jordan Research and Training Reactor (JRTR) for calculations of neutronic parameters
Keywords: JRTR; Neutronic; Keff; Criticality; Flux; Monte Carlo; Research reactor; Power peaking factor; MCNP6;
Biasing secondary particle interaction physics and production in MCNP6
Keywords: MCNP6; (a, n); Secondary particle biasing;
Applicability of LR-0 mock-up results to VVER-1000 reactor pressure vessel issues
Keywords: LR-0; VVER-1000; Neutron flux; Neutron spectra; MCNP6; TORT;
Review calculations for the OECD/NEA Burn-up Credit Criticality Safety Benchmark
Keywords: Burn-up credit; MCNP6; SERPENT2; CINDER; Waste disposal; Criticality safety;
Coupled MCNP6/CTF code: Development, testing, and application
Keywords: MCNP6; CTF; Internal coupling; Relaxation acceleration;
Nuclear excitation functions from 40 to 200 MeV proton irradiation of terbium
Keywords: Terbium; Proton irradiation; MCNP6; Radiotherapy; Radionuclide production
Features of MCNP6
Keywords: Monte Carlo; Particle transport; MCNP; MCNP6;
Application of Serpent 2 and MCNP6 to study different criticality configurations of a VVER-1000 mock-up
Keywords: LR-0 research reactor; MCNP6; Serpent 2; VVER-1000 mock-up;
MCNP6 moving objects part I: Theory
Keywords: MCNP6; Monte Carlo; Moving objects; Rigid-body kinematics; Rectilinear; Curvilinear; Translation; Rotation; Radiation transport; Fission; HEU; Pu; Beddingfield; SDEF; Delayed neutrons; Delayed gammas; DGNAA;
MCNP6 moving objects. Part II: Simulations
Keywords: MCNP6; Moving objects; Fission; HEU; Pu; Beddingfield; SDEF; Radiation transport; Delayed neutrons; Delayed gammas; Rigid-body kinematics; Translation; Rotation;
Performance upgrades to the MCNP6 burnup capability for large scale depletion calculations
Keywords: MCNP6; Monte Carlo linked burnup; Burnup
Monte Carlo calculation of the spatial response (Modulated Transfer Function) of a scintillation flat panel and comparison with experimental results
Keywords: X-ray imaging; Modulation Transfer Function (MTF); Slit method; Monte Carlo simulation; MCNP6; Flat panel scintillator;
Use of MOSFET dosimeters to validate Monte Carlo radiation treatment calculation in an anthropomorphic phantom
Keywords: Monte Carlo simulation; MCNP6; MOSFET detectors; Radiation treatment planning;
MCNP6 simulations of gamma line emissions from fission products and their comparisons to plutonium and uranium measurements
Keywords: MCNP6; Delayed gamma; Special nuclear materials;
Testing actinide fission yield treatment in CINDER90 for use in MCNP6 burnup calculations
Keywords: MCNP6; CINDER90; Fission yield; Burnup;
Total reaction cross sections in CEM and MCNP6 at intermediate energies
Keywords: Reaction cross sections; Monte Carlo; Transport codes; MCNP6; Cascade-Exciton Model (CEM);
SMITHERS: An object-oriented modular mapping methodology for MCNP-based neutronic-thermal hydraulic multiphysics
Keywords: Neutronic/thermal-hydraulic coupling; Operator splitting in reactor analysis; MCNP6; MONTEBURNS;