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Processing and benchmarking of evaluated nuclear data file/b-viii.0β4 cross-section library by analysis of a series of critical experimental benchmark using the monte carlo code MCNP(X) and NJOY2016
Nuclear Engineering and Technology
2017 .01
Criticality benchmarking of ENDF/B-Ⅷ.0 and JEFF-3.3 neutron data libraries with RMC code
Nuclear Engineering and Technology
2020 .09
The multigroup library processing method for coupled neutron and photon heating calculation of fast reactor
Nuclear Engineering and Technology
2024 .04
Analysis of VVER-1000 mock-up criticality experiments with nuclear data library ENDF/B-VIII.0 and Monte Carlo code MCS
Nuclear Engineering and Technology
2021 .01
Validation of MCS code for shielding calculation using SINBAD
Nuclear Engineering and Technology
2022 .09
Current Status of ACE Format Libraries for MCNP at Nuclear Data Center of KAERI
방사선방어학회지
2016 .01
Spent fuel characterization analysis using various nuclear data libraries
Nuclear Engineering and Technology
2022 .09
Analysis of the first core of the Indonesian multipurpose research reactor RSG-GAS using the Serpent Monte Carlo code and the ENDF/BVIII.0 nuclear data library
Nuclear Engineering and Technology
2020 .12
An inter-comparison between ENDF/B-VIII.0-NECP-Atlas and ENDF/B-VIII.0-NJOY results for criticality safety benchmarks and benchmarks on the reactivity temperature coefficient
Nuclear Engineering and Technology
2021 .08
Optimization of Target, Moderator, and Collimator in the Accelerator-based Boron Neutron Capture Therapy System: A Monte Carlo Study
Nuclear Engineering and Technology
2021 .06
Analysis of the CREOLE experiment on the reactivity temperature coefficient of the UO2 light water moderated lattices using Monte Carlo transport calculations and ENDF/B-VII.1 nuclear data library
Nuclear Engineering and Technology
2020 .01
Modelling atomic relaxation and bremsstrahlung in the deterministic code STREAM
Nuclear Engineering and Technology
2024 .02
Dynamic Monte Carlo transient analysis for the Organization for Economic Co-operation and Development Nuclear Energy Agency (OECD/NEA) C5G7-TD benchmark
Nuclear Engineering and Technology
2017 .01
Applicability of the Kr ? ko nuclear power plant core Monte Carlo model for the determination of the neutron source term
Nuclear Engineering and Technology
2021 .11
Sensitivity and Uncertainty quantification of neutronic integral data in the TRIGA Mark II Research Reactor
Nuclear Engineering and Technology
2022 .02
Monte Carlo simulation and optimization of neutron ray shielding performance of related materials
Nuclear Engineering and Technology
2024 .09
Copper neutron transport libraries validation by means of a 252 Cf standard neutron source
Nuclear Engineering and Technology
2021 .10
Neutron activation analysis: Modelling studies to improve the neutron flux of Americiume-Beryllium source
Nuclear Engineering and Technology
2017 .06
An Assessment of the Secondary Neutron Dose in the Passive Scattering Proton Beam Facility of the National Cancer Center
Nuclear Engineering and Technology
2017 .06
Thermal neutron albedo and flux for different geometries neutron guide
Nuclear Engineering and Technology
2019 .01
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