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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
Simulation of low-enriched uranium burnup in Russian VVER-1000 reactors with the Serpent MonteCarlo code
Nuclear Engineering and Technology
2021 .09
Interpretation of two SINBAD photon-leakage benchmarks with nuclear library ENDF/B-VIII.0 and Monte Carlo code MCS
Nuclear Engineering and Technology
2020 .01
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
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
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
Criticality benchmarking of ENDF/B-Ⅷ.0 and JEFF-3.3 neutron data libraries with RMC code
Nuclear Engineering and Technology
2020 .09
Overcoming the challenges of Monte Carlo depletion: Application to a material-testing reactor with the MCS code
Nuclear Engineering and Technology
2020 .09
Development and validation of isotope prediction module for VVER spent nuclear fuel analysis
Nuclear Engineering and Technology
2024 .05
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
Validation of UNIST Monte Carlo code MCS for criticality safety calculations with burnup credit through MOX criticality benchmark problems
Nuclear Engineering and Technology
2021 .01
Validation of nuclide depletion capabilities in Monte Carlo code MCS
Nuclear Engineering and Technology
2020 .09
Validation of UNIST Monte Carlo code MCS using VERA progression problems
Nuclear Engineering and Technology
2020 .01
Spent fuel characterization analysis using various nuclear data libraries
Nuclear Engineering and Technology
2022 .09
Application of a new neutronics/thermal-hydraulics coupled code for steady state analysis of light water reactors
Nuclear Engineering and Technology
2020 .08
ASSESSMENT OF THE TiO2/WATER NANOFLUID EFFECTS ON HEAT TRANSFER CHARACTERISTICS IN VVER-1000 NUCLEAR REACTOR USING CFD MODELING
Nuclear Engineering and Technology
2015 .01
Sensitivity and Uncertainty quantification of neutronic integral data in the TRIGA Mark II Research Reactor
Nuclear Engineering and Technology
2022 .02
Simulations of BEAVRS benchmark cycle 2 depletion with MCS/CTF coupling system
Nuclear Engineering and Technology
2020 .01
Development and verification of a Monte Carlo two-step method for lead-based fast reactor neutronics analysis
Nuclear Engineering and Technology
2023 .06
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