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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
Current Status of ACE Format Libraries for MCNP at Nuclear Data Center of KAERI
방사선방어학회지
2016 .01
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
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
Spent fuel characterization analysis using various nuclear data libraries
Nuclear Engineering and Technology
2022 .09
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
Copper neutron transport libraries validation by means of a 252 Cf standard neutron source
Nuclear Engineering and Technology
2021 .10
Sensitivity and Uncertainty quantification of neutronic integral data in the TRIGA Mark II Research Reactor
Nuclear Engineering and Technology
2022 .02
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
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
Simulation of low-enriched uranium burnup in Russian VVER-1000 reactors with the Serpent MonteCarlo code
Nuclear Engineering and Technology
2021 .09
Influence of nuclear data library on neutronics benchmark of China experimental fast reactor start-up tests
Nuclear Engineering and Technology
2022 .10
Atomic displacement cross-sections for neutron irradiation of materials from Be to Bi calculated using the arc-dpa model
Nuclear Engineering and Technology
2019 .01
A Lattice-Based Monte Carlo Evaluation of Canada Deuterium Uranium-6 Safety Parameters
Nuclear Engineering and Technology
2016 .06
회전 변조 시준기(RMC) 데이터와 영상화 알고리즘 시뮬레이션
대한전자공학회 학술대회
2015 .11
The impact of fuel depletion scheme within SCALE code on the criticality of spent fuel pool with RBMK fuel assemblies
Nuclear Engineering and Technology
2022 .12
Fission counter array for pulse-mode measurements of high-flux and high-energy neutrons
Nuclear Engineering and Technology
2024 .09
Study on (n,p) reactions of 58,60,61,62,64 Ni using new developed empirical formulas
Nuclear Engineering and Technology
2020 .01
Neutronics analysis of TRIGA Mark II research reactor
Nuclear Engineering and Technology
2018 .01
Convolutional Neural Network(CNN)을 이용한 회전 변조 시준기(RMC) 영상화 알고리즘
대한전자공학회 학술대회
2018 .06
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