張 承賢

An analytical comparison of the multivariate normal model and the Reed–McCann procedure 

Reliability Engineering and System Safety 265 2026年  

A study on quantification for the impact of stress and time margin on human error probability assessment in nuclear power plant using fuzzy inference 

Nuclear Engineering and Technology 57(5) 2025年  

Dynamic probabilistic risk assessment of seismic-induced flooding in pressurized water reactor by seismic, flooding, and thermal-hydraulics simulations 

Journal of Nuclear Science and Technology 60(4) 2023年  

Simulation-based dynamic probabilistic risk assessment of an internal flooding-initiated accident in nuclear power plant using THALES2 and RAPID 

Proceedings of the Institution of Mechanical Engineers Part O Journal of Risk and Reliability 237(5) 2023年  

Analysis of steam generator tube rupture accidents for the development of mitigation strategies 

Nuclear Engineering and Technology 54(1) 2022年  

Quasi-Monte Carlo sampling method for simulation-based dynamic probabilistic risk assessment of nuclear power plants 

Journal of Nuclear Science and Technology 59(3) 2022年  

Qualitative Risk Analysis of Primary Containment Vessel Venting using STAMP/STPA,格納容器ベントにおける STAMP/STPA を用いた定性的システムリスク分析 

Transactions of the Atomic Energy Society of Japan 21(4) 2022年  

Application of continuous Markov-chain Monte-Carlo method to multi-unit risk evaluations considering interdependence of accident progression among multiple units 

Journal of Nuclear Science and Technology 58(12) 2021年  

Development of source term evaluation method applicable to dynamic PRA,ダイナミック PRA に適用可能なソースターム評価手法の開発 

Transactions of the Atomic Energy Society of Japan 19(1) 2020年  

A novel method for risk-informed decision-making under non-ideal Instrumentation and Control conditions through the application of Bayes’ Theorem 

Reliability Engineering and System Safety 188 2019年  

Experimental and statistical study on fracture boundary of non-irradiated Zircaloy-4 cladding tube under LOCA conditions 

Journal of Nuclear Materials 499 2018年  

Dynamic scenario quantification for level 2 PRA of sodium-cooled fast reactor based on continuous Markov chain and Monte Carlo method coupled with meta-model of thermal–hydraulic analysis 

Journal of Nuclear Science and Technology 55(8) 2018年  

Uncertainty quantification of fracture boundary of pre-hydrided Zircaloy-4 cladding tube under LOCA conditions 

Nuclear Engineering and Design 331 2018年  

Risk Assessment Strategy for Decommissioning of Fukushima Daiichi Nuclear Power Station 

Nuclear Engineering and Technology 49(2) 2017年  

A Numerical Design and Feasibility Study of Self-Wastage Experiment Using Simulant Material in a Sodium Fast Reactor 

Nuclear Engineering and Technology 48(2) 2016年  

Numerical approach for quantification of selfwastage phenomena in sodium-cooled fast reactor 

Nuclear Engineering and Technology 47(6) 2015年