지원사업
학술연구/단체지원/교육 등 연구자 활동을 지속하도록 DBpia가 지원하고 있어요.
커뮤니티
연구자들이 자신의 연구와 전문성을 널리 알리고, 새로운 협력의 기회를 만들 수 있는 네트워킹 공간이에요.
이용수
등록된 정보가 없습니다.
논문 유사도에 따라 DBpia 가 추천하는 논문입니다. 함께 보면 좋을 연관 논문을 확인해보세요!
Circumferential steady-state creep test and analysis of Zircaloy-4 fuel cladding
Nuclear Engineering and Technology
2021 .07
Impact of hydrogen on rupture behaviour of Zircaloy-4 nuclear fuel cladding during loss-of-coolant accident: a novel observation of failure at multiple locations
Nuclear Engineering and Technology
2021 .02
냉각재 상실사고 통합 실험장치 설계 및 제작
대한기계학회 춘추학술대회
2017 .11
Deep neural network based prediction of burst parameters for Zircaloy-4 fuel cladding during loss-ofcoolant accident
Nuclear Engineering and Technology
2020 .11
Burst criterion for Indian PHWR fuel cladding under simulated loss-of-coolant accident
Nuclear Engineering and Technology
2019 .01
Influence of hydrogen concentration on burst parameters of Zircaloy-4 cladding tube under simulated loss-of-coolant accident
Nuclear Engineering and Technology
2020 .09
Thermal creep effects of aluminum alloy cladding on the irradiation-induced mechanical behavior in U – 10Mo/Al monolithic fuel plates
Nuclear Engineering and Technology
2020 .01
Stress-Based Criterion for the Creep Response of Spent Nuclear Fuel Rod
한국방사성폐기물학회 학술대회
2018 .01
Analytical criteria for fuel fragmentation and burst FGR during a LOCA
Nuclear Engineering and Technology
2020 .10
Thermal creep behavior of CZ cladding under biaxial stress state
Nuclear Engineering and Technology
2020 .12
Effect of Creep Deformation on Ductility of Simulated Spent Fuel Cladding
한국방사성폐기물학회 학술대회
2018 .01
Development of mechanistic cladding rupture model for severe accident analysis and application in PHEBUS FPT3 experiment
Nuclear Engineering and Technology
2022 .01
실험을 통한 플라스틱 크리프 거동 특성 연구
한국정밀공학회 학술발표대회 논문집
2015 .05
Axial strength of Zircaloy-4 samples with reduced thickness after a simulated loss of coolant accident
Nuclear Engineering and Technology
2021 .07
Modified projection model-based constant-stress creep curve for alloy 690 steam generator tube material
Nuclear Engineering and Technology
2022 .03
Investigating creep behavior of Ni–Cr–W alloy pressurized tube at 950 °C by using in-situ creep testing system
Nuclear Engineering and Technology
2020 .01
Laser cladding 표면의 젖음성 평가
한국트라이볼로지학회 학술대회
2019 .04
3D Digital Image Correlation 기법을 활용한 핵연료 피복관 고온 대변형 다차원 측정 실험
대한기계학회 춘추학술대회
2018 .12
Creep Rupture Characteristics of Cladding Tubes of FC92B and FC92N: Candidate Alloys of SFR Fuel Cladding Tube Materials
Metals and Materials International
2022 .06
Creep strain modeling for alloy 690 SG tube material based on modifi ed theta projection method
Nuclear Engineering and Technology
2022 .05
0