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논문 기본 정보
- 자료유형
- 학술대회자료
- 저자정보
- 발행연도
- 2019.1
- 수록면
- 84 - 85 (2page)
이용수
이 논문의 연구 히스토리 (4)
2018
Effect of Hoop Stress on the Hydride Reorientation of Non-irradiated Zircaloy-4 Cladding Tube : 90 MPa, 120 MPa & 150 MPa
Kim
,
DaeHo
,
Hong
외 9명
한국방사성폐기물학회 학술대회
2018.01
학술대회자료
Effect of Cooling Rate on the Hydride Reorientation of Non-irradiated Zircaloy-4 Cladding Tube : 0.5℃/min vs. 0.5℃/hr
Kim
,
DaeHo
,
Hong
외 9명
한국방사성폐기물학회 학술대회
2018.01
학술대회자료
Effect of Thermal Cycling on the Hydride Reorientation of Non-irradiated Zircaloy-4 Cladding Tube : 1 Cy, 3 Cy & 10 Cy
Kim
,
DaeHo
,
Hong
외 9명
한국방사성폐기물학회 학술대회
2018.01
학술대회자료
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Effect of Hoop Stress on the Hydride Reorientation of Non-irradiated Zircaloy-4 Cladding Tube : 90 MPa, 120 MPa & 150 MPa
한국방사성폐기물학회 학술대회
2018 .01
-
Effect of Thermal Cycling on the Hydride Reorientation of Non-irradiated Zircaloy-4 Cladding Tube : 1 Cy, 3 Cy & 10 Cy
한국방사성폐기물학회 학술대회
2018 .01
-
Hydrogen Content Effects on Delayed Hydride Cracking in Zircaloy-4 Cladding
한국방사성폐기물학회 학술대회
2018 .01
-
Effects of Hydride Re-orientation and Hydride Rim on Fracture Energy of Zircaloy-4 Cladding
한국방사성폐기물학회 학술대회
2018 .01
-
Micro-cantilever Tests on Hydride Blister and Zirconium Matrix of Zircaloy-4 Cladding Tube
한국방사성폐기물학회 학술대회
2019 .01
-
비조사 Zircaloy-4 피복관의 링압축 고온 특성 평가
대한기계학회 춘추학술대회
2024 .11
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Effects of Zr-hydride distribution of irradiated Zircaloy-3 cladding in RIA-simulating pellet-clad mechanical interaction testing
Nuclear Engineering and Technology
2018 .01
-
Measurement of Threshold Stress Intensity Factor of Delayed Hydride Cracking for Unirradiated Zircaloy-4 Cladding
한국방사성폐기물학회 학술대회
2019 .01
-
Temperature-dependent axial mechanical properties of Zircaloy-4 with various hydrogen amounts and hydride orientations
Nuclear Engineering and Technology
2022 .05
-
The effect of neutron irradiation on hydride reorientation and mechanical property degradation of zirconium alloy cladding
Nuclear Engineering and Technology
2017 .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
-
Allowable peak heat-up cladding temperature for spent fuel integrity during interim-dry storage
Nuclear Engineering and Technology
2017 .01
-
Effect of a surface oxide-dispersion-strengthened layer on mechanical strength of Zircaloy-5 tubes
Nuclear Engineering and Technology
2018 .01
-
Theoretical models of threshold stress intensity factor and critical hydride length for delayed hydride cracking considering thermal stresses
Nuclear Engineering and Technology
2018 .01
-
Three-dimensional Numerical Simulation of Hydrogen-induced Multi-field Coupling Behavior in Cracked Zircaloy Cladding Tubes
Nuclear Engineering and Technology
2019 .01
-
Methodology for numerical evaluation of fracture resistance under pinch loading of spent nuclear fuel cladding containing reoriented hydrides
Nuclear Engineering and Technology
2024 .06
-
Methodology of Delayed Hydride Cracking Assessment of Spent Fuel Cladding
한국방사성폐기물학회 학술대회
2017 .01
-
Understanding the role of hydrogen on creep behaviour of Zircaloy-4 cladding tubes using nanoindentation
Nuclear Engineering and Technology
2020 .09
-
Fatigue Life Characterization of Simulated Fuel Cladding to Hydride Contents for Normal Transportation Integrity
한국방사성폐기물학회 학술대회
2019 .01
-
Effect of Inhomogeneous Nucleation of Hydride at α/β Phase Boundary on Microstructure Evolution of Zr–2.5 wt%Nb Pressure Tube
Metals and Materials International
2019 .01
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