Youho Lee

Seoul National University — South Korea

Materials Science · Physical Sciences

21h-index1.5kcitations123works2.52yr avg
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Research Topics

Nuclear Materials and Properties(89), Nuclear reactor physics and engineering(61), Fusion materials and technologies(39), Nuclear Engineering Thermal-Hydraulics(15), Nuclear and radioactivity studies(14)

Publications123 total

An integral mechanistic model for the degradation and protection loss of Cr-coated zirconium alloys in steam oxidation environments
Corrosion Science·2025
Development of a 2D axisymmetric FDM structural analysis model for coated cladding
Nuclear Engineering and Technology·2025Open Access
Fracture parameter estimation for crack simulation models of spent nuclear fuel cladding with hydrides under pinch load
Nuclear Engineering and Design·2025Open Access
Missing half the story: Impact of secondary hydriding on post-LOCA ductility of Cr-coated zirconium alloy accident tolerant fuel cladding
Journal of Nuclear Materials·2025
Assessment of the impact of Zr-Cr eutectic reaction on the structural integrity of Cr-coated ATF cladding
Journal of Nuclear Materials·2025· 3 cited
Microstructural analysis of interconnected δ-hydride structures in a zirconium alloy
Acta Materialia·2025· 1 cited
Post-ballooning and burst steam oxidation of accident tolerant zirconium alloy cladding with cracked chromium coating
Journal of Nuclear Materials·2025· 2 cited
How brittle is 'too brittle'?: Comparative mechanical assessment of silicon carbide and post-quench steam-oxidized zirconium alloy cladding
Journal of Nuclear Materials·2025
Microstructure-informed modeling of radial hydride precipitation in reactor-grade PRXA Zirconium alloy cladding tube
Journal of Nuclear Materials·2025
Parabolic oxidation for various chromium-coated Zr-Nb Alloy Claddings
Journal of Nuclear Materials·2025· 5 cited
Full-core fuel analysis of a soluble boron-free SMR: Pellet-cladding interaction issue and enhancing fuel safety through loading pattern design
Nuclear Engineering and Technology·2025Open Access
Mechanisms of steam oxidation-induced degradation of chromium coating on zirconium alloys at high temperatures
Corrosion Science·2025· 10 cited
GIFT-1.0: Advanced light water reactor fuel performance code
Nuclear Engineering and Technology·2025· 3 citedOpen Access
Parabolic Oxidation Behavior of Various Chromium-Coated Zr-Nb Alloy Claddings
SSRN Electronic Journal·2025· 1 citedOpen Access
Mechanisms of Steam Oxidation-Induced Degradation of Chromium Coating on Zirconium Alloys at High Temperatures
SSRN Electronic Journal·2025Open Access
Missing Half the Story: Impact of Secondary Hydriding on Post-Loca Ductility of Cr-Coated Zirconium Alloy Accident Tolerant Fuel Cladding
SSRN Electronic Journal·2025Open Access
An Integral Mechanistic Model for the Degradation and Protection Loss of Cr-Coated Zirconium Alloys in Steam Oxidation Environments
SSRN Electronic Journal·2025Open Access
Assessment of the Impact of Zr-Cr Eutectic Phase on the Structural Integrity of Cr-Coated Atf Cladding
SSRN Electronic Journal·2025Open Access
Post-Ballooning and Burst Steam Oxidation of Accident Tolerant Zirconium Alloy Cladding with Cracked Chromium Coating
SSRN Electronic Journal·2025Open Access
Microstructure-Informed Modeling of Radial Hydride Precipitation in Reactor-Grade Prxa Zirconium Alloy Cladding Tube
SSRN Electronic Journal·2025Open Access
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Frequent Co-authors

Maolong Liu(7), Dongju Kim(7), Jeong Ik Lee(6), Mujid S. Kazimi(6), Nicholas R. Brown(5), Thomas McKrell(5), Soon Lee(4), Kurt A. Terrani(4), Joo‐Hee Kang(4), Hee Cheon No(3), Mingfu He(3), Hyunwoo Yook(3), Kyunghwan Keum(3), Jinsu Kim(3), Dahyeon Woo(3), Dongyeon Kim(3), Seong Gu Kim(2), Edward D. Blandford(2), Colby Jensen(2), Sangbum Kim(2)