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이유호 교수

Youho Lee

서울대학교 · 재료과학

연구실 소개

이 교수의 연구실은 원자로 내부에서의 연료 피복재 거동을 중심으로, 고온에서의 산화 거동, 피로 및 연성 저하, 그리고 손상 거동을 기계적·열적 모델링과 실험을 병행하여 연구하고 있습니다. 특히 크롬 도금된 지르칼로이 및 SiC 피복재의 고온 스팀 산화 거동, 수소화로 인한 연성 저하, 그리고 LOCA 상황에서의 거동을 분석하여 안전성 향상을 목표로 하고 있습니다. 연구는 실험적 검증과 수치 모델링을 융합하여 실증적인 원자력 소재 기술 개발에 기여하고 있습니다.

피복재 안전성SiC 산화 거동수소화 연성 저하LOCA 시험산화 모델링

연구 현황

논문 수
123
총 인용 수
1,541
최근 5년 논문
54
주요 분야
재료과학

연구 성과 추이

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5개년 연도별 논문 게재 수
54총합
2022
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2026
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주요 논문

15
1
논문|인용수 91·2014
Structural assessment of intermediate printed circuit heat exchanger for sodium-cooled fast reactor with supercritical CO2 cycle
Youho Lee, Jeong Ik Lee
SJR Q1FWCI 3.3Annals of Nuclear Energy
Ceramics and CompositesMaterials Science
2
논문|인용수 88·2014
A structural model for multi-layered ceramic cylinders and its application to silicon carbide cladding of light water reactor fuel
Youho Lee, Mujid S. Kazimi
SJR Q1FWCI 6.7Journal of Nuclear Materials
Ceramics and CompositesMaterials Science
3
논문|인용수 59·2017
Mechanical analysis of surface-coated zircaloy cladding
Youho Lee, Jeong Ik Lee, Hee Cheon No
SJR Q2FWCI 2.6Nuclear Engineering and TechnologyOA

A structural model for stress distributions of coated Zircaloy subjected to realistic incore pressure difference, thermal expansion, irradiation-induced axial growth, and creep has been developed in this study. In normal operation, the structural integrity of coating layers is anticipated to be significantly challenged with increasing burnup. Strain mismatch between the zircaloy and the coated layer, due to their different irradiation-induced axial growth, and creep deformation are found to be t

Materials ChemistryMaterials Science
4
논문|인용수 55·2018
Comparison of steady and transient flow boiling critical heat flux for FeCrAl accident tolerant fuel cladding alloy, Zircaloy, and Inconel
Soon Lee, Maolong Liu, Nicholas R. Brown, Kurt A. Terrani, Edward D. Blandford, Heng Ban, Colby Jensen, Youho Lee
SJR Q1FWCI 2.9International Journal of Heat and Mass TransferOA
Materials ChemistryMaterials Science
5
논문|인용수 48·2013
Safety Assessment of SiC Cladding Oxidation under Loss-of-Coolant Accident Conditions in Light Water Reactors
Youho Lee, Thomas McKrell, Chao Yue, Mujid S. Kazimi
SJR Q2FWCI 2.9Nuclear Technology

An experimental assessment was conducted of the silicon carbide (SiC) cladding oxidation rate in steam under conditions that are representative of loss-of-coolant accidents in light water reactors (LWRs). SiC oxidation tests were performed with monolithic alpha-phase tubular samples at atmospheric pressure for steam temperatures of 1140°C and 1500°C and a Reynolds number range of 40 to 330. Linear weight loss of SiC samples due to boundary layer controlled reaction of silica scale (SiO2 volatili

Materials ChemistryMaterials Science
6
7
논문|인용수 46·2021
Hydride embrittlement resistance of Zircaloy-4 and Zr-Nb alloy cladding tubes and its implications on spent fuel management
Sangbum Kim, Joo‐Hee Kang, Youho Lee
SJR Q1FWCI 2.4Journal of Nuclear MaterialsOA

In the spent fuel storage phase, nuclear fuel cladding is subjected to increased embrittlement owing to a large amount of hydride precipitation. This study compares the differences in the hydrogen-induced cladding embrittlement of cold work stress-relief annealed (CWSR or SRA) Zircaloy–4 and Zr-Nb alloy cladding with ring compression test at the temperature of the spent fuel pool, which is approximately 40 °C. Experiments demonstrate that an abrupt ductile to brittle (DTB) transition occurs at t

Materials ChemistryMaterials Science
8
논문|인용수 45·2021
Post-LOCA ductility of Cr-coated cladding and its embrittlement limit
Hyunwoo Yook, Koroush Shirvan, Bren Phillips, Youho Lee
SJR Q1FWCI 1.9Journal of Nuclear MaterialsOA

Post-Loss of Coolant Accident (LOCA) ductility assessments of Cr-coated Zircaloy cladding was conducted in compliance with the United States Nuclear Regulatory Commission (U.S.NRC)’s guidelines. The Equivalent Cladding Reacted (ECR) limit for Cr-coated cladding oxidized on both sides, representing the embrittlement process of an outer-side coated cladding near the burst hole followed by ballooning, is found to be 13.8% at steam oxidation temperature of 1204°C. Ring Compression Test (RCT) induces

Materials ChemistryMaterials Science
9
논문|인용수 32·2016
Design optimization of multi-layer Silicon Carbide cladding for light water reactors
Youho Lee, Hee Cheon No, Jeong Ik Lee
SJR Q1FWCI 2.4Nuclear Engineering and Design
Ceramics and CompositesMaterials Science
10
논문|인용수 31·2018
Development of effective thermal conductivity model for particle-type nuclear fuels randomly distributed in a matrix
Maolong Liu, Youho Lee, D.V. Rao
SJR Q1FWCI 1.3Journal of Nuclear MaterialsOA
Materials ChemistryMaterials Science
11
논문|인용수 30·2021
TRANOX: Model for non-isothermal steam oxidation of zircaloy cladding
Dongju Kim, Hyunwoo Yook, Kyunghwan Keum, Youho Lee
SJR Q1FWCI 1.8Journal of Nuclear MaterialsOA

TRANOX-1.0 (TRANsient OXidation), a mechanistic model that calculates oxygen distribution for isothermal and non-isothermal transients (1000–1250 °C) has been developed. TRANOX solves radial transient diffusion equation using Finite Difference Method (FDM) with spatial phase changes for each time step. Diffusion coefficients of each resulting phase were obtained by searching a combination that best guarantees Equivalent Cladding Reacted (ECR) and α-Zr(O) thickness simultaneously. The model has b

Materials ChemistryMaterials Science
12
논문|인용수 27·2013
THERMAL SHOCK FRACTURE OF SILICON CARBIDE AND ITS APPLICATION TO LWR FUEL CLADDING PERFORMANCE DURING REFLOOD
Youho Lee, Thomas McKrell, Mujid S. Kazimi
SJR Q2FWCI 2.9Nuclear Engineering and TechnologyOA

SiC has been under investigation as a potential cladding for LWR fuel, due to its high melting point and drastically reduced chemical reactivity with liquid water, and steam at high temperatures. As SiC is a brittle material its behavior during the reflood phase of a Loss of Coolant Accident (LOCA) is another important aspect of SiC that must be examined as part of the feasibility assessment for its application to LWR fuel rods. In this study, an experimental assessment of thermal shock performa

Ceramics and CompositesMaterials Science
13
논문|인용수 25·2023
Diffusion of chromium of Cr-coated Zircaloy accident tolerant fuel cladding: Model development and experimental validation
Dongju Kim, Youho Lee
SJR Q1FWCI 2.4Surface and Coatings Technology
Materials ChemistryMaterials Science
14
논문|인용수 25·2015
Failure probabilities of SiC clad fuel during a LOCA in public acceptable simple SMR (PASS)
Youho Lee, Ho Sik Kim, Hee Cheon No
SJR Q1FWCI 1.7Nuclear Engineering and Design
Materials ChemistryMaterials Science
15
논문|인용수 25·2023
Characterization of eutectic reaction of Cr and Cr/CrN coated zircaloy accident tolerant fuel cladding
Dongju Kim, Martin Ševeček, Youho Lee
SJR Q2FWCI 2.4Nuclear Engineering and TechnologyOA

Eutectic reactions of five kinds of Cr-coated Zr alloy cladding with different base materials (Zr–Nb–Sn alloy or Zr–Nb alloy), different coating thicknesses (6∼22.5 μm), and different coating materials (Cr single layer or Cr/CrN bilayer) were studied using Differential Scanning Calorimetry (DSC). The DSC experiments demonstrated that the onset temperatures of the Cr single layer coated specimens were almost identical to ∼1308 °C, regardless of base materials or coating thicknesses. This study de

Materials ChemistryMaterials Science

대표 연구 분야

Materials ChemistryAerospace EngineeringMechanical EngineeringCeramics and CompositesElectrical and Electronic EngineeringApplied Mathematics

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