Youho Lee
서울대학교 핵공학과 · 재료과학
이 교수의 연구실은 nuclear fuel cladding의 안전성과 내구성을 향상시키기 위한 고도화된 재료 과학 및 열역학적 분석을 중심으로 활동하고 있습니다. 주요 연구 방향은 Zircaloy 및 SiC 기반 클래딩 재료의 고온 산화 거동, 피로 및 연성 저하 현상, 특히 임계 열부하(CHA) 및 LOCA(냉각수 손실 사고) 조건에서의 거동을 기계학습과 실험적 평가를 융합하여 규명하는 것입니다. 또한 고배율 연료의 수명 연장과 습기 저장, 건식 저장 조건에서의 수소화합물 응집에 의한 취성 전이 메커니즘에 대한 미세구조적 분석도 핵심 과제입니다.
표시된 성과는 수집된 데이터 기준으로 산출되며, 일부 차이가 있을 수 있습니다.
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
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
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
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
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
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
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
