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Upcomming...

2024

Changhyun Jo, Dahyeon Woo, *Youho Lee. Development of an advanced hydride reorientation model for Zircaloy cladding and its experimental validation. (Accepted, Journal of Nuclear Materials)

2024

2024

Donghyeon Son, *Youho Lee. Experimental Investigation and Model Development for Recrystallization and Grain Growth Behavior of Reactor-Grade Zr-Nb-Sn Alloy at the Temperatures 500-700°C. ​(Submitted, under review)

Prince Setia, Dongju Kim, *Youho Lee, Novel Method for Accurate Stored Energy Measurement in Deformed Reactor-Grade Structural Materials. (Submitted, under review)

2024

SungHoon Joung, Hyunwoo Yook, Dongju Kim, *Youho Lee, Exploring the Peak Cladding Temperature Limit of Cr-Coated ATF by assessing the impact of the Zr-Cr eutectic on structural integrity of cladding. (Submitted, under review)

2024

Kyuseok Shim, Hyuntaek Rho, Chansoo Lee, Changhyun Jo, *Youho Lee, GIFT-1.0: Advanced Light Water Reactor Fuel Performance Code. (Submitted, under review)

2024

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Integral LOCA experiments to study FFRD behavior of high burnup nuclear fuels. 

Nuclear Engineering and Design

Hyunwoo Yook, Sunghoon Joung, Chansoo Lee 
Youho Lee*

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Ho-A Kim, Ju-Seong Kim, Yundong Lee, Youho Lee, Yong-Soo Kim, Joo-Hee Kang,Sangtae Kim

Microstructure and mechanical properties of hydride blisters formed on Zircaloy-4 claddings, Journal of Nuclear Materials

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SungHoon Joung, Jinsu Kim, Martin Ševeček, Juri Stuckert, Youho Lee

Post-quench ductility limits of coated ATF with various base zirconium-based alloysand coating designs, 

Journal of Nuclear Materials

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Chansoo Lee, *Youho Lee

Development of a coupled fuel (GIFT) and thermal (COBRA-SFS) analysis code for dry storage analysis and its application for the spent fuel safety assessments, Nuclear Engineering and Design

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Jinsu Kim, Chung Yong Lee, Hyuntaek Rho, Hun Jang, *Youho Lee

Elucidating changes in thermal creep rate of Zircaloy Accident Tolerant Fuel (ATF) cladding with thin chromium coating via experiment and mechanical analysis, Journal of Nuclear Materials

2023

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Jiwon Park, Kil-Won Moon, Chang-Seok Oh, Joo-Hee Kang, Jeki Jung, Youho Lee, Dongju Kim

Determination of onset temperature of melting in binary alloys using root test in differential scanning calorimetry, 

Journal of Thermal Analysis and Calorimetry, In Press

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Jinsu Kim, *Youho Lee

Effect of Cr coating on the mechanical integrity of Accident Tolerant Fuel cladding under ring compression test, Journal of Nuclear Materials

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Dongju Kim, *Youho Lee

Chromium Diffusion Model for Coated Zircaloy Accident Tolerant Fuel Cladding: Development and Experimental Validation, Surface and Coatings Technology

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Sang-Uk Lee, Yuyeon Kim, Youho Lee, Kyoung-Jae Chung

A plasma-based Pressure Pulse Generator for Simulating Pellet-Cladding Mechanical Interaction Druing Reactivity-Initiated Accident, Review of Scientific Instruments

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Dongju Kim, Martin Sevecek, *Youho Lee

Characterization of Eutectic Reaction of Cr and Cr/CrN coated Zircaloy Accident Tolerant Fuel Cladding, 

Nuclear Engineering and Technology, In Press

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Dahyeon Woo, *Youho Lee

Understanding the mechanical integrity of Zircaloy cladding with various radial and circumferential hydride morphologies via image analysis, 

Journal of Nuclear Materials

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Dongju Kim, Martin Steinbrück, Mirco Grosse, Chongchong Tang, Youho Lee

Eutectic Reaction and Oxidation Behavior of Cr-coated Zircaloy-4 Accident-tolerant Fuel Cladding under Various Heating Rates, Journal of Nuclear Materials

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Sangbum Kim, Donghyeon Son, Joo-Hee Kang, *Youho Lee

Sangbum Kim, Donghyeon Son, Joo-Hee Kang, Youho Lee*. Recrystallization and grain growth of Zr-Nb-Sn alloy in 400-500°C and effect on hydride embrittlement, Journal of Nuclear Materials

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Sangbum Kim, Joo-Hee Kang, *Youho Lee

Suppressed hydride precipitation in the welding zone of Zircaloy cladding tube, Journal of Nuclear Materials

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Chansoo Lee, *Youho Lee

Simulation of hydrogen diffusion along the axial direction in Zirconium cladding tube during dry Storage, Journal of Nuclear Materials

2022

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Dongyeon Kim, Dahyeon Woo, *Youho Lee

Radial hydride fraction with various rod internal pressures and hydrogen contents for Zr-Nb alloy cladding tube 

Journal of Nuclear Materials

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Dahyeon Woo, *Youho Lee

Spent fuel simulation during dry storage via FRAPCON-4.0 Modification: comparison between PWR and SMR and discharge burnup effect, 

Nuclear Engineering and Technology

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Hyunwoo Yook, *Youho Lee

Post-LOCA ductility assessment of Zr-Nb Alloy from 1100°C to 1300°C to explore variable peak cladding temperature and equivalent cladding reacted safety criteria, 

Journal of Nuclear Materials

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Donguk Kim, Dongyeon Kim, Dahyeon Woo, *Youho Lee

Development of an image analysis code for hydrided Zircaloy using Dijkstra’s algorithm and sensitivity analysis of radial hydride continuous path, 

Journal of Nuclear Materials

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Dongju Kim, Hun Jang, Daegyun Ko *Youho Lee

Study of high burnup effect on steam oxidation of Zircaloy and its regulatory implications via the development of pre-transient oxide model of TRANOX, 

Journal of Nuclear Materials

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Dongyeon Kim, Joo-Hee Kang, *Youho Lee

Accurate prediction of threshold stress for hydride reorientation in Zircaloy-4 with directly measured interface orientation relationship, 

Materialia

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Sangbum Kim, Joo-Hee Kang, *Youho Lee

Hydride embrittlement resistance of Zircaloy-4 and Zr-Nb alloy cladding tubes and its implications on spent fuel management, 

Journal of Nuclear Materials

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Hyunwoo Yook, Koroush Shirvan, Bren Phillips, *Youho Lee

Post-LOCA Ductility of Cr-coated cladding and its embrittlement limit

Journal of Nuclear Materials

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Hyuntaek Rho, *Youho Lee

 Development of a 2D Axisymmetric SiC Cladding Mechanical Model and its Applications for Steady-State and LBLOCA Analysis, 

Journal of Nuclear Materials

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​Shinhyo Bang, Ho-a Kim, Jae-soo Noh, Donguk Kim, Kyunghwan Keum, *Youho Lee

Temperature-dependent axial mechanical properties of Zircaloy-4 with various hydrogen amounts and hydride orientations, 

Nuclear Enginnering and Technology

2021

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Jaewon Choi, Ilsoon Hwang, *Youho Lee

Flow Accelerated Corrosion of Stainless Steel 316L by a Rotating Disk in Lead-Bismuth Eutectic Melt, 

Journal of Minerals, Metals & Materials Society (TMS)

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Dongju Kim, Hyunwoo Yuk, Kyunghwan Keum, *Youho Lee

TRANOX: Model for Non-Isothermal Steam Oxidation of Zircaloy Cladding, 

Journal of Nuclear Materials

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Shinhyo Bang, *Youho Lee

The Statistical Ductility of Zircaloy-4 and its Regulatory Implications, 

Journal of Nuclear Materials

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​Maolong Liu, Lang Wang, Youho Lee

Diagnosis of break size and location in LOCA and SGTR accidents using support vector machine, 

Journal of Nuclear Materials

2020

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​Kyunghwan Keum, *Youho Lee

Effect of Cooling Rate on the Residual Ductility of Post-LOCA Zircaloy-4 Cladding, 

Journal of Nuclear Materials

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Amir Ali, Hyun-Gil Kim, Khalid Hattar, Samuel Briggs, Dong Jun Park, Jung Hwang Park, *Youho Lee

Ion irradiation effects on Cr-coated zircaloy-4 surface wettability and pool boiling critical heat flux, 

Nuclear Engineering and Design

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Soon Lee, Youho Lee (co-first author), Nicholas Brown, Kurt Terrani

Elucidating the Impact of Flow on Material-Sensitive Critical Heat Flux and Boiling Heat Transfer Coefficients: An Experimental Study with Various Materials, 

International Journal of Heat and Mass Transfer

2019

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Seong Gu Kim, Yacine Addad, Maolong Liu, Jeong Ik Lee, *Youho Lee

Computational Investigation into Heat Transfer Coefficients of Randomly Packed Pebbles in Flowing FLiBe, 

International Journal of Heat and Mass Transfer

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Mingfu He, *Youho Lee

Revisiting heater size sensitive pool boiling critical heat flux using neural network modeling: Heater length of the half of the Rayleigh-Taylor Instability Wavelength maximizes CHF, 

Thermal Science and Engineering Progress

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Soon Lee, Maolong Liu, Nicholas Brown, Kurt, Terrani, *Youho Lee

Effect of Heater Material and Thickness on the Steady-state Flow Boiling Critical Heat Flux, 

Nuclear Technology (Invited for the Nuclear Technology special issue of Advanced Thermal Hydraulics 2018)

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Robby Christian, Youho Lee, Hyun Gook Kang

Comparison of Experimental and Simulated Critical Heat Flux Tests with Various Cladding Alloys: Sensitivity of Iron-Chromium-Aluminum (FeCrAl) to Heat Transfer Coefficients and Material Properties, 

Nuclear Engineering and Design

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Robby Christian, Youho Lee, Hyun Gook Kang

Emergency Core Cooling System Performance Criteria for Multi-Layered Silicon Carbide Nuclear Fuel Cladding, 

Nuclear Engineering and Design

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Dongjune Chang, Maolong Liu, *Youho Lee

Accident Diagnosis of a PWR Fuel Pin during Unprotected Loss of Flow Accident with Support Vector Machine, 

Nuclear Engineering and Design

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Mingfu He, *Youho Lee

Application of Deep Belief Network for Critical Heat Flux Prediction on Microstructure Surfaces, 

Nuclear Technology (Invited for the Nuclear Technology special issue of Advanced Thermal Hydraulics 2018)

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Maolong Liu, Jared Thurgood, Dasari Rao, *Youho Lee

Development of A Two-Regime Heat Conduction Model for TRISO-based Nuclear Fuels, 

Journal of Nuclear Materials

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Soon Lee, Maolong Liu, Nicholas Brown, Kurt Terrani, Colby Jensen, Heng Ban, Edward Blandford, *Youho Lee

Comparison of steady and transient flow boiling critical heat flux for FeCrAl accident tolerant fuel cladding alloy, Zircaloy, and Inconel, 

International Journal of Heat and Mass Transfer

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Maolong Liu, Saetbyul Jung, Hyungdae Kim, *Youho Lee

Experimental and Analytical Investigation into Boiling Induced Thermal Stress: Its Impact on the Stress State of Oxide Scales of Nuclear Component, 

Nuclear Engineering and Design

2018

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Mingfu He, Youho Lee*

Application of machine learning for prediction of critical heat flux: Support vector machine for data-driven CHF look-up table construction based on sparingly distributed training data points, 

Nuclear Engineering and Design

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Maolong Liu, Youho Lee*, Dasari V. Rao

Development of Effective Thermal Conductivity Model for Particle-Type Nuclear Fuels Randomly Distributed in a Matrix, 

Journal of Nuclear Materials

2017

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Youho Lee, Jeong Ik Lee, Hee Cheon NO

Mechanical Analysis of Surface-Coated Zircaloy Cladding, 

Nuclear Engineering and Technology

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Youho Lee, Hee Cheon NO, Jeong Ik Lee

Design Optimization of Multi-Layer Silicon Carbide Cladding for Light Water Reactors, 

Nuclear Engineering and Design

2016

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Youho Lee, Bokyung Kim, Hee Cheon NO

Improving Safety Margin of LWRs by Rethinking the Emergency Core Cooling System Criteria and Safety System Capacity, 

Nuclear Engineering and Design

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Seong Gu Kim, Youho Lee, Yoonhan Ahn, Jeong Ik Lee

CFD aided approach to heat exchangers for supercritical CO2 Brayton cycle application, 

Annals of Nuclear Energy

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Youho Lee, Jeong Ik Lee, Hee Cheon NO

Impacts of Transient Heat Transfer Modelling on Prediction of Advanced Cladding Fracture during LWR LBLOCA, 

Nuclear Engineering and Design

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Youho Lee, Thomas J. McKrell, Mujid S. Kazimi

Oxidation Behavior of Sintered Tubular Silicon Carbide in Pure Steam ІІ: Weight-Loss Correlation Developments, 

Ceramics International

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Youho Lee, Thomas J. McKrell, Aline Montecot, Michael Pantano, Yann Song, Mujid S. Kazimi

Oxidation Behavior of Sintered Tubular Silicon Carbide in Pure Steam І: Experiments, 

Ceramics International

2015

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Youho Lee, Thomas J. McKrell, Mujid S. Kazimi

Thermal Shock Fracture of Hot Silicon Carbide Immersed in Water, 

Journal of Nuclear Materials

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Youho Lee, Ho Sik Kim,
Hee Cheon NO

Failure probabilities of SiC Clad Fuel during a LOCA in Public Acceptable Simple SMR (PASS), 

Nuclear Engineering and Design

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Seongmin Son, Youho Lee, Jeong Ik, Lee

Development of an Advanced Printed Circuit Heat Exchanger Analysis Code for Realistic Flow Path Configurations near Header Regions, 

International Journal of Heat and Mass Transfer

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Youho Lee, Jeong Ik Lee

A Structural Model for Multi-Layered Ceramic Cylinders and its Application to Silicon Carbide Cladding of Light Water Reactor Fuel, 

Journal of Nuclear Materials

2014

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Youho Lee, Jeong Ik Lee

Structural Assessment of Intermediate Printed Circuit Heat Exchanger for Sodium-Cooled Fast Reactor with Supercritical CO2 Cycle, 

Annals of Nuclear Energy

2013

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Youho Lee, Thomas J. McKrell, Mujid S. Kazimi

Thermal Shock Fracture of Silicon Carbide and Its Application to LWR Fuel Cladding Performance during Reflood, 

Nuclear Engineering and Technology

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Youho Lee, Thomas J. McKrell,
Chao Yue, Mujid S. Kazimi

Safety Assessment of SiC Cladding Oxidation Under Loss of Coolant Accident (LOCA) Conditions in LWRs, 

Nuclear Technology

2010

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Youho Lee, Jeong Ik Lee,
Hee Cheon NO

A Point Model for the Design of a Sulfur Trioxide. Decomposer for the SI Cycle and Comparison with a CFD Model, 

International Jouranl of Hydrogen Energy

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