Distinguished Speakers Series presents Dr. Youho Lee: Comprehensive Studies of Zirconium Hydrides in Nuclear Fuel Cladding

Presentation Abstract

Comprehensive Studies of Zirconium Hydrides in Nuclear Fuel Cladding
 

Zirconium hydrides have been extensively studied as a primary degradation mechanism of zirconium alloy cladding in nuclear reactors. However, prior EBSD investigations have primarily focused on recrystallized material, leaving a critical gap concerning hydride characteristics in commercially used cladding materials. These materials typically possess high levels of stored energy through high dislocation densities and very fine grains, making EBSD characterization particularly challenging. This gap serves as an origin of the disconnect between laboratory research and industrial applications. Due to this disparity, the regulatory framework and associated limits for radial hydride reorientation in zirconium alloy fuel cladding remain largely empirical and unchanged, despite decades of significant research in this field.

Our study successfully overcomes these technical hurdles, allowing a detailed microstructural investigation of interconnected hydride morphology, interface types, and crystallographic orientation relationships in a widely used reactor-grade zirconium alloy, all of which collectively provide critical new insights into the hydride precipitation mechanisms. This finding challenges the prevailing assumptions about hydride interfaces in thermodynamic models of hydride reorientation. Incorporating this new microstructural information, an advanced hydride reorientation model has been developed, as well as hydrogen diffusion and precipitation kinetics. This model has been validated against a wide range of experimental results.

Additionally, embrittlement of zirconium alloys was assessed for various hydride morphologies, concentrations, and relative fractions of radial and circumferential hydrides, leading to the development of a toughness model for these alloys. The developed hydride reorientation and toughness models have been incorporated into the fuel simulation code GIFT, which is used to assess the structural integrity of zirconium alloy cladding subjected to long-term dry storage and hydride reorientation risks due to Pellet Cladding Mechanical Interaction (PCMI) during steady-state operations of large water-cooled reactors and Small Modular Reactors (SMRs).

About the Speaker

Dr. Youho Lee
Associate Professor
Department of Nuclear Engineering, Seoul National University

Professor Youho Lee specializes in nuclear fuel materials, fuel performance modeling, and reactor safety and design. His research focuses on high-performance light water reactor fuel materials, including Accident Tolerant Fuels (ATFs) and high burnup fuels, fuel performance code development, Small Modular Reactor (SMR) and fuel design, long-term storage of spent fuels, hydride embrittlement in zirconium alloys, and advanced reactors’ fuel materials.

Professor Lee earned his B.S. from the Korea Advanced Institute of Science and Technology (KAIST) in 2009, and his M.S. and Ph.D. in nuclear engineering from the Massachusetts Institute of Technology (MIT) in 2011 and 2013, respectively. He currently serves as an associate professor in the Department of Nuclear Engineering at Seoul National University (SNU). Prior to joining SNU, he worked as an assistant professor at the University of New Mexico in the U.S. from 2016 to 2019. In 2024, he chaired the national committee for the preliminary feasibility study on Advanced Fuel Technology Development of the Ministry of Trade, Industry and Energy. He is currently staying at MIT as a visiting scholar for sabbatical year.