Comprehensive Assessment of Cr-coated Accident Tolerant Fuel Cladding Performance in Large PWRs and Implications for SMR Applications

Abstract

After more than a decade of research into the performance of Cr-coated Accident Tolerant Fuel (ATF) cladding, the nuclear industry is now moving toward commercialization, supported by an improved understanding of its overall behavior. In several respects—such as steady-state corrosion resistance [1], steam oxidation resistance [2,3], generally crack-resistant coating behavior at elevated temperatures relevant to normal operation, and slightly reduced inward cladding creep during steady-state [4]—Cr-coated cladding demonstrates desirable performance. It also exhibits comparable fuel ballooning and burst behavior to uncoated zirconium alloy cladding [5,6], and only limited additional oxidation through cracked coatings near burst openings [6], illuminating the promise of the concept envisioned at its inception.

However, recent findings also highlight inherent limitations that may constrain the achievable operational and safety benefits. These include: (i) degradation of coating protectiveness due to Zr diffusion from the substrate, creating oxygen ingress paths through the Cr layer [3]; (ii) formation of a eutectic phase at around 1330 °C, which degrades oxidation resistance in steam and causes a sharp increase in hydrogen generation rate, thereby limiting any meaningful increase in the peak cladding temperature limit [7,8]; and (iii) significant loss of post-LOCA ductility due to oxygen uptake and secondary hydriding near the burst opening, leaving the cladding vulnerable to traditional inner-wall oxidation mechanisms [9,10]. Furthermore, Cr-coating does not mitigate over-pressurization or Fuel Fragmentation, Relocation, and Dispersal (FFRD), which remain key safety concerns for high-burnup fuels in the 24-month cycles sought by the industry [5]. In addition, the potential burnup extensions are competed by modern zirconium alloys, whose excellent corrosion resistance and resulting reduced hydride embrittlement [11, 12] is sufficient to support discharge burnup extension of large PWRs (~75 MWd/kgU).

These concerns highlight the importance of rethinking the synergy between advanced fuel materials and reactor design, specifically aligning power density with intrinsic fuel material limits. In line with it, low power density Small Modular Reactors (SMRs) provide operating conditions in which the advantages of Cr-coated ATF can be most effectively realized. Operating at lower linear heat rates giving lower fuel temperature and offering more graceful accident scenarios with no risk of fuel burst, SMRs can fully capitalize on the superior corrosion resistance of Cr coatings. This could enable ultra-long cycles and burnup extension with LEU+ fuels, aligning with operational and economic incentives. We believe that such applications deserve increased attention and could define a future strategic direction for SMR and Cr-coated ATF deployment.

References
[1]    Kyuseok Shim, Hyuntaek Rho, Chansoo Lee, Changhyun Jo, Youho Lee. GIFT-1.0: Advanced Light Water Reactor Fuel Performance Code. Nuclear Engineering and Technology, Volume 57, Issue 9, September, 103567. 2025
[2]    Hyeongtak Kang, Dongju Kim, Martin Ševeček, Youho Lee, Parabolic Oxidation Behavior of Various Chromium-coated Zr-Nb Alloy Claddings. Journal of Nuclear Materials, 615, 155946. 2025
[3]    Dongju Kim, Youho Lee, Mechanisms of Steam Oxidation-induced Degradation of Chromium Coating on Zirconium Alloys at High Temperatures. Corrosion Science, Volume 254, 113055. 2025
[4]    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, 592, 154947. 2024
[5]    Hyunwoo Yook, Sunghoon Joung, Chansoo Lee, Youho Lee. Integral LOCA experiments to study FFRD behavior of high burnup nuclear fuels. Nuclear Engineering and Design, 429, 113633. 2024
[6]    Hyunwoo Yook, Sunghoon Joung, Youho Lee, Post-Ballooning and Burst Steam Oxidation of Accident Tolerant Zirconium Alloy Cladding with Cracked Chromium Coating. Journal of Nuclear Materials, 616, 156095. 2025
[7]    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 the Structural Integrity of Cladding. Journal of Nuclear Materials, 155577. 2024
[8]    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, 55, 3535-3542.
[9]    Hyunwoo Yook, Koroush Shirvan, Bren Phillips, Youho Lee. Post-LOCA Ductility of Cr-coated cladding and its embrittlement limit. Journal of Nuclear Materials, 153354. 2022
[10]    SungHoon Joung, Jinsu Kim, Martin Ševeček, Juri Stuckert, Youho Lee. Post-quench ductility limits of coated ATF with various base zirconium-based alloys and coating designs. Journal of Nuclear Materials, 591, 154915. 2024 
[11]    Changhyun Jo, Donghyeon Son, Youho Lee, Microstructure-informed modeling of radial hydride precipitation in reactor-grade PRXA Zirconium alloy cladding tube. Journal of Nuclear Materials, 156014. 2025
[12]    Dahyeon Woo, Changhyun Jo, Joo-Hee Kang, Jarugula Rajesh, Michael Preuss, Youho Lee, Microstructural analysis of interconnected δ-hydride structures in a zirconium alloy. Acta Materialia, 301, 121473. 2025
 

Biography

Youho Lee, Associate Professor in the Department of Nuclear Engineering at Seoul National University (SNU), specializes in nuclear fuel materials and nuclear reactor engineering. His research focuses on advanced light water reactor (LWR) fuels, including Accident Tolerant Fuels (ATFs), LEU+, high burnup fuels, SMR design and analysis, zirconium alloys, and nuclear fuel cycle policy. He is currently spending a sabbatical year at the Massachusetts Institute of Technology (MIT).

He served as the General Secretary of the Korean Nuclear Society from 2021 to 2023 and has been serving as Secretary of Public Relations since 2023. In 2024, he chaired the committee for the preliminary feasibility study on Advanced Fuel Technology Development of the Ministry of Trade, Industry and Energy. He is also currently a member of the IAEA Programme Committee for the Technical Meeting on Advanced Technology Fuels.

Prior to joining SNU, Lee was an Assistant Professor in the Department of Nuclear Engineering at the University of New Mexico in the United States from 2016 to 2019. Youho Lee earned his BS in nuclear and quantum engineering from the Korea Advanced Institute of Science and Technology (KAIST) in 2009, and his MS and PhD in nuclear engineering from MIT in 2011 and 2013, respectively.