• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.232.2-232.2
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.234.1-234.1
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.216.2-216
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.250-250
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.363.1-363.1
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.310.1-310.1
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• 2003

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.1-13
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• 2020

At present, the Department of Energy (DOE) in Unite State are directing the efforts of developing accident tolerant fuel (ATF) technology. As the first barrier of nuclear fuel system, the material selection of fuel rod cladding for ATFs is a basic but very significant issue for the development of this concept. The advanced cladding is attractive for providing much stronger oxidation resistance and better in-pile behavior under sever accident conditions (such as SBO, LOCA) for giving more coping time and, of course, at least an equivalent performance under normal condition. In recent years, many researches on in-plie or out-pile physical properties of some suggested cladding materials have been conducted to solve this material selection problem. Base on published literatures, this paper introduced relevant research backgrounds, objectives, research institutions and their progresses on several main potential claddings include triplex SiC, FeCrAl and MAX phase material Ti3SiC2. The physical properties of these claddings for their application in ATF area are also reviewed in thermohydraulic and mechanical view for better understanding and simulating the behaviors of these new claddings. While most of important data are available from publications, there are still many relevant properties are lacking for the evaluations.

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.135-147
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• 2020

A methodology for evaluation of mechanical and thermal effects of localized non-axisymmetric oxidation in zircaloy claddings on LWR fuel reliability is proposed. To this end, the basic capabilities of the FALCON fuel behaviour code are used. Examples of methodology application to adjustment of selected operational limits for modern BWR fuel rods, to capture effects of the excess local oxidation, are presented. Specifically, the limiting rod internal pressure for the onset of cladding lift-off is reduced, depending on initial excess oxidation spot sizes. Also, the power limits for Anticipated Operational Occurrences are adjusted, to preclude fuel melting and cladding failure due to PCMI and PCI-SCC in the affected fuel rods.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.1
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• pp.115-147
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• 2023

The Fukushima accident in 2011 revealed some major flaws in traditional nuclear fuel materials under accidental conditions. Thus, the focus of research has shifted toward "accident tolerant fuel" (ATF). The aim of this approach is to develop fuel material solutions that lead to improved reactor safety. The application of protective coatings on the surface of nuclear fuel cladding has been proposed as a near-term solution within the ATF framework. Many coating materials are being developed and evaluated. In this article, an overview of different zirconium-based alloys currently in use in the nuclear industry is provided, and their performances in normal and accidental conditions are discussed. Coating materials proposed by different institutions and organizations, their performances under different conditions simulating nuclear reactor environments are reviewed. The strengths and weaknesses of these coatings are highlighted, and the challenges addressed by different studies are summarized, providing a basis for future research. Finally, technologies and methods used to synthesize thin-film coatings are outlined.

• Corrosion Science and Technology
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• v.6 no.4
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• pp.170-176
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• 2007

Iodine-induced stress corrosion cracking (ISCC) properties and the associated ISCC process of Zircaloy-4 and an Nb-containing advanced nuclear fuel cladding were evaluated. An internal pressurization test with a pre-cracked specimen was performed with a stress-relieved (SR) or recrystallized (RX) microstructure at $350^{\circ}C$, in an iodine environment. The results showed that the $K_{ISCC}$ of the SR and RX Zircaloy-4 claddings were 3.3 and 4.8MPa\;m^{0.5}, respectively. And the crack propagation rate of the RX Zircaloy-4 was 10 times lower than that of the SR one. The chemical effect of iodine on the crack propagation rate was very high, which was increased $10^4$ times by iodine addition. Main factor affecting on the micro-crack nucleation was a pitting formation and its agglomeration along the grain boundary. However, this pitting formation on the grain-boundary was suppressed in the case of an Nb addition, which resulted in an increase of the ISCC resistance when compared to Zircaloy-4. Crack initiation and propagation mechanisms of fuel claddings were proposed by a grain boundary pitting model and a pitting assisted slip cleavage model and they showed reasonable results.