Development of Self-Healing Zirconium-Silicide Coatings for Improved Performance Zirconium-Alloy Fuel Cladding

Technical Report · Sat Mar 31 00:00:00 EDT 2018

DOI:https://doi.org/10.2172/1430630· OSTI ID:1430630

Sridharan, Kumar ^[1]; Mariani, Robert ^[2]; Bai, Xianming ^[2]; Xu, Peng ^[3]; Lahoda, Ed ^[3]

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1. Univ. of Wisconsin, Madison, WI (United States)
2. Idaho National Lab. (INL), Idaho Falls, ID (United States)
3. Westinghouse Electric Company LLC, Cranberry Township, PA (United States)

Zirconium-alloy fuel claddings have been used successfully in Light Water Reactors (LWR) for over four decades. However, under high temperature accident conditions, zirconium-alloys fuel claddings exhibit profuse exothermic oxidation accompanied by release of hydrogen gas due to the reaction with water/steam. Additionally, the ZrO₂ layer can undergo monoclinic to tetragonal to cubic phase transformations at high temperatures which can induce stresses and cracking. These events were unfortunately borne out in the Fukushima-Daiichi accident in in Japan in 2011. In reaction to such accident, protective oxidation-resistant coatings for zirconium-alloy fuel claddings has been extensively investigated to enhance safety margins in accidents as well as fuel performance under normal operation conditions. Such surface modification could also beneficially affect fuel rod heat transfer characteristics. Zirconium-silicide, a candidate coating material, is particularly attractive because zirconium-silicide coating is expected to bond strongly to zirconium-alloy substrate. Intermetallic compound phases of zirconium-silicide have high melting points and oxidation of zirconium silicide produces highly corrosion resistant glassy zircon (ZrSiO₄) and silica (SiO₂) which possessing self-healing qualities. Given the long-term goal of developing such coatings for use with nuclear reactor fuel cladding, this work describes results of oxidation and corrosion behavior of bulk zirconium-silicide and fabrication of zirconium-silicide coatings on zirconium-alloy test flats, tube configurations, and SiC test flats. In addition, boiling heat transfer of these modified surfaces (including ZrSi₂ coating) during clad quenching experiments is discussed in detail.

• Research Organization: Univ. of Wisconsin, Madison, WI (United States); Idaho National Lab. (INL), Idaho Falls, ID (United States)
• Sponsoring Organization: USDOE Office of Nuclear Energy (NE)
• DOE Contract Number: NE0008300
• OSTI ID: 1430630
• Report Number(s): DOE-WISC-0008300
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
Accident Tolerant Fuel Cladding
Zirconium-Silicide
Surface Coating Technologies
Light Water Reactors