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

Abstract

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. Development of Self-Healing Zirconium-Silicide Coatings for Improved Performance Zirconium-Alloy Fuel Cladding DE-NE0008300 Final Report 5 was also confirmed that the theoretical composition coating enhanced oxidation protection and wear resistance. Distinctive oxide layers of ZrSi2 prepared at 1000 °C and 1400 °C in ambient air were subjected to a 3.9 MeV Si 2+ ions irradiation at 305 °C and their radiation responses were characterized and analyzed. Nanocomposite oxides consisting of ZrO2 nanocrystals embedded in amorphous SiO2 matrix formed on ZrSi2 surface after oxidation at 1000 °C. Radiation-induced phase mixing of the oxide phases and amorphization of ZrO2 was observed up to ~ 820 nm in depth, about one-third of the total radiation damaged region (2.5 µm in thickness) as estimated by SRIM calculation. The ion-mixing was facilitated with fine ZrO2 grains (< 25 nm) at low doses (5 to 10 dpa) but not in larger grains (> 30 nm) even at high damage levels (30 to 60 dpa). Qualitative explanation of the ion-mixing of the nanocomposite was proposed with respect to grain size of ZrO2 in SiO2 matrix. Polygonal crystalline ZrSiO4 grains in dual-layered oxide scale on ZrSi2 at 1400 °C were completely amorphized under the ion-irradiation. Given the high corrosion resistance of ZrSiO4 and immobil...