Oxidation of Advanced Zirconium Cladding Alloys in Steam at Temperatures in the Range of 600–1200 °C

Abstract

The oxidation kinetics of the classical pressurized water reactors (PWR) cladding alloy Zircaloy-4 have been extensively investigated over a wide temperature range from operational conditions to beyond design basis accident (BDBA) temperatures. In recent years, new cladding alloys optimized for longer operation and higher burn-up are used in Western light water reactors (LWR). This paper presents the results of thermo-gravimetric tests with Zircaloy-4 as the reference material, Duplex DX-D4, M5 Ò (both AREVA), ZIRLO TM (Westinghouse), and the Russian E110 alloy. All materials were investigated in isothermal and transient tests in a thermal balance with steam furnace. Post-test analyses were performed by lightmicroscopy and neutron radiography for investigation of the hydrogen absorbed by the metal. Strong and varying differences (up to 800%) in oxidation kinetics between the alloys were found at up to 1000°C, where the breakaway effect plays a role. Less but significant differences (ca. 30%) were observed at 1100 and 1200°C. Generally, the M5 Ò alloy revealed the lowest oxidation rate over the temperature range investigated whereas the behavior of the other alloys was considerably dependent on temperature. A strong correlation was found between oxide scale structure and amount of absorbed hydrogen.

Air oxidation of Zircaloy-4, M5® and ZIRLO™ cladding alloys at high temperatures

“…3 impressively show the strong degradation of the cladding tube segments at 1273 K (as an example), which is much more severe than the degradation during steam oxidation [15]. At this temperature, the barrier effect is lost after one hour of oxidation at the latest.…”

“…This is the third paper of a series containing results of comparative investigations of various cladding alloys. The oxidation of zirconium alloys in oxygen [18] and in steam [15] has been published elsewhere.…”

Cold spray deposition of Ti2AlC coatings for improved nuclear fuel cladding

“…However, there are a few well-acknowledged limitations of zirconium-alloy claddings, such as the formation of a thick oxide layer and associated hydriding effects and susceptibility to fretting wear at grid spacers [3,4]. In the event of a Loss of Coolant Accident (LOCA), zirconium alloys at elevated temperatures can oxidize via an exothermic reaction that is accompanied by hydrogen production as a result of reaction of zirconium with steam [5]. One approach that is being investigated to address this problem is to develop oxidation resistant coatings for zirconium-alloy cladding that can provide the necessary protection during an off-normal high temperature or LOCA conditions, while potentially enhancing performance under normal operating conditions.…”

High-Temperature Oxidation of Zircaloy-4 in Oxygen–Nitrogen Mixtures

“…[2][3][4][5][6][7]. The oxidation in steam, the most relevant atmosphere in nuclear power plants, is described by cubic (for temperatures below 1000°C) or parabolic (at temperatures beyond 1000°C) kinetics determined by the diffusion of oxygen (vacancies) through the oxide scale and in the metal.…”