Corrosion Resistance of Traditional and Advanced Fuel Rod Cladding Materials for Water-Cooled Reactors

Abstract

The available literature experimental data on corrosion resistance of traditional and advanced fuel rod cladding materials for water-cooled reactors are summarized. A review of zirconium alloys, which have proven themselves in operation for more than half a century, is presented. As noted, the research work is constantly being carried out to improve zirconium alloys by optimizing their composition, in particular, the amount of tin, niobium, iron and oxygen, as well as development of the new alloys. First of all, the direction of these works is stimulated by stringent nuclear energy requirements, including maximum safety, efficiency and environmental friendliness. At the same time, in the last decade, one of the main goals of researchers around the world is the development of nuclear fuel systems, which tolerate severe accidents. Another trigger for this was the accident in 2011 in Japan at the Fukushima-1 NPP. As the most optimal possible solution, it is considered the surface modification of zirconium alloys by the development of chromium coatings. Such coatings provide an increased corrosion resistance and wear resistance, as well as hydrogen pickup reduced at operating temperatures of the primary coolant and in emergencies. A more radical way to increase the fuel rod cladding accident resistance is to replace the zirconium alloy with another one. The best candidates are FeCrAl alloys and duplex stainless steels (DSS), whose corrosion resistance can be 50 times greater than that of zirconium alloys in loss-of-coolant accident (LOCA) conditions. Unfortunately, under the nominal water-cooled reactor operating conditions, a long-term operation such as claddings will lead to the corrosion product formation and its removal to the coolant followed by their activation and formation of deposits in the core and steam generator. This will certainly entail an increase in the radiation-dose rate from the primary circuit equipment. Considering the traditional and advanced water-cooled reactors claddings, which tolerate severe accident scenarios, an optimized zirconium alloy with chromium coating can be considered as the most advanced one. The corrosion resistance of such claddings is at least five times higher compared to traditional zirconium alloys both under normal operating conditions and severe accidents, and will not cause significant neutron absorption, coolant activation or deposit formation in the primary circuit.

Keywords

• zirconium alloys
• fuel rod
• nuclear fuel
• water-cooled reactor
• corrosion