The main phenomena limiting cladding endurance are waterside corrosion and hydrogen pick-up. The latter is closely linked to corrosion since the hydrogen absorbed by the cladding stems mainly from the corrosion process itself. Cladding corrosion is primarily a concern for PWRs because of their higher coolant temperature.

Waterside corrosion will make the cladding wall thinner and thus reduce its load-bearing capability. A thin oxide layer is in fact protective, but if it grows too much, it will crack and eventually spall. This happens when the oxide thickness

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9.13 Fuel centre temperature response to rod overpressure and clad lift-off.

approaches 100 pm, and regulation usually limits cladding corrosion to this amount (NEA, 2003).

Some fraction (10-20% for Ziy-4) of the hydrogen generated in the oxidation process is picked up and diffuses into the metal matrix. At the operating temperature of the cladding (PWR: about 350 °C), the solubility limit of hydrogen in Zircaloy is about 100 ppm. For concentrations above this amount, hydrogen precipitates as hydride plates (ZrH^ 66). In cladding tubes, hydrides will normally have a circumferential orientation. A high concentration of hydrides will cause embrittlement and weaken the load-bearing capability of the cladding. Regulation usually limits the average hydrogen concentration to 500-600 ppm (the corresponding average oxide thickness is approximately 50-60 pm).

Since the oxide of zirconium has a much lower thermal conductivity than the metal, a spot with somewhat lower temperature will arise where the oxide layer has spalled. Hydrogen that was picked up tends to migrate to these cold spots and may reach concentrations of several thousand ppm, causing severe hydriding and weakening of the cladding. It has been found that spalling with consequential hydride concentration and material embrittlement considerably lowers the failure limit in reactivity insertion accidents (Papin, 2003; Vitanza, 2007).

The nuclear industry has responded to these issues by improving the original Zry-4 (which itself is improved over Zry-2 regarding corrosion and hydrogen pick-up) and developing new alloys with better corrosion resistance.

228 Nuclear fuel cycle science and engineering