Thermal-hydraulic transients in LWRs can result in short-term dry-out at the fuel rod surface with subsequent temperature increases of the cladding. In order to assess post dry-out fuel performance, it is necessary to know what effect such transient heating has on the microstructural and mechanical properties of irradiated Zircaloy cladding, and for this reason a series of dry-out experiments were carried out at the Halden Reactor Project.

Two fresh and six pre-irradiated (22 to 40 MWd/kgU) fuel rod segments were individually exposed to reduced or no-flow conditions in a heated light water loop within the Halden reactor. Surface thermocouples were used to monitor clad temperature during the dry-out events together with clad elongation detectors. Dry-out occurred over the upper region of each rod, but due to poor thermal contact between some of the thermocouples and the cladding, six of the rods were more severely tested than planned. These six rods developed peak clad temperatures (PCTs) in the range 950 to 1200°C, while PCTs of 750 to 850°C occurred in the other two rods as planned.

Clad surface condition and fuel rod dimensions were assessed post dry-out followed by destructive PIE to investigate clad microstructure and mechanical properties. Some rods exhibited clad collapse into pellet-pellet interfaces, and in the most severely tested rods the clad had undergone α to β phase transformation. The material with a quenched, former β-grain structure exhibited reduced UTS and brittle fracture in testing at room temperature, but did not fail in-pile. Significant improvement in ductility was observed in clad that had been exposed to the less severe transients, where a small α-phase grain structure was retained.