[en] Light water reactor cores may be subjected to thermal-hydraulic transients resulting in inadequate core cooling for short periods of time. In PWRs this would result in departure from nucleate boiling (DNB) and in BWRs the result would be short term dry-out at the fuel rod surface. Both situations lead to transitory temperature increases of the cladding and it is a safety requirement that after such an event reasonable fuel performance must be maintained up to the subsequent shutdown. In order to be able to assess post dry-out fuel performance, it is necessary to know what effect the transient heating has on the mechanical properties of irradiated cladding. To this end some out-of-pile experiments with pre-irradiated BWR fuel rod cladding have been earned out, which have yielded valuable information. In order to expand this database, and investigate any changes to the fuel as well as the clad, a series of dry-out experiments have been carried out at the Halden Project. The aim was to expose fresh and pre-irradiated fuel rods to short-term in-pile dry-outs of the type anticipated to occur after a pump trip in a BWR. This in-pile exposure was then to be followed by post irradiation examination (PIE) directed at discovering the microstructural changes induced in the fuel rod by an actual dry-out event and the consequence of these changes to fuel rod mechanical integrity. Further details are presented in the paper (author) (ml)