Thermal annealing of irradiated prehydrided zircaloy-4 materials

Hydrogen-assisted irradiation growth is an important issue for reliability and performance of nuclear materials in nuclear fuel cladding and fuel assembly components. The mechanisms of irradiation growth are not entirely clear and the role of hydrogen and hydrides are less so. There is evidence that breakaway irradiation growth is strongly influenced by the formation and dynamics of dislocation loops that vary with alloy composition, heat treatment and hydrogen content. To study this phenomenon, thermal annealing of eight Zircaloy-4 specimens irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory was performed using various techniques. Transmission electron microscopy (TEM) with a hot stage was used for the quantification of <a> and <c> dislocation loop population densities and size distributions at a range of temperatures from 25 to 800^oC. Differential scanning calorimetry (DSC) and dilatometry were performed by dynamically heating and cooling from 30 to 750ºC at rates of 5, 10, and 20 K per minute (Kpm) to determine the behavior of dislocation loop annealing. The results are discussed in the context of the larger AsTeR program that will include higher exposure (~20 dpa) and different alloys when they become available in 2021.