Zircaloy Cladding Rupture During Repository Storage

Creep, a potential degradation mechanism of Zircaloy cladding after repository disposal of spent nuclear fuel, is investigated. The deformation and fracture map methodology is used to predict maximum allowable initial storage temperatures to achieve a 1000-yr life without rupture as a function of spent-fuel history. Maximum allowable temperatures are 340°C (613 K) for typically stressed rods (70 to 100 MPa) and 300°C (573 K) for highly stressed rods (140 to 160 MPa).