The extended use of Zircaloy cladding in light water reactors degrades its mechanical properties by a combination of irradiation embrittlement, coolant-side oxidation, hydrogen pickup, and hydride formation. The hydrides are usually concentrated in the form of a dense layer or rim near the cooler outer surface of the cladding, Utilizing plane-strain ring-stretch tests to approximate the loading path in a reactivity-initiated accident (RIA) transient, we examined the influence of a hydride rim on the fracture behavior of unirradiated Zircaloy-4 cladding at room temperature and 300°C. Failure is sensitive to hydride-rim thickness such that cladding tubes with a hydride-rim thickness > 100 μm (≈700 wppm total hydrogen) exhibit brittle behavior, while those with a thickness <90 μm (≈600 wppm) remain ductile. The mechanism of failure is identified as strain-induced crack initiation within the hydride rim and failure within the uncracked ligament due to either a shear instability or damage-induced fracture. We also report some preliminary results of the uniaxial tensile behavior of low-Sn Zircaloy-4 cladding tubes in a cold-worked, stress-relieved condition in the transverse (hoop) direction at strain rates of 0.001/s and 0.2/s and temperatures of 26 to 400°.
|Original language||English (US)|
|Number of pages||17|
|Journal||ASTM Special Technical Publication|
|State||Published - 2002|
|Event||Zirconium in the Nuclear Industry: Thirteenth International Symposium - Annecy, France|
Duration: Jun 10 2001 → Oct 14 2001
All Science Journal Classification (ASJC) codes