Embrittlement behavior of zircaloy-4 cladding during oxidation and water quench

• Published in: Nuclear engineering and design 2005-01, Vol.235 (1), p.67-75
• Main Authors: Kim, Jun Hwan, Lee, Myoung Ho, Choi, Byoung Kwon, Jeong, Yong Hwan
• Format: Article
• Language: English
• Subjects: Applied sciences; Controled nuclear fusion plants; Energy; Energy. Thermal use of fuels; Exact sciences and technology; Fission nuclear power plants; Fuels; Installations for energy generation and conversion: thermal and electrical energy; Nuclear fuels
• ISSN: 0029-5493; 1872-759X
• DOI: 10.1016/j.nucengdes.2004.08.030

Simulated loss of coolant accident (LOCA) tests on Zircaloy-4 cladding were carried out to evaluate the thermal shock property during the injection of emergency core coolant. A Zircaloy-4 specimen was oxidized in a steam environment between 1000 and 1250°C followed by a flooding of the cooling water. After the test, the ductility of the thermally embrittled specimen was measured by a ring-compression test and a microstructural analysis was carried out. The results showed that the threshold equivalent cladding reacted (ECR) value to cause a failure was higher than the conventional 17% criterion calculated by the Baker–Just equation. A residual metal thickness under 0.3mm as well as a ring-compression ductility below 0.2mm for a fracture is effective to assess the thermal shock embrittlement of Zircaloy-4 in an axially unrestrained, or even in a restrained condition.