Title of article :
Failure behavior of Zircaloy-4 cladding after oxidation and water quench
Author/Authors :
Kim، نويسنده , , Jun Hwan and Lee، نويسنده , , Myoung Ho and Choi، نويسنده , , Byoung-Kwon and Jeong، نويسنده , , Yong Hwan، نويسنده ,
Issue Information :
روزنامه با شماره پیاپی سال 2007
Abstract :
Simulated LOCA (loss of coolant accident) tests and subsequent mechanical tests on Zircaloy-4 cladding were carried out to evaluate the failure behavior of the cladding. Zircaloy-4 claddings were oxidized in a steam environment from 900 to 1250 °C for a given time period followed by a flooding of cool water to simulate LOCA tests. After the simulated LOCA test, the ductility of the oxidized cladding was evaluated by mechanical tests such as ring compression test and 3-point bend test. Evaluation of the absorbed contents such as hydrogen and oxygen were also carried out. The results showed that Zircaloy-4 cladding failed during thermal shock when the ECR (equivalent cladding reacted) value exceeded 20%. Lower boundary of brittle failure at thermal shock corresponds to 20% of ECR line calculated by the Baker–Just equation regardless of test temperature. On the other hand, boundary of ductile failure by the mechanical test did not followed after the ECR line. It rapidly decreased above 1000 °C to show that all Zircaloy-4 claddings behaved brittle fracture above 1150 °C when it oxidized at 300 s. Microstructural analysis revealed that boundary of ductile failure by the mechanical test fitted well when the absorbed oxygen content inside the prior-β layer was below 0.5 wt%.
Journal title :
Journal of Nuclear Materials
Journal title :
Journal of Nuclear Materials