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

被引:19
|
作者
Kim, JH [1 ]
Lee, MH [1 ]
Choi, BK [1 ]
Jeong, YH [1 ]
机构
[1] Korea Atom Energy Res Inst, Zirconium Fuel Cladding Team, Taejon 305600, South Korea
关键词
D O I
10.1016/j.nucengdes.2004.08.030
中图分类号
TL [原子能技术]; O571 [原子核物理学];
学科分类号
0827 ; 082701 ;
摘要
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 degreesC 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 0.3 mm as well as a ring-compression ductility below 0.2 mm for a fracture is effective to assess the thermal shock embrittlement of Zircaloy-4 in an axially unrestrained, or even in a restrained condition. (C) 2004 Elsevier B.V. All rights reserved.
引用
收藏
页码:67 / 75
页数:9
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