锆合金包壳I-SCC性能评价  

Evaluation on I-SCC Properties of Zirconium Cladding

作　　者：闫萌[1] 王朋飞[1] 洪晓峰[1] 梁波[1] 戴训[1] Yan Meng Wang Pengfei Hong Xiaofeng Liang Bo Dai Xun(Science and Technology on Reactor Fuel and Materials Laboratory, Nuclear Power Institute of China, Chengdu, 610213, China)

机构地区：[1]中国核动力研究设计院反应堆燃料及材料重点实验室,成都610213

出　　处：《核动力工程》2017年第5期138-140,共3页Nuclear Power Engineering

摘　　要：对N36、Zr-4、X锆合金包壳管环形试样在350、400℃下施加周向拉伸载荷,研究N36锆合金包壳管在10~2 Pa、10~3 Pa、10~4 Pa碘分压、Zr-4及X试样在102Pa碘分压下的碘致应力腐蚀开裂行为。研究发现:在350、400℃下以最大载荷为指标时,N36、Zr-4及X试样在一定碘分压环境中均会发生不同程度的碘致应力腐蚀开裂,断裂能量迅速下降;在相同试验条件下,N36试样的最大载荷和断裂能量下降最慢。Hoop tensile test were performed on zirconium （such as N36, Zr-4 and X） ring specimens at 350% and 400% to study the iodine induced stress corrosion cracking （I-SCC） behavior of zirconium cladding in 102 Pa, 103 Pa and 104 Pa iodine partial pressure environment. Different levels of I-SCC happened with N36, Zr-4 and X samples at special iodine partial pressure with the maximum load, and the fracture energy dropped down quickly as result, and N36 maximum load and fracture energy dropped slower at same test situation.

关 键 词：碘致应力腐蚀开裂 锆合金包壳 周向拉伸 

分 类 号：TG174.34[金属学及工艺—金属表面处理]