Comparison of ODSCC behavior between Alloy 600 and Alloy 690 under secondary side environments

机译：二次侧环境下Alloy 600和690 Alloy之间的ODSCC行为比较

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摘要

Alloy 690TT are usually used as steam generator tubing materials in newly constructing nuclear power plants because of excellent stress corrosion cracking(SCC)resistance in normal primary and secondary side bulk water conditions.

机译：690TT合金通常在新建的核电站中用作蒸汽发生器的管道材料，因为在正常的一次和二次侧大容量水条件下，其具有出色的抗应力腐蚀开裂（SCC）性能。

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《5th International Symposium on Materials and Reliability in Nuclear Power Plants amp; Symposium on Water Chemistry and Corrosion in Nuclear Power Plants in Asia-2017 amp; 3rd Asian Forum on Material Aging I》|2017年|16|共1页
会议地点 Shenyang (CN)
作者
Dong-Jin Kim; Eun Hee Lee; Hong Pyo Kim; Yun Soo Lim; Seong Sik Hwang; Sung-Woo Kim; Min Jae Choi; Jong Yeon Lee;
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Nuclear Materials Research Division,Korea Atomic Energy Research Institute,111,Daedeok-daero 989beon-gil,Yuseong-gu,Daejeon 34057,Korea;

Nuclear Materials Research Division,Korea Atomic Energy Research Institute,111,Daedeok-daero 989beon-gil,Yuseong-gu,Daejeon 34057,Korea;

Nuclear Materials Research Division,Korea Atomic Energy Research Institute,111,Daedeok-daero 989beon-gil,Yuseong-gu,Daejeon 34057,Korea;

Nuclear Materials Research Division,Korea Atomic Energy Research Institute,111,Daedeok-daero 989beon-gil,Yuseong-gu,Daejeon 34057,Korea;

Nuclear Materials Research Division,Korea Atomic Energy Research Institute,111,Daedeok-daero 989beon-gil,Yuseong-gu,Daejeon 34057,Korea;

Nuclear Materials Research Division,Korea Atomic Energy Research Institute,111,Daedeok-daero 989beon-gil,Yuseong-gu,Daejeon 34057,Korea;

Nuclear Materials Research Division,Korea Atomic Energy Research Institute,111,Daedeok-daero 989beon-gil,Yuseong-gu,Daejeon 34057,Korea;

Nuclear Materials Research Division,Korea Atomic Energy Research Institute,111,Daedeok-daero 989beon-gil,Yuseong-gu,Daejeon 34057,Korea;

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正文语种 eng
中图分类电工技术;原子能技术;
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专利

1. Influence of pH on the chemical and structural properties of the oxide films formed on 316L stainless steel, alloy 600 and alloy 690 in high temperature aqueous environments [French] [J] . Dupin M., Gosser P., Walls MG., Annales de Chimie-Science des Materiaux . 2002,第1期

机译：pH对在高温水性环境中在316L不锈钢，600合金和690合金上形成的氧化膜的化学和结构性质的影响[法语]
2. Crack growth and cracking behavior of Alloy 600/182 and Alloy 690/152 welds in simulated PWR primary water [J] . Yun Soo Lim, Dong Jin Kim, Sung Woo Kim, Nuclear engineering and technology . 2019,第1期

机译：合金600/182和690/152合金焊缝在模拟PWR初级水中的裂纹扩展和开裂行为
3. Oxidation products of INCONEL alloys 600 and 690 in pressurized water reactor environments and their role in intergranular stress corrosion cracking [J] . J. B. Ferguson, Hugo F. Lopez Metallurgical and Materials Transactions A . 2006,第8期

机译：压水反应堆环境中INCONEL合金600和690的氧化产物及其在晶间应力腐蚀开裂中的作用
4. Comparison of ODSCC behavior between Alloy 600 and Alloy 690 under secondary side environments [C] . Dong-Jin Kim, Eun Hee Lee, Hong Pyo Kim, Asian forum on material aging issues in nuclear system . 2017

机译：二次侧环境合金600和合金690之间ODSCC行为的比较
5. Oxidation products of Inconel alloys 600 and 690 in hydrogenated steam environments and their role in stress corrosion cracking. [D] . Ferguson, J. Bryce. 2006

机译：Inconel合金600和690在氢化蒸汽环境中的氧化产物及其在应力腐蚀开裂中的作用。
6. The Evolution of Fretting Wear Behavior and Damage Mechanism in Alloy 690TT with Cycle Number [O] . Long Xin, Yongming Han, Ligong Ling, 2020

机译：具有循环次数的690TT合金微动磨损行为及损伤机理的演变
7. Electrochemical Properties of Oxide Films Formed on Cold Worked Alloy600 and Alloy690 in Simulated PWR Primary Water Environments [O] . Whee-Sung Kim, Hiroaki Tsuchiya, Masugu Sato, 2015

机译：在仿真PWWR初级水环境中在冷加工合金600和合金690上形成的氧化膜的电化学性能
8. FRACTURE BEHAVIOR OF ALLOY 600, ALLOY 690, EN82H WELDS AND EN52 WELDS IN WATER [R] . 2000

机译：水中合金600，合金690，EN82H焊缝和EN52焊缝的断裂行为

Comparison of ODSCC behavior between Alloy 600 and Alloy 690 under secondary side environments

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