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南韩高等工程研究院和汉阳大学材料科学与工程系对用于核燃料包套的锆-4合金表面进行了预氧化处理,并研究了处理后的合金在360℃下的腐蚀性能。他们将常规的锆-4合金用20φ%氧气加80φ%氩气的混合气体分别在700℃和1100℃下退火以形成预氧化膜,随后将形成预氧化膜的合金在装有
The South Korean Institute of Advanced Engineering and Department of Materials Science and Engineering, Hanyang University, pre-oxidized zirconium-4 alloy surfaces for nuclear fuel cladding and investigated the corrosion behavior of the treated alloys at 360 ° C. They annealed the conventional zirconium-4 alloy with a mixed gas of 20 phi oxygen and 80 phi argon gas at 700 ° C. and 1100 ° C. to form a pre-oxidized film, respectively, and then the alloy forming the pre-oxidized film