高温含铅碱液中690TT合金氧化膜的耐蚀性能Corrosion resistance of oxide films formed on alloy 690TT in high temperature lead-containing caustic solution
胡轶嵩,王俭秋,柯伟,韩恩厚
摘要(Abstract):
利用静态高压釜,在330℃的10%NaOH+10 g/L PbO腐蚀介质中,对690TT合金进行5~60 d的浸泡实验,结果表明:690TT合金氧化膜由NiO、NiFe2 O4和NiCr2 O4构成。合金的氧化膜分层,靠近基体的腐蚀产物为混杂的富Cr和富Ni氧化物,是沿晶腐蚀和表面均匀腐蚀综合作用所致。中间层主要以NiCr2O4为主,外层是NiFe2O4以及NiO。氧化膜同时具有n型和p型半导体特征,内层富Cr的氧化物为p型半导体,而外层富Fe的氧化物为n型半导体。
关键词(KeyWords): 690TT合金;氧化膜;半导体;沿晶腐蚀
基金项目(Foundation): 国家“973”项目(G2006CB605002)
作者(Author): 胡轶嵩,王俭秋,柯伟,韩恩厚
DOI: 10.13289/j.issn.1009-6264.2011.04.026
参考文献(References):
- [1]Was G S.Grain-boundary chemistry and intergranular fractrue in austenitic nickel-base alloys-a review[J].Corrosion,1990,46(4):319-330.
- [2]华慧中,黄春波,吕战鹏,等.800、600和690合金的铅致应力腐蚀断裂[J].腐蚀与防护,2001,22(11):483-488.Hua H Z,Huang C B,LüZ P,et al.Lead-induced stress corrosion cracking of alloys 800,600 and 690[J].Corrosion and Protection,2001,22(11):483-488.
- [3]Miglin B P,Sarver J M.Preliminary studies of lead stress corrosion cracking of alloy 690[C]//Proceedins of the 4th International Symposium onEnvironmental Degradation of Materials in Nuclear Power Systems-Water Reactors,Jekyll Island,GA:1989:7-18.
- [4]Staehle R W,Gorman J A.Quantitative assessment of submodes of stress corrosion cracking on the secondary side of steam generator tubing inpressurized water reactors:Part 1[J].Corrosion,2003,59(11):931-994.
- [5]Staehle R W,Gorman J A.Quantitative assessment of submodes of stress corrosion cracking on the secondary side of steam generator tubing inpressurized water reactors:Part 2[J].Corrosion,2004,60(1):5-63.
- [6]Staehle R W,Gorman J A.Quantitative assessment of submodes of stress corrosion cracking on the secondary side of steam generator tubing inpressurized water reactors:Part 3[J].Corrosion,2004,60(2):115-180.
- [7]Ziemniak S E,Hanson M.Corrosion behavior of NiCrFe alloy 600 in high temperature,hydrogenated water[J].Corrosion Science,2006,48:498-521.
- [8]Belo M Da Chuha,Hakiki N E,Ferreira M G S.Semiconducting properties of passive films formed on nickel-base alloys type alloy 600:Influenceof the alloying elements[J].Electrochimical Acta,1999,44:2473-2481.
- [9]Panter J,Viguier B,Cloue J M,et al.Influence of oxide films on primary water stress corrosion cracking initiation of alloy 600[J].Journal ofNuclear Materials,2006,348:213-221.
- [10]Montemor M F,Ferreira M G S,Walls M,et al.Influence of pH on properties of oxide films formed on type 316L stainless steel,alloy 600,andalloy 690 in high-temperature aqueous environments[J].Corrosion,2003,59(1):11-21.
- [11]Lemire R J,McRae G A.The corrosion of alloy 690 in high-temperature aqueous media-thermodynamic considerations[J].Journal of NuclearMaterials,2001,294:141-147.
- [12]Miglin B P,Sarver J M,Psaila-Dombrowksi M J,et al.Lead assisted stress corrosion cracking of nuclear steam generator tube materials[C]//Proceedings of Improving the Understanding and Control of Corrosion on the Secondary Side of Steam Generators.Airlie:NACE,1995:305-320.
- [13]Robertson J.The mechanism of high temperature aqueous corrosion of stainless steels[J].Corrosion Science,1991,32(4):443-465.
- [14]Chen C M,Aral K,Theus G J.Computer Calculated Potential pH Diagrams to 300℃[M].Volumes 1-3,NP-3137,EPRI,Palo Alto,1983:1-3.
- [15]Chuha Belo M Da,Rondot B,Compere C,et al.Chemical composition and semiconduction behaviour of stainless steel passive films in contact withartificial seawater[J].Corrosion Science,1998,40(2-3):481-494.
- [16]Chuha Belo M Da,Walls M,Hakiki N E,et al.Composition,structrue and properties of the oxide films formed on the stainless steel 316L in aprimary type PWR environment[J].Corrosion Science,1998,40(2-3):447-463.
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