期刊文献+

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
下载PDF
导出
摘要 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。 Immersion tests were carried out in 10% NaOH + 10 g/L PbO solution at 330 ℃ in a static autoclave.The results show that the lamellar oxide films of alloy 690TT consist of NiO,NiFe2 O4 and NiCr2 O4.Internal corrosion products closing to substrate formed due to intergranular attack and general corrosion contain mixed nickel-rich oxide and chromium-rich oxide.The intermediate layer of oxide film is NiCr2 O4 spinel oxide and the outer layer consists of NiFe2 O4 and NiO.The oxide film,which consists of an internal chromium-rich oxide layer and an external iron-rich oxide layer,characterizes p-type and n-type semiconductivity,respectively.
出处 《材料热处理学报》 EI CAS CSCD 北大核心 2011年第4期116-121,共6页 Transactions of Materials and Heat Treatment
基金 国家"973"项目(G2006CB605002)
关键词 690TT合金 氧化膜 半导体 沿晶腐蚀 alloy 690TT oxide film semiconductor intergranular attack
  • 相关文献

参考文献16

  • 1Was G S. Grain-boundary chemistry and intergranular fractrue in austenitic nickel-base alloys-a review [ J]. Corrosion, 1990, 46 (4) :319 -330.
  • 2华惠中,黄春波,吕战鹏,杨武.800、600和690合金的铅致应力腐蚀破裂[J].腐蚀与防护,2001,22(11):483-488. 被引量:13
  • 3Miglin B P, Sarver J M. Preliminary studies of lead stress corrosion cracking of alloy 690 [ C ]//Proceedins of the 4th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, Jekyll Island, GA: 1989:7 -18.
  • 4Staehle R W, Gorman J A. Quantitative assessment of submodes of stress corrosion cracking on the secondary side of steam generator tubing in pressurized water reactors : Part 1 [ J ]. Corrosion, 2003, 59 ( 11 ) : 931 - 994.
  • 5Staehle R W, Gorman J A. Quantitative assessment of submodes of stress corrosion cracking on the secondary side of steam generator tubing in pressurized water reactors : Part 2 [ J ]. Corrosion, 2004, 60 ( 1 ) : 5 - 63.
  • 6Staehle R W, Gorman J A. Quantitative assessment of submodes of stress corrosion cracking on the secondary side of steam generator tubing in pressurized water reactors: Part 3[J]. Corrosion, 2004, 60(2) : 115 -180.
  • 7Ziemniak S E, Hanson M. Corrosion behavior of NiCrFe alloy 600 in high temperature, hydrogenated water [ J]. Corrosion Science, 2006, 48:498 - 521.
  • 8Belo M Da Chuha, Hakiki N E, Ferreira M G S. Semieondueting properties of passive films formed on nickel-base alloys type alloy 600: Influence of the alloying elements [ J]. Electroehimieal Acta, 1999, 44 : 2473 - 2481.
  • 9Panter J, Viguier B, Cloue J M, et al. Influence of oxide films on primary water stress corrosion cracking initiation of alloy 600 [ J]. Journal of Nuclear Materials, 2006, 348 : 213 - 221.
  • 10Montemor 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, and alloy 690 in high-temperature aqueous environments [ J ]. Corrosion, 2003, 59 ( 1 ) : 11 - 21.

二级参考文献29

  • 1[1]Sarver J M. Information on lead concentration for BWI water chemistry manual[R]. Internal Report, B&W25211-002,1994.
  • 2[2]Agrawal A K, Paine J P N. Lead cracking of alloy 600-A review[A]. Proc 4th Iht Symp on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors [C]. Jekyll Insland GA:1989.
  • 3[3]Miglin B P, Sarver J M. Investigation of lead as a cause of stress corrosion at support plate interatctions[R]. EPRI NP-7367-S, 1991.
  • 4[4]Sakai T,Senjuh T, Aoki K, et al. Lead-induced stress corrosion cracking of alloy 600 and 690 in high temperature water[A]. Proc 5th Int Symp on Environmental Degradation of Materials in Nuclear Power SystemsWater Reactors[C]. Monterey, CA:1991,764.
  • 5[5]Miglin B P, Sarver J M, Pasaila-Dombrowski M J, et al. Lead assisted stress corrosion crocking of nuclear steam generator tubing materials[A]. Proc of Improving the Understanding and Control of Corrosion on the Secondary Side of Steam Generators[C]. Airlie,VA: 1995,95.
  • 6[6]Hayner G, Frye C, Theus G, et al. Examination of tubes removed from st. lucie unit 1 and investigation of causes of the corrosion[A]. Proc of the Third International Symposium on Environment Degradation of Materials in Nuclear Power Systems-Water Reactors[C].TMS, Traverse City, MI: 1987. 449.
  • 7[7]King P J, Gonzalez F, Brown J. Stress corrosion cracking experience in steam generators at bruce NGS[A]. Proc of the 62th Iht Symp on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors[C]. San Diego, CA: 1993. (Minerals Metals Materials Society, Warrendale, PA) 233.
  • 8[8]Rocher A,Nordmann F. Transport of lead in secondary system of PWR plarts: laboratory and plant investigations[A]. 6th Conf on Water Chemistry of Nuclear Reactor System, Bournemouth [C]. British Nuclear Energy Soc. , London: 1992. 249.
  • 9[9]Agrawal A L Lead cracking of alloy 600[R]. Paper Presented at EPRI Meeting on Lead SCC of Alloy 600,Charlotte NC.. 1988.
  • 10[10]Nordmann F,Cattant F, Comby R. Corrosion intergranulatire cote secondaire dea tubes de generateurs de vapeur francais[R]. Fonteraud 1 International Symposium, 1990.

共引文献12

相关作者

内容加载中请稍等...

相关机构

内容加载中请稍等...

相关主题

内容加载中请稍等...

浏览历史

内容加载中请稍等...
;
使用帮助 返回顶部