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锆合金中的氢对其在400℃过热蒸汽中耐腐蚀性能的影响 被引量:1

Effect of Hydrogen in Zirconium Alloys on Their Corrosion Resistance in Superheated Steam at 400 ℃
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摘要 对Zr-4、C7、SZA-1、SZA-4、SZA-5和SZA-6锆合金试样进行了气相渗氢并达到了60~260μg/g的氢含量。通过高压釜腐蚀试验研究了锆合金中的氢对以上6种锆合金试样在400℃、10. 3 MPa压力的过热蒸汽中耐腐蚀性能的影响。用扫描电镜观察试样氧化膜的显微组织。结果表明,6种锆合金渗氢试样在400℃、10. 3 MPa压力的过热蒸汽中腐蚀后,渗氢和未渗氢试样的腐蚀增重和氧化膜显微组织没有明显差别,含量低于260μg/g的氢对Zr-4、C7、SZA-1、SZA-4、SZA-5和SZA-6锆合金在400℃、10. 3 MPa压力的过热蒸汽中耐腐蚀性能的影响不大。 Zr-4,C7,SZA-1,SZA-4,SZA-5 and SZA-6 zirconium alloy specimens were impregnated with hydrogen in a gaseous medium to hydrogen contents of 60 to 260μg/g.The effect of hydrogen in zirconium alloys on their corrosion resistance was investigated by corrosion tests in superheated steam at temperature of 400℃and pressure of 10.3 MPa in a high-pressure autoclave.The microstructure of oxide film on the specimens was observed by SEM.The results showed that no obvious differences in the corrosion weight gains and in the microstructures of the oxide film were observed for the hydrogen-impregnated and un-impregnated specimens after being corroded in the superheated steam at temperature of 400℃and pressure of 10.3 MPa,and that the hydrogen content less than 260μg/g in the zirconium alloys had a little effect on their corrosion resistance in the superheated steam at the temperature and pressure stated above.
作者 毛亚婧 段文荣 姚美意 周邦新 张金龙 Mao Yajing;Duan Wenrong;Yao Meiyi;Zhou Bangxin;Zhang Jinlong(Institute of Materials,Shanghai University,Shanghai 200072,China)
出处 《上海金属》 CAS 北大核心 2018年第6期1-6,共6页 Shanghai Metals
基金 国家自然科学基金(No.51471102)
关键词 锆合金 耐腐蚀性能 腐蚀增重 zirconium alloy hydrogen corrosion resistance weight gain
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  • 1周邦新,李聪,黄德诚.Zr(Fe,Cr)_2金属间化合物的氧化[J].核动力工程,1993,14(2):149-153. 被引量:14
  • 2周邦新,李强,姚美意,刘文庆,褚玉良.锆-4合金在高压釜中腐蚀时氧化膜显微组织的演化[J].核动力工程,2005,26(4):364-371. 被引量:46
  • 3周邦新,李强,刘文庆,姚美意,褚于良.水化学及合金成分对锆合金腐蚀时氧化膜显微组织演化的影响[J].稀有金属材料与工程,2006,35(7):1009-1016. 被引量:42
  • 4Zhou B X,Li Q, Yao M Y, et al. Effect of water chemistry and composition on microstructural evolution of oxide on Zr-Alloys, zirconium in the nuclear industry: 15^th International Symposium[EB/OL]. Sunriver Oregon US, 2007, June, 24- 28. (paper ID JAI 10112, available online at www. astrn, org).
  • 5Yilmazbayhan A, Motta A T, Comstock R J, et al. Structure of zirconium alloy oxides formed in prre water studied with synchrotron radiation and optical microscopy:relation to corrosion rate[J]. J Nuel Mater [J]. 2004,324:6-22.
  • 6Yilmazbayhan A, Breval E, Motta A T, et al. Transmission electron microscopy examination of oxide layer formed on Zr alloys[J]. J Nucl Mater, 2006,349:265 -281.
  • 7Bryner J S. The cyclic nature of corrosion of Zircaloy-4 in 633K water[J]. J Nucl Mater. , 1979,82:84-88.
  • 8Comstock R J, Schoenberger G, Sabol G P. Influence of processing variables and alloy chemistry on the corrosion behavior of ZIRLO nuclear fuel cladding, Zirconium in the nuclear industry. Eleventh International Symposium[C]//ASTM STP 1295, Bradley E R and Sabol G P, Eds. American Society for Testing and Materials, 1996 : 710-725.
  • 9Cox B. Some thoughts on the mechanisms of in-reactor corrosion of zirconium alloys[J]. J Nucl Mater, 2005,336 : 331-368.
  • 10IAEA-TECDOC-996. Waterside corrosion of zirconium alloys in nuclear power plants[J].IAEA, Vienna, ISSN 1011-4289,1998.

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