通过慢应变速率拉伸和高温电化学试验相结合的方法,研究了外加电位对压水堆核电机组安全端 508III-52M-690合金异种材料焊接接头在含氯离子的高温高压水中应力腐蚀开裂(SCC)倾向的影响规律.结果表明,在温度为300℃高温高压水环境下,当氯离子的含量为50mg/kg、除氧时,焊接接头的SCC敏感性随电极电位升高而增大,即随着溶解氧浓度增加而增加.存在一个介于-500~-400mV(相对标准氢电极)的SCC临界电位,低于该电位时,焊接接头SCC敏感性较小,未见明显沿晶开裂,断裂为由力学性能主导的塑性开裂,与焊接接头不同冶金组织的硬度密切相关,硬度越低,越容易断裂,断裂位置均为硬度最低的52Mb处;高于临界电位时,SCC敏感性急剧增加,并出现明显的沿晶开裂和穿晶开裂断口,断裂为腐蚀主导的脆性开裂,断裂位置均为腐蚀性能最差的低合金钢 508III 热影响区.同时发现,焊接接头中52Mb对接焊和 508III 钢之间的热影响区对SCC最敏感.
Effect of applied potential on stress corrosion cracking (SCC) behavior of 508III-52M-690 dissi-milar weld joint of safe-end was researched with slow strain rate tensile (SSRT) tests and high temperature electrochemistry tests in high temperature (300℃) water containing 50mg/kg chloride. The results revealed that the SCC susceptibility increases dramatically with the applied potential, then the potential above a critical value exists between -500 and -400mV (versus standard hydrogen electrode). The SCC susceptibility is low and no obvious intergranular or transgranular stress corrosion cracks can be found when the applied potential below the critical value which corresponds to deoxygenated water chemistry. It means the fracture is dominated by mechanical properties and closely relates to the hardness distribution of welded joint. The lower the hardness is, the more easily fractures occur. Therefore, all ductile fractures are located at 52Mb (butt welded) with the lowest hardness. While, when electrode potential is higher than the critical potential, all brittle fractures are located at 508III heat affected zone (508III HAZ) with the lowest corrosion resistance, where exhibits significant SCC behavior with large area intergranular and transgranular stress corrosion cracks. Hence, the dissolved oxygen concentration need be controlled strictly to make sure the corrosion potential is below the critical potential. Besides, the 52Mb and 508III HAZ of this dissimilar weld joint are the most venerable sites to crack which need to be paid high attention to during the operation.
[1]HWANG S S. Review of PWSCC and mitigation management strategies of Alloy 600 materials of PWRs[J]. Journal of Nuclear Materials, 2013, 443(1/2/3): 321-330.
[2]SERIES I N E. Stress corrosion cracking in light water reactors: Good practices and lessons learned. No. NPT-3.13[M]. Vienna: International Atomic Energy Agency, 2011.
[3]XU H, MAHMOUD S, NANA A, et al. A new modeling method for natural PWSCC cracking simulation in a dissimilar metal weld[J]. International Journal of Pressure Vessels and Piping, 2014, 116 (4): 20-26.
[4]KANG S S, HWANG S S, KIM H P, et al. The experience and analysis of vent pipe PWSCC (primary water stress corrosion cracking) in PWR vessel head penetration[J]. Nuclear Engineering and Design, 2014, 269(4): 291-298.
[5]DU D, CHEN K, YU L, et al. SCC crack growth rate of cold worked 316L stainless steel in PWR environment[J]. Journal of Nuclear Materials, 2015, 456: 228-234.
[6]BOSCH R W, FRON D, CELIS J P, Electroche-mistry in light water reactors: Reference electrodes, measurement, corrosion and tribocorrosion issuesed[M]. New York: CRC Press, 2007.
[7]ANDRESEN P L, MARTIN M M, AHLUWALIA K. SCC of alloy 690 and its weld metals in high temperature water[C]∥Busby J T, Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors. Cham: Springer, 2011: 161-178.
[8]KANE R D. Slow strain rate testing for the evaluation of environmentally induced cracking[M]. Philadelphia: ASTM International, 1993.
[9]BERTINI L, SANTUS C, VALENTINI R, et al. New high concentration-high temperature hydrogenation method for slow strain rate tensile tests[J]. Materials Letters, 2007, 61(11): 2509-2513.
[10]MTHIS K, PRCHAL D, NOVOTN R, et al. Acoustic emission monitoring of slow strain rate tensile tests of 304L stainless steel in supercritical water environment[J]. Corrosion Science, 2011, 53(1): 59-63.
[11]LONG B. DAI Y, BALUC N. Investigation of liquid LBE embrittlement effects on irradiated ferritic/martensitic steels by slow-strain-rate tensile tests[J]. Journal of Nuclear Materials, 2012, 431(1): 85-90.
[12]彭德全, 胡石林, 张平柱, 等. 氧氯协同对304L不锈钢在高温高压硼锂水中应力腐蚀开裂的影响[J]. 稀有金属材料与工程, 2014, 43(1): 178-183.
PENG Dequan, HU Shilin, ZHANG Pingzhu, et al. Effect of oxygen and chloride cooperation on stress corrosion cracking of 304L stainless steel in high temperature and high pressure water containing boric acid and lithium ion[J]. Rare Metal Materials and Engineering, 2014, 43(1): 178-183.
[13]INDIG M E. 1990 speller award lecture: Technology transfer: Aqueous electrochemical measurements room temperature to 290 ℃[J]. Corrosion, 1990, 46(8): 680-686.
[14]LIN C C, SMITH F R, ICHIKAWA N, et al. Electrochemical potential measurements under simulated BWR water chemistry conditions[J]. Corrosion, 1992, 48(1): 16-28.
[15]SHOJI T, TAKAHASHI H, AIZAWA S, et al. Effects of sulfate contamination, sulfur in steel and strain rate on critical cracking potential for SCC of pressure vessel steels in pressurized high temperature waters[C]∥Theus G J. Proceedings of the Third International Symposium on Environmental Degradation of Materials in Nuclear Power Systems. Pennsylvania: The Metallurgical Society, Inc, 1988: 251-258.
[16]ZHOU X Y, CHEN J. Stress corrosion cracking of iron base alloys in high temperature water[C]∥Theus G J. Proceedings of Eighth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems. Illinois: American Nuclear Society, 1997: 953-959.
[17]ANDRESEN P L, YOUNG L M. Characterization of the roles of electrochemistry, convection and crack chemistry in stress corrosion cracking[C]∥Airey G, Andresen P, Brown J. Seventh International Symposium on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors. Houston, TX: NACE, 1995: 579-596.
[18]ANDRESEN P L, MORRA M M. IGSCC of non-sensitized stainless steels in high temperature water[J]. Journal of Nuclear Materials, 2008, 383: 97-111.
[19]FORD F P. Quantitative prediction of environmentally assisted cracking[J]. Corrosion, 1996, 52(5): 375-395.