金属学报  2011, Vol. 47 Issue (7): 790-796    DOI: 10.3724/SP.J.1037.2011.00162

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核级不锈钢高温水腐蚀疲劳机制及环境疲劳设计模型

吴欣强1, 2), 徐松1, 2), 韩恩厚1, 2), 柯伟1, 2)

1) 中国科学院金属研究所金属腐蚀与防护国家重点实验室, 沈阳 110016 2) 中国科学院金属研究所辽宁省核电材料安全与评价技术重点实验室, 沈阳 110016

CORROSION FATIGUE OF NUCLEAR--GRADE STAINLESS STEEL IN HIGH TEMPERATURE WATER AND ITS ENVIRONMENTAL FATIGUE DESIGN MODEL

WU Xinqiang1, 2),  XU Song1, 2), HAN En-Hou1, 2), KE Wei1, 2)

1) State Key Laboratory for Corrosion and Protection, Institute of Metal Research, Chinese Academy of Sciences, Shenyang 110016 2) Liaoning Key Laboratory for Safety and Assessment Technique of Nuclear Material, Institute of Metal Research, Chinese Academy of Sciences, Shenyang 110016

吴欣强 徐松 韩恩厚 柯伟. 核级不锈钢高温水腐蚀疲劳机制及环境疲劳设计模型[J]. 金属学报, 2011, 47(7): 790-796.
. CORROSION FATIGUE OF NUCLEAR--GRADE STAINLESS STEEL IN
HIGH TEMPERATURE WATER AND ITS ENVIRONMENTAL FATIGUE DESIGN MODEL[J]. Acta Metall Sin, 2011, 47(7): 790-796.

摘要 参考文献 相关文章 Metrics 全文: PDF(1482 KB)   摘要: 通过模拟核电高温高压循环水腐蚀疲劳实验, 研究了国产核级不锈钢的环境疲劳损伤行为与失效机制; 评价了影响不锈钢高温高压水疲劳寿命的环境和载荷等因素, 建立了一个植入环境损伤效应的疲劳设计模型, 给出了便于工程应用的核级不锈钢的环境疲劳设计曲线. 关键词 ： 核级不锈钢,  高温高压水,  腐蚀疲劳,  环境致裂,  设计模型     Abstract：The high safety of light water reactor nuclear power plants (NPPs) requires very strict design standard and service property of pressure boundary components materials. The service degradation and life assessment of the components materials primarily depend on the understanding of environmentally assisted failure mechanism, accumulation of service property data and construction of evaluation models. Currently domestic NPPs are relying on foreign design, operation and life assessment standards. However, recent experimental data indicate that even the ASME design fatigue code may be deficient in safety margin under certain conditions of loading and environment. In the present work, based on the corrosion fatigue tests in simulated NPPs' high temperature pressurized water, the corrosion fatigue behavior and environmentally assisted failure mechanism of domestic nuclear-grade stainless steel have been investigated. The factors affecting fatigue life of nuclear grade stainless steel in high temperature water were evaluated. A design fatigue model was constructed by taking environmental degradation effects into account and the corresponding design curves were given for the convenience of engineering applications. Key words： nuclear-grade stainless steel    high temperature pressurized water    corrosion fatigue    environmentally assisted cracking    design model 收稿日期: 2011-03-23      基金资助: 国家重点基础研究发展计划项目2011CB610506和国家科技重大专项项目2011ZX06004--009资助 作者简介: 吴欣强, 男, 1971年生, 研究员 服务 把本文推荐给朋友 加入引用管理器 E-mail Alert RSS 作者相关文章 吴欣强 44.8K 链接本文: https://www.ams.org.cn/CN/10.3724/SP.J.1037.2011.00162      或      https://www.ams.org.cn/CN/Y2011/V47/I7/790 [1] Cahn R W, Haasen P, Kramer E J. Mater Sci Technol, 1994: 10B: 38 [2] Wu X Q, Han E H, Ke W, Katada Y. Nucl Eng Des, 2007; 237: 1452 [3] Wu X Q, Kataday Y. Corros Sci, 2005; 47: 1415 [4] Raman S G S, Kitsunai Y. Eng Fail Anal, 2004; 11: 293 [5] Chopra O K, Shack W J. ASME PVP, 2003; 453: 71 [6] Chopra O K, Shack W J. Nucl Eng Des, 1998; 184: 49 [7] Higuchi M, Iida K, Asada Y. In: Van Der Sluys WA, Piascik R S, Zauierucha R, eds., Effects of Strain Rate Change on Fatigue Life of Carbon Steel in High–Temperature Water, ASTM STP 1298, Philadelphia: American Society for Testing and Materials, 1997: 216 [8] H¨anninen H, Cullen W, Kemppainen M. Corrosion, 1990; 46: 563 [9] Kuniya J, Anzai H, Masaoka I. Corrosion, 1992; 48: 419 [10] Atkinson J D, Yu J. Fatigue Fract Eng Mater Struct, 1997; 20: 1 [11] Chopra O K, Shack W J. J Pressure Vessel Technol Trans ASME, 1999; 121: 49 [12] Ford F P. Corrosion, 1996; 52: 375 [13] Maiya P S. J Pressure Vessel Technol Trans ASME, 1987; 109: 116 [14] Xu S, Wu X Q, Han E H, Ke W, Katada Y. Mater Sci Eng, 2008; A49: 16 [15] Laird C, Smith G C. Philo Mag, 1962; 8: 847 [16] Laird C, Finney J M, Wilsdorf D K. Mater Sci Eng, 1981; 50: 127 [17] Langer B F. ASME J Basic Eng, 1962; 84: 389 [1] 林晓冬, 马海滨, 任啟森, 孙蓉蓉, 张文怀, 胡丽娟, 梁雪, 李毅丰, 姚美意. Fe13Cr5Al4Mo合金在高温高压水环境中的腐蚀行为[J]. 金属学报, 2022, 58(12): 1611-1622. [2] 谭季波, 王翔, 吴欣强, 韩恩厚. 316LN不锈钢管状试样高温高压水的腐蚀疲劳行为[J]. 金属学报, 2021, 57(3): 309-316. [3] 郦晓慧, 王俭秋, 韩恩厚, 郭延军, 郑会, 杨双亮. 690合金在模拟核电高温高压水中的电化学及原位划伤行为研究[J]. 金属学报, 2020, 56(11): 1474-1484. [4] 王俭秋, 黄发, 柯伟. Inconel 690TT和Incoloy 800MA蒸汽发生器管材在高温高压水中的腐蚀行为研究*[J]. 金属学报, 2016, 52(10): 1333-1344. [5] 吴欣强, 谭季波, 徐松, 韩恩厚, 柯伟. 核级低合金钢高温水腐蚀疲劳机制及环境疲劳设计模型[J]. 金属学报, 2015, 51(3): 298-306. [6] 张利涛,王俭秋. 国产锻造态核级管材316L不锈钢在高温高压水中的应力腐蚀裂纹扩展行为[J]. 金属学报, 2013, 49(8): 911-916. [7] 郦晓慧 王俭秋 韩恩厚 柯伟. 核级商用690合金和800合金在模拟压水堆核电站一回路高温高压水中的腐蚀行为研究[J]. 金属学报, 2012, 48(8): 941-950. [8] 匡文军 吴欣强 韩恩厚. 闭塞区中Ni2+对核级304不锈钢高温水氧化行为的影响[J]. 金属学报, 2011, 47(7): 927-931. [9] 钟祥玉 吴欣强 韩恩厚. 核级不锈钢和铁素体-马氏体耐热钢在400℃/25 MPa超临界水中的腐蚀行为[J]. 金属学报, 2011, 47(7): 932-938. [10] 韩恩厚. 核电站关键材料在微纳米尺度上的环境损伤行为研究------进展与趋势[J]. 金属学报, 2011, 47(7): 769-776. [11] 杨波 李谋成 姚美意 周邦新 沈嘉年. 高温高压水环境中锆合金腐蚀的原位阻抗谱特征[J]. 金属学报, 2010, 46(8): 946-950. [12] 常; 红; 韩恩厚; 王俭秋; 柯伟 . 阴极极化对LY12CZ铝合金腐蚀疲劳寿命的影响[J]. 金属学报, 2005, 41(5): 556-560 . [13] 方丙炎; 韩恩厚; 王俭秋; 柯伟 . 应变速率对管线钢近中性pH值环境敏感开裂的影响[J]. 金属学报, 2005, 41(11): 1174-1182 . [14] 杨继红; 贾维平; 李守新 . [013]取向Cu单晶体在NaCl水溶液中腐蚀疲劳位错结构的观察[J]. 金属学报, 2004, 40(1): 99-102 . [15] 张波; 张正贵 . LY12CZ铝合金铆接联接件在3.5aCl溶液中的腐蚀疲劳失效行为研究[J]. 金属学报, 2000, 36(8): 842-846 . Viewed Full text Abstract Cited   Shared      Discussed