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概述了压水堆核电厂典型的结构材料种类与腐蚀类型,并以此为基础介绍了常见的腐蚀防护设计手段及腐蚀老化管理的理念和方法,对明确压水堆核电厂设备/部件、材料、环境、腐蚀、防护、老化管理间的相互关系具有参考价值,为确保机组的安全与经济运行提供重要保障.

参考文献

[1] S.J. Zinkle;G.S. Was.Materials challenges in nuclear energy[J].Acta materialia,20133(3):735-758.
[2] 龚嶷;窦一康.美国核电厂GALL报告解读[J].核安全,2014(2):88-94.
[3] 丁亚平;徐雪莲.压水堆核电厂长寿命化的腐蚀损伤问题[J].腐蚀与防护,2001(11):489-493,497.
[4] Kyu-Tae Kim.Evolutionary developments of advanced PWR nuclear fuels and cladding materials[J].Nuclear engineering and design,2013Oct.(Oct.):59-69.
[5] Steinbrück, M.;B?ttcher, M..Air oxidation of Zircaloy-4, M5? and ZIRLO? cladding alloys at high temperatures[J].Journal of Nuclear Materials: Materials Aspects of Fission and Fusion,20112(2):276-285.
[6] 韩向臻;攸国顺;孙微.第三代反应堆AP1000和EPR的堆芯核设计[J].中国科技信息,2013(3):50,53.
[7] 胡源;顾楠.核级锆材:重复建设需深思[J].中国有色金属,2015(15):42-43.
[8] 赵文金;周邦新;苗志;彭倩;蒋有荣;蒋宏曼;庞华.我国高性能锆合金的发展[J].原子能科学技术,2005(z1):2-9.
[9] 周军;李中奎.轻水反应堆(LWR)用包壳材料研究进展[J].中国材料进展,2014(9):554-559.
[10] I. S. Hwang;I. -G. Park.Control of alkaline stress corrosion cracking in pressurized-water reactor steam generator tubing[J].Corrosion: The Journal of Science and Engineering,19996(6):616-625.
[11] 龚嶷;徐雪莲.压水堆核电厂蒸汽发生器老化机理及其影响因素[J].腐蚀与防护,2014(2):163-168,174.
[12] D.L. Harrod;R.E. Gold;R.J. Jacko.Alloy Optimization for PWR Steam Generator Heat-Transfer Tubing[J].JOM,20017(7):14-17.
[13] L. Tan;T.R. Allen;Y. Yang.Corrosion behavior of alloy 800H (Fe-21Cr-32Ni) in supercritical water[J].Corrosion Science: The Journal on Environmental Degradation of Materials and its Control,20112(2):703-711.
[14] Tucker, J. D.;Miller, M. K.;Young, G. A..Assessment of thermal embrittlement in duplex stainless steels 2003 and 2205 for nuclear power applications[J].Acta materialia,2015:15-24.
[15] 王西涛;李时磊.核电用钢的研究现状及发展趋势[J].新材料产业,2014(7):2-8.
[16] Yueling Guo;En-Hou Han;Jianqiu Wang.Effects of Forging and Heat Treatments on the Microstructure and Oxidation Behavior of 316LN Stainless Steel in High Temperature Water[J].材料科学技术(英文版),2015(4):403-412.
[17] 卢华兴.AP1000核电站主管道国产化研制进展[J].上海金属,2010(04):29-32.
[18] 李光福;方可伟;许君;杨武.异材焊接件A508Ⅲ-52M-316L基本材料在高温水环境中的电化学特性[J].腐蚀与防护,2014(12):1177-1181.
[19] 李江;吴欣强;韩恩厚;柯伟.核电焊接结构材料腐蚀失效研究现状与进展[J].腐蚀科学与防护技术,2014(1):1-7.
[20] 丁杰;张志明;王俭秋;韩恩厚;唐伟宝;张茂龙;孙志远.三代核电接管安全端异种金属焊接接头的显微表征[J].金属学报,2015(4):425-439.
[21] H.T. Wang;G.Z. Wang;F.Z. Xuan.Fracture mechanism of a dissimilar metal welded joint in nuclear power plant[J].Engineering failure analysis,2013:134-148.
[22] Wang, H.T.;Wang, G.Z.;Xuan, F.Z.;Liu, C.J.;Tu, S.T..Local mechanical properties of a dissimilar metal welded joint in nuclear powersystems[J].Materials Science & Engineering, A. Structural Materials: Properties, Misrostructure and Processing,2013:108-117.
[23] 孟晶.福岛第一核电站核污染事故的原因及其影响本刊记者与上海核工程设计研究院前总工程师蔡剑平访谈[J].电力与能源,2011(02):101-104.
[24] Y. Gong;J. Cao;X.-H. Meng;Z.-G. Yang.Pitting corrosion on 316L pipes in terephthalic acid (TA) dryer[J].Materials and Corrosion,200911(11):899-908.
[25] 汪长春;王成铭;郑文远.大亚湾和岭澳核电站海水冷却系统的腐蚀与控制[J].电力安全技术,2009(2):18-21.
[26] Jian Xu;Xinqiang Wu;En-Hou Han.Acoustic emission response of sensitized 304 stainless steel during intergranular corrosion and stress corrosion cracking[J].Corrosion Science: The Journal on Environmental Degradation of Materials and its Control,2013Aug.(Aug.):262-273.
[27] 夏爽;周邦新;陈文觉.690合金的晶界特征分布及其对晶间腐蚀的影响[J].电子显微学报,2008(6):461-468.
[28] 高文娇;谭华;韩冬;邓博;李劲;蒋益明.退火温度对Incoloy800合金晶间腐蚀敏感性的影响[J].材料热处理学报,2012(2):1-6.
[29] Benoit Ter-Ovanessian;Julien Deleume;Jean-Marc Cloue;Eric Andrieu.Quantitative assessment of intergranular damage due to PWR primary water exposure in structural Ni-based alloys[J].Corrosion Science: The Journal on Environmental Degradation of Materials and its Control,2013Feb.(Feb.):11-19.
[30] Lee, Tae Hyun;Hwang, Il Soon;Kim, Hong Deok;Kim, Ji Hyun.TECHNIQUES FOR INTERGRANULAR CRACK FORMATION AND ASSESSMENT IN ALLOY 600 BASE AND ALLOY 182 WELD METALS[J].Nuclear engineering and technology: An International Journal of the Korean Nuclear Scoeity,20151(1):102-114.
[31] 刘晓强;孟凡江;徐雪莲;汪家梅;杨晨;张乐福.690合金晶间腐蚀化学浸泡试验方法的适用性[J].腐蚀与防护,2016(3):236-240.
[32] 刘肖;赵建仓;王淦刚;朱平;迟鸣声.核电厂管道及焊接接头失效案例综述[J].失效分析与预防,2013(5):300-305.
[33] R.W. Staehle;J.A. Gorman.Quantitative Assessment of Submodes of Stress Corrosion Cracking on the Secondary Side of Steam Generator Tubing in Pressurized Water Reactors: Part 1[J].Corrosion: The Journal of Science and Engineering,200311(11):931-994.
[34] Qunjia Peng;Juan Hou;Yoichi Takeda;Tetsuo Shoji.Effect of chemical composition on grain boundary microchemistry and stress corrosion cracking in Alloy 182[J].Corrosion Science: The Journal on Environmental Degradation of Materials and its Control,2013Feb.(Feb.):91-99.
[35] Du, Donghai;Chen, Kai;Yu, Lun;Lu, Hui;Zhang, Lefu;Shi, Xiuqiang;Xu, Xuelian.SCC crack growth rate of cold worked 316L stainless steel in PWR environment[J].Journal of Nuclear Materials: Materials Aspects of Fission and Fusion,2015:228-234.
[36] 马成;彭群家;韩恩厚;柯伟.核电结构材料应力腐蚀开裂的研究现状与进展[J].中国腐蚀与防护学报,2014(1):37-45.
[37] Gary S. Was;Yugo Ashida;Peter L. Andresen.Irradiation-assisted stress corrosion cracking[J].Corrosion reviews,20111/2(1/2):7-49.
[38] Peter L Andresen.Stress Corrosion Cracking of Current Structural Materials in Commercial Nuclear Power Plants[J].Corrosion: The Journal of Science and Engineering,201310(10):1024-1038.
[39] Fanjiang Meng;Zhanpeng Lu;Tetsuo Shoji;Jianqiu Wang;En-hou Han;Wei Ke.Stress corrosion cracking of uni-directionally cold worked 316NG stainless steel in simulated PWR primary water with various dissolved hydrogen concentrations[J].Corrosion Science: The Journal on Environmental Degradation of Materials and its Control,20118(8):2558-2565.
[40] Z. G. Yang;Y. Gong;J. Z. Yuan.Failure analysis of leakage on titanium tubes within heat exchangers in a nuclear power plant Part I: Electrochemical corrosion[J].Materials and Corrosion,20121(1):7-17.
[41] Y. Gong;Z. G. Yang;J. Z. Yuan.Failure analysis of leakage on titanium tubes within heat exchangers in a nuclear power plant Part II: Mechanical degradation[J].Materials and Corrosion,20121(1):18-28.
[42] Wu, H. C.;Yang, B.;Wang, S. L.;Zhang, M. X..Effect of oxidation behavior on the corrosion fatigue crack initiation and propagation of 316LN austenitic stainless steel in high temperature water[J].Materials Science & Engineering, A. Structural Materials: Properties, Misrostructure and Processing,2015:176-183.
[43] 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(3):298-306.
[44] R. Barry Dooley.Flow-Accelerated Corrosion in Fossil and Combined Cycle/HRSG Plants[J].PowerPlant Chemistry: The Journal of All Power Plant Chemistry Areas,20082(2):68-89.
[45] R. B. Dooley;V. K. Chexal.Flow-accelerated corrosion of pressure vessels in fossil plants[J].International Journal of Pressure Vessels and Piping,20002/3(2/3):85-90.
[46] 束国刚;薛飞;遆文新;汪小龙;陆念文;刘鹏;戴忠华.核电厂管道的流体加速腐蚀及其老化管理[J].腐蚀与防护,2006(2):72-76.
[47] 张桂英;顾宇;邵杰.核电站汽水管道流动加速腐蚀的影响因素分析及对策[J].动力工程学报,2012(2):170-176.
[48] 刘春波;郑玉贵.核电行业中流动促进腐蚀的模型和数值模拟研究进展[J].腐蚀科学与防护技术,2008(6):436-439.
[49] H.P. Rani;T. Divya;R.R. Sahaya;Vivekanand Kain;D.K. Barua.CFD study of flow accelerated corrosion in 3D elbows[J].Annals of nuclear energy,2014Jul.(Jul.):344-351.
[50] 龚嶷;徐雪莲.流动加速腐蚀评价程序CHECWORKS及其核电厂老化管理应用[J].腐蚀与防护,2014(5):401-406.
[51] V. I. Baranenko;A. V. Kumov;Yu. A. Yanchenko;A. S. Prokopenko.Development of Software Tools for Calculating Corrosion-Erosion Wear of Pipelines at Nuclear Power Stations[J].Thermal engineering,200712(12):981-988.
[52] V. I. Baranenko;A. A. Prosvirnov;S. V. Evropin;A. A. Aref'ev;V. A. Yurmanov;O. M. Gulina.Development of Software Means and Normative Documentation on the Flow-Accelerated Corrosion of Pipelines of Nuclear Power Plants[J].Thermal engineering,20125(5):378-383.
[53] 朱隽;徐雪莲;石秀强.核电厂水化学参数ASTM和GB分析方法比较[J].腐蚀与防护,2012(10):841-844,848.
[54] 段振刚;潘向烽;张乐福;王力;徐雪莲;石秀强.压水堆一回路水中锌含量对镍基690合金氧化膜的影响[J].腐蚀与防护,2014(4):348-351.
[55] 孟凡江;王俭秋;韩恩厚;庄子哲雄;柯伟.690TT合金划痕显微组织及划伤诱发的应力腐蚀[J].金属学报,2011(7):839-846.
[56] 刘晓强;徐雪莲;孟凡江;石秀强.非能动核电站安全壳涂层的设计与可靠性分析[J].涂料工业,2015(4):74-78.
[57] 窦一康.核电厂生命周期全过程的老化管理[C].2011年全国失效分析学术会议论文集,2011:10-14.
[58] 龚嶷;崔满满;窦一康;韩镇辉;石秀强;邹建平.核电厂运行许可证延续(OLE)安全监管的对策[J].核安全,2015(1):1-11.
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