张乐福 研究员
所在系所核科学与工程学院
办公电话**
通讯地址机动行政办公楼(A楼)332室
电子邮件lfzhang@sjtu.edu.cn
个人主页暂无


教育背景 1993－1997华中理工大学机械工程二系博士
1990－1993华中理工大学机械工程二系硕士
1986－1990华中理工大学机械工程二系学士

工作经历 1997－2001，华中理工大学，材料科学与工程学院，讲师
2001－2004，日本岩手大学工学部，JSPS博士后研究员
2004－2017，上海交通大学机械与动力工程学院，副教授
2018 - 现在， 上海交通大学机械与动力工程学院，研究员

出访及挂职经历 2014-2017 上海新闵(东台)重型锻造公司科技副总

研究方向 水冷堆核电站材料腐蚀性能评价
材料应力腐蚀/腐蚀疲劳机理研究
核电站设备老化机理与失效分析
核反应堆水化学

科研项目 2007―2011 国家973项目“超临界水堆关键科学问题的基础研究”，研究骨干
2009―2011 国家自然科学基金项目“镍基690 合金晶界结构优化与耐腐蚀性研究”，负责人
2009―2010 上海核工程研究设计院项目“核电厂二回路碳钢管道流动加速腐蚀性能实验”，负责人
2010―2013 大型先进压水堆核电站重大专项重大共性技术及关键设备、材料研究课题“压水堆核电材料环境相容性研究”子课题“焊接工艺及焊后热处理方法对管道焊接件环境相容性的影响”，负责人
2010―2015 大型先进压水堆核电站重大专项重大共性技术及关键设备、材料研究课题“蒸汽发生器690合金U形管的研制和应用性能研究”，研究骨干
2014-2015 大型先进压水堆核电站重大专项“CAP1400水化学研究”
2014-2015 大型先进压水堆核电站重大专项“690合金管在高温水环境下微动磨损试验”
2015-2016 日立公司“泵、阀材料气蚀机理及可靠性研究”
2017-2018 海洋动力装备构件腐蚀疲劳机理研究，科技部973项目
2017-2020 应力-腐蚀-热时效耦合环境核电关键材料服役行为高通量评价技术，科技部重点研发
2018-2022 晶间碳化物对镍基690合金应力腐蚀开裂的影响机理研究，自然科学基金
2019-2022 超临界水冷堆材料与化学研发，科技部重点研发国际合作项目

代表性论文专著 2019
[1]Kai Chen, Jiamei Wang, et al. Effect of intergranular carbides on the cracking behavior of cold worked Alloy 690 in subcritical and supercritical water, Corrosion Science Online 108313.
[2]Kai Chen, Jiamei Wang, et al. Comparison of the stress corrosion cracking growth behavior of cold worked Alloy 690 in subcritical and supercritical water, Journal of Nuclear Materials 520 (2019) 235-244.
[3]Jiamei Wang, Haozhan, Su, Kai Chen*, Donghai Du*, Lefu Zhang*, Yongduo Sun, Corrosion fatigue crack growth behavior of alloy 52?M in high-temperature water, Journal of Nuclear Materials 528 (2020) 151848.
[4]Zhao Shen*, David Tweddle, Mark Thomas Lapington, Benjamin Jenkins, Donghai Du, Lefu Zhang, Michael P. Moodya, Sergio Lozano-Perez, Observation of internal oxidation in a 20% cold-worked Fe-17Cr-12Ni stainless steel through high-resolution characterization, Scripta Materialia 173 (2019) 144-148.
[5]Jiamei Wang, Haozhan, Su, Kai Chen*, Donghai Du, Lefu Zhang, Zhao Shen, Effect of δ-ferrite on the stress corrosion cracking behavior of 321 stainless steel, Corrosion Science 158 (2019) 108079.
[6]Donghai Du*, Jiamei Wang, Kai Chen, Lefu Zhang*, Stress corrosion cracking behavior of warm forged 316L stainless steel at different orientations, Journal of Nuclear Materials 522 (2019) 220-225.
[7]Lefu Zhang, Farzin Arjmand*, Water chemistry, solution resistivity, and oxygen reduction reaction in simulated primary coolant of pressurized water reactors, Materials and Corrosion 70 (2019) 1179-1191.
[8]Farzin Arjmand, Jiamei Wang, Lefu Zhang*, Zinc addition and its effect on the corrosion behavior of a 30% cold forged Alloy 690 in simulated primary coolant of pressurized water reactors, Journal of Alloys and Compounds 791 (2019) 1176-1192.
[9]Xianglong Guo*, Yi Fan, Wenhua Gao, Rui Tang, Kai Chen, Zhao Shen, Lefu Zhang*, Corrosion resistance of candidate cladding materials for supercritical water reactor, Annals of Nuclear Energy 127 (2019) 351-363.
[10]Ping Lai, Hao Zhang, Lefu Zhang, Qifeng Zeng, Junqiang Lu, Xianglong Guo*, Effect of micro-arc oxidation on fretting wear behavior of zirconium alloy exposed to high temperature water, Wear 424-425 (2019) 53-61.
[11]Zhao Shen*, Donghai Du*, Lefu Zhang, Sergio Lozano-Perez, An insight into PWR primary water SCC mechanisms by comparing surface and crack oxidation, Corrosion Science, 148 (2019) 213-227.
[12]Zhao Shen, Kai Chen, Xianglong Guo, Lefu Zhang*, A study on the corrosion and stress corrosion cracking susceptibility of 310-ODS steel in supercritical water, Journal of Nuclear Materials 514 (2019) 56-65.
[13]Donghai Du, Jiamei Wang, Kai Chen*, Lefu Zhang*, Peter L. Andresen, Environmentally assisted cracking of forged 316LN stainless steel and its weld in high temperature water, Corrosion Science 147 (2019) 69-80.
2018
[1].Lefu ZHANG, Jiamei WANG, Farzin ARJMAND, Electrochemical Behavior of Alloy 825 at High Temperature Pressurized Water Chloride Solutions (30°C to 280°C), Corrosion 74 (2018) 1245-1258
[2].Kai Chen, Jiamei Wang, Donghai Du*, Xianglong Guo, Lefu Zhang*. Characterizing the effects of in-situ sensitization on stress corrosion cracking of austenitic steels in supercritical water, Scripta Materialia 158 (2019) 66-70.
[3].Kai Chen, Jiamei Wang, et al. Investigation on the stress corrosion crack initiation and propagation behavior of Alloy 600 in high temperature water, Corrosion 74 (2018) 1395-1405.
[4].Kai Chen, Jiamei Wang, Donghai Du, Xianglong Guo, Peter L. Andresen, Lefu Zhang*. Stress corrosion crack growth behavior of 310S stainless steel in supercritical water, Corrosion 74 (2018) 776-787.
[5].Kai Chen, Jiamei Wang, Donghai Du, Peter L. Andresen, Lefu Zhang*. dK/da effects on the SCC growth rates of nickel base alloys in high-temperature water, Journal of Nuclear Materials 503 (2018) 13-21.
[6].Xianglong Guo, Kai Chen, Wenhua Gao, Zhao Shen, Lefu Zhang*. Corrosion behavior of alumina-forming and oxide dispersion strengthened austenitic 316 stainless steel in supercritical water. Corrosion Science 138 (2018) 297-306.
[7].Xianglong Guo, Ping Lai, Lichen Tang, Kai Chen, Lefu Zhang. Time-dependent wear behavior of alloy 690 tubes fretted against 405 stainless steel in high temperature argon and water. Wear, in press.
[8].Xianglong Guo, Wenhua Gao, Kai Chen, Zhao Shen, Lefu Zhang*. Corrosion and stress corrosion cracking susceptibility of 347H stainless steel in supercritical water. Corrosion, 2018, 74(1): 227-236
[9].Xianglong Guo*, Ping Lai, Lichen Tang, Junqiang Lu, Jiamei Wang, Lefu Zhang*. Fretting wear of alloy 690 tube mated with different materials in high temperature water, Wear 400-401 (2018) 119-126.
[10].Kai Chen, Donghai Du, Wenhua Gao, Xianglong Guo, Lefu Zhang*, Peter L. Andresen. Effect of cold work on the stress corrosion cracking behavior of Alloy 690 in supercritical water environment, Journal of Nuclear Materials 498 (2018) 117-128.
[11].Lefu Zhang, Ping Lai, Qingdong Liu, Qifeng Zeng, Junqiang Lu, Xianglong Guo*. Fretting wear behavior of zirconium alloy in B-Li water at 300℃. Journal of Nuclear Materials 499 (2018) 401-409.
[12].Ping Lai, Xiaochuan Gao, Lichen Tang, Xianglong Guo*, Lefu Zhang. Effect of temperature on fretting wear behavior and mechanism of alloy 690 in water, Nuclear Engineering and Design 327 (2018) 51-60
[13].Zhang L, Wang J. et al. High Temperature Electrochemical Corrosion Behavior of Fe-Cr-Ni Alloys in Simulated PWR water. CORROSION, 74(2018) 415-423.

教学工作 1.课程名称： 核反应堆材料与水化学
授课对象： 本科生
学时数： 36学时
学分：2学分
2.课程名称： 核工程导论
授课对象： 本科生
学时数： 18学时
学分：1学分
3. 课程名称： 核反应堆材料
授课对象： 硕士/博士研究生
学时数： 54学时
学分：3学分

软件版权登记及专利 1字梁式高压釜上试验机压力平衡装置授权日
2017.08.01国家发明专利授权
专利授权号：ZL 2014 1 **.6
2一种在线检测金属减薄速率的试样及方法授权日
2017．07．11国家发明专利授权
专利授权号：ZL 2014 1 **.2

学术兼职 能源行业核电标准化技术委员会(NEA/TC2)委员
国际第四代核能论坛(GIF)超临界水冷堆(SCWR)技术委员会中国副代表
中国核学会教育咨询委员会委员