蒸汽发生器焊接堵管残余应力数值模拟及寿命预测
Numerical Simulation of Residual Stress and Life Prediction of Welding Blocked Pipes of Reactor Steam Generator
- 2023年53卷第5期 页码:12-20
DOI: 10.7512/j.issn.1001-2303.2023.05.02
扫 描 看 全 文
扫 描 看 全 文
杨二娟,李勇,米紫昊,等.蒸汽发生器焊接堵管残余应力数值模拟及寿命预测[J].电焊机,2023,53(5):12-20.
YANG Erjuan, LI Yong, MI Zihao, et al.Numerical Simulation of Residual Stress and Life Prediction of Welding Blocked Pipes of Reactor Steam Generator[J].Electric Welding Machine, 2023, 53(5): 12-20.
对于蒸汽发生器焊接堵管工艺,焊接接头的残余应力为应力腐蚀的发生孕育了环境,而且相比压水堆核电站,第四代先进核反应堆服役温度显著提高,造成更加明显的蠕变现象。采用热弹塑性有限元,基于Nims数据库和ASME标准中提供的数据,对于两种不同焊接结构,计算了使用Incoloy 800H材料的高温端和T22材料的低温端的传热管-堵头环焊的残余应力,并基于线性损伤累积准则,预测了蠕变-疲劳交互作用下焊接堵管的寿命。结果表明,最大残余拉应力点出现在焊缝的根部,开坡口结构的残余应力高于不开坡口结构的残余应力;相较于Nims数据库提供的数据,ASME标准中计算得出的寿命较为保守,开坡口结构的寿命低于不开坡口结构的寿命。
For the welded reactor steam generator, on the one hand, the residual stress of the welded joint provides conditions for stress corrosion; on the other hand, the increase of service temperature in the advanced reactors will cause more obvious creep phenomenon. Based on the data provided in the Nims database and ASME standards, for two different welding structures, this paper calculates the residual stress of heat transfer tube plug circumferential welding at the high temperature end of Incoloy 800H material and the low temperature end of T22 material, and predicts the life of welded pipe plug under the creep fatigue interaction based on the linear damage accumulation criterion. The results show that the point of maximum residual tensile stress occurs at the root of the weld, and the residual stress of the grooved structure is higher than that of the non-grooved structure. Compared with the data provided by the Nims database, the life calculated by the ASME standard is more conservative. The life of the grooved structure is lower than that of the non-grooved structure.
蒸汽发生器残余应力蠕变-疲劳寿命预测
steam generatorresidual stresscreep-fatiguelife expectancy
高建新. 高温气冷堆在石化行业耦合应用的思考与建议[J]. 炼油技术与工程, 2022, 52(08): 50-55.
GAO J X. Reflections and suggestions on coupled application of high temperature gas-cooled reactors in petrochemical industry[J]. Refinery Technology and Engineering, 2022, 52(08): 50-55.
陈瑜. 党建促进国家科技重大专项“开花结果”[N]. 科技日报, 2021-12-21(1).
刘福广, 杨二娟, 李勇, 等. 核电蒸汽发生器堵管技术研究进展及其在高温气冷堆的应用前景[J]. 热力发电, 2022, 51(06): 74-81.
LIU F G, YANG E J, LI Y, et al. Research progress of nuclear steam generator plugging technology and its application prospect in high temperature gas-cooled reactor[J]. Thermal Power Generation, 2022, 51(06): 74-81.
韩建成, 李巨峰, 吴志军, 等. 核电站蒸汽发生器用Incoloy800H传热管抗晶间腐蚀性能研究[J]. 热力发电, 2012, 41 (01): 50-52.
HAN J C, LI J F, WU Z J, et al. Research on Intergranular Corrosion Resistance of Incoloy800H Heat Tr-ansfer Tubes for Nuclear Power Plant Steam Generators[J]. Thermal Power Generation, 2012, 41(01):50-52.
罗飞华. 浅析核电厂蒸汽发生器传热管降质及堵管技术[J]. 科技视界, 2017(35): 90-91.
LUO F H. Brief Analysis on Degradation and Blocking Technology of Heat Transfer Tube of Steam Generator in Nuclear Power Plant[J]. Science and Technology Vision, 2017(35): 90-91.
Epri. Steam generator progress report[R]. Palo Alto, CA, 2017.
Ling X, Huang S Y. Discussion on replacement of steam generator[J]. Nuclear Power Engineering,2004, 25(3): 267-269.
郭城. 核电厂蒸汽发生器传热管断裂事故运行管理[J].核动力工程, 2013, 34(2): 107-110.
GUO C. Operation research on steam generator tube rupture accident in PWR NPPs[J]. Nuclear Power Engineering, 2013, 34(2): 107-110.
程檀. 蒸汽发生器传热管机械堵头的安装及拆除[J]. 设备管理与维修, 2020(21): 124-126.
CHENG T. Installation and Removal of Mechanical Plugs for Heat Transfer Tubes of Steam Generators [J]. Equipment Management and Maintenance, 2020(21): 124-126.
李翠翠, 和广庆. 核电蒸汽发生器胀接接头质量事故分析及应对措施研究[J]. 压力容器, 2019, 36(11): 70-73.
LI C C, HE G Q. Analysis of Quality Accidents of Expansion Joints of Nuclear Power Steam Generators and Research on Countermeasures[J]. Pressure Vessel, 2019, 36(11): 70-73.
CHENG W C, SAKURAHARA T, ZHANG S, et al. Review and categorization of existing studies on the estimation of probabilistic failure metrics for reactor coolant pressure boundary piping and steam generator tubes in nuclear power plants[J]. Progress in Nuclear Energy, 2020(118): 103-105.
闫宗宝. AP1000核电蒸汽发生器管子与管板胀/焊过程模拟与接头性能研究[D]. 上海: 华东理工大学, 2011.
YAN Z B. Simulation of expansion/welding process of AP1000 nuclear steam generator tube and tube sheet and research on joint performance[D]. Shanghai: East China University of Science and Technology, 2011.
章贵和, 邓小云, 徐晓, 等. 蒸汽发生器焊接堵管堵头的设计与评价[J]. 原子能科学技术, 2016, 50(07): 1270-1274.
ZHANG G H, DENG X Y, XU X, et al. Design and Evaluation of Welding Plug for Steam Generator[J]. Atomic Energy Science and Technology, 2016, 50(07):1270-1274.
李明, 林健, 雷永平, 等.核电蒸汽发生器传热管/管板接头传热管内壁的焊接残余应力分布[J]. 机械工程材料, 2019, 43(01): 82-86.
LI M, LIN J, LEI Y P, et al. Welding residual stress distribution on inner wall of heat transfer tube/tube-sheet joint of nuclear power steam generator[J]. Mechanical Engineering Materials, 2019, 43(01): 82-86.
张俊宝, 葛可可, 徐连勇. 核电站蒸汽发生器异种钢焊接残余应力研究[J]. 热力发电, 2017, 46(03): 39-44.
ZHANG J B, GE K K, XU L Y. Study on Welding Residual Stress of Dissimilar Steel for Steam Generator in Nuclear Power Plant[J]. Thermal Power Generation, 2017, 46(03): 39-44.
王润梓. 基于能量密度耗散准则的蠕变—疲劳寿命预测模型及应用[D]. 上海:华东理工大学,2019.
WANG R Z. Creep-fatigue life prediction model based on energy density dissipation criterion and its application[D]. Shanghai: East China University of Science and Technology, 2019.
Bahn C B, Majumdar S, Kasza K E, et al. Leak behavior of steam generator tube-to-tubesheet joints under creep condition: Experimental study[J]. International Journal of Pressure Vessels and Piping, 2013, 101: 55-63.
陈宝文, 毛欢, 孔翔程, 等. 全厂断电引发的严重事故下蒸汽发生器传热管蠕变失效风险研究[J]. 原子能科学技术, 2014,48(06): 1026-1030.
CHEN B W, MAO H, KONG X C, et al. Research on the Creep Failure Risk of Heat Transfer Tubes of Steam Generators in Serious Accidents Caused by Power Outages in the Whole Plant[J]. Atomic Energy Science and Technology, 2014,48(06): 1026-1030.
杨磊, 姜维维, 郝亚雷. 小型压水堆严重事故下一回路承压管道蠕变预测分析模型开发[J]. 核技术, 2017, 40(05): 73-78.
YANG L, JIANG W W, HE Y L. Development of a Creep Prediction Analysis Model for the Next-Circuit Pressurized Pipeline in a Severe Accident of a Small Pressurized Water Reactor[J]. Nuclear Technology, 2017, 40(05): 73-78.
刘桐.高温气冷堆蒸汽发生器管板的热应力与疲劳分析[D].北京:清华大学, 2009.
LIU T. Thermal stress and fatigue analysis of high temperature gas-cooled reactor steam generator tube sheets [D]. Beijing: Tsinghua University, 2009.
赵雷, 冯国才, 徐连勇, 等. 新型马氏体耐热钢蠕变-疲劳性能与寿命预测[J]. 焊接学报, 2022, 43(05): 1-7.
ZHAO L, FENG G C, XU L Y, et al. Creep-fatigue properties and life prediction of new martensitic heat-resistant steel[J]. Journal of Welding, 2022, 43(05): 1-7.
美国机械工程师学会. ASME锅炉及压力容器规范第Ⅲ卷第1册NH分卷(2004版)[S]. 上海发电设备成套设计研究院, 上海核工程研究设计院, 译. 上海: 上海科学技术文献出版社, 2007.
Wang Y, Shi L, Han C, et al. Creep rupture mechanisms and life prediction of IN617 for VHTR applications[J]. Materials Science and Engineering A, 2021, 812: 141151.
Manson S S. Behavior of materials under conditions of thermal stress[R]. TN 2933: NASA. 1954.
杨世铭, 陶文铨. 传热学(第4版)[M]. 北京: 高等教育出版社, 2006: 37-45.
相关作者
相关机构