不同焊接方法下燃料元件包壳用ODS钢焊接接头界面演化规律
Interface Evolution of ODS Steel Welded Joints for Fuel Element Cladding Using Different Welding Methods
- 2024年54卷第5期 页码:39-45
纸质出版日期: 2024-05-25
DOI: 10.7512/j.issn.1001-2303.2024.05.05
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纸质出版日期: 2024-05-25 ,
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杨灿湘,魏连峰,郑勇,等.不同焊接方法下燃料元件包壳用ODS钢焊接接头界面演化规律[J].电焊机,2024,54(5):39-45.
YANG Canxiang, WEI Lianfeng, ZHENG Yong, et al.Interface Evolution of ODS Steel Welded Joints for Fuel Element Cladding Using Different Welding Methods[J].Electric Welding Machine, 2024, 54(5): 39-45.
ODS钢具有优异的高温力学性能、抗辐照性能和抗热蠕变性能等,是最有潜力的下一代核反应堆包壳候选材料之一。ODS钢的焊接技术主要包括熔焊、钎焊、压力焊等,根据核燃料元件包壳结构,选择电子束焊接、旋转摩擦焊及脉冲电流辅助扩散焊三种焊接方法进行对比研究,并对ODS钢焊接接头微观形貌演化进行分析,揭示了最优焊接方法及界面形貌演化规律。结果表明,电子束焊接及旋转摩擦焊接工艺下,焊接接头的晶界处均有Y
2
O
3
析出,而采用脉冲电流辅助扩散焊无氧化物析出和团聚,对于Fe-Cr系ODS钢有突出优势。因此,脉冲电流辅助扩散焊在抑制ODS钢中纳米氧化物的析出、团聚和减少接头变形等方面具有显著优势,是一种适合ODS钢焊接的高质量焊接方法。
ODS steel has excellent high-temperature mechanical properties
radiation resistance
and thermal creep resistance
and is one of the most potential candidate materials for the next generation nuclear reactor fuel element cladding. The welding technology of ODS steel mainly includes fusion welding
brazing
pressure welding
etc. According to the cladding structure of nuclear fuel elements
three welding methods
namely electron beam welding
rotary friction welding
and pulse current assisted diffusion welding
are selected for comparative study
and their microscopic morphology evolution is analyzed to reveal the optimal welding method and the evolution law of interface morphology. The results show that Y
2
O
3
precipitates at the grain boundaries of the welded joints under electron beam welding and
rotary friction welding processes
and pulse current assisted diffusion welding has no oxide precipitation and agglomeration. Pulse current assisted diffusion welding has prominent advantages for Fe-Cr ODS steels.
ODS钢核燃料元件焊接电子束焊接旋转摩擦焊接脉冲电流辅助扩散焊接
oxide dispersion strengthened steelnuclear fuel element weldingelectron beam weldingrotary friction weldingpulsed current assisted diffusion welding
Hoffelner W. Damage assessment in structural metallic materials for advanced nuclear plants[J].Journal of Materials Science,2010,45(9):2247-2257.
Murty K L,Charit I. Structural materials for Gen-IV nuclear reactors:Challenges and opportunities[J]. Journal of Nuclear Materials,2008,383(1-2):189-195.
Kurtz R J,Odette G R. Overview of Reactor Systems and Operational Environments for Structural Materials in Fusion Reactors[M]. Elsevier,2019.
王晓丁,李太斌,孙磊,等. 低碳经济下我国新能源产业的现状及展望[J]. 新型工业化,2021,11(05):20-21+24.
WANG X D,LI T B,SUN L,et al. Current Situation and Prospect of China's New Energy Industry in a Low Carbon Economy[J]. New Industrialization, 2021,11(05):20-21+24.
郭天超,孙善星,张文娟.“碳中和”目标下 核能积极有序发展策略研究[J]. 中国能源,2021,43(05):44-50.
GUO T C,SUN S X,ZHANG W J. Research on the strategy of positive and orderly development of nuclear energy under the goal of "Carbon Neutralization"[J]. China Energy,2021,43(05):44-50.
李明洋. 通过调控纳米析出相制备新型核聚变堆用热沉材料和结构材料[D]. 北京:北京科技大学,2021.
LI M Y. Preparation of new heat sink materials and structural materials for nuclear fusion reactors by regulating nano precipitates[D]. Beijing:Beijing University of Science and Technology, 2021.
程心雨,刘荣正,刘马林,等. 碳化物陶瓷材料在核反应堆领域应用现状[J]. 科学通报, 2021, 66(24):3154-3170.
CHENG X Y,LIU R Z,LIU M L,et al. Application status of carbide ceramic materials in the field of nuclear reactors[J]. Science Bulletin,2021,66(24):3154-3170.
Kim T K,Noh S,Kang S H,et al. Current Status and Future Prospective of Advanced Radiation Resistant Oxide Dispersion Strengthened Steel (ARROS) Development for Nuclear Reactor System Applications[J]. Nuclear Engineering & Technology,2016,48(2):572-594.
Susila P,Sturm D,Heilmaier M,et al. Microstructural studies on nanocrystalline oxide dispersion strengthened austenitic (Fe-18Cr-8Ni-2W-0.25Y2O3) alloy synthesized by high energy ball milling and vacuum hot pressing[J]. Journal of Materials Science,2010,45(17):4858-4865.
Kim T K,Chang S B,Kim D H,et al. Microstructural Observation and Tensile Isotropy of an Austenitic ODS Steel[J]. Nuclear Engineering & Technology,2008,40(4):305-310.
Zinkle S J,Boutard J L,Hoelzer D T. Development of next generation tempered and ODS reduced activation ferritic/martensitic steels for fusion energy applications[J]. Nuclear Fusion,2017,57(9):092005.
Ts A,Yuan W A,Jc B,et al. Characterization of polyhedral nano-oxides and helium bubbles in an annealed nanostructured ferritic alloy[J]. Acta Materialia, 2020,183:484-492.
Yvon P,Flem M L,Cabet C,et al. Structural materials for next generation nuclear systems:Challenges and the path forward[J]. Nuclear Engineering and Design,2015,294:161-169.
Benjamin J S. Dispersion strengthened superalloys by mechanical alloying[J]. Metallurgical Transactions,1970,1(10):2943-2951.
Chen Y,Sridharan K,Ukai S,et al. Oxidation of 9Cr oxide dispersion strengthened steel exposed in supercritical water[J].Journal of Nuclear Materials,2007,371(1-3):118-128.
Oksiuta Z,Baluc N. Microstructure and Charpy impact properties of 12-14Cr oxide dispersion-strengthened ferritic steels[J]. Journal of Nuclear Materials,2008,374(1-2):178-184.
Narita T,Ukai S, Ohtsuka S,et al. Effect of tungsten addition on microstructure and high temperature strength of 9CrODS ferritic steel[J].Journal of Nuclear Materials,2011,417(1-3):158-161.
Brocq M L,Legendre F,Mathon M H,et al. Influence of ball-milling and annealing conditions on nanocluster characteristics in oxide dispersion strengthened steels[J]. Acta Materialia,2012,60(20):7150-7159.
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