Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components最新文献_第4页

Application and Prospect of Additive Manufacturing Technology in Nuclear Fuel Assembly Fabrication 增材制造技术在核燃料组件制造中的应用与展望

Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components Pub Date : 2022-08-08 DOI: 10.1115/icone29-91618

Jie Ding, Yixiong Zheng, Chen Ye, Cong Li

引用次数: 0

Non-Destructive Examination of PWR Steam Generator Tube-Tube Plate Welds After Repairing Melting Damage 压水堆蒸汽发生器管板焊缝熔损修复后的无损检测

Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components Pub Date : 2022-08-08 DOI: 10.1115/icone29-92978

Yan Zhang, JianLei Zhang, Zhonghe Cheng

{"title":"Non-Destructive Examination of PWR Steam Generator Tube-Tube Plate Welds After Repairing Melting Damage","authors":"Yan Zhang, JianLei Zhang, Zhonghe Cheng","doi":"10.1115/icone29-92978","DOIUrl":"https://doi.org/10.1115/icone29-92978","url":null,"abstract":"\u0000 The tube-tube plate weld is the boundary weld that separates the primary and secondary circuits of the Pressurized Water Reactor (PWR) Steam Generator and is the weakest link of the Steam Generator. If the weld leaks, the media that may be contaminated by radiation in the primary circuit may flow into the secondary circuit, resulting in a nuclear leakage event. Therefore, the welding quality of the weld is directly related to the safety of the nuclear power site and requires special attention during the manufacturing process. The tube-tube plate weld of the PWR Steam Generator generally uses non-filled wire self-fusion welding technology. In addition to common welding defects such as pores, incomplete penetration, and cracks during the welding process, melting damage may sometimes occur due to improper operation. Since the melt loss repair involves both the tube plate surfacing layer and the tube-to-tube plate sealing weld, the repair process is much more complicated. There is little experience in the successful implementation of non-destructive inspection after melting damage repair. How to remove the defects? Which non-destructive examination methods should be selected in the repair process to ensure the defect removal effect? How to weld repair after the defect removal? Which nondestructive examination techniques should be selected during the repair process to ensure the repair effect? How to conduct final non-destructive testing after repair to confirm Rework quality? These are urgent issues. In this paper, by using tube-tube plate weld samples to carry out melting damage repair simulation experiments. After repairing, the penetration examination and the radiographic examination have been successfully implemented, and good results have been achieved.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"7 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"122381104","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 1

Research of Dissimilar Steel Welding for HTR-PM SG Heat Exchange Tube HTR-PM SG换热管异种钢焊接研究

Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components Pub Date : 2022-08-08 DOI: 10.1115/icone29-92837

Mingtao Zheng, Li Wang

{"title":"Research of Dissimilar Steel Welding for HTR-PM SG Heat Exchange Tube","authors":"Mingtao Zheng, Li Wang","doi":"10.1115/icone29-92837","DOIUrl":"https://doi.org/10.1115/icone29-92837","url":null,"abstract":"\u0000 The modular High-Temperature Gas-Cooled Reactor nuclear power plant has inherent safety, simple system, high power generation efficiency and wide use. Incoloy800H and 2.25Cr-1Mo heat exchange tubes are used for High-Temperature Gas-Cooled Reactor (HTR-PM) Steam Generator. The steam generator uses the Incoloy800H and 2.25Cr-1Mo heat exchange tubes to transfer the high temperature helium medium to the water medium. The dissimilar steel welds are designed between Incoloy800H and 2.25Cr-1Mo heat exchange tubes. The heat exchange tubes of HTR-PM Steam Generator with small OD, large thickness, and design structure complicated. The dissimilar steel welds between two kinds of heat exchange tubes with limited welding operation space and welding seam quality requirements high. During the HTR-PM Steam Generator operation process, Incoloy800H heat exchange tube, 2.25Cr-1Mo heat exchange tube and the dissimilar steel welds between Incoloy800H and 2.25Cr-1Mo heat exchange tube are service in high temperature environment for a long time, until the steam generator fails. The surface forming quality, conventional properties and high temperature stress-rupture of dissimilar steels weld is required. The quality of dissimilar steel weld of Incoloy800H and 2.25Cr-1Mo heat exchange tubes is directly related to the safe operation of steam generator. Through systematic research including tube-to-tube joint type, the welding material, welding parameters, heat treatment process etc. The results show that using appropriate GTAW process with Nickel base alloy consumable inserts, welded joints can meet all design requirements such as strict inner and outside surface forming quality, conventional performance and the strength of stress-rupture test in HTR-PM Steam Generator service environment. This process has been used in the product manufacturing of demonstration projects, and provides good demonstration experience for similar application scenarios in the industry.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"2 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"114768087","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

The Effect of Nozzle Safe End Hot/Cold Wire Tungsten Inert Gas Buttering on Microstructure and Property 喷嘴安全端热/冷钨丝惰性气体涂敷对组织和性能的影响

Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components Pub Date : 2022-08-08 DOI: 10.1115/icone29-89075

Chengdong Yang, Maolong Zhang, Xiangkun Ru, Weibao Tang

{"title":"The Effect of Nozzle Safe End Hot/Cold Wire Tungsten Inert Gas Buttering on Microstructure and Property","authors":"Chengdong Yang, Maolong Zhang, Xiangkun Ru, Weibao Tang","doi":"10.1115/icone29-89075","DOIUrl":"https://doi.org/10.1115/icone29-89075","url":null,"abstract":"\u0000 Hot wire tungsten inert gas (HWTIG) and cold wire tungsten inert gas (CWTIG) have been used for nozzle safe-end buttering of nuclear island critical equipment, the effect of the HWTIG and CWTIG buttering on microstructure and property has been investigated. Research results revealed that the grain orientation of HWTIG buttering layer is mostly concentrated in (0 0 1) direction and the average grain size is 118.375μm, while the grain orientation of the CWTIG dispersed distribution in (0 0 1), (1 1 1) and (1 0 1) direction, and the grain size is 158.479μm. Furthermore, the grain boundary misorientation angle of HWTIG buttering layer has higher occurrence frequency between 10° and 30° high-angle boundary, whereas the CWTIG has higher occurrence frequency within 30° and 60°. Moreover, the microstructure of HWTIG buttering layer has less chromium-rich M23C6-type carbide compared to CWTIG, and the M23C6-type carbides discontinuous distribution on the grain boundary of HWTIG buttering layer. The buttering layer of HWTIG has lower ductility-dip cracking sensitivity than CWTIG.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"121734457","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Numerical Simulation on the Flow-Induced Vibration of Two Square Cylinders in Staggered Arrangement 交错排列两方圆柱流激振动的数值模拟

Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components Pub Date : 2022-08-08 DOI: 10.1115/icone29-92963

Yidong Wu, Li Shi, Xinxin Wu, Xiaoxin Wang

{"title":"Numerical Simulation on the Flow-Induced Vibration of Two Square Cylinders in Staggered Arrangement","authors":"Yidong Wu, Li Shi, Xinxin Wu, Xiaoxin Wang","doi":"10.1115/icone29-92963","DOIUrl":"https://doi.org/10.1115/icone29-92963","url":null,"abstract":"\u0000 Based on the design of the heat transfer tube bundles with square cross section of the steam generators in the high temperature reactor-pebble bed modules (HTR-PM), the flow-induced vibration (FIV) of two single-degree-of-freedom (SDOF) square cylinders in staggered arrangement were investigated to guide the design, processing and manufacture of heat transfer components. Numerical simulation on the FIV of two square cylinders in staggered arrangement was carried out at Reynolds number Re = 100 with the mass ratio of 20. The wake structures of two square cylinders in different arrangements were analyzed. Three distinct flow regimes, i.e. one single-street modes S-I and two double-street modes T-I and T-II, are identified based on Strouhal numbers and flow structures. The amplitude and frequency of FIV response and the lift coefficient of two square cylinders at Ur = 1–30 were obtained to analyze the FIV behavior. The results show that upstream cylinder in the arrangements of mode S-Ia① and S-Ia② is prone to galloping while downstream cylinder in the arrangements of mode S-Ia③ and T-I is prone to galloping. The VIV amplitude of downstream cylinder in most staggered arrangements is much larger than that of single square cylinder while the VIV of upstream cylinder is suppressed.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"56 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"122876317","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Research and Development of Wet Winding Motor Reactor Coolant Pump for Generation 3 Nuclear Power Plant 第三代核电站湿式缠绕电机反应堆冷却剂泵的研究与开发

Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components Pub Date : 2022-08-08 DOI: 10.1115/icone29-90741

Liu Tianbin, Ma Lin

{"title":"Research and Development of Wet Winding Motor Reactor Coolant Pump for Generation 3 Nuclear Power Plant","authors":"Liu Tianbin, Ma Lin","doi":"10.1115/icone29-90741","DOIUrl":"https://doi.org/10.1115/icone29-90741","url":null,"abstract":"\u0000 Within an eight-year period, starting in September 2011 and ending in July 2019, Shanghai Electric-KSB Nuclear Pumps & Valves Co., Ltd (hereinafter called SEC-KSB) has managed to develop a new pump type that can be used as reactor coolant pump (hereinafter called RCP) in generation 3 pressurized water reactor design. Wet winding motor reactor coolant pump (pump type RUV 650–870) is a customized and especially designed to meet the high safety criteria of generation 3 nuclear power plant, which is a hermetically sealed (without shaft seal system) type pump, and it is first time used as reactor coolant pump in pressurized water reactor in the worldwide. Started with the pump commissioning test in December 2014, total eight times high temperature and high-pressure full load testing had to be carried out in order to eliminate design or manufacturing flaws. In early 2019 within three months of intensive testing, under all the specified design transient scenarios, the RUV 650-870 pump had successfully passed all the qualification test items. On July 31. 2019 an expert qualification meeting was organized by China Machinery Industry Federation and China General Machinery Industry Association, the qualification committee states that the development of first set wet winding motor reactor coolant pump is successful in the worldwide, and the pump technical performance meet the requirements of the equipment specification, test specification and related industry standards, the RUV 650-870 pump can be used for nuclear power plant projects. On January 2022, the first production pump was delivered to nuclear power plant.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"44 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"125717498","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Manufacturing Technology of Mn-Ni-Mo Alloy Steel Heavy Forgings for HPR1000 Nuclear Power Plant HPR1000核电站用Mn-Ni-Mo合金钢重型锻件制造技术

Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components Pub Date : 2022-08-08 DOI: 10.1115/icone29-91711

Z. Zhifeng, Li Xiang, Wang Xiaofang

{"title":"Manufacturing Technology of Mn-Ni-Mo Alloy Steel Heavy Forgings for HPR1000 Nuclear Power Plant","authors":"Z. Zhifeng, Li Xiang, Wang Xiaofang","doi":"10.1115/icone29-91711","DOIUrl":"https://doi.org/10.1115/icone29-91711","url":null,"abstract":"\u0000 HPR1000 nuclear power technology is the mainstream technology of commercial pressurized water nuclear power in China, which is adopted by most nuclear power plants under construction or planned. Mn-Ni-Mo alloy steel heavy forgings are widely used in the equipment of pressurized water nuclear reactor. In order to achieve the aim of safe operation for at least 60 years, the technical specifications of Mn-Ni-Mo alloy steel heavy forgings for HPR1000 are high and strict. Due to the high technology and difficult manufacturing, only a few factories have the manufacturing capacity of forgings with stable quality worldwide, which can’t meet the continuous growth demand of Chinese market. Therefore, whether Mn-Ni-Mo alloy steel heavy forgings can be delivered on time has become an important factor restricting the construction cycle of HPR1000 nuclear power plants.\u0000 In order to meet the market demand, Shanghai Electric SHMP Casting & Forging Co., Ltd (SCF) has developed the manufacturing technology of heavy forgings for HPR1000. According to the technical specifications of HPR1000, SCF has carried out systematic research in smelting, forging and heat treatment technologies. By 2020, SCF has realized the mass production of heavy forgings for HPR1000 nuclear power plant.\u0000 This paper reports new technologies obtained by SCF in the development of Mn-Ni-Mo alloy steel heavy forgings for HPR1000 nuclear power plant.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"76 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"115211072","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Study on the Application of Electromagnetic Stepping CRDM at Tilted Environment 电磁步进CRDM在倾斜环境中的应用研究

Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components Pub Date : 2022-08-08 DOI: 10.1115/icone29-90150

Zhaohui Lu, Zewen Li, Jianming Zhou, Qingsong Liu

引用次数: 0

The Research on Welding Procedure and Performance of Zr-3 Thick Plate Zr-3厚板焊接工艺及性能研究

Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components Pub Date : 2022-08-08 DOI: 10.1115/icone29-92037

Yu Gao, Xiaoru Liang, Yuxin Wang, Lifeng Guo

{"title":"The Research on Welding Procedure and Performance of Zr-3 Thick Plate","authors":"Yu Gao, Xiaoru Liang, Yuxin Wang, Lifeng Guo","doi":"10.1115/icone29-92037","DOIUrl":"https://doi.org/10.1115/icone29-92037","url":null,"abstract":"\u0000 Due to the excellent corrosion resistance and small absorption interface of thermal neutrons, Zr has been widely used in nuclear industry. As an important manufacturing process, the welding of Zr is usually performed by manual tungsten inert-gas arc welding (TIG). Due to the low deposition efficiency of tungsten inert-gas arc welding (TIG), it is more suitable for thin plates. If this method is used for thick plates, it will increase the cost of time, manpower and material resources. Therefore, for thick plates, an advanced and efficient welding method is required to improve the welding quality and construction efficiency. In this paper, a combination of plasma arc welding (PAW) for root bead and tungsten inert-gas arc welding (TIG) for filling and cap beads was successfully developed for 20mm Zr-3 plate, with specifically designed clamp and gas shielding devices based on the welding characteristics of Zr. Different post-weld heat treatment was performed for the weld pieces. Moreover, the mechanical properties and corrosion resistance of welded joints were evaluated. The results demonstrates that welded joints which meet the requirements specified in code and standard and engineering requirements can be obtained by developed welding process and welding system; the influence of different heat treatment on mechanical properties and corrosion resistance of welded joints was not obvious.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"18 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"132802373","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Research on Measurement Method of a Device for Measuring the Perpendicularity of Tubesheet Hole 管板孔垂直度测量装置的测量方法研究

Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components Pub Date : 2022-08-08 DOI: 10.1115/icone29-93045

Xianghua Xu, Guihong Hu

{"title":"Research on Measurement Method of a Device for Measuring the Perpendicularity of Tubesheet Hole","authors":"Xianghua Xu, Guihong Hu","doi":"10.1115/icone29-93045","DOIUrl":"https://doi.org/10.1115/icone29-93045","url":null,"abstract":"\u0000 The tubesheet is the key part of the steam generator, and the quality of the tubesheet hole is the key characteristic of the steam generator. The inspection of tubesheet holes includes defects on the surface of the tubesheet holes and the geometric dimensions of the tubesheet holes. The geometric dimensions of the tubesheet hole mainly include hole diameter, perpendicularity, hole bridge, position and roughness, etc. The perpendicularity of tubesheet hole is an important characteristic, which directly affects the size of tubesheet secondary hole bridge and heat exchange tube installation. This paper describes a device for measuring the perpendicularity of tubesheet hole, discusses the principle and measurement method of the tubesheet hole perpendicularity measuring device, and analyzes the error and proposes measures to improve the measurement accuracy of the device. Analysis of data processing provides a basis for the measurement device and measurement method to meet the requirements of specifications. The results show that the designed measuring device to measure perpendicularity has the advantages of simple operation, simple measurement and high accuracy of measurement results. It is beneficial to ensure the quality of heat exchange tube installation of the steam generator, which can speed up the manufacturing progress of the steam generator.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"43 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"114774428","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0