Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components最新文献_第3页

Research on Key Assembling Technology of Tube Bundle Assembly of Hualong-1 Type Steam Generator 华龙1型蒸汽发生器管束总成关键装配技术研究

Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components Pub Date : 2022-08-08 DOI: 10.1115/icone29-92130

Ye Wanbing, Liu Yuanbin, Chen Yaomao, Zhang Sanjun, Wang Aili

{"title":"Research on Key Assembling Technology of Tube Bundle Assembly of Hualong-1 Type Steam Generator","authors":"Ye Wanbing, Liu Yuanbin, Chen Yaomao, Zhang Sanjun, Wang Aili","doi":"10.1115/icone29-92130","DOIUrl":"https://doi.org/10.1115/icone29-92130","url":null,"abstract":"\u0000 Fangchenggang Phase II nuclear power project adopts the third generation advanced nuclear power technology of Hualong-1. The steam generator tube bundle assembly of the project adopts HL1000 type, which was manufactured for the first time in China. The tube bundle assembly is composed of three parts: tubes, support assembly of straight section of tube bundle and anti-vibration bar assembly of U tube bend. The structure of tube bundle assembly is novel, involving several kinds of parts and components, high assembly accuracy requirements and strict assembly sequence. During manufacturing of the tube bundle assembly, key manufacturing technologies including the installation and alignment of 9 pieces of Lattice Grid Assembly (diamond shaped holes) and tubesheet holes, anti-sagging control of tube bend section, integrated connection of anti-vibration components and anti-offset of anti-vibration components were successfully in command. The development of HL1000 tube bundle components was successfully completed, the independent research and manufacturing of products was realized, and the blanks of this new reactor manufacturing technology in China was filled in. This paper summarizes and analyzes the key assembly technology of HL1000 tube bundle assembly, so as to provide experience for the design and manufacture of steam generators of other reactor.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"3 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":"131542651","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 Martensitic Stainless Steel Used on the Latch Housing of European Pressurized Reactor (EPR) Control Rod Drive Mechanism (CRDM) 欧洲压水堆(EPR)控制棒驱动机构(CRDM)闩锁壳用马氏体不锈钢研究

Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components Pub Date : 2022-08-08 DOI: 10.1115/icone29-91786

Cuizhu He, Yinhui Lan

{"title":"Research on Martensitic Stainless Steel Used on the Latch Housing of European Pressurized Reactor (EPR) Control Rod Drive Mechanism (CRDM)","authors":"Cuizhu He, Yinhui Lan","doi":"10.1115/icone29-91786","DOIUrl":"https://doi.org/10.1115/icone29-91786","url":null,"abstract":"\u0000 As one of the most important equipment for reactivity control, Control Rod Drive Mechanism (CRDM), which is widely used in Pressurized Water Reactor (PWR) nuclear power plant, has a series of important security functions. As part of the reactor pressure boundary, materials for European Pressurized Reactor (EPR) CRDM pressure housing needs to have relatively high strength and sufficient toughness (i.e., appropriate yield ratio). As an important part of the magnetic circuit, martensitic stainless steel used for latch housing (a part of pressure housing) needs to have good magnetic properties. It is difficult for the yield ratio and the magnetic properties of the materials to meet the specification requirements simultaneously. Based on the design specification, this research refers to the relevant international manufacturing experience and combines the current manufacture with engineering practice. The manufacturing process and key parameters of the material for latching housing of CRDM are established. The martensitic stainless steel which meets the design requirements has been successfully prepared. The comprehensive properties (e.g. yield ratio and magnetic properties) of the material are improved as well. The performance indicators have reached international advanced level. The successful development of the martensitic steel for EPR CRDM latch housing provides strong technical support for the construction of nuclear power projects at home and abroad.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"29 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":"114646806","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 Study on Welding Deformation of Upper Support Structure in 600MW Sodium-Cooled Fast Reactor 600MW钠冷快堆上部支撑结构焊接变形数值研究

Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components Pub Date : 2022-08-08 DOI: 10.1115/icone29-91200

W. Dong, S. Lu

{"title":"Numerical Study on Welding Deformation of Upper Support Structure in 600MW Sodium-Cooled Fast Reactor","authors":"W. Dong, S. Lu","doi":"10.1115/icone29-91200","DOIUrl":"https://doi.org/10.1115/icone29-91200","url":null,"abstract":"\u0000 The upper support structure of 600 MW sodium-cooled fast reactor is welded by inner and outer cylinders, upper and lower tube-sheets and multiple rib plates. The structure has a large size and many welding passes, which poses a great challenge to the control of welding deformation. It is an effective method to control welding deformation of structural parts by using numerical simulation method to predict welding deformation and optimize welding process. In this work, the welding process of typical T-joint is numerically simulated and tested firstly. The reliability of the simulated results is verified by comparing the temperature field, and the inherent strain of typical joint is obtained. Then, the inherent strain method is used to calculate the welding deformation of typical T-joint. The welding deformation accuracy is compared and analyzed by using thermoelastic-plastic method and inherent strain method, respectively. Finally, the welding process of the upper support structure is simulated by the inherent strain method. The influence of the designed welding sequence on the welding deformation of the upper support structure is studied, and the shrinkage allowance are obtained. The research results have been verified by field application, which provides technical support for the construction of 600 MW sodium-cooled fast reactor.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"5 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":"125320586","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 SA508Gr.4N Steel Ultra-Thick Forgings With High Strength and Toughness for Nuclear Engineering SA508Gr的研发核工程用高强度高韧性4N钢超厚锻件

Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components Pub Date : 2022-08-08 DOI: 10.1115/icone29-92329

Dongxu Du, D. Zhao, Liguo Ren, Wenhui Zhang, Jiaju Li, Xi-kou He

{"title":"Research and Development of SA508Gr.4N Steel Ultra-Thick Forgings With High Strength and Toughness for Nuclear Engineering","authors":"Dongxu Du, D. Zhao, Liguo Ren, Wenhui Zhang, Jiaju Li, Xi-kou He","doi":"10.1115/icone29-92329","DOIUrl":"https://doi.org/10.1115/icone29-92329","url":null,"abstract":"\u0000 Due to low hardenability, the conventional SA508Gr.3 steel cannot meet the design requirements of increasing wall thickness of forgings for nuclear power pressure equipment. In this paper, the hot processing and manufacturing technology of SA508Gr.4N steel ultra-thick forgings for the new generation of nuclear reactor pressure vessel was introduced. The engineering development of nuclear SA508Gr.4N steel forgings with wall thickness from about 700mm to 1000mm was realized by China First Heavy Industry (CFHI). The largest forging was made of 228 tons ingot and the final maximum dimensions reached about ϕ4000mm × 1000mm. The samples were taken anatomically from the forging at various thickness and radial positions, and subjected to the tensile, Charpy impact and Drop-weight tests. As a result, the full-section property of the forging met the requirements of the ASME code. Compared with nuclear SA508Gr.3 steel forgings, the strength and toughness of SA508Gr.4N forgings were greatly improved. Meanwhile, based on the fact that the actual property value is much higher than the requirements of ASME code, higher design property requirements of SA508Gr.4N steel forging was proposed. In addition, the nuclear neutron-irradiation-embrittlement data of SA508Gr.4N steel, which were cut from the actual engineering forgings and experimented in a high flux engineering test reactor (HFETR), was obtained. The main research and development goals for engineering application of SA508Gr.4N steel in the future were also discussed.","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":"128304697","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

Microstructure and Hardness Investigation on 25Cr2Ni2MoV Steel Welding Joint of Nuclear Power Steam Turbine Low Pressure Welded Rotor and Engineering Applications 核电汽轮机低压焊接转子25Cr2Ni2MoV钢焊接接头显微组织、硬度研究及工程应用

Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components Pub Date : 2022-08-08 DOI: 10.1115/icone29-90504

X. Huo, Xia Liu

{"title":"Microstructure and Hardness Investigation on 25Cr2Ni2MoV Steel Welding Joint of Nuclear Power Steam Turbine Low Pressure Welded Rotor and Engineering Applications","authors":"X. Huo, Xia Liu","doi":"10.1115/icone29-90504","DOIUrl":"https://doi.org/10.1115/icone29-90504","url":null,"abstract":"\u0000 Large Nuclear power steam turbine Low pressure (LP) rotor composed of several shafts and discs with 25Cr2Ni2MoV steel was manufactured by gas tungsten-arc welding (GTAW) for backing welding and submerged-arc welding (SAW) for filling welding. In this study, the welding joint is investigated and it is found that the main microstructure of weld metal and base metal is typically bainite, while tempered martensite is observed in heat affected zone (HAZ) by over-quenching imposed by welding cycling and post weld heat treatment (PWHT). Compared with the weld metal of SAW, the microstructure of GTAW is relatively fine and homogeneous, due to its lower heat input. The hardness test indicates that the maximum hardness of HV400 occurs in HAZ because of tempered martensite. Weld metal of SAW has the hardness of HV270, which is a little bit lower than GTAW’s on account of fine grain. Based on the researches, the trial welded rotor and production rotor have been successfully produced, relevant inspections and tests were conducted, the results are well consistently with original research.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"74 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":"121029050","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

Modal Simulation and Verification on Stator End Winding of Large Nuclear Power Turbine Generator 大型核动力汽轮发电机定子端部绕组模态仿真与验证

Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components Pub Date : 2022-08-08 DOI: 10.1115/icone29-91595

Xike Yang, Jian Zhang

{"title":"Modal Simulation and Verification on Stator End Winding of Large Nuclear Power Turbine Generator","authors":"Xike Yang, Jian Zhang","doi":"10.1115/icone29-91595","DOIUrl":"https://doi.org/10.1115/icone29-91595","url":null,"abstract":"\u0000 The 1300 MW nuclear power turbine generator is a four-pole generator with a rated speed of 1500 r/min. Due to its large capacity and low speed, the stator core and stator end winding bear huge electromagnetic torque. At the same time, because of the large heat generated by the stator winding during operation, the stator end winding is subjected to a large thermal expansion force. All these loads may have a serious impact on the stator end winding. In this paper, the three dimensional parametric finite element method and modal optimization algorithm are used to obtain the finite element model of each component of the stator end winding. Finally, the modal natural frequency of the overall end winding structure is obtained, so that the four-lobe modal natural frequency avoids the excitation frequency of 100 Hz, and the accuracy of the simulation mode is verified by modal test. The stator end winding modal simulation provides a reference for the stator end winding structure design of the large nuclear power turbine generator, ensures the good dynamic characteristics of the end winding structure, which provides a guarantee for the safe and stable operation of the stator end winding.","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":"131131109","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 Manufacturing Process of Large Super Alloy X-750 Forgings for Nuclear Equipment 核设备用大型高温合金X-750锻件制造工艺研究

Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components Pub Date : 2022-08-08 DOI: 10.1115/icone29-91791

Mei-fang Chen, Sheng-qiang Cao, K. Dong

{"title":"Study on the Manufacturing Process of Large Super Alloy X-750 Forgings for Nuclear Equipment","authors":"Mei-fang Chen, Sheng-qiang Cao, K. Dong","doi":"10.1115/icone29-91791","DOIUrl":"https://doi.org/10.1115/icone29-91791","url":null,"abstract":"\u0000 X-750 is a nickel-chromium based super alloy with high strength, high toughness, excellent creep resistance and good stress relaxation resistance at high temperature. It can be used in the manufacture of holddown springs for reactor vessel internals (RVI). Because of high content of alloying elements and the large size of ingot for the RVI holddown spring, observations by optical microscopy (OM) and scanning electron microscopy (SEM) showed that there was obvious dendrite segregation and serious carbonitride net in the as-cast electroslag remelting (ESR) ingot. Before forging, homogenization treatment was carried out for the ESR ingot to reduce element segregation, and promote the secondary carbonitrides dissolve into the alloy, which could significantly improve the malleability of the billet and reduce forging cracks. The homogenization process adopted in this study was annealing at 1220°C for 70 hours. In addition, the final forging temperature also played a key role in the control of formability of the billet and grain size of the forging. The carbonitrides or grain size of both the as-cast and forged samples annealed at 900°C∼1200°C for 10h were observed, and also charpy impact tests were carried out on the forged and annealed samples. The results showed that the final forging temperature for upsetting and drawing process should be no less than 1100°C, and the final forging temperature for forming process could be reduced to 1000°C. After solution and aging treatment, the comprehensive mechanical properties of the forging were excellent, with KV impact value at room temperature as high as 92J, yield strength of tensile test at 350°C up to 696MPa, and tensile strength up to 1019MPa.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"38 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":"132325764","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

Performance Evaluation of Boron Doped NMCP Used in Neutron Radiography 硼掺杂NMCP在中子照相中的性能评价

Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components Pub Date : 2022-08-08 DOI: 10.1115/icone29-92974

Jinhao Tan, Jianrong Zhou, Yushou Song, Wenqin Yang, Jianjin Zhou, Yuan-guang Xia, Hui Liu, Shulin Liu, B. Yan, Songlin Wang, Zhi-jia Sun, Yuan-Bai Chen

{"title":"Performance Evaluation of Boron Doped NMCP Used in Neutron Radiography","authors":"Jinhao Tan, Jianrong Zhou, Yushou Song, Wenqin Yang, Jianjin Zhou, Yuan-guang Xia, Hui Liu, Shulin Liu, B. Yan, Songlin Wang, Zhi-jia Sun, Yuan-Bai Chen","doi":"10.1115/icone29-92974","DOIUrl":"https://doi.org/10.1115/icone29-92974","url":null,"abstract":"\u0000 Neutron radiography plays an important role in the development of safer and more efficient nuclear power plants by providing valuable information for the irradiated nuclear fuel, nuclear waste containers and other critical components. Neutron sensitive microchannel plate (nMCP) doped with boron and gadolinium is a popular neutron imaging detector. It has been successfully employed in neutron resonances imaging of irradiated nuclear fuel, providing the possibility in mapping the elemental composition of reactor components. However, the exiting of gadolinium in nMCP degrades the spatial resolution and neutron-gamma discrimination capability, because of the long range and low energy deposition of the electrons generated by gadolinium. nMCP doped with boron has advantages on spatial resolution and neutron-gamma discrimination capability. In this paper, performances of boron doped nMCP were studied. According to the simulation, the theoretical detection efficiency of nMCP doped with 20mol% 10B2O3 could reach 34%, and the spatial resolution can be close to the pore size at the same time. Moreover, it is expected to achieve good neutron-gamma discrimination ability. The detection efficiency for nMCP which doped with 10mol% natB2O3 was experimented detailly at CSNS. Results show that the detection efficiency is 4.55% for 2.9 Å neutron and 7.26% for 4.8 Å neutron, which agrees well with simulation results. These results indicate that the excellent performance of boron doped nMCP is possible to be achieved. In the future, nMCP doped with 20mol% 10B2O3 will be fabricated and it will be a promising scheme for the application of neutron radiography in nuclear reactors.","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":"132276222","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

Manufacturing Process of Fuel Salt Discharge Tank for 2Mwt Liquid Thorium Based Molten Salt Reactor 2Mwt液钍基熔盐堆燃料盐排放罐制造工艺

Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components Pub Date : 2022-08-08 DOI: 10.1115/icone29-91521

Yan Wang, Hongmei Shi

{"title":"Manufacturing Process of Fuel Salt Discharge Tank for 2Mwt Liquid Thorium Based Molten Salt Reactor","authors":"Yan Wang, Hongmei Shi","doi":"10.1115/icone29-91521","DOIUrl":"https://doi.org/10.1115/icone29-91521","url":null,"abstract":"\u0000 At present, the research on Thorium Based Molten Salt Reactor (TMSR) mainly focuses on molten salt-molten salt heat exchanger, molten salt-air heat exchanger, high-temperature molten salt pump, gas-liquid separator and Hastelloy structure and performance, and there is no research on the manufacturing process of fuel salt discharge tank. The fuel salt discharge tank mainly performs the function of fuel loading and unloading, that is, loading and unloading the fuel salt through the gas pressure. The difference gas pressure between the core and the fuel salt discharge tank make the fuel flow into the reactor pressure vessel or flow out of the reactor pressure vessel. The fuel salt discharge tank also has the function of storing the molten salt fuel and preventing fuel leakage when the core is overheated. It exists as a safety facility and has a very important impact on the normal and safety operation of the reactor system. Based on the requirements of standards and specifications, the main manufacturing process of fuel salt discharge tank were studied, technological processes such as tank material selection, head forming, barrel rolling, welding and pneumatic test were introduced. The fuel salt discharge tank of liquid TMSR meeting the requirements of standards and specifications was successfully developed.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"12 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":"115604854","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

Analytical Study for Low Ground Contact Ratio of Buildings Due to the Basemat Uplift Using a Three-Dimensional Finite Element Model 基于三维有限元模型的地基隆起低地面接触比分析研究

Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components Pub Date : 2022-08-08 DOI: 10.1115/icone29-93870

Byunghyun Choi, A. Nishida, T. Shiomi, Manabu Kawata, Yinsheng Li

{"title":"Analytical Study for Low Ground Contact Ratio of Buildings Due to the Basemat Uplift Using a Three-Dimensional Finite Element Model","authors":"Byunghyun Choi, A. Nishida, T. Shiomi, Manabu Kawata, Yinsheng Li","doi":"10.1115/icone29-93870","DOIUrl":"https://doi.org/10.1115/icone29-93870","url":null,"abstract":"\u0000 The basemat uplift is a phenomenon that the bottom of the basemat of a building partially rises from the ground due to overturning moments and vertical motions during earthquake. The degree of the basemat uplift can be indicated using a ground contact ratio which is defined as a ratio of the contact area of the bottom of the basemat to its entire area. The problem is that, at a large earthquake, the basemat uplift becomes large, so-called low ground contact ratio state, and the basemat falls resultant to recontact between the basemat and the ground, and large acceleration response occur on the floor of the building. It is a crucial aspect in the seismic evaluation of a nuclear facility building. It affects not only structural integrity of the building but also the response of the equipment installed in the building. However, the building behavior under the low ground contact ratio state lacks sufficient study. In this study, we conducted seismic response analyses for the building using a three-dimensional finite element model and simulated shaking table experiments focused on the basemat uplift and confirmed the validity of this analysis method. Since the basemat uplift is a strong non-linear phenomenon, we conducted computer simulations under the same analysis conditions with three different analysis codes, namely E-FrontISTR, FINAS/STAR, and TDAPIII, and compared the results. We investigated the influence on the structural response caused by the difference of the adhesive force of the basemat and the bearing ground at the low ground contact ratio state. In addition, we studied the influence of numerical parameters to the structural response through sensitivity analyses. This paper reports the analysis results and the insights obtained from our investigations.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"22 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":"121290233","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