Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components最新文献

Evaluation on Weldability of Nickel-Base Alloy Weld Materials With High Chromium Content 高铬镍基合金焊接材料的可焊性评价

Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components Pub Date : 2022-08-08 DOI: 10.1115/icone29-89096

Y. Katayama, T. Hayashi, Daiki Tanaka, Hidenori Takahashi

{"title":"Evaluation on Weldability of Nickel-Base Alloy Weld Materials With High Chromium Content","authors":"Y. Katayama, T. Hayashi, Daiki Tanaka, Hidenori Takahashi","doi":"10.1115/icone29-89096","DOIUrl":"https://doi.org/10.1115/icone29-89096","url":null,"abstract":"Nickel (Ni)-base alloy 82 weld material is widely used for the primary reactor components of boiling water reactor (BWR). Although alloy 82 has shown excellent service performance for BWR application in terms of material reliability, several experiences have recently been reported regarding stress corrosion cracking (SCC) in welds associated with alloy 82. The present study aimed at enhancing SCC resistance of the weld material by developing and evaluating alloy 82 with increased chromium (Cr) content of 30%, 33% and 36%. Weld joint specimens were prepared by tungsten inert gas (TIG) welding with the prototyped wires of alloy 82 with increased Cr content. The mechanical properties and weldability were investigated using the weld joint specimens. For comparison, the conventional alloy 82 and alloy 52 were also tested. The mechanical properties of alloy 82 with increased Cr content were equal to or better than those of the conventional alloy 82. Varestraint test and bending test were carried out to determine the weldability. In the Varestraint test, the susceptibility to solidification cracking and ductility-dip cracking were evaluated. The temperature ranges of the ductility-dip cracking of the alloy 82 with increased Cr content was about the same as that of the conventional alloy 82. As a result of these tests, it was confirmed that the alloy 82 with increased Cr content has the same weldability as that of the conventional alloy 82 and was applicable for BWR components.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"103 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":"123805711","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 Crack Resistance of Nuclear Grade Nickel-Based Alloy Welding Materials 核级镍基合金焊接材料抗裂性能研究

Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components Pub Date : 2022-08-08 DOI: 10.1115/icone29-89321

Ying Ao, C. Xu, G. He, Zhanpeng Lv, Ping Zhu, Maolong Zhang, J. Zou

{"title":"Research on Crack Resistance of Nuclear Grade Nickel-Based Alloy Welding Materials","authors":"Ying Ao, C. Xu, G. He, Zhanpeng Lv, Ping Zhu, Maolong Zhang, J. Zou","doi":"10.1115/icone29-89321","DOIUrl":"https://doi.org/10.1115/icone29-89321","url":null,"abstract":"\u0000 This paper mainly introduces the anti-crack performance of the nuclear grade nickel-based alloy welding materials produced by Baowu Special Metallurgy Co., Ltd. The research materials include welding wires (ERNiCrFe-7A), covered electrodes (ENiCrFe-7) and welding strips (EQNiCrFe-7A) used in PWR nuclear power plants. Evaluation of the anti-crack performance of the deposited metal is mainly carried out from the following three aspects: In order to study the thermal crack (crystal cracks) sensitivity of Baowu nickel-based alloy welding materials, the adjustable constraint test method is used and the results are compared with those of imported commercial welding materials; Based on STF (strain-to-fracture test) experimental method, the sensitivity of ductility dip cracking (DDC) of Baowu and imported nickel-based alloy welding materials is studied by thermal simulation equipment; The stress corrosion cracking (SCC) performance evaluation of Baowu welding materials is carried out by stress corrosion crack propagation experiment in simulated PWR primary water and compared with imported materials respectively. According to the above research, the crystal crack resistance, DDC resistance and SCC resistance of Baowu nickel-based alloy welding materials have reached the level of international high-level commercial welding material, which will further provide experimental data support for the engineering applicability and performance reliability of domestic nickel-based welding materials.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"27 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":"127679765","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 of the Standards Compilation of Welding Material for Carbon Steel in Pressurized Water Reactor Nuclear Power Plants 压水堆核电站碳钢焊接材料标准编制研究

Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components Pub Date : 2022-08-08 DOI: 10.1115/icone29-92165

Bai Lin, Liang Xiaoru

{"title":"Research of the Standards Compilation of Welding Material for Carbon Steel in Pressurized Water Reactor Nuclear Power Plants","authors":"Bai Lin, Liang Xiaoru","doi":"10.1115/icone29-92165","DOIUrl":"https://doi.org/10.1115/icone29-92165","url":null,"abstract":"\u0000 Carbon steel welding is a major manufacturing method for nuclear power construction. For the moment, there are two sets of standards of welding material for carbon steel which used in pressurized water reactor nuclear power plants of China. The coexistence of standards of two different technical routes brings unnecessary cost and waste to pressurized water reactor nuclear power plant manufacturing. With the rapid development of pressurized water reactor nuclear power plants in China, the research of developing a set of unified standards for carbon steel electrodes and rods for gas shielded arc welding and carbon steel covered electrodes is necessary. At present, two main standards will be revised and published. Through the interpretation of “Welding material for pressurized water reactor nuclear power plants — Part 1 Carbon steel covered electrodes” and “Welding material for pressurized water reactor nuclear power plants — Part 6 Carbon steel electrodes and rods for gas shielded arc welding” and the introduction to their compilation thoughts and technical points, the article guides engineers to understand the compilation thoughts of these two standards quickly. The standards specify classification, technical requirements, test methods and testing of carbon steel electrodes and rods for carbon steel in pressurized water reactor nuclear power plants.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"119 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":"125990532","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

High Temperature Gas Cooled Reactor Steam Generator Research on Stress-Rupture Test of High Temperature Materials and Welded Joints 高温气冷堆蒸汽发生器高温材料及焊接接头应力断裂试验研究

Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components Pub Date : 2022-08-08 DOI: 10.1115/icone29-92956

Shimei Bao

{"title":"High Temperature Gas Cooled Reactor Steam Generator Research on Stress-Rupture Test of High Temperature Materials and Welded Joints","authors":"Shimei Bao","doi":"10.1115/icone29-92956","DOIUrl":"https://doi.org/10.1115/icone29-92956","url":null,"abstract":"\u0000 The outlet design temperature of the steam generator of high temperature gas-cooled reactor is 570°C, and the design temperature of the high temperature components have reached or exceeded the creep temperature of the material. In the creep temperature range, the allowable stress of the material mainly depends on its high temperature stress-rupture strength. Therefore, it is necessary to High-temperature stress-rupture test studies are carried out on materials and welded joints to verify whether their stress-rupture strength can meet the design requirements. High-temperature stress-rupture tests were carried out on 9 kinds of high-temperature materials and 5 kinds of welded joints by designing the non-standard specimen for heat exchange tubes, reasonably setting test stress points, parallel samples and sampling positions.14 double logarithmic curves of stress-rupture time were drawn, and the long-term high-temperature stress-rupture strength of high-temperature materials and welded joints was extrapolated, Providing important data support for the safety evaluation and life evaluation to the steam generator of high temperature gas-cooled reactor.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"16 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":"133670210","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 the Design Method of Nuclear Heat Exchanger Tube Based on Multi-Objective Optimization 基于多目标优化的核换热管设计方法研究

Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components Pub Date : 2022-08-08 DOI: 10.1115/icone29-90436

G. Cao, L. Dai, Z. X. Zhao, T. He

引用次数: 0

Investigation on Irradiation Resistance of Anti-Galling Lubricant for High Temperature Fasteners of Nuclear Power Plant 核电站高温紧固件防磨损润滑剂耐辐照性能研究

Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components Pub Date : 2022-08-08 DOI: 10.1115/icone29-92166

Xiao Qiang Liu, Ming Jiao, Xiuqiang Shi, Yichen Bao

{"title":"Investigation on Irradiation Resistance of Anti-Galling Lubricant for High Temperature Fasteners of Nuclear Power Plant","authors":"Xiao Qiang Liu, Ming Jiao, Xiuqiang Shi, Yichen Bao","doi":"10.1115/icone29-92166","DOIUrl":"https://doi.org/10.1115/icone29-92166","url":null,"abstract":"\u0000 Lubricant grease is a special material commonly used in the field of nuclear power plant. It is widely used in the sliding fit and press fit parts of bolts, studs and valves to prevent thread wear, seizing as well as stress corrosion cracking of fasteners during assembly, service and disassembly. In this paper, the performance of lubricant materials such as wear resistance and high temperature oxidation resistance under irradiation environment were studied. The results show that the lubricant grease does not change significantly and there is no grease stratification after irradiation with 0.1MGy integrated irradiation dose. However, after irradiation with 10MGy integrated irradiation dose, the irradiation has an obvious destructive effect on the oxidation resistance performance of the grease, and the grease appears obvious stratification. With the increase of irradiation dose, the thermal decomposition temperature first decreases and then increases. Under 0.1MGy irradiation dose, the thermal decomposition loss of the sample changes little, but at 10MGy irradiation dose, a large amount of mass loss occurred, up to more than 80%. The four-ball friction experiment shows that its extreme pressure resistance improves with the increase of irradiation dose, and the sintering load increases. These data provide technical guidance for the long-term reliability of nuclear power fastener materials in service.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"42 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":"128422274","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 Fatigue Analysis Method of Nuclear Equipment Under Ocean Conditions 海洋条件下核设备疲劳分析方法研究

Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components Pub Date : 2022-08-08 DOI: 10.1115/icone29-92711

Guangdong Liu, He Zhu, Zhui Zhang, G. Zhang, Guitao Cao, Lin Tan, J. Qi

{"title":"Research on Fatigue Analysis Method of Nuclear Equipment Under Ocean Conditions","authors":"Guangdong Liu, He Zhu, Zhui Zhang, G. Zhang, Guitao Cao, Lin Tan, J. Qi","doi":"10.1115/icone29-92711","DOIUrl":"https://doi.org/10.1115/icone29-92711","url":null,"abstract":"\u0000 Under the conditions of the marine environment (wind, wave, current), the floating platform on the ocean undergoes heeling, trimming, rolling, pitching, swaying, surging, heaving movements. The floating reactor on the floating platform within the design life, in addition to withstand the low-frequency fatigue caused by primary loop temperature and pressure, it also has to withstand the inertial force generated by the sway of the wave for the service period. The inertial force generated by the wave is a constantly changing dynamic load. Alternating stresses inside of the structure may cause fatigue damage to the structure. In order to estimate the fatigue life of floating reactor equipment under various ocean conditions, a finite element analysis model of equipment was established to carry out research for fatigue analysis methods under sway loads, meanwhile the operating temperature and pressure loads were considered. The probability of sway at each angle obeys the Rayleigh distribution, through the analysis of the status for the platform, it can estimate the number of swings at each angle. Based on Miner linear cumulative damage theory, the cumulative fatigue usage factors meet the requirements of the ASME BPVC III NB-3220 code. This analysis method provides a reference for the fatigue analysis of the floating reactor nuclear equipment.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"2014 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":"132258426","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 Crown Predictive Model of Cold-Rolled Zirconium Alloy Strip Sheet Based on Machine Learning Algorithm 基于机器学习算法的冷轧锆合金带材凸度预测模型

Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components Pub Date : 2022-08-08 DOI: 10.1115/icone29-92252

Yuan Cao, Jianguo Cao, Yingqi Gao, Ben Wang, Pengfei Zhang, Fang Li, Bo Gao

引用次数: 0

High-Temperature Deformation Constitutive Relationship of Zr-Sn-Nb Alloy for Nuclear Power 核电用Zr-Sn-Nb合金高温变形本构关系

Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components Pub Date : 2022-08-08 DOI: 10.1115/icone29-92123

Ying Gao, Ben Wang, Jian-Wu Cao, Qianqian Luo, Yuan Cao, Zhihao Zhang, Lian Wang, Bo Gao

{"title":"High-Temperature Deformation Constitutive Relationship of Zr-Sn-Nb Alloy for Nuclear Power","authors":"Ying Gao, Ben Wang, Jian-Wu Cao, Qianqian Luo, Yuan Cao, Zhihao Zhang, Lian Wang, Bo Gao","doi":"10.1115/icone29-92123","DOIUrl":"https://doi.org/10.1115/icone29-92123","url":null,"abstract":"\u0000 Zr-Sn-Nb zirconium alloys are widely used in the field of nuclear power instead of traditional zirconium alloys and can be used as structural materials for nuclear power fuel assemblies. In this paper, the constitutive relation of high temperature plastic deformation of Zr-Sn-Nb alloy for nuclear power is described. Gleeble-3800 thermal simulation tester was used to conduct compression tests at different temperatures (450∼700°C) and deformation rates (0.01∼10s−1), stress-strain curves under different deformation conditions were drawn based on stress-strain data. According to the analysis of flow stress curve types, Zr-Sn-Nb alloy shows obvious temperature sensitivity and strain rate sensitivity. Under the condition of low temperature and high strain rate, the power dissipation efficiency is calculated according to Murty instability discriminant. Zirconium alloy material shows flow instability due to the formation of adiabatic shear band. In order to control the successful rolling of microstructure, careful process design must be carried out to avoid the flow instability zone. Based on the types of stress-strain curves, the curves are divided into work hardening stage and dynamic recovery (DRV)/ dynamic recrystallization (DRX) stage. A piecewise constitutive model is established, which can accurately predict the machining properties of Zr-Sn-Nb alloy for nuclear power.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"19 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":"126374748","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

Optimization Configuration and Design Considerations of Spent Fuel Storage and Transport Casks for Nuclear Power Plants in China 中国核电厂乏燃料储运桶的优化配置与设计考虑

Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components Pub Date : 2022-08-08 DOI: 10.1115/icone29-92905

Fengwu Liu, Shiyuan Li, Lijun Wang, Shunping Jing, Chao Liu, Xiaoqing Li

{"title":"Optimization Configuration and Design Considerations of Spent Fuel Storage and Transport Casks for Nuclear Power Plants in China","authors":"Fengwu Liu, Shiyuan Li, Lijun Wang, Shunping Jing, Chao Liu, Xiaoqing Li","doi":"10.1115/icone29-92905","DOIUrl":"https://doi.org/10.1115/icone29-92905","url":null,"abstract":"\u0000 With the rapid development of nuclear power in China, the amount of spent fuel keeps increasing, which makes the demand for safe and suitable casks to store and transport spent fuel more prominent. Due to China’s national policy of “closed nuclear fuel cycle” and the uncertainties in the construction of reprocessing facilities, the configuration of spent fuel storage and transport cask has been directly related to the safety of nuclear power plants operation. Based on the demand of spent fuel storage and transport in China, this paper analyzes the requirement of different modes of storage and transportation and different loading contents for casks, such as 12ft / 14ft spent fuel loading, compatibility with short-term storage and transportation, and off-site transportation of dry storage canisters, etc. In order to ensure safety and control cost, this paper also analyzed the applicable standards, transport restrictions and optimization design objective for the optimal design of spent fuel storage and transport cask, and puts forward the optimal configuration and design method of the storage and transport cask based on the analysis of key concerns, which could provide reference for the configuration and design of spent fuel storage cask.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"130 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":"129814173","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