Bolted-flange connected structure is widely used on Primary Equipment of Nuclear Power Plant and it is one of the most important assemblies of Primary Coolant Loop. Special attentions should be paid for the design and validation of this kind of structure for high risks of leakage and structure integrity of the equipment. In this paper, preload of manway bolt for Pressurizer of Nuclear Power Plant under hydrotest condition was calculated theoretically. Bolt-load distribution for the bolted-flange connected structure was simulated by Finite Element Analysis method with the above preload applied. Bolt load loss and asymmetric distribution during the test was presented resulting from the simulation for saddle-shaped manway flange. Assessment was carried out according to ASME in the aspect of stress in the bolt shank, pressure on the gasket and rotating angle of the flange (sealing verification) with the corrected preload applied. The simulation results were validated by 1:1 prototype test carrying out for the Pressurizer and they were consistent with the test results. The value of coefficient Ks for bolt preload design formula was confirmed by the test. Key aspects of influence on manway bolt preload design were summarized and full design formula as well as value of the design preload for hydrotest condition were given.
{"title":"Design and Validation on Preload of Manway Bolt for Pressurizer of Nuclear Power Plant Under Hydrotest Condition","authors":"Tao Chen, Guangming Xiong, Guanhua Fu, Xiaolong Zhao, Defu Wang, Yongqiang Zhang, Enming Liang","doi":"10.1115/icone29-91714","DOIUrl":"https://doi.org/10.1115/icone29-91714","url":null,"abstract":"\u0000 Bolted-flange connected structure is widely used on Primary Equipment of Nuclear Power Plant and it is one of the most important assemblies of Primary Coolant Loop. Special attentions should be paid for the design and validation of this kind of structure for high risks of leakage and structure integrity of the equipment. In this paper, preload of manway bolt for Pressurizer of Nuclear Power Plant under hydrotest condition was calculated theoretically. Bolt-load distribution for the bolted-flange connected structure was simulated by Finite Element Analysis method with the above preload applied. Bolt load loss and asymmetric distribution during the test was presented resulting from the simulation for saddle-shaped manway flange. Assessment was carried out according to ASME in the aspect of stress in the bolt shank, pressure on the gasket and rotating angle of the flange (sealing verification) with the corrected preload applied. The simulation results were validated by 1:1 prototype test carrying out for the Pressurizer and they were consistent with the test results. The value of coefficient Ks for bolt preload design formula was confirmed by the test. Key aspects of influence on manway bolt preload design were summarized and full design formula as well as value of the design preload for hydrotest condition were given.","PeriodicalId":422334,"journal":{"name":"Volume 12: Innovative and Smart Nuclear Power Plant Design","volume":"105 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":"134563331","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}
Shuzhi Chen, Meiliang Huang, Qimeng Shen, Jie Chen, Xiaoqiang Liu, Junli Wang, Weidong Wang, Yueming Fu
To achieve the in-depth combination of informatization and intelligence has become a development trend in the nuclear power industry, and the digitization/intelligence of instrumentation is a big part. Based on the technical status of nuclear power instruments and industrial integrated circuits, this paper analyzes the feasibility and suggestions of intelligent/digital instrument application for nuclear power plant by combing the environmental conditions of nuclear power, intelligence/digitization of typical meters, testing and verification etc. basic work.
{"title":"Feasibility Analysis of Intelligent Instrument Application In Nuclear Power","authors":"Shuzhi Chen, Meiliang Huang, Qimeng Shen, Jie Chen, Xiaoqiang Liu, Junli Wang, Weidong Wang, Yueming Fu","doi":"10.1115/icone29-93298","DOIUrl":"https://doi.org/10.1115/icone29-93298","url":null,"abstract":"\u0000 To achieve the in-depth combination of informatization and intelligence has become a development trend in the nuclear power industry, and the digitization/intelligence of instrumentation is a big part. Based on the technical status of nuclear power instruments and industrial integrated circuits, this paper analyzes the feasibility and suggestions of intelligent/digital instrument application for nuclear power plant by combing the environmental conditions of nuclear power, intelligence/digitization of typical meters, testing and verification etc. basic work.","PeriodicalId":422334,"journal":{"name":"Volume 12: Innovative and Smart Nuclear Power Plant Design","volume":"117 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":"117126799","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}
With the development of a new generation of information technology, the monitoring systems in the nuclear power field are also facing the technological revolution of digital reform. Based on the example of Thorium Molten Salt Reactor-Solid Fuel (TMSR-SF0), this paper introduces the overall structure, modeling and mapping, and digital visualization expression means, then a complete development scheme of the real-time online 3D monitoring system is proposed. Firstly, the digital mapping model of the molten salt reactor was established, and the model docking and virtual scene rendering were completed in the Unity engine. Secondly, the back-end server using Experimental Physical and Industrial Control System (EPICS) and a proxy server using Node.js are established, and spatio-temporal data association is realized through Node-EPICS event driver and Socket.io. Finally, the MySql database interface and xCharts visualization framework are introduced to complete data storage and expression. Verified by practice, The 3D monitoring system is feasible in design and operation. Compared with the previous monitoring system, the data update period is increased to 20ms, and it has the whole process of data collection, network communication, graph element dynamic display, and other functions, which can promote the operation management and monitoring of nuclear power plants, and provide a reference for the digital transformation of monitoring technology in the nuclear power field.
随着新一代信息技术的发展，核电领域的监测系统也面临着数字化改革的技术革命。以钍熔盐堆-固体燃料(TMSR-SF0)为例，介绍了系统的总体结构、建模与制图、数字化可视化表达手段，提出了实时在线三维监测系统的完整开发方案。首先，建立熔盐堆的数字映射模型，在Unity引擎中完成模型对接和虚拟场景渲染;其次，建立基于实验物理与工业控制系统(Experimental Physical and Industrial Control System, EPICS)的后端服务器和基于Node.js的代理服务器，并通过Node-EPICS事件驱动和Socket.io实现时空数据关联。最后，介绍了MySql数据库接口和xCharts可视化框架，完成了数据的存储和表达。实践证明，该三维监控系统在设计和运行上是可行的。与以往的监测系统相比，数据更新周期提高到20ms，具有数据采集全过程、网络通信、图形元素动态显示等功能，可促进核电站的运行管理和监测，为核电领域监测技术的数字化转型提供参考。
{"title":"Digital Twin-Driven Development of Online Monitoring and Data Management Systems in TMSR-SF0","authors":"Wen-Qian Liu, Li Han, Li Huang","doi":"10.1115/icone29-92770","DOIUrl":"https://doi.org/10.1115/icone29-92770","url":null,"abstract":"\u0000 With the development of a new generation of information technology, the monitoring systems in the nuclear power field are also facing the technological revolution of digital reform. Based on the example of Thorium Molten Salt Reactor-Solid Fuel (TMSR-SF0), this paper introduces the overall structure, modeling and mapping, and digital visualization expression means, then a complete development scheme of the real-time online 3D monitoring system is proposed. Firstly, the digital mapping model of the molten salt reactor was established, and the model docking and virtual scene rendering were completed in the Unity engine. Secondly, the back-end server using Experimental Physical and Industrial Control System (EPICS) and a proxy server using Node.js are established, and spatio-temporal data association is realized through Node-EPICS event driver and Socket.io. Finally, the MySql database interface and xCharts visualization framework are introduced to complete data storage and expression. Verified by practice, The 3D monitoring system is feasible in design and operation. Compared with the previous monitoring system, the data update period is increased to 20ms, and it has the whole process of data collection, network communication, graph element dynamic display, and other functions, which can promote the operation management and monitoring of nuclear power plants, and provide a reference for the digital transformation of monitoring technology in the nuclear power field.","PeriodicalId":422334,"journal":{"name":"Volume 12: Innovative and Smart Nuclear Power Plant Design","volume":"55 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":"122061641","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}
The reactor system has a large scale of components. And the system dynamic analysis model for loss of coolant accidents (LOCA) contains many non-linear factors. The transient calculation analysis takes a long time and the convergence is difficult to guarantee. In this paper, two different analysis models (complete three-loop model and single-loop model) of the reactor loop system are established for comparative analysis. The analysis shows that the calculation error of the main position load in the two different model is within 10%. The decoupling influence of the surge tube and the reactor inlet and outlet on the system dynamic characteristics is within an acceptable range. When the single-loop model is used, it is necessary to decouple the reactor pressure vessel and pressurizer. And the simplified boundary is adapted to simulate its effect on the whole loop. The single-loop dynamic analysis model based on the decoupling criterion can greatly reduce the calculation scale and meet the calculation accuracy, which can be used to calculate the LOCA dynamic characteristics of the reactor system quickly. The model can also be used for seismic analysis of reactor system. And using this model for reactor system dynamic analysis can greatly improve the calculation efficiency.
{"title":"Research on LOCA Dynamic Analysis Model of Reactor System","authors":"Shuai Liu","doi":"10.1115/icone29-92491","DOIUrl":"https://doi.org/10.1115/icone29-92491","url":null,"abstract":"\u0000 The reactor system has a large scale of components. And the system dynamic analysis model for loss of coolant accidents (LOCA) contains many non-linear factors. The transient calculation analysis takes a long time and the convergence is difficult to guarantee. In this paper, two different analysis models (complete three-loop model and single-loop model) of the reactor loop system are established for comparative analysis. The analysis shows that the calculation error of the main position load in the two different model is within 10%. The decoupling influence of the surge tube and the reactor inlet and outlet on the system dynamic characteristics is within an acceptable range. When the single-loop model is used, it is necessary to decouple the reactor pressure vessel and pressurizer. And the simplified boundary is adapted to simulate its effect on the whole loop. The single-loop dynamic analysis model based on the decoupling criterion can greatly reduce the calculation scale and meet the calculation accuracy, which can be used to calculate the LOCA dynamic characteristics of the reactor system quickly. The model can also be used for seismic analysis of reactor system. And using this model for reactor system dynamic analysis can greatly improve the calculation efficiency.","PeriodicalId":422334,"journal":{"name":"Volume 12: Innovative and Smart Nuclear Power Plant Design","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":"132182600","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}
With rapid development of informatization and digitalization technologies, digital transformation is pushing ahead in industry around the world, especially in nuclear power operation and maintenance. In the full life-cycle digital transformation plan of nuclear power enterprises, the design institute has accumulated a large number of digital design results, such as the three-dimensional model data of nuclear power plants. These 3D models are handed over to the operation organizations to build digital, visual and intelligent operation and maintenance solutions. Without digital models, field staffs have to check the paper procedures and technical data before or even during the on-site inspection and maintenance preparation, however, this is adverse for field work. The application of 3D digital auxiliary tools and platforms can effectively improve management efficiency and reduce potential risks in nuclear power operation and maintenance. This paper proposed a digitalization and visualization method to improve the digital management level, and developed a 3D visual operation management platform for Daya Bay nuclear power plant. With the platform, nuclear power operation organizations could handle the daily maintenance work in an efficient way, so that to improve the operation management level.
{"title":"Research and Development of Three-Dimensional Digital Operation Management Platform for NPPs","authors":"Hao Wang, Licheng Tian, Jian Lin, Huimin Huang","doi":"10.1115/icone29-92994","DOIUrl":"https://doi.org/10.1115/icone29-92994","url":null,"abstract":"\u0000 With rapid development of informatization and digitalization technologies, digital transformation is pushing ahead in industry around the world, especially in nuclear power operation and maintenance. In the full life-cycle digital transformation plan of nuclear power enterprises, the design institute has accumulated a large number of digital design results, such as the three-dimensional model data of nuclear power plants. These 3D models are handed over to the operation organizations to build digital, visual and intelligent operation and maintenance solutions. Without digital models, field staffs have to check the paper procedures and technical data before or even during the on-site inspection and maintenance preparation, however, this is adverse for field work. The application of 3D digital auxiliary tools and platforms can effectively improve management efficiency and reduce potential risks in nuclear power operation and maintenance. This paper proposed a digitalization and visualization method to improve the digital management level, and developed a 3D visual operation management platform for Daya Bay nuclear power plant. With the platform, nuclear power operation organizations could handle the daily maintenance work in an efficient way, so that to improve the operation management level.","PeriodicalId":422334,"journal":{"name":"Volume 12: Innovative and Smart Nuclear Power Plant Design","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":"130451288","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}
This research aims to investigate the differences in structural integrity assessment results between R5 Volume 2/3 procedures and a detailed inelastic analysis. The membrane sidewall/roof bimetallic stub tube subjected to creep-fatigue loading is selected as an example. Based on a linear elastic analysis, the maximum equivalent stress is found to be at the weldment located at the top of the stub tube. The R5 Volume 2/3 procedures enhance the start of a dwell stress and inelastic strain ranges by applying a weld strain enhancement factor and a stress concentration factor, and the assessment predicts a total creep-fatigue damage of 54% for a 45 year life. As a comparison, a detailed inelastic analysis has been conducted for the same bi-metallic stub tube using an incremental, step-by-step analysis. Temperature dependent material properties are used to simulate stress-strain response and a user subroutine is employed for creep deformation. Assessment results for the detailed inelastic analysis confirm the weldment is similarly affected as indicated by the elastic analysis results but the comparable total predicted creep-fatigue damage is less than 35%. Cold eye review of both assessment results concluded that the significant deviation is attributed to the additional factors applied for the assessment of weldments, including weldment geometry, flaw effects and weld micro-cracking, as provided in the R5 Volume 2/3 procedures.
{"title":"Comparative Study on Creep-Fatigue Damage Assessment Between R5 Procedures and a Full Inelastic Analysis","authors":"A. Hurst, Louis C. W. Chang, N. Cho","doi":"10.1115/icone29-91707","DOIUrl":"https://doi.org/10.1115/icone29-91707","url":null,"abstract":"\u0000 This research aims to investigate the differences in structural integrity assessment results between R5 Volume 2/3 procedures and a detailed inelastic analysis. The membrane sidewall/roof bimetallic stub tube subjected to creep-fatigue loading is selected as an example. Based on a linear elastic analysis, the maximum equivalent stress is found to be at the weldment located at the top of the stub tube. The R5 Volume 2/3 procedures enhance the start of a dwell stress and inelastic strain ranges by applying a weld strain enhancement factor and a stress concentration factor, and the assessment predicts a total creep-fatigue damage of 54% for a 45 year life. As a comparison, a detailed inelastic analysis has been conducted for the same bi-metallic stub tube using an incremental, step-by-step analysis. Temperature dependent material properties are used to simulate stress-strain response and a user subroutine is employed for creep deformation. Assessment results for the detailed inelastic analysis confirm the weldment is similarly affected as indicated by the elastic analysis results but the comparable total predicted creep-fatigue damage is less than 35%. Cold eye review of both assessment results concluded that the significant deviation is attributed to the additional factors applied for the assessment of weldments, including weldment geometry, flaw effects and weld micro-cracking, as provided in the R5 Volume 2/3 procedures.","PeriodicalId":422334,"journal":{"name":"Volume 12: Innovative and Smart Nuclear Power Plant Design","volume":"26 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":"126599106","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}
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