Volume 1: Nuclear Plant Operation and Maintenance, Engineering and Modification, Operation Life Extension (OLE), and Life Cycle最新文献_第9页

On Line Monitoring of Sensor Failure in Nuclear Power Plant Based on Sequential Probability Ratio Test 基于序贯概率比试验的核电站传感器故障在线监测

Volume 1: Nuclear Plant Operation and Maintenance, Engineering and Modification, Operation Life Extension (OLE), and Life Cycle Pub Date : 2022-08-08 DOI: 10.1115/icone29-89215

Fangxiaozhi Yu, Xintong Ma, Zhuang Miao, Min Zhang

{"title":"On Line Monitoring of Sensor Failure in Nuclear Power Plant Based on Sequential Probability Ratio Test","authors":"Fangxiaozhi Yu, Xintong Ma, Zhuang Miao, Min Zhang","doi":"10.1115/icone29-89215","DOIUrl":"https://doi.org/10.1115/icone29-89215","url":null,"abstract":"\u0000 The safe and reliable operation of nuclear power plant is inseparable from the integrity of the control system function, and the integrity of the sensor function is the basis for the I&C system to monitor the operation status of nuclear power plant and trigger automatic action. However, there are some shortcomings in the traditional threshold alarm method of nuclear power plant: for example, it is difficult to identify the tiny sensor failure (such as tiny drift) earlier by this method. The threshold of this method is fixed and cannot be adjusted according to the actual situation (such as transient change). This paper introduces the on-line monitoring technology of sensor failure, which plays an important role in reducing the probability of unplanned shutdown and unit design and operation cost. In the future, the on-line monitoring technology of sensor failure has a good application prospect, and the relevant research is of great significance.\u0000 This paper mainly studies an overall technical scheme of sensor failure monitoring in nuclear power plant. In this paper, the data of the PWR nuclear power plant simulation simulator are used, and the sequential probability ratio test (SPRT) model, a typical failure diagnosis algorithm, is used to establish the sensor condition monitoring model, and the performance of the monitoring model are tested. The results show that the SPRT sensor monitoring model established in this paper has high accuracy, and can effectively monitor the occurrence of tiny drift phenomenon of sensors, and preliminarily verify the feasibility of on-line monitoring technology of sensor failure in PWR nuclear power plant.","PeriodicalId":110070,"journal":{"name":"Volume 1: Nuclear Plant Operation and Maintenance, Engineering and Modification, Operation Life Extension (OLE), and Life Cycle","volume":"194 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":"116282963","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

Cause Analysis and Treatment Measures for Corrosion of Warm-Up Pipe of Hualong no.1 Main Feed Water Pump System 华龙1号暖管腐蚀原因分析及处理措施主给水泵系统

Volume 1: Nuclear Plant Operation and Maintenance, Engineering and Modification, Operation Life Extension (OLE), and Life Cycle Pub Date : 2022-08-08 DOI: 10.1115/icone29-93343

Zhiming Wu, Yixin Ling, Xing Liu, Dongdong Fei, Xiaoyu Hou, Xue Wen

{"title":"Cause Analysis and Treatment Measures for Corrosion of Warm-Up Pipe of Hualong no.1 Main Feed Water Pump System","authors":"Zhiming Wu, Yixin Ling, Xing Liu, Dongdong Fei, Xiaoyu Hou, Xue Wen","doi":"10.1115/icone29-93343","DOIUrl":"https://doi.org/10.1115/icone29-93343","url":null,"abstract":"\u0000 The Main Feedwater System of Fuqing nuclear power plant is composed of three electric pump sets arranged in parallel with 50% capacity to provide the required Feedwater for the secondary side of steam generator. Under normal conditions, two pumps are in operation, one is standby, and three pumps can be switched at will. When one of the two running electric feed water pump sets trips, the electric feed water pump set in standby state will start quickly. If the equipment has problems, it will seriously affect the safe and reliable operation of the unit.\u0000 After the leakage defect of the warm-up pump pipe downstream of the warm-up pump stop valve in train C of the main feed water pump system of Hualong unit was found, through the disassembly inspection, it was found that there were many surface corrosion pits in the downstream pipeline of the throttle orifice plate. In this paper, through the physical and chemical analysis of the pipeline composition, organization, non-metallic inclusions, surface quality, macro and micro morphology, etc, At the same time, combined with the operating conditions of the pipeline, analyze the causes of the defects, judge the defects caused by the medium vaporization caused by the excessive throttling of the throttle orifice, and give effective treatment measures. Through these relevant measures, the corrosion rate of the warm-up pump pipe is reduced and the operation reliability of the important equipment of the unit is improved.","PeriodicalId":110070,"journal":{"name":"Volume 1: Nuclear Plant Operation and Maintenance, Engineering and Modification, Operation Life Extension (OLE), and Life Cycle","volume":"31 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":"115931249","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

Comparative Model of Nuclear Power Plant General Layout Scheme 核电站总平面布置方案的比较模型

Volume 1: Nuclear Plant Operation and Maintenance, Engineering and Modification, Operation Life Extension (OLE), and Life Cycle Pub Date : 2022-08-08 DOI: 10.1115/icone29-93554

Chong Guang Huang, Yan Wu

{"title":"Comparative Model of Nuclear Power Plant General Layout Scheme","authors":"Chong Guang Huang, Yan Wu","doi":"10.1115/icone29-93554","DOIUrl":"https://doi.org/10.1115/icone29-93554","url":null,"abstract":"\u0000 Nuclear power plant sites are scarce resources for nuclear energy development. It is necessary to optimize the general layout scheme of nuclear power plant to reduce the project cost and operation cost. In the feasibility study stage of the project, the determination of the general layout scheme of nuclear power plant can provide design input for the land acquisition negotiation, geological prospecting, environmental assessment, site safety analysis and other topics. Then the rationality of the nuclear power plant project is proved more strongly.\u0000 The work content of general drawing runs through the whole process of power plant engineering construction, involving many aspects such as seismic geology, hydrometeorology, and bulky transportation. Among them, the comparison and selection of the general layout scheme of nuclear power plant is a key step of the demonstration in the early stage of the project, and it needs to be iterated with several topics in the early stage to be reasonably determined. Therefore, it is very important to find a method to compare and choose the optimal general layout scheme of power plant.\u0000 In this paper, a fuzzy comprehensive evaluation method based on AHP is proposed to establish an evaluation index system for the general layout of nuclear power plants. By comprehensive analysis of the main influencing factors of the project such as the self-human condition, three overall layout schemes are quantitatively evaluated from each dimension, and the optimal scheme is obtained.","PeriodicalId":110070,"journal":{"name":"Volume 1: Nuclear Plant Operation and Maintenance, Engineering and Modification, Operation Life Extension (OLE), and Life Cycle","volume":"9 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":"130561892","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

A Proof-of-Concept Application of Fast Fracture Evaluation for the Primary Loop Section in Nuclear Power Plant 核电一次回路快速断裂评价的概念验证应用

Volume 1: Nuclear Plant Operation and Maintenance, Engineering and Modification, Operation Life Extension (OLE), and Life Cycle Pub Date : 2022-08-08 DOI: 10.1115/icone29-93082

Zihao Zheng, Wenxin Qiu, Yang Dai, Jiesheng Min

{"title":"A Proof-of-Concept Application of Fast Fracture Evaluation for the Primary Loop Section in Nuclear Power Plant","authors":"Zihao Zheng, Wenxin Qiu, Yang Dai, Jiesheng Min","doi":"10.1115/icone29-93082","DOIUrl":"https://doi.org/10.1115/icone29-93082","url":null,"abstract":"\u0000 The primary loop in nuclear power plants (NPPs) consists of piping and pressure vessel, which operate under a high level of pressure and temperature. During the long-term service, these sections need to have their integrity analyzed to check whether they might be prone to degradation, especially when exposed to radiation. In-service codes and standards such as the RSE-M and the ASME Section VIII enclose rules for the analysis of defects in piping systems. Although geometries for pipes are well-known and the fracture analyses methods are all today quite mature, it still takes an amount of time for structural engineers to rebuild models of different defects on piping section and the evaluation results varies with different modelling methods.\u0000 Hence, this paper provides a proof-of-concept tool for fast fracture evaluation, specific to the brittle fracture induced by thermal fatigue and radiation in the primary loop. In this application, the procedure of finite element method (FEM) analysis and fracture-related criteria is well integrated, based on the open-source FEM code code_aster and multi-tool platform SALOME. It automates several steps of primary loop section simulation, including mesh generation, calculation file configuration, post-processing and reporting results.\u0000 Furthermore, the verification and validation tests are built to ensure the robustness and reliability of the software and productivity.","PeriodicalId":110070,"journal":{"name":"Volume 1: Nuclear Plant Operation and Maintenance, Engineering and Modification, Operation Life Extension (OLE), and Life Cycle","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":"128411623","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

Rapid Assessment Method for Corrosion Flaws of Seawater Cooling Pipes in Nuclear Power Plants 核电站海水冷却管腐蚀缺陷快速评估方法

Volume 1: Nuclear Plant Operation and Maintenance, Engineering and Modification, Operation Life Extension (OLE), and Life Cycle Pub Date : 2022-08-08 DOI: 10.1115/icone29-92871

Zhichao Han, Haijun Chang, Mengling Shen, Zhishen Rong, Zhengpeng Li

{"title":"Rapid Assessment Method for Corrosion Flaws of Seawater Cooling Pipes in Nuclear Power Plants","authors":"Zhichao Han, Haijun Chang, Mengling Shen, Zhishen Rong, Zhengpeng Li","doi":"10.1115/icone29-92871","DOIUrl":"https://doi.org/10.1115/icone29-92871","url":null,"abstract":"\u0000 The corrosion of the fluid to the pipes could damage the pipe structure, and then affect the safe operation of the whole pipeline system in the nuclear power plant. In order to assess the influence of the corrosion flaws on pipe integrity, it is necessary to perform the stress analysis and calculate the remaining strength of the pipes containing flaws. In this paper, the corrosion flaws in the seawater cooling pipes are considered as the volumetric type flaws. The remaining strength of the pipes containing the flaws has been calculated, the minimum required wall thickness of the pipes has been obtained, and the flaws have been assessed according to SY/T 6477 Standard. The finite element method has been used to analyze the pipes containing corrosion flaws. Stress distribution of the pipes has been obtained under the pressure and nozzle loading. Based on the RCC-M Standard, the strength assessment has been conducted. Results showed that the stress calculated from SY/T 6477 Standard is close to that from the finite element method, and it can be considered that the assessment method summarized from SY/T 6477 Standard is suitable for the current pipes containing flaws under the nozzle loadings. Therefore, this method can be used to assess the corrosion flaws of the seawater cooling pipes rapidly, and provide advices for the monitoring and repair of the pipes containing flaws in nuclear power plants.","PeriodicalId":110070,"journal":{"name":"Volume 1: Nuclear Plant Operation and Maintenance, Engineering and Modification, Operation Life Extension (OLE), and Life Cycle","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":"129908340","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 Reactor and Turbine Matching Key Technology of CPR1000 Nuclear Power Plant for Industrial Steam Supply CPR1000型工业供汽核电站堆机匹配关键技术研究

Volume 1: Nuclear Plant Operation and Maintenance, Engineering and Modification, Operation Life Extension (OLE), and Life Cycle Pub Date : 2022-08-08 DOI: 10.1115/icone29-92527

Zhou Mingjia, Wang Kai, Zhang Wei, Zhuang Jianmin, Mao Yulong

引用次数: 0

Optimization of Transient Control Method for RCV System RCV系统暂态控制方法的优化

Volume 1: Nuclear Plant Operation and Maintenance, Engineering and Modification, Operation Life Extension (OLE), and Life Cycle Pub Date : 2022-08-08 DOI: 10.1115/icone29-93279

Y. Huang, Zhi Lin Chen, Xin Chen, Zheng Gu, Hanming Chen, R. Zhang

{"title":"Optimization of Transient Control Method for RCV System","authors":"Y. Huang, Zhi Lin Chen, Xin Chen, Zheng Gu, Hanming Chen, R. Zhang","doi":"10.1115/icone29-93279","DOIUrl":"https://doi.org/10.1115/icone29-93279","url":null,"abstract":"\u0000 This paper provides an optimization scheme for the steam chamber of the pressurizer of the nuclear power unit. The scheme includes: if the operation information of the pressurizer meets the triggering conditions of the chamber, raise the charging flow and increase the discharge flow of the RCV system; Adjust the opening of the low-pressure control valve according to the measured pressure downstream of the letdown orifice; Close the connection control valve; If the discharge flow drops to the first preset flow, reduce the opening of the low-pressure control valve; Control the pressurizer to reduce the rising speed of charging flow; If the discharge flow drops to the second preset flow, stop the opening measurement adjustment of the low-pressure control valve and increase the opening measurement of the connecting control valve; If the letdown flow rises to the third preset flow, stop the opening adjustment of the connecting control valve, reduce the charging flow to the fourth preset flow, and control the pressurizer to perform cavity building operation. The method can effectively control the temperature rise of the fluid and the charging speed exceeding 107 °C. The service life of the nuclear power unit is improved, and the design of increasing the letdown flow to the third preset flow and reducing the charging flow to the fourth preset flow makes the letdown flow and charging flow meet the flow state of the pressurizer cavity, and ensures the stability of the pressurizer cavity establishment.","PeriodicalId":110070,"journal":{"name":"Volume 1: Nuclear Plant Operation and Maintenance, Engineering and Modification, Operation Life Extension (OLE), and Life Cycle","volume":"86 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":"126161616","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