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

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The Effect of Specimen Size and Surface Roughness on Tensile Strength of Nuclear Graphite 试样尺寸和表面粗糙度对核石墨抗拉强度的影响
Yuxin Guan, Yingjie Zhan, Shuiyong Wang, Qingwu Wang, Wen-xin Ti, Bing Gong
{"title":"The Effect of Specimen Size and Surface Roughness on Tensile Strength of Nuclear Graphite","authors":"Yuxin Guan, Yingjie Zhan, Shuiyong Wang, Qingwu Wang, Wen-xin Ti, Bing Gong","doi":"10.1115/icone29-91966","DOIUrl":"https://doi.org/10.1115/icone29-91966","url":null,"abstract":"\u0000 The tensile strength of nuclear graphite is an important parameter in evaluating and calculating the structural strength of High Temperature Reactor-Pebbled Modules (HTR-PM). The tensile strength of nuclear graphite with varied sizes and surface roughness was tested, and the size effect tensile strength formula was created. The results reveal that when the size of nuclear graphite increases, its tensile strength falls. However, in the 4 mm to 8 mm range, the size impact is less noticeable. The Weibull modulus m0.95 value is calculated based on the experimental data, and the size effect formula is established. The difference between the estimated value and the experimental value computed using the size effect formula is less than 10%. Tensile strength may be calculated using the size effect formula for nuclear graphite specimens of various sizes. The impact of surface roughness on tensile strength is linked to the grinding direction and specimen size. After a certain level of surface roughness, the tensile strength diminishes when the surface scratches are perpendicular to the tensile force. The stronger the effect, the smaller the specimen size. When surface scratches are parallel to the tensile tension, the tensile strength does not vary much as the surface roughness increases.","PeriodicalId":110070,"journal":{"name":"Volume 1: Nuclear Plant Operation and Maintenance, Engineering and Modification, Operation Life Extension (OLE), and Life Cycle","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":"123529850","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
Analysis and Mitigation Strategy on the Decreasing Opening of Thermal Pressurizer Pilot Operated Safety Relief Valves 热稳压器先导式安全阀开度减小的分析及缓解策略
Wensheng Hu, Chaofan Guo, Haisheng Zhao, Zhaoji Ruan, Jidong Min, Pencheng Du, Xinxin Wang
{"title":"Analysis and Mitigation Strategy on the Decreasing Opening of Thermal Pressurizer Pilot Operated Safety Relief Valves","authors":"Wensheng Hu, Chaofan Guo, Haisheng Zhao, Zhaoji Ruan, Jidong Min, Pencheng Du, Xinxin Wang","doi":"10.1115/icone29-92099","DOIUrl":"https://doi.org/10.1115/icone29-92099","url":null,"abstract":"\u0000 The pressurizerpilot operated safety relief valves (abbr. POSRV) protect the reactor coolant system against overpressure situation under accident conditions. It can be used us safety valves, relief valves or isolation valves. It’s one of the most important valves in PWR. The POSRV of a nuclear power unit adopts the thermal pilot structure design which is the first application of the third generation nuclear power unit. The POSRV composed of main valve and metallic pilot detector with special sealing design which can withstand the high temperature and high pressure medium in the primary circuit. During the first cycle operation of the nuclear unit, the opening of the isolation valve of the POSRV decreases, the discharge line temperature of metallic pilot detector continuous rise and the leakage rate of the primary coolant circuit increases abnormally, which bring great retreat risk to the unit. Combined with the structural principle of the POSRV and the change trend of key parameters, the analysis shows that the reason for the decrease of the valve opening is the leakage of the three-way sealing surface of the metallic pilot detector, which leads to the introduction of pressurized medium into the valve head of the main valve, then reduces the valve opening and affects the leakage rate of the primary coolant circuit. By formulating mitigation strategies such as energizing the solenoid valve to change the pilot valve state, draining the leakage medium in the primary coolant circuit and the unit operation decision-making scheme, the valve availability and unit safety are ensured, which provides a reference for similar equipment problems.","PeriodicalId":110070,"journal":{"name":"Volume 1: Nuclear Plant Operation and Maintenance, Engineering and Modification, Operation Life Extension (OLE), and Life Cycle","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":"129624387","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
Seismic Analysis of Fuel Transfer Tube Gate Valve 燃油输送管闸阀的地震分析
Su Ziwei, Zhang Xin
{"title":"Seismic Analysis of Fuel Transfer Tube Gate Valve","authors":"Su Ziwei, Zhang Xin","doi":"10.1115/icone29-89869","DOIUrl":"https://doi.org/10.1115/icone29-89869","url":null,"abstract":"\u0000 The Gate valve is an important functional part of the Fuel Transfer system, in order to achieve complete independence of Fuel Transfer unit, the scientific research and manufacture of the gate valve is imperative. According to the objectives, the overall structure design, component function design and material selection are completed, and then manufacture scientific research prototype, carry out a series of performance tests such as strength test, sealing test and whole machine life test, etc. Among them, it is an important step to verify the integrity of the Gate valve under seismic conditions, so in accordance with the design structure, establish the 3D geometric model, divide mesh, creat the finite element model, and apply the boundary conditions, the numerical simulation for Gate valve was presented by using ANSYS software.\u0000 The results show that under a combination of loads of various working conditions, including SL-1 abnormal condition or SL-2 accident condition, the stress of each component of Gate valve body, Support, Operating mechanism and Driving rod is less than the specified limit and meets the requirements of the judgement criterion; so the equipment achieve complete independence of Fuel Transfer unit, and be with reliable and stable performance, which can effectively guarantee the safe operation of 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":"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":"127036576","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 Study of Removal of Secondary Neutron Source of HPR-1000 HPR-1000二次中子源的去除研究
Huiwen Xiao, Xiang Li, Guoming Liu, Yicheng Zhang, Zeng Shao, Min Xu, Xuan Yi, Haifeng Yang
{"title":"The Study of Removal of Secondary Neutron Source of HPR-1000","authors":"Huiwen Xiao, Xiang Li, Guoming Liu, Yicheng Zhang, Zeng Shao, Min Xu, Xuan Yi, Haifeng Yang","doi":"10.1115/icone29-93024","DOIUrl":"https://doi.org/10.1115/icone29-93024","url":null,"abstract":"\u0000 When the reactor was in the subcritical state, the criticality safety of the core was monitored by the source range detector which was placed out of the core. According to the “Initial Loading Test for Pressurized Water Reactor of Nuclear Power Plant” (NB/T 20434-2017RK), the recommended value of count rate of the detector was larger than 0.5 cps. But during loading and other subcritical reactor conditions, the number of neutrons in the core is relatively low, causing that the count rate of the ex-core source range detectors could not reach 0.5 cps. Thus, secondary neutron source was placed in the following cycle of reactor to increase the neutron count rate of the ex-core source range detectors to 0.5 cps to meet the standard. In fact, the use of secondary neutron sources raises a number of problems, such as: 1) the production of radioactive and other tritium from secondary neutron sources, the total production of which is a limiting criterion for the construction of nuclear power plants; 2) the increased procurement costs of secondary neutron sources; and 3) the vulnerability to damage due to the long presence of secondary neutron sources in the core, increase the radioactivity level of the primary circuit. Because of the use of low leakage loading fuel management in HPR-1000, the spent fuel assemblies were placed closer to the source range detector and generated more neutrons, it is possible for the ex-core source range detectors to meet the counting requirement of 0.5 cps regardless of whether the core is in the subcritical state. This paper studied the removal of secondary neutron source in HPR-1000. The neutron source intensity of spent fuel assemblies and neutron count rate in fuel loading process were calculated in a conservative method with three dimension monte carlo code. The calculated results had shown that the removal of secondary neutron source in HPR-1000 satisfies safety requirements.","PeriodicalId":110070,"journal":{"name":"Volume 1: Nuclear Plant Operation and Maintenance, Engineering and Modification, Operation Life Extension (OLE), and Life Cycle","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":"130000934","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
Design Verification of Minimum Flow Valve of the Main Feed Water Pump Based on the Full-Range Simulation Platform of CPR 1000 基于CPR 1000全范围仿真平台的主给水泵最小流量阀设计验证
Zhenhua Zhang, Bo Zhang, Chaoying Zheng
{"title":"Design Verification of Minimum Flow Valve of the Main Feed Water Pump Based on the Full-Range Simulation Platform of CPR 1000","authors":"Zhenhua Zhang, Bo Zhang, Chaoying Zheng","doi":"10.1115/icone29-90543","DOIUrl":"https://doi.org/10.1115/icone29-90543","url":null,"abstract":"\u0000 When the flow of the main feed water pump of CPR1000 unit is low, that is, when the unit is operating at low load, in order to protect the main feed water pump from damage, two low-flow protection recirculation pipelines are installed at the pressure stage pump outlet, and the two lines are installed with two minimum flow regulating valves. These two valves generally have two designs: proportional type and on-off type. The paper simulates the response of the minimum flow valve of main feed pump to water level regulation of steam generator at low load based on the full-range simulation platform of CPR1000 nuclear power unit. By analyzing the response curve of the water level of the steam generator, the advantages and disadvantages of the different valve designs are verified separately. In the proportional design, the valve position of the minimum flow valve is easy to oscillate with the fluctuation of the flow signal, which will affect the stability of the steam generator feed water control system and then affect the operation of the water level regulation system of the steam generator. At the same time, the proportional minimum flow control valve has the control dead zone. Under the transient conditions with frequent disturbance, the response speed of the minimum flow valve does not match the response of the water level adjustment of the steam generator, resulting in the unstable water level of the steam generator. In the on-off design, the stability of water level regulation will have certain advantages. After comparing the influence of the two valve designs on the water level regulation of the steam generator and analyzing the limitation of the proportional design and the importance of minimum flow valve, two improvements schemes are given for the design of the minimum continuous regulating flow valve.","PeriodicalId":110070,"journal":{"name":"Volume 1: Nuclear Plant Operation and Maintenance, Engineering and Modification, Operation Life Extension (OLE), and Life Cycle","volume":"17 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":"128973309","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
Deuterization Process and Result Analysis of Resin in Heavy Water Purification System of a Reactor 反应器重水净化系统中树脂的氘化过程及效果分析
Wen Juan Zhang, Fengxia Xu, Xiao Hua
{"title":"Deuterization Process and Result Analysis of Resin in Heavy Water Purification System of a Reactor","authors":"Wen Juan Zhang, Fengxia Xu, Xiao Hua","doi":"10.1115/icone29-91360","DOIUrl":"https://doi.org/10.1115/icone29-91360","url":null,"abstract":"\u0000 This paper introduces in detail the whole process of resin replacement of heavy water purification system in a reactor reconstruction project in Algeria, including old resin discharge, pipeline cleaning, new resin filling, new resin deuterization, water quality detection, heavy water storage and continuous operation of the system. By sorting out deuterization data and tracking and analyzing the water quality detection results, it is considered that the resin deuterization process of the heavy water purification system of a reactor is smooth. After deuterization, the effluent concentrations of the three resin columns of the heavy water purification system meet the operation requirements. The whole deuterization process tries to save precious high concentration heavy water for Algeria. After deuterization, the resin column is connected to the heavy water purification system. After continuous operation and sampling detection, it is determined that the purification effect is ideal and the pH value is not affected. The resin of the heavy water purification system is successfully replaced and can be put into use. At the same time, aiming at some problems in the deuterization process, this paper analyzes the causes in detail, summarizes the experience and lessons, and gives optimization suggestions and precautions in the deuterization process.","PeriodicalId":110070,"journal":{"name":"Volume 1: Nuclear Plant Operation and Maintenance, Engineering and Modification, Operation Life Extension (OLE), and Life Cycle","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":"125471830","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 Intelligent Classification Method of Primary Loop Transient in Nuclear Power Plant 核电站一次回路暂态智能分类方法研究
R. Zhang, Zhi Lin Chen, Y. Huang, Xin Chen
{"title":"Research on Intelligent Classification Method of Primary Loop Transient in Nuclear Power Plant","authors":"R. Zhang, Zhi Lin Chen, Y. Huang, Xin Chen","doi":"10.1115/icone29-92874","DOIUrl":"https://doi.org/10.1115/icone29-92874","url":null,"abstract":"\u0000 The primary circuit transient of nuclear power plant is one of the important references for characterizing the fatigue of primary circuit materials. For the classification of transients, the commonly used method is to determine the category of transients according to the change in value and the operating operation of the unit. This paper forms a complete set of digital transient characteristics through the study of the thermal parameters behind various transients and the exploration of the law of transient occurrence. Combined with new computer technologies such as machine learning, the automatic processing of transient classification is realized. It provides ideas for the key links in the development of the whole-process automation platform for transient management.","PeriodicalId":110070,"journal":{"name":"Volume 1: Nuclear Plant Operation and Maintenance, Engineering and Modification, Operation Life Extension (OLE), and Life Cycle","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":"123362779","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
Effect of Carbon Content on the Resistance to Intergranular Corrosion of Alloy 800H 碳含量对800H合金抗晶间腐蚀性能的影响
Qian Zhong, Man Wang, Wenxiang Sun, Junwei Wu, Jiong Qian, Sen-Sen Ning
{"title":"Effect of Carbon Content on the Resistance to Intergranular Corrosion of Alloy 800H","authors":"Qian Zhong, Man Wang, Wenxiang Sun, Junwei Wu, Jiong Qian, Sen-Sen Ning","doi":"10.1115/icone29-91676","DOIUrl":"https://doi.org/10.1115/icone29-91676","url":null,"abstract":"\u0000 The effect of carbon content on the intergranular corrosion performance of Alloy 800H was studied by thermodynamic software simulation calculation combined with double-loop electrochemical potentiodynamic reactivation (DL-EPR), sulfuric acid-copper sulfate test, and microscopic morphology observation (optical microscope, scanning electron microscope). The results show that the main precipitation phases of Alloy 800H are α-Cr phase, M(C,N), M23C6, η phase and σ phase. At 500∼720°C, the higher the carbon content and the more M23C6 phase, the higher the susceptibility to intergranular corrosion. In the solid solution state, the Alloy 800H has low sensitivity to intergranular corrosion, and the carbon content has no obvious effect on its resistance to intergranular corrosion. In the sensitized state, carbides are continuously precipitated at the grain boundaries, and the intergranular corrosion susceptibility of the alloy increases rapidly.","PeriodicalId":110070,"journal":{"name":"Volume 1: Nuclear Plant Operation and Maintenance, Engineering and Modification, Operation Life Extension (OLE), and Life Cycle","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":"121692843","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
Multiplication Factor Prediction Feasibility Study for the BEAVRS Core Based on the LSTM Algorithm 基于LSTM算法的BEAVRS堆芯倍增因子预测可行性研究
Ren Chang-an, Lei Ji-chong, Yang Xiao-hua
{"title":"Multiplication Factor Prediction Feasibility Study for the BEAVRS Core Based on the LSTM Algorithm","authors":"Ren Chang-an, Lei Ji-chong, Yang Xiao-hua","doi":"10.1115/icone29-90164","DOIUrl":"https://doi.org/10.1115/icone29-90164","url":null,"abstract":"\u0000 This paper focuses on exploring the feasibility of the LSTM (Long Short-Term Memory) algorithm in deep learning for effective multiplication factor keff prediction at the core level, modeled by BEAVRS (Benchmark for Evaluation And Validation of Reactor Simulations) core first cycle loading with keff of operating at full power for 0–300 days was used as the study subject. The first 65% of the dataset is the training and validation set, and the last 35% of the dataset is the prediction target. The training and alignment results of the physical parameters of the components were obtained using the DRAGON4.1 and DONJON4.1 codes, and the LSTM algorithm in deep learning was applied. By adjusting the number of LSTM cells, L2 regularization parameters, optimizer type, and other parameter coefficients in the algorithm, the results showed that the absolute error of the predicted core effective multiplication factor keff could be made within 2 pcm by adjusting the appropriate parameters, which validated the successful application of machine learning to transport equations.","PeriodicalId":110070,"journal":{"name":"Volume 1: Nuclear Plant Operation and Maintenance, Engineering and Modification, Operation Life Extension (OLE), and Life Cycle","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":"122395461","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 Design About Generating or Removing Vapor Space Of The Primary Loop Pressurizer 主回路稳压器产生或去除蒸汽空间的优化设计
F. Ke
{"title":"Optimization Design About Generating or Removing Vapor Space Of The Primary Loop Pressurizer","authors":"F. Ke","doi":"10.1115/icone29-91960","DOIUrl":"https://doi.org/10.1115/icone29-91960","url":null,"abstract":"\u0000 In the second generation CPR nuclear power station, due to the limitations of design and equipment, it is not possible to generate or remove vapor space of pressurizer simultaneously with the temperature rise or fall of the primary loop coolant. But in the third generation HUALONG-1 nuclear power station, the obstacles on the equipment have been resolved and the foundation is already in place to solve the above technical problem. If we can improve the scheme appropriately, the primary loop coolant is heated or cooled while generating or removing vapor space of the pressurizer, which will be realized firstly. There are two main benefits: 1, It can reduce outage duration and improve economic efficiency. 2, It is conducive to decrease radioactive sources during the outage condition.","PeriodicalId":110070,"journal":{"name":"Volume 1: Nuclear Plant Operation and Maintenance, Engineering and Modification, Operation Life Extension (OLE), and Life Cycle","volume":"13 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":"133603316","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
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