The internal conditions and surface defects of key equipment in nuclear power plant (NPP) are mainly detected though non-destructive testing (NDT) technology, which is the key basis for maintenance and diagnosis of nuclear power main equipment. At present, there are many types and huge quantities of NDT data. Due to the lack of unified information management means, it depends more on personal skills and experience. On the other hand, discrete data files and 2D detection data can’t provide the staff with intuitive and stereoscopic visual effects, which affects the prediction and analysis of the staff to a certain extent. In view of the above problems, this paper deeply analyzes the characteristics of nuclear power NDT data, and proposes a 3D visualization method for ultrasonic defects of NPP reactor pressure vessel (RPV) based on digital twin. It also forms a 3D visualization digital twin system, which overlays and maps the 3D digital model of RPV with the reconstructed 3D defects, while providing the necessary 3D interaction. Through the research results of this paper, the digital and intelligent management methods of NDT data in NPP have been improved, and an outstanding 3D basis for prediction and analysis of NDT defects in NPP has been provided.
{"title":"Research and Application of 3d Visualization Digital Twin System for Ultrasonic Testing of Key Equipment in NPP","authors":"Zheng Yang, Yi Zhang, Hao Wang, Ya-nan Zhang","doi":"10.1115/icone29-91533","DOIUrl":"https://doi.org/10.1115/icone29-91533","url":null,"abstract":"\u0000 The internal conditions and surface defects of key equipment in nuclear power plant (NPP) are mainly detected though non-destructive testing (NDT) technology, which is the key basis for maintenance and diagnosis of nuclear power main equipment. At present, there are many types and huge quantities of NDT data. Due to the lack of unified information management means, it depends more on personal skills and experience. On the other hand, discrete data files and 2D detection data can’t provide the staff with intuitive and stereoscopic visual effects, which affects the prediction and analysis of the staff to a certain extent. In view of the above problems, this paper deeply analyzes the characteristics of nuclear power NDT data, and proposes a 3D visualization method for ultrasonic defects of NPP reactor pressure vessel (RPV) based on digital twin. It also forms a 3D visualization digital twin system, which overlays and maps the 3D digital model of RPV with the reconstructed 3D defects, while providing the necessary 3D interaction. Through the research results of this paper, the digital and intelligent management methods of NDT data in NPP have been improved, and an outstanding 3D basis for prediction and analysis of NDT defects in NPP has been provided.","PeriodicalId":422334,"journal":{"name":"Volume 12: Innovative and Smart Nuclear Power Plant Design","volume":"30 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":"125202094","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}
Jianfei Hou, Xun Zhang, Heino Zimmermann, Zhaohui Xiang
In order to avoid sudden fault and damage of equipment which will affect reliability and safety of the nuclear power plant, it is necessary to monitor the actual thermal performance of the nuclear power plant and analyse the thermal performance trends of thermal circuits and important equipment which is the basis of fault predicts. Based on the mature and widely applied EBSILON software, thermodynamic model of a nuclear power plant in operation is built including the primary loop, second and third loop and important equipment such as steam turbines, condenser, deaerator, low-pressure and high-pressure heaters and main feed water pumps. In particular, hundreds of instruments are taken into account in the thermodynamic model and all the measured data during operation of the nuclear power plant is imported into the instruments. Thermal performance of the nuclear power plant is evaluated and analyzed in different seasons and results displays the performance factors of important equipment compared to the design performance. In addition, a data reconciliation calculation is taken out to check the correctness of measurements using the massive redundant measured data of the plant and results shows some instruments may have potential measurement fault. Current work lays a foundation for the future research of fault predicts and diagnoses of the nuclear power plant.
{"title":"Thermal Performance Monitoring and Analysis of Nuclear Power Plant","authors":"Jianfei Hou, Xun Zhang, Heino Zimmermann, Zhaohui Xiang","doi":"10.1115/icone29-91886","DOIUrl":"https://doi.org/10.1115/icone29-91886","url":null,"abstract":"\u0000 In order to avoid sudden fault and damage of equipment which will affect reliability and safety of the nuclear power plant, it is necessary to monitor the actual thermal performance of the nuclear power plant and analyse the thermal performance trends of thermal circuits and important equipment which is the basis of fault predicts. Based on the mature and widely applied EBSILON software, thermodynamic model of a nuclear power plant in operation is built including the primary loop, second and third loop and important equipment such as steam turbines, condenser, deaerator, low-pressure and high-pressure heaters and main feed water pumps. In particular, hundreds of instruments are taken into account in the thermodynamic model and all the measured data during operation of the nuclear power plant is imported into the instruments. Thermal performance of the nuclear power plant is evaluated and analyzed in different seasons and results displays the performance factors of important equipment compared to the design performance. In addition, a data reconciliation calculation is taken out to check the correctness of measurements using the massive redundant measured data of the plant and results shows some instruments may have potential measurement fault. Current work lays a foundation for the future research of fault predicts and diagnoses of the nuclear power plant.","PeriodicalId":422334,"journal":{"name":"Volume 12: Innovative and Smart Nuclear Power Plant Design","volume":"111 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":"126075215","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}
Pengbin Zhang, Kunpeng Xing, Tao Zou, Zhijia Yang, Jingyang Wang, Hongrui Wang
Nuclear power is one of several significant clean energy and its production capacity is increasing rapidly. This paper proposed a novel PID self-tuning method based on closed-loop identification used in nuclear power plants to deal with the turbine’s load tracking problem. A setpoint-shifting method is adopted to achieve the informativity of the closed-loop system, and the PID parameters are designed through an optimization method. For validating the proposed method, this paper modeled a pressurized-water reactor (PWR) nuclear power plant, a complex system consisting of three circuits, through the first principle. The primary circuit contains three main devices: a nuclear reactor, a pressurizer, and a steam generator. The models of these three devices are provided by previous literature separately. The whole model is generated through mass and energy balance. This mechanism model is used as the plant model, while the identified model is used as the control model. The initial PID parameters are set through open-loop identification using an external excitation. Then the proposed method works online to track the turbine’s load in time. The effectiveness of the proposed method is illustrated by simulations.
{"title":"A Novel PID Self-Tuning Method Based on Closed-Loop Identification Used in Nuclear Power Plants","authors":"Pengbin Zhang, Kunpeng Xing, Tao Zou, Zhijia Yang, Jingyang Wang, Hongrui Wang","doi":"10.1115/icone29-92728","DOIUrl":"https://doi.org/10.1115/icone29-92728","url":null,"abstract":"\u0000 Nuclear power is one of several significant clean energy and its production capacity is increasing rapidly. This paper proposed a novel PID self-tuning method based on closed-loop identification used in nuclear power plants to deal with the turbine’s load tracking problem. A setpoint-shifting method is adopted to achieve the informativity of the closed-loop system, and the PID parameters are designed through an optimization method. For validating the proposed method, this paper modeled a pressurized-water reactor (PWR) nuclear power plant, a complex system consisting of three circuits, through the first principle. The primary circuit contains three main devices: a nuclear reactor, a pressurizer, and a steam generator. The models of these three devices are provided by previous literature separately. The whole model is generated through mass and energy balance. This mechanism model is used as the plant model, while the identified model is used as the control model. The initial PID parameters are set through open-loop identification using an external excitation. Then the proposed method works online to track the turbine’s load in time. The effectiveness of the proposed method is illustrated by simulations.","PeriodicalId":422334,"journal":{"name":"Volume 12: Innovative and Smart Nuclear Power Plant Design","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":"122383637","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}
Generally, the reactor vessel is cylindrical having a hemispherical lower end. The core barrel is connected to the interior walls of the reactor vessel at or adjacent to the area where the cylindrical and hemispherical portions of the reactor vessel meet. Below the main core support, i.e. the core barrel that is capped at its lower end with the lower core support, the hemispherical vessel defines a lower head or lower plenum. A generally annular downcomer surrounds the reactor core barrel between the core barrel and an inner wall of the reactor vessel. Cooling fluid, typically water, is pumped into this annular downcomer. The coolant fluid circulates downward into the lower plenum. The hemispherical shape of the lower plenum assists in evenly circulating the coolant fluid therein. A plurality of reactor core coolant inlet openings are located on the underside of the lower core support plate. Coolant flows from the lower plenum, into the core coolant inlet openings and upwardly into the core to cool the fuel assemblies. With the advent of larger plants with larger cores it became evident that further means were necessary to improve the distribution of coolant flow in the lower plenum to assure uniform coolant flow and pressure were maintained across all of the reactor core coolant inlet openings in the lower core support plate. Non-uniform coolant pressure or flow causes uneven coolant flow into the core, which results in uneven cooling of the fuel assemblies of the core. Uneven fuel assembly cooling may force the entire core to be derated to accommodate “hot assembly” locations. Non-uniform coolant flow and pressure may result in vortices or other flow disruptions to form in the coolant fluid circulating in the lower plenum. According to patent filing history, dozens of devices have been proposed to uniform the distribution of coolant flow. Those device could be categorized several kinds. As coolant flow in the fringe of low core plate, firstly the cylinder shape has been applied to prevent the vortex. This biggest vortex is formed by that the downflow from core barrel has 180° direction diverting with the upflow into LCP, diverting between the high speed flows. In order to maintain adequate and uniform cooling throughout the core, it is important that a uniform coolant flow and pressure be maintained across all of the reactor core coolant inlet openings in the lower core support plate. The second way to is put some device in the bottom of lower plenum to prevent vortex. The third is changing flow distribution of LCP by setting some flow limiting devices.
{"title":"Evolution of the Evening Method of Coolant Flow in the Large Core Based on Patent History","authors":"Wenchi Yu, Hao Fu","doi":"10.1115/icone29-88894","DOIUrl":"https://doi.org/10.1115/icone29-88894","url":null,"abstract":"\u0000 Generally, the reactor vessel is cylindrical having a hemispherical lower end. The core barrel is connected to the interior walls of the reactor vessel at or adjacent to the area where the cylindrical and hemispherical portions of the reactor vessel meet. Below the main core support, i.e. the core barrel that is capped at its lower end with the lower core support, the hemispherical vessel defines a lower head or lower plenum. A generally annular downcomer surrounds the reactor core barrel between the core barrel and an inner wall of the reactor vessel. Cooling fluid, typically water, is pumped into this annular downcomer. The coolant fluid circulates downward into the lower plenum. The hemispherical shape of the lower plenum assists in evenly circulating the coolant fluid therein. A plurality of reactor core coolant inlet openings are located on the underside of the lower core support plate. Coolant flows from the lower plenum, into the core coolant inlet openings and upwardly into the core to cool the fuel assemblies.\u0000 With the advent of larger plants with larger cores it became evident that further means were necessary to improve the distribution of coolant flow in the lower plenum to assure uniform coolant flow and pressure were maintained across all of the reactor core coolant inlet openings in the lower core support plate.\u0000 Non-uniform coolant pressure or flow causes uneven coolant flow into the core, which results in uneven cooling of the fuel assemblies of the core. Uneven fuel assembly cooling may force the entire core to be derated to accommodate “hot assembly” locations. Non-uniform coolant flow and pressure may result in vortices or other flow disruptions to form in the coolant fluid circulating in the lower plenum. According to patent filing history, dozens of devices have been proposed to uniform the distribution of coolant flow. Those device could be categorized several kinds. As coolant flow in the fringe of low core plate, firstly the cylinder shape has been applied to prevent the vortex. This biggest vortex is formed by that the downflow from core barrel has 180° direction diverting with the upflow into LCP, diverting between the high speed flows. In order to maintain adequate and uniform cooling throughout the core, it is important that a uniform coolant flow and pressure be maintained across all of the reactor core coolant inlet openings in the lower core support plate. The second way to is put some device in the bottom of lower plenum to prevent vortex. The third is changing flow distribution of LCP by setting some flow limiting devices.","PeriodicalId":422334,"journal":{"name":"Volume 12: Innovative and Smart Nuclear Power Plant Design","volume":"733 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":"123859623","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}
A typical northern nuclear power plant and a typical southern nuclear power plant are selected to optimize the cold end of the circulating water system with one machine and two pumps in an expanded unit system. And study the optimal operation mode of the circulating pump and the cold end configuration scheme with the minimum annual cost in this water supply mode. It is discussed how to combine the site conditions of the nuclear power plant, the price of raw materials and the main cold-end equipment of the steam turbine, the cost of the main structures of the water supply and drainage system and the related annual operating costs, power generation income and other factors, to seek the optimal solution through economic and technical comparison of multiple schemes. That is to select the optimal matching combination scheme of the configuration of the cooling water system such as the configuration scheme of the low-pressure cylinder of the steam turbine, the area of the condenser, the flow rate of the cooling water pump, and the cooling water pipe trench, and determine the economical and reasonable design parameters of the equipment at the cold end. By comparing the cold end optimization results of the constant speed scheme of 1 machine with 2 pumps, the frequency conversion scheme of 1 machine with 2 pumps and the expansion unit system scheme of 1 machine with 2 pumps, the frequency conversion scheme of 1 machine with 2 pumps is superior in terms of technology and economy and has a great advantage for the northern site. For the southern site, due to the higher water temperature and smaller variation, the advantage is less.
{"title":"Research on Cold End Optimization of Water Supply Mode of Expanded Unit System of Circulating Water System in Coastal Nuclear Power Plant","authors":"Fangyi Lin, Jiale Jian, Fang Wang, Jia Yang","doi":"10.1115/icone29-92613","DOIUrl":"https://doi.org/10.1115/icone29-92613","url":null,"abstract":"\u0000 A typical northern nuclear power plant and a typical southern nuclear power plant are selected to optimize the cold end of the circulating water system with one machine and two pumps in an expanded unit system. And study the optimal operation mode of the circulating pump and the cold end configuration scheme with the minimum annual cost in this water supply mode. It is discussed how to combine the site conditions of the nuclear power plant, the price of raw materials and the main cold-end equipment of the steam turbine, the cost of the main structures of the water supply and drainage system and the related annual operating costs, power generation income and other factors, to seek the optimal solution through economic and technical comparison of multiple schemes. That is to select the optimal matching combination scheme of the configuration of the cooling water system such as the configuration scheme of the low-pressure cylinder of the steam turbine, the area of the condenser, the flow rate of the cooling water pump, and the cooling water pipe trench, and determine the economical and reasonable design parameters of the equipment at the cold end. By comparing the cold end optimization results of the constant speed scheme of 1 machine with 2 pumps, the frequency conversion scheme of 1 machine with 2 pumps and the expansion unit system scheme of 1 machine with 2 pumps, the frequency conversion scheme of 1 machine with 2 pumps is superior in terms of technology and economy and has a great advantage for the northern site. For the southern site, due to the higher water temperature and smaller variation, the advantage is less.","PeriodicalId":422334,"journal":{"name":"Volume 12: Innovative and Smart Nuclear Power Plant Design","volume":"257 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":"114462944","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 RAM system is the reactor control rod drive mechanism power supply system, which is the only power supply designed for the control rod drive mechanism. The RAM (control rod Power System) system of the Nuclear Power Plant only has two synchronous generators (RAM001/002AP) operating in parallel. On November 14, 2019,when one unit has an excitation system failure, due to the unreasonable setting of the protection value of the two units, the over-current protection of the unfailed unit precedes the loss of field protection action of the field lost unit, resulting the two generators in the RAM system tripping one after the other, therefore causing CRDM losing all power supply, resulting in the control rod dropping off and finally causing the automatic shutdown of the reactor, bringing serious consequences to the reactor and power grid. Therefore, the setting of synchronous generator protection in the RAM system is very important for the safe and stable operation of nuclear power plant. This report carries out the digital modeling and simulation of the RAM system of the Nuclear Power Plants, simulates various faults of RAM system synchronous generator RAM001/002AP and provides the important basis for verifying the settings of over-current protection and loss of excitation protection.
{"title":"Optimization the Protection Configuration of the RAM Excitation System of the Nuclear Power Plant","authors":"Wang Lin, Zhao Yan-jun, Zhang Xing-Zhen","doi":"10.1115/icone29-92110","DOIUrl":"https://doi.org/10.1115/icone29-92110","url":null,"abstract":"\u0000 The RAM system is the reactor control rod drive mechanism power supply system, which is the only power supply designed for the control rod drive mechanism. The RAM (control rod Power System) system of the Nuclear Power Plant only has two synchronous generators (RAM001/002AP) operating in parallel. On November 14, 2019,when one unit has an excitation system failure, due to the unreasonable setting of the protection value of the two units, the over-current protection of the unfailed unit precedes the loss of field protection action of the field lost unit, resulting the two generators in the RAM system tripping one after the other, therefore causing CRDM losing all power supply, resulting in the control rod dropping off and finally causing the automatic shutdown of the reactor, bringing serious consequences to the reactor and power grid. Therefore, the setting of synchronous generator protection in the RAM system is very important for the safe and stable operation of nuclear power plant. This report carries out the digital modeling and simulation of the RAM system of the Nuclear Power Plants, simulates various faults of RAM system synchronous generator RAM001/002AP and provides the important basis for verifying the settings of over-current protection and loss of excitation protection.","PeriodicalId":422334,"journal":{"name":"Volume 12: Innovative and Smart Nuclear Power Plant Design","volume":"826 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":"116142475","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 adsorption reaction between oxygen (O2) molecule and ferrum (Fe) (110) crystal surface in the oxidation process of Fe surface was studied by using the first-principles method. The differential charge density analysis of the adsorption sites of oxygen molecule on Fe (110) crystal surface, the calculation of adsorption energy at different sites and the analysis of electronic density of states showed that the stable adsorption position of oxygen molecule was parallel to Fe (110) crystal surface, and the oxygen atom tended to adsorb at the triangular gap of Fe atoms. The electronic structure of the adsorption system showed that the 2p electron orbital of oxygen atom plays a major role in the adsorption, and only O-Fe electron interaction exists when oxygen molecule is adsorbed in the parallel orientation, which makes the whole Fe (110) crystal surface lose electrons, increase the system potential and the risk of electrochemical corrosion. The research conclusions can provide theoretical support for the further insight in the oxidation corrosion mechanism of nuclear metal surface.
{"title":"First-Principles Study on Adsorption Reaction of Oxygen Molecules on Fe (110) Crystal Surface","authors":"Zeng Xiaochuan, Li Xuejun, He Cuizhu, Hu Qiaodan","doi":"10.1115/icone29-92890","DOIUrl":"https://doi.org/10.1115/icone29-92890","url":null,"abstract":"\u0000 The adsorption reaction between oxygen (O2) molecule and ferrum (Fe) (110) crystal surface in the oxidation process of Fe surface was studied by using the first-principles method. The differential charge density analysis of the adsorption sites of oxygen molecule on Fe (110) crystal surface, the calculation of adsorption energy at different sites and the analysis of electronic density of states showed that the stable adsorption position of oxygen molecule was parallel to Fe (110) crystal surface, and the oxygen atom tended to adsorb at the triangular gap of Fe atoms. The electronic structure of the adsorption system showed that the 2p electron orbital of oxygen atom plays a major role in the adsorption, and only O-Fe electron interaction exists when oxygen molecule is adsorbed in the parallel orientation, which makes the whole Fe (110) crystal surface lose electrons, increase the system potential and the risk of electrochemical corrosion. The research conclusions can provide theoretical support for the further insight in the oxidation corrosion mechanism of nuclear metal surface.","PeriodicalId":422334,"journal":{"name":"Volume 12: Innovative and Smart Nuclear Power Plant Design","volume":"9 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":"127376958","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}
Nowadays, 3D modeling and virtual reality technology has become a new method to enhance the digitization level in nuclear power engineering field. Reactor pressure vessel head handling, refueling and main equipment installation are the important process flows in nuclear power plant. Traditional measures without digitization and visualization, which need much more experience and technical skills to conduct the staffs on-site construction work. In that case, there are more risks when they do some complicated operations especially the new ones they never know before. This paper presents a technical method to design an immersive virtual environment, which contains the whole 3D model data of the nuclear island buildings in nuclear power plant to simulate the key processing flows. Meanwhile, the paper develops an interactive VR software system to help training workers about their on-site works in a virtual way before they solve on-site problems. On the other hand, another application of this paper’s work is design verification of the process flows themselves, so that the simulation results can tell designer what and where is the optimizing goal. The achievements of this paper have already been applied in the actual HPR1000 nuclear power project of CGN, and bring considerable efficiency return in construction field.
{"title":"Research on Immersive Simulation of Key Equipment Process Flow in Nuclear Power Plant Based on VR Technology","authors":"Yi Zhang, Zhiguo Zhou, Hao Wang, Zheng Yang","doi":"10.1115/icone29-93221","DOIUrl":"https://doi.org/10.1115/icone29-93221","url":null,"abstract":"\u0000 Nowadays, 3D modeling and virtual reality technology has become a new method to enhance the digitization level in nuclear power engineering field. Reactor pressure vessel head handling, refueling and main equipment installation are the important process flows in nuclear power plant. Traditional measures without digitization and visualization, which need much more experience and technical skills to conduct the staffs on-site construction work. In that case, there are more risks when they do some complicated operations especially the new ones they never know before. This paper presents a technical method to design an immersive virtual environment, which contains the whole 3D model data of the nuclear island buildings in nuclear power plant to simulate the key processing flows. Meanwhile, the paper develops an interactive VR software system to help training workers about their on-site works in a virtual way before they solve on-site problems. On the other hand, another application of this paper’s work is design verification of the process flows themselves, so that the simulation results can tell designer what and where is the optimizing goal. The achievements of this paper have already been applied in the actual HPR1000 nuclear power project of CGN, and bring considerable efficiency return in construction field.","PeriodicalId":422334,"journal":{"name":"Volume 12: Innovative and Smart Nuclear Power Plant Design","volume":"15 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":"126029964","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}
After a nuclear accident, the nuclear power plant will monitor and analyze the unit status, environmental characteristics and accident process, determine the emergency state classification under the nuclear accident conditions, and respond to the emergency state hierarchically, to reduce the impact of the accident and ensure the safety of personnel. For determining emergency state classification of nuclear power plant cold shutdown and refueling shutdown operation mode, reactor vessel water level is an important basis for judging. The Three Mile Island nuclear accident made the industry realize the necessity of monitoring the primary water load, especially the water level in the Reactor vessel. Reactor vessel water level provides an essential basis for monitoring the core cooling state after an accident, to ensure that the core cooling state can be diagnosed correctly in time, accurately and conveniently, thus providing a criterion for determining the emergency state level, and then selecting the appropriate accident operation strategy. Based on the demand analysis of emergency state classification based on reactor vessel water level, combined with the current situation of reactor vessel measurement in Reactor vessel and the setting of peripheral dose rate monitoring channels, this paper supplements the water level measurement method under potential water loss conditions, realizes the full-range measurement of Validity water level in Reactor vessel, evaluates Validity water level Validity in Validity, and then efficiently determines the emergency state classification, providing Validity basis for emergency response.
{"title":"Study on Reactor Vessel Water Level Measurement Method for Emergency Decision-Making of Nuclear Power Plant","authors":"Yanming Shi, Zhen-Ying Wang","doi":"10.1115/icone29-92804","DOIUrl":"https://doi.org/10.1115/icone29-92804","url":null,"abstract":"\u0000 After a nuclear accident, the nuclear power plant will monitor and analyze the unit status, environmental characteristics and accident process, determine the emergency state classification under the nuclear accident conditions, and respond to the emergency state hierarchically, to reduce the impact of the accident and ensure the safety of personnel. For determining emergency state classification of nuclear power plant cold shutdown and refueling shutdown operation mode, reactor vessel water level is an important basis for judging. The Three Mile Island nuclear accident made the industry realize the necessity of monitoring the primary water load, especially the water level in the Reactor vessel. Reactor vessel water level provides an essential basis for monitoring the core cooling state after an accident, to ensure that the core cooling state can be diagnosed correctly in time, accurately and conveniently, thus providing a criterion for determining the emergency state level, and then selecting the appropriate accident operation strategy.\u0000 Based on the demand analysis of emergency state classification based on reactor vessel water level, combined with the current situation of reactor vessel measurement in Reactor vessel and the setting of peripheral dose rate monitoring channels, this paper supplements the water level measurement method under potential water loss conditions, realizes the full-range measurement of Validity water level in Reactor vessel, evaluates Validity water level Validity in Validity, and then efficiently determines the emergency state classification, providing Validity basis for emergency response.","PeriodicalId":422334,"journal":{"name":"Volume 12: Innovative and Smart Nuclear Power Plant Design","volume":"20 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":"126257209","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 paper proposes a more stringent method for customizing project rules. This method customizes the comprehensive management of the project and component reference database on the digital plant design platform based on general design codes, standards, item classification principles in nuclear engineering, digitalization requirements in reactor design, plant layout, project management, and material procurement and construction, etc. To improve the correlation between design specifications and digital power plants, suggest enhancing the data consistency among different design disciplines. Standardize the three-dimensional layout design of nuclear power plants, and ensure the consistency between the digital power plant model and the natural power plant. The rules are sorted out, analyzed, and transformed systematically in this paper. These rules include the naming and classification principles of items in nuclear power projects, model data composition structure, essential attribute content, component selection filters, material performance, model parameters, output content format, vital requirements in plant layout for reactor design, etc. This paper will finally form a systematic rule customization scheme through refinement and improvement. The parameters, such as process, operating conditions, materials, fluids, specifications, safety, quality assurance, seismic and radioactivity levels, as well as item naming rules, project database, component reference database, three-dimensional modeling, information integration, attribute inheritance, data extraction, and other regulations. This scheme can make the three-dimensional arrangement more standard, the operation steps more concise, and significantly reduce the designer’s attribute range of manual input. It can effectively promote the accurate and appropriate expression of the process and instrumentation design scheme in the reactor building model. Significantly shorten the project design cycle. Data integration and transmission between rules enable system attributes to be deeply inherited and automatic checking and judging of operating conditions parameters and pressure and temperature limits in physical properties of component materials. This scheme can make the three-dimensional layout more standard, the operation steps more concise, significantly reduce the attributes range of manual input by the designer, and obtain better application feedback in automatic drawing and material reports. These rules provide more comprehensive data support for coupling experiments, data integration, process simulation, and digital handover of different disciplines and depths.
{"title":"Research on the Impact of Advanced Rule Design System on the Digitalization of Reactor Building Model","authors":"Jin Wang, Hongyi Yang","doi":"10.1115/icone29-92995","DOIUrl":"https://doi.org/10.1115/icone29-92995","url":null,"abstract":"\u0000 This paper proposes a more stringent method for customizing project rules. This method customizes the comprehensive management of the project and component reference database on the digital plant design platform based on general design codes, standards, item classification principles in nuclear engineering, digitalization requirements in reactor design, plant layout, project management, and material procurement and construction, etc. To improve the correlation between design specifications and digital power plants, suggest enhancing the data consistency among different design disciplines. Standardize the three-dimensional layout design of nuclear power plants, and ensure the consistency between the digital power plant model and the natural power plant. The rules are sorted out, analyzed, and transformed systematically in this paper. These rules include the naming and classification principles of items in nuclear power projects, model data composition structure, essential attribute content, component selection filters, material performance, model parameters, output content format, vital requirements in plant layout for reactor design, etc. This paper will finally form a systematic rule customization scheme through refinement and improvement. The parameters, such as process, operating conditions, materials, fluids, specifications, safety, quality assurance, seismic and radioactivity levels, as well as item naming rules, project database, component reference database, three-dimensional modeling, information integration, attribute inheritance, data extraction, and other regulations. This scheme can make the three-dimensional arrangement more standard, the operation steps more concise, and significantly reduce the designer’s attribute range of manual input. It can effectively promote the accurate and appropriate expression of the process and instrumentation design scheme in the reactor building model. Significantly shorten the project design cycle. Data integration and transmission between rules enable system attributes to be deeply inherited and automatic checking and judging of operating conditions parameters and pressure and temperature limits in physical properties of component materials. This scheme can make the three-dimensional layout more standard, the operation steps more concise, significantly reduce the attributes range of manual input by the designer, and obtain better application feedback in automatic drawing and material reports. These rules provide more comprehensive data support for coupling experiments, data integration, process simulation, and digital handover of different disciplines and depths.","PeriodicalId":422334,"journal":{"name":"Volume 12: Innovative and Smart Nuclear Power Plant Design","volume":"27 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":"130331054","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}
京公网安备 11010802042870号
京ICP备2023020795号-1
扫码关注我们
求助内容：
应助结果提醒方式：