The steam generator (SG) channel head forms part of the reactor coolant pressure boundary and is of high nuclear safety duty. The channel head failure is considered intolerable and there are no reasonably practicable physical provisions available to prevent its failure. Therefore the channel head is classified as either an Incredibility of Failure (IoF) or High Integrity (HI) component and it requires additional analyses and assessments beyond the design code requirements to achieve and demonstrate its structural integrity. The hydrodynamic effects of the primary coolant in the annulus around the drain plug of the SG channel head are not very well understood, but are exacerbated by high flow rates in its immediate vicinity due to typical design details. Vibration of the drain plug due to coolant flow may result in fatigue induced failure of the channel head drain penetration weld. In the study presented here, random vibration analyses and a fatigue assessment have been carried out for a submerged drain plug in a pressurized water reactor (PWR) SG channel head. A finite element (FE) model of the drain plug submerged in water coolant has been developed. Modal analyses confirmed that the natural frequency of the submerged drain plug is significantly reduced by the large hydrodynamic added mass from the surrounding fluid. The fatigue evaluation undertaken using ASME III fatigue curve concluded that the fatigue life usage due to the vibration of the drain plug is negligible even after an extended plant life. Therefore the coolant flow-induced drain plug vibration is not a threat to the channel head integrity.
{"title":"Fatigue Risk Evaluation of a Pressure Vessel Plug Subject to Flow Induced Vibration","authors":"Robert X. Wang, L. Chang, Tom Hurst, A. Hurst","doi":"10.1115/icone2020-16650","DOIUrl":"https://doi.org/10.1115/icone2020-16650","url":null,"abstract":"\u0000 The steam generator (SG) channel head forms part of the reactor coolant pressure boundary and is of high nuclear safety duty. The channel head failure is considered intolerable and there are no reasonably practicable physical provisions available to prevent its failure. Therefore the channel head is classified as either an Incredibility of Failure (IoF) or High Integrity (HI) component and it requires additional analyses and assessments beyond the design code requirements to achieve and demonstrate its structural integrity.\u0000 The hydrodynamic effects of the primary coolant in the annulus around the drain plug of the SG channel head are not very well understood, but are exacerbated by high flow rates in its immediate vicinity due to typical design details. Vibration of the drain plug due to coolant flow may result in fatigue induced failure of the channel head drain penetration weld. In the study presented here, random vibration analyses and a fatigue assessment have been carried out for a submerged drain plug in a pressurized water reactor (PWR) SG channel head. A finite element (FE) model of the drain plug submerged in water coolant has been developed. Modal analyses confirmed that the natural frequency of the submerged drain plug is significantly reduced by the large hydrodynamic added mass from the surrounding fluid. The fatigue evaluation undertaken using ASME III fatigue curve concluded that the fatigue life usage due to the vibration of the drain plug is negligible even after an extended plant life. Therefore the coolant flow-induced drain plug vibration is not a threat to the channel head integrity.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":"84 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74091470","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 communication system of nuclear power plants in China is not a safety class system, but it plays an important role in the safe operation of nuclear power plants. Under emergency state, the communication system is a prerequisite for accident management. In order to ensure communication on-site and off-site, diverse communication sub-systems are designed throughout the nuclear power plant, including various communication means for voice, data and images. For an advanced generation II pressurized water reactor (PWR) nuclear power plant (NPP) in China, there are various subsystems, including normal telephone system, safety telephone system, grid telephone system and so on. Although NPPs have designed diverse communication sub-systems, there is not any clear classification of the sub-systems, which is not enough for the reliability of communication sub-systems under accident conditions. Therefore, it can hardly ensure effective communications between different emergency response organizations and this will influence the mitigation of the accident. In order to identify the importance of different communication sub-systems, to optimize the design of communication system, and to improve the reliability and efficiency of nuclear power plant communication system, it’s necessary to analyze the function and operation of each sub-system, as well as to develop the classification method of nuclear power plant communication system. Considering the availability and reliability of onsite and offsite communication under emergency conditions, slightly considering economic issue, this paper determines 7 assessment factors and develops a set of scoring methods for communication system classification. On this basis, this paper completes the classification of the communication system for an advanced generation II PWR NPP, which provides a reference for communication system classification and provides the technical basis for design modification of the communication system.
{"title":"Classification Analysis of Communication System of Nuclear Power Plant","authors":"Yu Yun, Zheng Shen, Liu Jing","doi":"10.1115/icone2020-16235","DOIUrl":"https://doi.org/10.1115/icone2020-16235","url":null,"abstract":"\u0000 The communication system of nuclear power plants in China is not a safety class system, but it plays an important role in the safe operation of nuclear power plants. Under emergency state, the communication system is a prerequisite for accident management. In order to ensure communication on-site and off-site, diverse communication sub-systems are designed throughout the nuclear power plant, including various communication means for voice, data and images. For an advanced generation II pressurized water reactor (PWR) nuclear power plant (NPP) in China, there are various subsystems, including normal telephone system, safety telephone system, grid telephone system and so on.\u0000 Although NPPs have designed diverse communication sub-systems, there is not any clear classification of the sub-systems, which is not enough for the reliability of communication sub-systems under accident conditions. Therefore, it can hardly ensure effective communications between different emergency response organizations and this will influence the mitigation of the accident. In order to identify the importance of different communication sub-systems, to optimize the design of communication system, and to improve the reliability and efficiency of nuclear power plant communication system, it’s necessary to analyze the function and operation of each sub-system, as well as to develop the classification method of nuclear power plant communication system.\u0000 Considering the availability and reliability of onsite and offsite communication under emergency conditions, slightly considering economic issue, this paper determines 7 assessment factors and develops a set of scoring methods for communication system classification. On this basis, this paper completes the classification of the communication system for an advanced generation II PWR NPP, which provides a reference for communication system classification and provides the technical basis for design modification of the communication system.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":"22 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78265740","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}
E. Porcheron, C. Dazon, Yohan Leblois, T. Gelain, C. Chagnot, I. Doyen, C. Journeau, C. Suteau, D. Roulet
Assessing the production and dispersion of aerosols carrying contamination during Fukushima fuel debris retrieval is IRSN’s contribution to a research project undertaken by a French consortium jointly with ONET Technologies and CEA on behalf of METI and managed by the Mitsubishi Research Institute (Georges et al. 2017). The objective is to obtain quantified data for evaluating the risk of disseminating contamination when implementing cutting tools such as laser or others such as mechanical one, over the next few years, in the process of decommissioning the damaged reactors at the Fukushima-Daiichi plant.
评估福岛燃料碎片回收过程中携带污染物的气溶胶的产生和扩散是IRSN对一个研究项目的贡献，该项目由法国财团与ONET Technologies和CEA代表日本经济产业省共同承担，由三菱研究所管理(Georges et al. 2017)。目标是获得量化数据，以便在今后几年内，在关闭福岛第一核电站受损反应堆的过程中，在使用激光等切割工具或其他诸如机械切割工具时，评估传播污染的风险。
{"title":"Aerosol Source Terms Characterization During Cutting of Fuel Debris Simulants With Different Tools in the Context of Fukushima Daiichi Decommissioning","authors":"E. Porcheron, C. Dazon, Yohan Leblois, T. Gelain, C. Chagnot, I. Doyen, C. Journeau, C. Suteau, D. Roulet","doi":"10.1115/icone2020-16181","DOIUrl":"https://doi.org/10.1115/icone2020-16181","url":null,"abstract":"\u0000 Assessing the production and dispersion of aerosols carrying contamination during Fukushima fuel debris retrieval is IRSN’s contribution to a research project undertaken by a French consortium jointly with ONET Technologies and CEA on behalf of METI and managed by the Mitsubishi Research Institute (Georges et al. 2017). The objective is to obtain quantified data for evaluating the risk of disseminating contamination when implementing cutting tools such as laser or others such as mechanical one, over the next few years, in the process of decommissioning the damaged reactors at the Fukushima-Daiichi plant.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":"5 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82326290","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}
Yoshiteru Komuro, Atsushi Kodama, Y. Kondo, Koichi Tanimoto, T. Hibiki
Two-phase flows are observed in various industrial plants and piping systems. Understanding two-phase flow behaviors such as flow patterns and unsteady void fraction in horizontal and vertical pipes are crucial in improving plant safety. Notably, the flow patterns observed in a large diameter pipe (approx. 4–6 in or larger) are significantly different from those observed in a medium diameter pipe. In a vertical large diameter pipe, no slug flow is observed due to the instantaneous slug bubble breakup caused by the surface instability. Besides, in a horizontal pipe, flow regime transition from stratification of liquid and gas to slug (plug) flow that induces unsteady flow should be taken into account. From this viewpoint, it is necessary to predict the flow regime in horizontal and vertical large diameter pipes with some elbows and to evaluate the unsteady flow regime. In this study, the simulation method based on the two-fluid model is developed. The two-fluid model is considered the most accurate model because the governing equations for mass, momentum, and energy transfer are formulated for each phase. When using the two-fluid model, some constitutive equations should be given in computing the momentum transfer between gas and liquid phases. In this study, several state-of-art constitutive equations of the bubble diameter, the interfacial drag force and non-drag forces such as the lift force and the bubble-bubble collision force, are implemented in the platform of ANSYS FLUENT. The developed simulation method is validated with visualization results and force from an air-water flow at the elbow of the piping system.
{"title":"Development on Simulation Method for Two-Phase Flow in Large Diameter Pipes With 90 Degree Elbows","authors":"Yoshiteru Komuro, Atsushi Kodama, Y. Kondo, Koichi Tanimoto, T. Hibiki","doi":"10.1115/icone2020-16747","DOIUrl":"https://doi.org/10.1115/icone2020-16747","url":null,"abstract":"\u0000 Two-phase flows are observed in various industrial plants and piping systems. Understanding two-phase flow behaviors such as flow patterns and unsteady void fraction in horizontal and vertical pipes are crucial in improving plant safety. Notably, the flow patterns observed in a large diameter pipe (approx. 4–6 in or larger) are significantly different from those observed in a medium diameter pipe. In a vertical large diameter pipe, no slug flow is observed due to the instantaneous slug bubble breakup caused by the surface instability. Besides, in a horizontal pipe, flow regime transition from stratification of liquid and gas to slug (plug) flow that induces unsteady flow should be taken into account. From this viewpoint, it is necessary to predict the flow regime in horizontal and vertical large diameter pipes with some elbows and to evaluate the unsteady flow regime. In this study, the simulation method based on the two-fluid model is developed. The two-fluid model is considered the most accurate model because the governing equations for mass, momentum, and energy transfer are formulated for each phase. When using the two-fluid model, some constitutive equations should be given in computing the momentum transfer between gas and liquid phases. In this study, several state-of-art constitutive equations of the bubble diameter, the interfacial drag force and non-drag forces such as the lift force and the bubble-bubble collision force, are implemented in the platform of ANSYS FLUENT. The developed simulation method is validated with visualization results and force from an air-water flow at the elbow of the piping system.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":"35 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79074511","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}
During a loss of vacuum accident (LOVA), dust particles that will be present in the future tokamak ITER are likely to be resuspended, inducing a risk for explosion and airborne contamination. Evaluating the particle resuspension/deposition and resulting airborne concentration in case of a LOVA is therefore a major issue and it can be investigated by using a CFD code. To this end, this article presents the implementation of a resuspension model in a CFD code (ANSYS CFX) and its application to an air ingress in a vacuum toroidal vessel with a volume comparable to ITER one. In the first part of the article, the Rock’n Roll model and its operational version with the Biasi’s correlation is presented. The second part of the article will be devoted to the implementation of the Rock’n’Roll model in ANSYS CFX for constant friction velocities and its adaptation to non-constant friction velocities. Finally, the paper presents the simulations obtained on the particle resuspension for an air ingress scenario in a large vacuum vessel. This case is particularly interesting and non-intuitive because as the initial pressure is reduced, the particle behavior is different from that at atmospheric pressure. Further, a competition between airflow forces and gravitational force occurs, due to the low pressure environment, potentially restricting the resuspension, and the pressure influence also has to be taken into account in the particle transport and deposition (Nerisson, 2011). Three particle diameters were studied allowing to show the evolution of the resuspension with this parameter and to calculate dust resuspension rates and airborne fractions during the air ingress.
在失去真空事故(LOVA)期间，未来托卡马克ITER中存在的粉尘颗粒可能会重新悬浮，从而引发爆炸和空气污染的风险。因此，在LOVA的情况下，评估颗粒的再悬浮/沉积以及由此产生的空气中浓度是一个主要问题，可以通过使用CFD代码进行研究。为此，本文提出了在CFD代码(ANSYS CFX)中实现重悬浮模型，并将其应用于容积与ITER相当的真空环形容器的进气口。在文章的第一部分，Rock 'n Roll模型及其与Biasi相关的操作版本被提出。文章的第二部分将致力于在ANSYS CFX中实现恒摩擦速度的Rock 'n 'Roll模型及其对非恒摩擦速度的适应。最后，本文给出了大型真空容器空气进入情况下粒子再悬浮的模拟结果。这种情况特别有趣，也不直观，因为当初始压力降低时，粒子的行为与大气压下的不同。此外，由于低压环境，气流力和重力之间的竞争可能会限制再悬浮，并且在颗粒的运输和沉积中也必须考虑压力的影响(Nerisson, 2011)。研究了三种颗粒直径，允许用该参数显示再悬浮的演变，并计算了空气进入过程中的粉尘再悬浮率和空气中的分数。
{"title":"Implementation of a Particle Resuspension Model in a CFD Code: Application to an Air Ingress Scenario in a Vacuum Toroidal Vessel","authors":"T. Gelain, L. Ricciardi, F. Gensdarmes","doi":"10.1115/icone2020-16139","DOIUrl":"https://doi.org/10.1115/icone2020-16139","url":null,"abstract":"\u0000 During a loss of vacuum accident (LOVA), dust particles that will be present in the future tokamak ITER are likely to be resuspended, inducing a risk for explosion and airborne contamination. Evaluating the particle resuspension/deposition and resulting airborne concentration in case of a LOVA is therefore a major issue and it can be investigated by using a CFD code. To this end, this article presents the implementation of a resuspension model in a CFD code (ANSYS CFX) and its application to an air ingress in a vacuum toroidal vessel with a volume comparable to ITER one.\u0000 In the first part of the article, the Rock’n Roll model and its operational version with the Biasi’s correlation is presented.\u0000 The second part of the article will be devoted to the implementation of the Rock’n’Roll model in ANSYS CFX for constant friction velocities and its adaptation to non-constant friction velocities.\u0000 Finally, the paper presents the simulations obtained on the particle resuspension for an air ingress scenario in a large vacuum vessel. This case is particularly interesting and non-intuitive because as the initial pressure is reduced, the particle behavior is different from that at atmospheric pressure. Further, a competition between airflow forces and gravitational force occurs, due to the low pressure environment, potentially restricting the resuspension, and the pressure influence also has to be taken into account in the particle transport and deposition (Nerisson, 2011).\u0000 Three particle diameters were studied allowing to show the evolution of the resuspension with this parameter and to calculate dust resuspension rates and airborne fractions during the air ingress.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":"68 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78092215","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}
Takayuki Suzuki, H. Yoshida, Naoki Horiguchi, Sota Yamamura, Y. Abe
In the severe accident (SA) of nuclear reactors, fuel and components melt, and melted materials fall to a lower part of a reactor vessel. In the lower part of a reactor vessel, in some sections of the SAs, it is considered that there is a water pool. Then, the melted core materials fall into a water pool in the lower plenum as a jet. The molten material jet is broken up, and heat transfer between molten material and coolant may occur. This process is called a fuel-coolant interaction (FCI). FCI is one of the important phenomena to consider the coolability and distribution of core materials. In this study, the numerical simulation of jet breakup phenomena with a shallow pool was performed by using the developed method (TPFIT). We try to understand the hydrodynamic interaction under various, such as penetration, reach to the bottom, spread, accumulation of the molten material jet. Also, we evaluated a detailed jet spread behavior and examined the influence of lattice resolution and the contact angle. Furthermore, the diameters of atomized droplets were evaluated by using numerical simulation data.
{"title":"Numerical Simulation of Liquid Jet Behavior in Shallow Pool by Interface Tracking Method","authors":"Takayuki Suzuki, H. Yoshida, Naoki Horiguchi, Sota Yamamura, Y. Abe","doi":"10.1115/icone2020-16213","DOIUrl":"https://doi.org/10.1115/icone2020-16213","url":null,"abstract":"\u0000 In the severe accident (SA) of nuclear reactors, fuel and components melt, and melted materials fall to a lower part of a reactor vessel. In the lower part of a reactor vessel, in some sections of the SAs, it is considered that there is a water pool. Then, the melted core materials fall into a water pool in the lower plenum as a jet. The molten material jet is broken up, and heat transfer between molten material and coolant may occur. This process is called a fuel-coolant interaction (FCI). FCI is one of the important phenomena to consider the coolability and distribution of core materials.\u0000 In this study, the numerical simulation of jet breakup phenomena with a shallow pool was performed by using the developed method (TPFIT). We try to understand the hydrodynamic interaction under various, such as penetration, reach to the bottom, spread, accumulation of the molten material jet. Also, we evaluated a detailed jet spread behavior and examined the influence of lattice resolution and the contact angle. Furthermore, the diameters of atomized droplets were evaluated by using numerical simulation data.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":"21 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81462087","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 radioactive aerosol removal equipment is used as one of the safety systems of nuclear reactors. In this equipment, microparticles of aerosol are removed through gas-liquid interfaces of two-phase flow. The mechanism related to the removal of microparticles through the gas-liquid interface is not precise; a numerical evaluation method of performance of aerosol removal equipment is not realized. Then, we have started to construct a numerical simulation method to simulate the removal of microparticles through gas-liquid interfaces. In this simulation method, a detailed two-phase flow simulation code TPFIT is used as the basis of this method. TPFIT adopts an advanced interface tracking method and can simulate interface movement and deformation directly. Also, to simulate the movement of particles, the Lagrangian particle tracking method is incorporated. By combining the interface tracking method, and the Lagrangian particle tracking method, the interaction between interfaces and microparticles can be simulated in detail. To solve the Lagrangian equations of particles, fluid properties and fluid velocity surrounding aerosol particles are evaluated by considering the relative position of particles and gas-liquid interface, to simulate particle movement near the interface. In this paper, we show an outline and preliminary results of this simulation method.
{"title":"Numerical Simulation of Microparticles Motion in Two-Phase Bubbly Flow","authors":"H. Yoshida, Shin-ichiro Uesawa","doi":"10.1115/icone2020-16393","DOIUrl":"https://doi.org/10.1115/icone2020-16393","url":null,"abstract":"\u0000 The radioactive aerosol removal equipment is used as one of the safety systems of nuclear reactors. In this equipment, microparticles of aerosol are removed through gas-liquid interfaces of two-phase flow. The mechanism related to the removal of microparticles through the gas-liquid interface is not precise; a numerical evaluation method of performance of aerosol removal equipment is not realized. Then, we have started to construct a numerical simulation method to simulate the removal of microparticles through gas-liquid interfaces. In this simulation method, a detailed two-phase flow simulation code TPFIT is used as the basis of this method. TPFIT adopts an advanced interface tracking method and can simulate interface movement and deformation directly. Also, to simulate the movement of particles, the Lagrangian particle tracking method is incorporated. By combining the interface tracking method, and the Lagrangian particle tracking method, the interaction between interfaces and microparticles can be simulated in detail. To solve the Lagrangian equations of particles, fluid properties and fluid velocity surrounding aerosol particles are evaluated by considering the relative position of particles and gas-liquid interface, to simulate particle movement near the interface. In this paper, we show an outline and preliminary results of this simulation method.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":"51 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88613571","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}
Linyu Lin, Paridhi Athe, P. Rouxelin, Truc-Nam Dinh, J. Lane
In this work, a Nearly Autonomous Management and Control (NAMAC) system is designed to diagnose the reactor state and provide recommendations to the operator for maintaining the safety and performance of the reactor. A three layer-hierarchical workflow is suggested to guide the design and development of the NAMAC system. The three layers in this workflow corresponds to knowledge base, digital twin developmental layer (for different NAMAC functions), and NAMAC operational layer. Digital twin in NAMAC is described as knowledge acquisition system to support different autonomous control functions. Therefore, based on the knowledge base, a set of digital twin models is trained to determine the plant state, predict behavior of physical components or systems, and rank available control options. The trained digital twin models are assembled according to NAMAC operational workflow to support decision-making process in selecting the optimal control actions during an accident scenario. To demonstrate the capability of the NAMAC system, a case study is designed, where a baseline NAMAC is implemented for operating a simulator of the Experimental Breeder Reactor II (EBR-II) during a single loss of flow accident. Training database for development of digital twin models is obtained by sampling the control parameters in the GOTHIC data generation engine. After the training and testing, the digital twins are assembled into a NAMAC system according to the operational workflow. This NAMAC system is coupled with the GOTHIC plant simulator, and a confusion matrix is generated to illustrate the accuracy and robustness of implemented NAMAC system. It is found that within the training databases, NAMAC can make reasonable recommendations with zero confusion rate. However, when the scenario is beyond the training cases, the confusion rate increases, especially when the scenarios are more severe. Therefore, a discrepancy checker is added to detect unexpected reactor states and alert operators for safety-minded actions.
{"title":"Development and Assessment of a Nearly Autonomous Management and Control System During a Single Loss of Flow Accident","authors":"Linyu Lin, Paridhi Athe, P. Rouxelin, Truc-Nam Dinh, J. Lane","doi":"10.1115/icone2020-16908","DOIUrl":"https://doi.org/10.1115/icone2020-16908","url":null,"abstract":"\u0000 In this work, a Nearly Autonomous Management and Control (NAMAC) system is designed to diagnose the reactor state and provide recommendations to the operator for maintaining the safety and performance of the reactor. A three layer-hierarchical workflow is suggested to guide the design and development of the NAMAC system. The three layers in this workflow corresponds to knowledge base, digital twin developmental layer (for different NAMAC functions), and NAMAC operational layer. Digital twin in NAMAC is described as knowledge acquisition system to support different autonomous control functions. Therefore, based on the knowledge base, a set of digital twin models is trained to determine the plant state, predict behavior of physical components or systems, and rank available control options. The trained digital twin models are assembled according to NAMAC operational workflow to support decision-making process in selecting the optimal control actions during an accident scenario.\u0000 To demonstrate the capability of the NAMAC system, a case study is designed, where a baseline NAMAC is implemented for operating a simulator of the Experimental Breeder Reactor II (EBR-II) during a single loss of flow accident. Training database for development of digital twin models is obtained by sampling the control parameters in the GOTHIC data generation engine. After the training and testing, the digital twins are assembled into a NAMAC system according to the operational workflow. This NAMAC system is coupled with the GOTHIC plant simulator, and a confusion matrix is generated to illustrate the accuracy and robustness of implemented NAMAC system. It is found that within the training databases, NAMAC can make reasonable recommendations with zero confusion rate. However, when the scenario is beyond the training cases, the confusion rate increases, especially when the scenarios are more severe. Therefore, a discrepancy checker is added to detect unexpected reactor states and alert operators for safety-minded actions.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":"40 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"84118833","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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