This paper describes the modeling of flow regimes beyond bubbly flows in a large diameter channel considering polydispersity and bubble induced turbulence using the Eulerian two-fluid approach. A two-bubble-group approach with two-group interfacial area transport equations (IATEs) is used to demonstrate flow phenomena in a large diameter pipe. Source and sink terms for mass and momentum exchanges between the two groups of bubbles and for bubble coalescence and breakup mechanisms are implemented. For predicting particle size and its distribution, S-Gamma (Sγ) model is used. The Sγ model with two-group IATEs are evaluated by comparing local distributions of void fractions and Sauter mean diameters with results of adaptive-multiple-size-group (AMUSIG) models and experimental dataset developed by Schlegel et al., (2012) for model validations. It shows that two-group IATEs with Sγ model predict reasonably accurate flow characteristics of beyond bubbly flow regimes, but also show shortcomings in their accuracies predicting local distributions, which imply that further studies for modeling of interfacial force are needed.
{"title":"CFD Analysis of S-Gamma Model Coupled With Two-Group Interfacial Area Transport Equations and AMUSIG Model for a Large Diameter Pipe","authors":"Sungje Hong, J. Schlegel, Subash Sharma","doi":"10.1115/icone2020-16645","DOIUrl":"https://doi.org/10.1115/icone2020-16645","url":null,"abstract":"\u0000 This paper describes the modeling of flow regimes beyond bubbly flows in a large diameter channel considering polydispersity and bubble induced turbulence using the Eulerian two-fluid approach. A two-bubble-group approach with two-group interfacial area transport equations (IATEs) is used to demonstrate flow phenomena in a large diameter pipe. Source and sink terms for mass and momentum exchanges between the two groups of bubbles and for bubble coalescence and breakup mechanisms are implemented. For predicting particle size and its distribution, S-Gamma (Sγ) model is used. The Sγ model with two-group IATEs are evaluated by comparing local distributions of void fractions and Sauter mean diameters with results of adaptive-multiple-size-group (AMUSIG) models and experimental dataset developed by Schlegel et al., (2012) for model validations. It shows that two-group IATEs with Sγ model predict reasonably accurate flow characteristics of beyond bubbly flow regimes, but also show shortcomings in their accuracies predicting local distributions, which imply that further studies for modeling of interfacial force are needed.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77747507","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, P. Rouxelin, Paridhi Athe, Truc-Nam Dinh, J. Lane
A critical component of the autonomous control system is the implementation of digital twin (DT) for diagnosing the conditions and prognosing the future transients of physical components or systems. The objective is to achieve an accurate understanding and prediction of future behaviors of the physical components or systems and to guide operating decisions by an operator or an autonomous control system. With specific requirements in the functional, interface, modeling, and accuracy, DTs are developed based on operational and simulation databases. As one of the modeling methods, data-driven methods have been used for implementing DTs since they have more adaptive forms and are able to capture interdependencies that can be overlooked in model-based DTs. To demonstrate the capabilities of DTs, a case study is designed for the control of the EBR-II sodium-cooled fast reactor during a single loss of flow accident, where either a complete or a partial loss of flow in one of the two primary sodium pumps is considered. Based on the definition of DTs and the design of autonomous control system, DTs for diagnosis and prognosis are implemented by training feedforward neural networks with suggested inputs, training parameters, and knowledge base. Furthermore, inspired by the validation and uncertainty quantification scheme for scientific computing, a list of sources of uncertainty in input variables, training parameters, and knowledge base is formulated. The objective is to assess qualitative impacts of different sources of uncertainty on the DT errors. It is found that the performance of DT for diagnosis and prognosis satisfies the acceptance criteria within the training databases. Meanwhile, the accuracy of DTs for diagnosis and prognosis is highly affected by multiple sources of uncertainty.
{"title":"Development and Assessment of Data-Driven Digital Twins in a Nearly Autonomous Management and Control System for Advanced Reactors","authors":"Linyu Lin, P. Rouxelin, Paridhi Athe, Truc-Nam Dinh, J. Lane","doi":"10.1115/icone2020-16813","DOIUrl":"https://doi.org/10.1115/icone2020-16813","url":null,"abstract":"\u0000 A critical component of the autonomous control system is the implementation of digital twin (DT) for diagnosing the conditions and prognosing the future transients of physical components or systems. The objective is to achieve an accurate understanding and prediction of future behaviors of the physical components or systems and to guide operating decisions by an operator or an autonomous control system. With specific requirements in the functional, interface, modeling, and accuracy, DTs are developed based on operational and simulation databases. As one of the modeling methods, data-driven methods have been used for implementing DTs since they have more adaptive forms and are able to capture interdependencies that can be overlooked in model-based DTs.\u0000 To demonstrate the capabilities of DTs, a case study is designed for the control of the EBR-II sodium-cooled fast reactor during a single loss of flow accident, where either a complete or a partial loss of flow in one of the two primary sodium pumps is considered. Based on the definition of DTs and the design of autonomous control system, DTs for diagnosis and prognosis are implemented by training feedforward neural networks with suggested inputs, training parameters, and knowledge base. Furthermore, inspired by the validation and uncertainty quantification scheme for scientific computing, a list of sources of uncertainty in input variables, training parameters, and knowledge base is formulated. The objective is to assess qualitative impacts of different sources of uncertainty on the DT errors. It is found that the performance of DT for diagnosis and prognosis satisfies the acceptance criteria within the training databases. Meanwhile, the accuracy of DTs for diagnosis and prognosis is highly affected by multiple sources of uncertainty.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"84888986","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}
As analysts still grapple with understanding core damage accident progression at Three Mile Island and Fukushima that caught the nuclear industry off-guard once too many times, one notices the very limited detail with which the large reactor cores of these subject reactors have been modelled in their severe accident simulation code packages. At the same time, modelling of CANDU severe accidents have largely borrowed from and suffered from the limitations of the same LWR codes (see IAEA TECDOC 1727) whose applications to PHWRs have poorly caught critical PHWR design specifics and vulnerabilities. As a result, accident management measures that have been instituted at CANDU PHWRs, while meeting the important industry objective of publically seeming to be doing something about lessons learnt from say Fukushima and showing that the reactor designs are oh so close to perfect and the off-site consequences of severe accidents happily benign. Integrated PHWR severe accident progression and consequence assessment code ROSHNI can make a significant contribution to actual, practical understanding of severe accident progression in CANDU PHWRs, improving significantly on the other PHWR specific computer codes developed three decades ago when modeling decisions were constrained by limited computing power and poor understanding of and interest in severe core damage accidents. These codes force gross simplifications in reactor core modelling and do not adequately represent all the right CANDU core details, materials, fluids, vessels or phenomena. But they produce results that are familiar and palatable. They do, however to their credit, also excel in their computational speed, largely because they model and compute so little and with such un-necessary simplifications. ROSHNI sheds most previous modelling simplifications and represents each of the 380 channels, 4560 bundle, 37 elements in four concentric ring, Zircaloy clad fuel geometry, materials and fluids more faithfully in a 2000 MW(Th) CANDU6 reactor. It can be used easily for other PHWRs with different number of fuel channels and bundles per each channel. Each of horizontal PHWR reactor channels with all their bundles, fuel rings, sheaths, appendages, end fittings and feeders are modelled and in detail that reflects large across core differences. While other codes model at best a few hundred core fuel entities, thermo-chemical transient behaviour of about 73,000 different fuel channel entities within the core is considered by ROSHNI simultaneously along with other 15,000 or so other flow path segments. At each location all known thermo-chemical and hydraulic phenomena are computed. With such detail, ROSHNI is able to provide information on their progressive and parallel thermo-chemical contribution to accident progression and a more realistic fission product release source term that would belie the miniscule one (100 TBq of Cs-137 or 0.15% of core inventory) used by EMOs now in Canada on recommendation
{"title":"Simulation of Severe Accident Progression Using ROSHNI: A New Integrated Simulation Code for PHWR Severe Accidents","authors":"S. Nijhawan, Y. Song","doi":"10.1115/icone2020-16633","DOIUrl":"https://doi.org/10.1115/icone2020-16633","url":null,"abstract":"\u0000 As analysts still grapple with understanding core damage accident progression at Three Mile Island and Fukushima that caught the nuclear industry off-guard once too many times, one notices the very limited detail with which the large reactor cores of these subject reactors have been modelled in their severe accident simulation code packages. At the same time, modelling of CANDU severe accidents have largely borrowed from and suffered from the limitations of the same LWR codes (see IAEA TECDOC 1727) whose applications to PHWRs have poorly caught critical PHWR design specifics and vulnerabilities. As a result, accident management measures that have been instituted at CANDU PHWRs, while meeting the important industry objective of publically seeming to be doing something about lessons learnt from say Fukushima and showing that the reactor designs are oh so close to perfect and the off-site consequences of severe accidents happily benign.\u0000 Integrated PHWR severe accident progression and consequence assessment code ROSHNI can make a significant contribution to actual, practical understanding of severe accident progression in CANDU PHWRs, improving significantly on the other PHWR specific computer codes developed three decades ago when modeling decisions were constrained by limited computing power and poor understanding of and interest in severe core damage accidents. These codes force gross simplifications in reactor core modelling and do not adequately represent all the right CANDU core details, materials, fluids, vessels or phenomena. But they produce results that are familiar and palatable. They do, however to their credit, also excel in their computational speed, largely because they model and compute so little and with such un-necessary simplifications.\u0000 ROSHNI sheds most previous modelling simplifications and represents each of the 380 channels, 4560 bundle, 37 elements in four concentric ring, Zircaloy clad fuel geometry, materials and fluids more faithfully in a 2000 MW(Th) CANDU6 reactor. It can be used easily for other PHWRs with different number of fuel channels and bundles per each channel. Each of horizontal PHWR reactor channels with all their bundles, fuel rings, sheaths, appendages, end fittings and feeders are modelled and in detail that reflects large across core differences. While other codes model at best a few hundred core fuel entities, thermo-chemical transient behaviour of about 73,000 different fuel channel entities within the core is considered by ROSHNI simultaneously along with other 15,000 or so other flow path segments. At each location all known thermo-chemical and hydraulic phenomena are computed. With such detail, ROSHNI is able to provide information on their progressive and parallel thermo-chemical contribution to accident progression and a more realistic fission product release source term that would belie the miniscule one (100 TBq of Cs-137 or 0.15% of core inventory) used by EMOs now in Canada on recommendation ","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82167985","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 solution of the highly complex unsteady high speed oscillating compressible flow field inside a cylindrical tube has been obtained numerically, assuming one dimensional, viscous, and heat conducting flow, by solving the appropriate fluid dynamic and energy equations. The tube is approximated by a right circular cylinder closed at one end with a piston oscillating at very high resonant frequency at the other end. An iterative implicit finite difference scheme is employed to obtain the solution. The scheme permits arbitrary boundary conditions at the piston and the end wall and allows assumptions for transport properties. The solution would also be valid for tapered tubes if the variations in the cross-sectional area are small. In successfully predicting the time dependent results, an innovative simple but stable solution of unsteady fluid dynamic and energy equations is provided here for wide ranging research, design, development, analysis, and industrial applications in solving a variety of complex fluid flow heat transfer problems. The method is directly applicable to pulsed or pulsating flow and wave motion thermal energy transport, fluid-structure interaction heat transfer enhancement, and fluidic pyrotechnic initiation devices. It can further be easily extended to cover muzzle blasts and nuclear explosion blast wave propagations in one dimensional and/or radial spherical coordinates with or without including energy generation / addition terms.
{"title":"Time-Dependent Solution of Unsteady Fluid Flow Equations for High Speed Oscillating Compressible Flows and Blast Wave Propagations","authors":"R. Sinha","doi":"10.1115/icone2020-16419","DOIUrl":"https://doi.org/10.1115/icone2020-16419","url":null,"abstract":"\u0000 A solution of the highly complex unsteady high speed oscillating compressible flow field inside a cylindrical tube has been obtained numerically, assuming one dimensional, viscous, and heat conducting flow, by solving the appropriate fluid dynamic and energy equations. The tube is approximated by a right circular cylinder closed at one end with a piston oscillating at very high resonant frequency at the other end. An iterative implicit finite difference scheme is employed to obtain the solution. The scheme permits arbitrary boundary conditions at the piston and the end wall and allows assumptions for transport properties. The solution would also be valid for tapered tubes if the variations in the cross-sectional area are small. In successfully predicting the time dependent results, an innovative simple but stable solution of unsteady fluid dynamic and energy equations is provided here for wide ranging research, design, development, analysis, and industrial applications in solving a variety of complex fluid flow heat transfer problems. The method is directly applicable to pulsed or pulsating flow and wave motion thermal energy transport, fluid-structure interaction heat transfer enhancement, and fluidic pyrotechnic initiation devices. It can further be easily extended to cover muzzle blasts and nuclear explosion blast wave propagations in one dimensional and/or radial spherical coordinates with or without including energy generation / addition terms.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79725713","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}
Sifan Peng, Yujia Liu, N. Gui, Xingtuan Yang, J. Tu, Shengyao Jiang
Graphite is widely used in nuclear reactors as moderator and structural material. Among present graphite preparation methods, air flow mill is considered to be qualified in the control of particle size and purity, and promising for future mass production. In this work, an opposed jet mill is designed to crush large graphite particles. The opposed jet mill accelerates the particles through two supersonic jet flows in opposite directions, and finally the particles collide in the crushing cavity. In order to estimate the performance of opposed jet mill, it is necessary to solve the coupling calculation of the compressible flow and the collision process of discrete particles. However, the research on calculating the compressible gas solid coupling problems is scarcely rare. In this paper, coupled CFD-DEM model is used to simulate the particle movement process with jet flows and accompanying jet in opposed jet mill. By comparing with experimental results, it is proved that these simulation results of the acceleration process of compressible gas through these nozzles and the collision process of the final two supersonic jet flows in the opposed-jet mill are accurate, with the accuracy model of the coupled CFD-DEM provided. The practice has proved that the contrastive flow mill has a broad application prospect in the production of graphite particles.
{"title":"CFD-DEM Simulations of Graphite Particle Collisions in Opposed Jet Mill","authors":"Sifan Peng, Yujia Liu, N. Gui, Xingtuan Yang, J. Tu, Shengyao Jiang","doi":"10.1115/icone2020-16340","DOIUrl":"https://doi.org/10.1115/icone2020-16340","url":null,"abstract":"\u0000 Graphite is widely used in nuclear reactors as moderator and structural material. Among present graphite preparation methods, air flow mill is considered to be qualified in the control of particle size and purity, and promising for future mass production. In this work, an opposed jet mill is designed to crush large graphite particles. The opposed jet mill accelerates the particles through two supersonic jet flows in opposite directions, and finally the particles collide in the crushing cavity. In order to estimate the performance of opposed jet mill, it is necessary to solve the coupling calculation of the compressible flow and the collision process of discrete particles. However, the research on calculating the compressible gas solid coupling problems is scarcely rare. In this paper, coupled CFD-DEM model is used to simulate the particle movement process with jet flows and accompanying jet in opposed jet mill. By comparing with experimental results, it is proved that these simulation results of the acceleration process of compressible gas through these nozzles and the collision process of the final two supersonic jet flows in the opposed-jet mill are accurate, with the accuracy model of the coupled CFD-DEM provided. The practice has proved that the contrastive flow mill has a broad application prospect in the production of graphite particles.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78286741","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}
Vertical swirling gas-liquid flow is a kind of complex two-phase flow containing a nonzero tangential velocity component in engineering applications. The accurate flow regime characterization, phase distribution information and pressure drop data about vertical swirling flow are the basis for the optimization of steam generator (SG), which can greatly reduce the cost and improve the safety of nuclear plants. To get these key parameters of swirling vertical flow, we have made a comprehensive visualization experiment in a vertical pipe with 30mm diameter and 5m length by high-speed camera. The experimental pipe is separated into swirling part and non-swirling part. We have set three observation section with different vertical heights in the swirling part. Changing the flow rate of water and gas, different swirling flow pattern photos can be captured by high-speed camera. Based on the photos of different positions and image-processing MATLAB code, we can get three flow regime maps and figure out the decaying law of swirling gas-liquid flow. The pressure drop can be recorded by rotameter at each position. The decaying law of pressrure drop can be concluded from it. These data can be a guide for designing gas-liquid separator in SG to improve the efficiency of nuclear plant.
{"title":"Experimental Study on Flow Patterns and Pressure Drop of Decaying Swirling Gas-Liquid Flow in a Vertical Pipe","authors":"Jiarong Zhang, Li Liu, L. Shuai, Gu Hanyang","doi":"10.1115/icone2020-16484","DOIUrl":"https://doi.org/10.1115/icone2020-16484","url":null,"abstract":"\u0000 Vertical swirling gas-liquid flow is a kind of complex two-phase flow containing a nonzero tangential velocity component in engineering applications. The accurate flow regime characterization, phase distribution information and pressure drop data about vertical swirling flow are the basis for the optimization of steam generator (SG), which can greatly reduce the cost and improve the safety of nuclear plants. To get these key parameters of swirling vertical flow, we have made a comprehensive visualization experiment in a vertical pipe with 30mm diameter and 5m length by high-speed camera. The experimental pipe is separated into swirling part and non-swirling part. We have set three observation section with different vertical heights in the swirling part. Changing the flow rate of water and gas, different swirling flow pattern photos can be captured by high-speed camera. Based on the photos of different positions and image-processing MATLAB code, we can get three flow regime maps and figure out the decaying law of swirling gas-liquid flow. The pressure drop can be recorded by rotameter at each position. The decaying law of pressrure drop can be concluded from it. These data can be a guide for designing gas-liquid separator in SG to improve the efficiency of nuclear plant.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79496202","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}
Hydrogen combustion and explosion is an important safety issue in nuclear power plants (NPPs) containment during postulated severe accidents or in utilization of hydrogen. It is significant to understand the hydrogen flow and distribution in space for mitigating hydrogen risk. In this paper, a numerical model to investigate hydrogen flow and distribution in a vessel is established using computational fluid dynamics (CFD) code GASFLOW. Hydrogen is simulated by helium which is used to study the hydrogen distribution. The k-ε turbulent model is selected to establish the numerical model, and the numerical model which has no obstacle inside the vessel or includes the obstacle inside is verified under medium momentum conditions of injected gas by comparing numerical results with experimental data. Regardless of the presence of the obstacle in the vessel, helium stratification occurs under all momentum condition of injected gas. When the obstacle is present, it blocks the flow path of the injected helium to the upper space, then the helium volume concentration in upper space is lower than the condition that there is no obstacle in the vessel. As the initial Froude number increases from 0.19 to 19.29, the flow mechanism around the obstacle under high momentum condition of injected gas is different from that under medium or low momentum conditions. Consequently, the boundary of the helium stratification moves down, and the distribution of helium looks more uniform in most area of the vessel for high momentum conditions of injected gas.
{"title":"Numerical Investigation of the Obstacle on Hydrogen Distribution in a Vessel","authors":"Tianlin Wang, L. Tong, Xuewu Cao","doi":"10.1115/icone2020-16273","DOIUrl":"https://doi.org/10.1115/icone2020-16273","url":null,"abstract":"\u0000 Hydrogen combustion and explosion is an important safety issue in nuclear power plants (NPPs) containment during postulated severe accidents or in utilization of hydrogen. It is significant to understand the hydrogen flow and distribution in space for mitigating hydrogen risk. In this paper, a numerical model to investigate hydrogen flow and distribution in a vessel is established using computational fluid dynamics (CFD) code GASFLOW. Hydrogen is simulated by helium which is used to study the hydrogen distribution. The k-ε turbulent model is selected to establish the numerical model, and the numerical model which has no obstacle inside the vessel or includes the obstacle inside is verified under medium momentum conditions of injected gas by comparing numerical results with experimental data. Regardless of the presence of the obstacle in the vessel, helium stratification occurs under all momentum condition of injected gas. When the obstacle is present, it blocks the flow path of the injected helium to the upper space, then the helium volume concentration in upper space is lower than the condition that there is no obstacle in the vessel. As the initial Froude number increases from 0.19 to 19.29, the flow mechanism around the obstacle under high momentum condition of injected gas is different from that under medium or low momentum conditions. Consequently, the boundary of the helium stratification moves down, and the distribution of helium looks more uniform in most area of the vessel for high momentum conditions of injected gas.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83366356","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}
When a pressurised pipe in a reactor coolant system breaks, it results in hydraulic loads on the reactor containment system and escaping fluid exerts a thrust force on the pipe. The double-ended guillotine break (DEGB) is generally the most onerous loss of coolant accident (LOCA) in design of a reactor coolant system [1]. In addition to the hydraulic loads, the continuing thrust force on the broken end of the pipe generates a rapidly accelerating rotational displacement of the pipe section on the break side of the plastic hinge, the phenomenon called pipe whip. The whipping pipe has the potential to damage objects within the hazard zone therefore must be assessed. Currently a 20bar threshold is used in nuclear power plant (NPP) design of light water reactors (LWRs) to delineate high energy lines (HEL) and moderate energy lines (MEL). The threshold is used as a process to establish the requirement for additional pipe whip assessments to be performed as part of HEL guillotine break analysis. It is currently argued that no such studies are required for MEL. However the basis of using the 20bar threshold for not carrying out pipe whip assessment of MELs is not well understood. The work presented here provides details of the finite element (FE) analyses undertaken to substantiate the 20 bar threshold used for the differentiation of HEL and MEL. Using the FE analysis method, a range of pipe characteristics, and pressures in the range of 10–50bar have been studied to determine whether a plastic hinge will occur, and whether pipe whip effects will be seen for that case. The FE results have also been used to assess the equivalent plastic strain criteria generally used to define the formation of a plastic hinge and initiation of pipe whip.
{"title":"Using Pipe Whip Analysis via the Finite Element Method to Underpin the Delineation Between High and Moderate Energy Lines","authors":"A. Hurst, P. Bansal","doi":"10.1115/icone2020-16474","DOIUrl":"https://doi.org/10.1115/icone2020-16474","url":null,"abstract":"\u0000 When a pressurised pipe in a reactor coolant system breaks, it results in hydraulic loads on the reactor containment system and escaping fluid exerts a thrust force on the pipe. The double-ended guillotine break (DEGB) is generally the most onerous loss of coolant accident (LOCA) in design of a reactor coolant system [1]. In addition to the hydraulic loads, the continuing thrust force on the broken end of the pipe generates a rapidly accelerating rotational displacement of the pipe section on the break side of the plastic hinge, the phenomenon called pipe whip. The whipping pipe has the potential to damage objects within the hazard zone therefore must be assessed. Currently a 20bar threshold is used in nuclear power plant (NPP) design of light water reactors (LWRs) to delineate high energy lines (HEL) and moderate energy lines (MEL). The threshold is used as a process to establish the requirement for additional pipe whip assessments to be performed as part of HEL guillotine break analysis. It is currently argued that no such studies are required for MEL. However the basis of using the 20bar threshold for not carrying out pipe whip assessment of MELs is not well understood.\u0000 The work presented here provides details of the finite element (FE) analyses undertaken to substantiate the 20 bar threshold used for the differentiation of HEL and MEL. Using the FE analysis method, a range of pipe characteristics, and pressures in the range of 10–50bar have been studied to determine whether a plastic hinge will occur, and whether pipe whip effects will be seen for that case. The FE results have also been used to assess the equivalent plastic strain criteria generally used to define the formation of a plastic hinge and initiation of pipe whip.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83418954","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 long-reach articulated manipulator that can deploy various sensors in a large but confined workspace is urgently needed in decommissioning tasks at Fukushima Daiichi Nuclear Power Stations. The most critical problem involves managing the large gravitational torques acting on the manipulator’s joints. In previous studies, several prototype models were developed; however, they were extremely heavy and bulky. This paper presents fundamental ideas for creating a lightweight super long-reach articulated manipulator. First, a mechanical structure could be utilized to support gravity. Second, a vertical force could be generated to compensate for gravity. Third, a tendon-driven mechanism could be utilized because it permits the installation of heavy actuators on the base. The tendons can transmit the actuator power to each joint. Thus, the weight of the arm is significantly reduced. We have discussed the advantages and drawbacks of each recommendation based on the hardware prototyping.
{"title":"Challenge to Investigation of Fuel Debris in RPV by an Advanced Super Dragon Articulated Robot Arm: Design and Prototyping of a Lightweight Super Long Reach Articulated Manipulator","authors":"G. Endo, H. Takahashi, H. Kikura","doi":"10.1115/icone2020-16834","DOIUrl":"https://doi.org/10.1115/icone2020-16834","url":null,"abstract":"\u0000 A long-reach articulated manipulator that can deploy various sensors in a large but confined workspace is urgently needed in decommissioning tasks at Fukushima Daiichi Nuclear Power Stations. The most critical problem involves managing the large gravitational torques acting on the manipulator’s joints. In previous studies, several prototype models were developed; however, they were extremely heavy and bulky. This paper presents fundamental ideas for creating a lightweight super long-reach articulated manipulator. First, a mechanical structure could be utilized to support gravity. Second, a vertical force could be generated to compensate for gravity. Third, a tendon-driven mechanism could be utilized because it permits the installation of heavy actuators on the base. The tendons can transmit the actuator power to each joint. Thus, the weight of the arm is significantly reduced. We have discussed the advantages and drawbacks of each recommendation based on the hardware prototyping.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90116495","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}
Yassine Serbouti, K. Kurihara, Y. Kometani, Masatoshi Itagaki, Makoto Tatemura
Control rod blades are comprised of a stainless steel sheath, which contains neutron absorber tubes (filled with boron carbide powder). During decommissioning, the first stage of size reduction consists of cutting the connector (bottom portion) of the control rod, while the second stage consists of separating the blades of the control rod by cutting through the tie rod. The last stage consists of segmenting the control rod blades by cutting through absorber tubes. In this study, the control rod blades segmentation (last stage of size reduction) is investigated using an actual control rod (unused). During the experiments, we used a forming press on the cut locations followed by a plasma arc cutting underwater. The purpose of this cutting technique is to minimize the scattering of boron carbides into water by using the stainless sheath melt to seal the absorber tubes. After the segmentation, we confirmed the sealing of the absorber tubes by visually examining the cut cross-sections. The water analysis showed that the boron carbide scattering was relatively low (only 0.07% of the total boron carbides was scattered). Finally, we confirmed that the off-gas emission is considerably reduced by using Argon plasma instead of Argon-Hydrogen plasma.
{"title":"Control Rod Blades Size Reduction Using Underwater Plasma Cutting and its Effects on Boron Carbide Powder Scattering","authors":"Yassine Serbouti, K. Kurihara, Y. Kometani, Masatoshi Itagaki, Makoto Tatemura","doi":"10.1115/icone2020-16414","DOIUrl":"https://doi.org/10.1115/icone2020-16414","url":null,"abstract":"\u0000 Control rod blades are comprised of a stainless steel sheath, which contains neutron absorber tubes (filled with boron carbide powder). During decommissioning, the first stage of size reduction consists of cutting the connector (bottom portion) of the control rod, while the second stage consists of separating the blades of the control rod by cutting through the tie rod. The last stage consists of segmenting the control rod blades by cutting through absorber tubes.\u0000 In this study, the control rod blades segmentation (last stage of size reduction) is investigated using an actual control rod (unused). During the experiments, we used a forming press on the cut locations followed by a plasma arc cutting underwater. The purpose of this cutting technique is to minimize the scattering of boron carbides into water by using the stainless sheath melt to seal the absorber tubes.\u0000 After the segmentation, we confirmed the sealing of the absorber tubes by visually examining the cut cross-sections. The water analysis showed that the boron carbide scattering was relatively low (only 0.07% of the total boron carbides was scattered). Finally, we confirmed that the off-gas emission is considerably reduced by using Argon plasma instead of Argon-Hydrogen plasma.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86243311","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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