Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518416
R. Bulmer
The Tokamak Physics Experiment (TPX) will have a poloidal field system capable of full inductive operation for approximately a 20-s flattop and, with superconducting toroidal and poloidal field coils and non-inductive current drive, it will be capable of true steady-state operation. The poloidal field design is based on the ideal MHD equilibrium model as implemented in the TEQ code developed at LLNL. The PF coils are arranged in an up-down symmetric configuration, external to the TF coils. The TPX diverted plasma will have an aspect ratio of 4.5 and is highly shaped with a nominal elongation of 2 and triangularity of approximately 0.8 as measured at the separatrix. The tokamak design is based on a high-current (q/sub /spl Psi//=3) plasma scenario and a low current scenario. Each scenario has an operational flexibility requirement which is defined as a region of plasma pressure and inductivity (/spl beta//sub N/-l/sub i/) space, where the plasma shape is constrained to keep the divertor configuration operational. Single-null plasma configurations are feasible, even with the same divertor hardware, by operating the PF coils asymmetrically. Recently applied optimization techniques have improved the capability of the PF system without additional cost.
{"title":"Tokamak Physics Experiment poloidal field design","authors":"R. Bulmer","doi":"10.1109/FUSION.1993.518416","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518416","url":null,"abstract":"The Tokamak Physics Experiment (TPX) will have a poloidal field system capable of full inductive operation for approximately a 20-s flattop and, with superconducting toroidal and poloidal field coils and non-inductive current drive, it will be capable of true steady-state operation. The poloidal field design is based on the ideal MHD equilibrium model as implemented in the TEQ code developed at LLNL. The PF coils are arranged in an up-down symmetric configuration, external to the TF coils. The TPX diverted plasma will have an aspect ratio of 4.5 and is highly shaped with a nominal elongation of 2 and triangularity of approximately 0.8 as measured at the separatrix. The tokamak design is based on a high-current (q/sub /spl Psi//=3) plasma scenario and a low current scenario. Each scenario has an operational flexibility requirement which is defined as a region of plasma pressure and inductivity (/spl beta//sub N/-l/sub i/) space, where the plasma shape is constrained to keep the divertor configuration operational. Single-null plasma configurations are feasible, even with the same divertor hardware, by operating the PF coils asymmetrically. Recently applied optimization techniques have improved the capability of the PF system without additional cost.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"134265715","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}
Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518414
H. Miura, S. Nishio, T. Suzuki
One of the most critical issues for fusion in-vessel components is to reduce the transient electromagnetic force due to plasma disruption and to establish suitable support structures. It is indicated that the component deflection is damped by the secondary induced eddy current to reduce the primary electromagnetic force (coupling effects of electromagnetic induction and mechanical deflection). In order to perform the structural design of components, we must fully understand the transient electromagnetic phenomena. For the better fusion reactor design, we've developed a computer code for 3-D thin shell structure with non-ferrous and elastic conductors. Here, we employed a finite element numerical model for mechanical deformation and a wire-grid numerical model for eddy currents. In order to verify the computer code and to understand the influence of support condition in consideration of above coupling effects, we performed some experimental and numerical studies. The experimental results agree well with the numerical results, and we could grasp the influence of support condition for the vibration phenomena in the magnetic field.
{"title":"Experimental and analytical study of the electromagnetomechanics in fusion reactors","authors":"H. Miura, S. Nishio, T. Suzuki","doi":"10.1109/FUSION.1993.518414","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518414","url":null,"abstract":"One of the most critical issues for fusion in-vessel components is to reduce the transient electromagnetic force due to plasma disruption and to establish suitable support structures. It is indicated that the component deflection is damped by the secondary induced eddy current to reduce the primary electromagnetic force (coupling effects of electromagnetic induction and mechanical deflection). In order to perform the structural design of components, we must fully understand the transient electromagnetic phenomena. For the better fusion reactor design, we've developed a computer code for 3-D thin shell structure with non-ferrous and elastic conductors. Here, we employed a finite element numerical model for mechanical deformation and a wire-grid numerical model for eddy currents. In order to verify the computer code and to understand the influence of support condition in consideration of above coupling effects, we performed some experimental and numerical studies. The experimental results agree well with the numerical results, and we could grasp the influence of support condition for the vibration phenomena in the magnetic field.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"134273848","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}
Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518369
R. Hong
DIII-D neutral beam ion sources are operated in the emission limited arc discharge regime. Filament temperature has been observed to play an important role in the ion source operation. Effects of the filament temperature on the arc discharge efficiency (Langmuir probe signal of the arc discharge/arc power) and arc efficiency (extracted beam current/arc power) have been experimentally measured. The results show that both efficiencies reach optimum values with respect to filament temperature. The optimum arc efficiency is independent of the beam energy, but the filament temperature at which optimum arc efficiency is obtained shifts to higher values with higher beam energy. Gas puff into the neutralizer is necessary to achieve optimum neutralization efficiency, but it also increases both particle collisions in the accelerator column and gas flow into the arc chamber. Tests show that arc efficiency decreases only slightly with increasing particle collisions, since loss in the extracted beam current is compensated for by the decreasing we power required to sustain the arc discharge with additional gas flow. The most significant effect of particle collisions is the sharp increase of the gradient grid current, which can cause damage to the grid.
{"title":"Efficiency of arc discharge and beam extraction of the DIII-D neutral beam ion source","authors":"R. Hong","doi":"10.1109/FUSION.1993.518369","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518369","url":null,"abstract":"DIII-D neutral beam ion sources are operated in the emission limited arc discharge regime. Filament temperature has been observed to play an important role in the ion source operation. Effects of the filament temperature on the arc discharge efficiency (Langmuir probe signal of the arc discharge/arc power) and arc efficiency (extracted beam current/arc power) have been experimentally measured. The results show that both efficiencies reach optimum values with respect to filament temperature. The optimum arc efficiency is independent of the beam energy, but the filament temperature at which optimum arc efficiency is obtained shifts to higher values with higher beam energy. Gas puff into the neutralizer is necessary to achieve optimum neutralization efficiency, but it also increases both particle collisions in the accelerator column and gas flow into the arc chamber. Tests show that arc efficiency decreases only slightly with increasing particle collisions, since loss in the extracted beam current is compensated for by the decreasing we power required to sustain the arc discharge with additional gas flow. The most significant effect of particle collisions is the sharp increase of the gradient grid current, which can cause damage to the grid.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"133277515","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}
Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518301
T. Gibney, D. Greenwood
Recent years have see an upsurge in inter-laboratory collaboration within the fusion community. Also, increasingly powerful and inexpensive workstations, linked together by Local Area Networks and through the Internet, have become available for data analysis. It has therefore become essential to provide a mechanism for getting data from a central facility computer to application programs running on a diverse assortment of workstations and larger computers, both locally and at remote sites. The Data Server (DS) protocol defines a client/server communication standard to help satisfy this need. The protocol, a joint development project between PPPL and ORNL, was designed specifically to handle the type of experimental data (separate shot numbers, diagnostic specific) generated by the fusion energy community. It provides a flexible, machine-independent standard for network access to remotely stored data. Subroutines within the local data-access library, while retaining the same outward appearance, are modified to include client code for network data access. Through normal library calls, a user application transparently connects with an experiment-specific Server via TCP/IP over a Local Area Network or through the Internet. The server program presents its local data to the network client following protocol standards. By simple, well-defined request/reply transactions, the client (library routine) gets the information necessary for its normal operation. The network communication is entirely hidden from the user.
{"title":"DS data server: sharing fusion-energy data and results","authors":"T. Gibney, D. Greenwood","doi":"10.1109/FUSION.1993.518301","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518301","url":null,"abstract":"Recent years have see an upsurge in inter-laboratory collaboration within the fusion community. Also, increasingly powerful and inexpensive workstations, linked together by Local Area Networks and through the Internet, have become available for data analysis. It has therefore become essential to provide a mechanism for getting data from a central facility computer to application programs running on a diverse assortment of workstations and larger computers, both locally and at remote sites. The Data Server (DS) protocol defines a client/server communication standard to help satisfy this need. The protocol, a joint development project between PPPL and ORNL, was designed specifically to handle the type of experimental data (separate shot numbers, diagnostic specific) generated by the fusion energy community. It provides a flexible, machine-independent standard for network access to remotely stored data. Subroutines within the local data-access library, while retaining the same outward appearance, are modified to include client code for network data access. Through normal library calls, a user application transparently connects with an experiment-specific Server via TCP/IP over a Local Area Network or through the Internet. The server program presents its local data to the network client following protocol standards. By simple, well-defined request/reply transactions, the client (library routine) gets the information necessary for its normal operation. The network communication is entirely hidden from the user.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"131689098","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}
Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518347
S. K. Ingram
In the existing budgetary environment, low-cost, high-payoff fusion experiments are increasingly attractive. IGNITEX, a homopolar generator powered single-turn tokamak, has the potential to achieve ignition at relatively low cost, A 60 MJ homopolar generator facility (with an easy upgrade path to 90 MJ) is in place at CEM-UT. Using this facility, CEM-UT researchers have produced a 20 T on-axis magnetic field in a 1/16 scale IGNITEX prototype. This paper presents the results of optimization studies conducted to determine the lowest-cost homopolar generator facility which can power an IGNITEX prototype of any chosen scale operating at 20 T on-axis.
{"title":"Optimization studies in IGNITEX","authors":"S. K. Ingram","doi":"10.1109/FUSION.1993.518347","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518347","url":null,"abstract":"In the existing budgetary environment, low-cost, high-payoff fusion experiments are increasingly attractive. IGNITEX, a homopolar generator powered single-turn tokamak, has the potential to achieve ignition at relatively low cost, A 60 MJ homopolar generator facility (with an easy upgrade path to 90 MJ) is in place at CEM-UT. Using this facility, CEM-UT researchers have produced a 20 T on-axis magnetic field in a 1/16 scale IGNITEX prototype. This paper presents the results of optimization studies conducted to determine the lowest-cost homopolar generator facility which can power an IGNITEX prototype of any chosen scale operating at 20 T on-axis.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"132093398","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}
Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518395
J. J. Harris, G. Campbell
A centralized graphical user interface has been added to the DIII-D neutral beam (NB) control systems for status monitoring and remote control applications. This user interface provides for automatic data acquisition, alarm detection and supervisory control of the four NB programmable logic controllers (PLC) as well as the mode control PLC. These PLCs are used for interlocking, control and status of the NB vacuum pumping, gas delivery, and water cooling systems as well as beam mode status and control. The system allows for both a friendly user interface as well as a safe and convenient method of communicating with remote hardware that formerly required interruption of machine operations to access. In the future, to enable high level of control of PLC subsystems, complete procedures can be written and executed at the touch of a screen control panel button. The system consists of an IBM compatible 486 computer running the FIX DMACS for Windows data acquisition and control operator interface software, a Texas Instruments/Siemens communication card and Phoenix Digital optical communications modules. Communication is achieved via the TIWAY (Texas Instruments) protocol link utilizing both fiber optic communications and a copper local area network (LAN). Hardware and software capabilities will be reviewed. Data and alarm reporting, extended monitoring and control capabilities will also be discussed.
集中式图形用户界面已添加到DIII-D中性波束(NB)控制系统中，用于状态监测和远程控制应用。该用户界面提供了四个NB可编程逻辑控制器(PLC)的自动数据采集、报警检测和监控以及PLC的模式控制。这些plc用于联锁，控制和NB真空泵，气体输送和水冷却系统的状态，以及光束模式的状态和控制。该系统既允许一个友好的用户界面，也允许一种安全方便的方法与以前需要中断机器操作才能访问的远程硬件进行通信。将来，为了实现PLC子系统的高水平控制，可以通过触摸屏幕控制面板按钮来编写和执行完整的程序。该系统由一台运行FIX DMACS for Windows数据采集和控制操作界面软件的IBM兼容486计算机、一张德州仪器/西门子通信卡和Phoenix数字光通信模块组成。通信通过TIWAY(德州仪器)协议链路实现，利用光纤通信和铜局域网(LAN)。将审查硬件和软件能力。数据和报警报告，扩展的监测和控制能力也将被讨论。
{"title":"DIII-D neutral beam control system operator interface","authors":"J. J. Harris, G. Campbell","doi":"10.1109/FUSION.1993.518395","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518395","url":null,"abstract":"A centralized graphical user interface has been added to the DIII-D neutral beam (NB) control systems for status monitoring and remote control applications. This user interface provides for automatic data acquisition, alarm detection and supervisory control of the four NB programmable logic controllers (PLC) as well as the mode control PLC. These PLCs are used for interlocking, control and status of the NB vacuum pumping, gas delivery, and water cooling systems as well as beam mode status and control. The system allows for both a friendly user interface as well as a safe and convenient method of communicating with remote hardware that formerly required interruption of machine operations to access. In the future, to enable high level of control of PLC subsystems, complete procedures can be written and executed at the touch of a screen control panel button. The system consists of an IBM compatible 486 computer running the FIX DMACS for Windows data acquisition and control operator interface software, a Texas Instruments/Siemens communication card and Phoenix Digital optical communications modules. Communication is achieved via the TIWAY (Texas Instruments) protocol link utilizing both fiber optic communications and a copper local area network (LAN). Hardware and software capabilities will be reviewed. Data and alarm reporting, extended monitoring and control capabilities will also be discussed.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"132180640","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}
Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518383
Y. Feldshteyn, S. Dinkevich, Ș. Stoenescu, J. Warren
The Tokamak Physics Experiment (TPX) machine is a compact steady state fusion device. The TPX vacuum vessel provides the vacuum boundary for the plasma and mechanical support for internal components. The vacuum vessel itself and internal components are subjected to large heat loads during bakeout and plasma operation. A dual function heating/cooling system has been designed to satisfy thermal requirements for the vacuum vessel and in-vessel components. A successful Conceptual Design Review (CDR) was held in March 1993. Important parameters presented at the CDR are described herein.
{"title":"TPX vacuum vessel heating and cooling system","authors":"Y. Feldshteyn, S. Dinkevich, Ș. Stoenescu, J. Warren","doi":"10.1109/FUSION.1993.518383","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518383","url":null,"abstract":"The Tokamak Physics Experiment (TPX) machine is a compact steady state fusion device. The TPX vacuum vessel provides the vacuum boundary for the plasma and mechanical support for internal components. The vacuum vessel itself and internal components are subjected to large heat loads during bakeout and plasma operation. A dual function heating/cooling system has been designed to satisfy thermal requirements for the vacuum vessel and in-vessel components. A successful Conceptual Design Review (CDR) was held in March 1993. Important parameters presented at the CDR are described herein.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"133897536","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}
Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518520
W. Beck
While bringing Alcator C-MOD on line, failure of a solder joint caused an open circuit in one of the PF coils located within the toroidal field magnet. A design review was conducted to analyze the failure and propose possible solutions. The exact reason for the failure was not determined, but the joint may have been weakened by high temperatures during bakeout of the vacuum vessel. Peeling forces also may have been induced by unforeseen temperature gradients and/or magnetic loads. Significant design changes, which are limited to highly stressed PF coils located within the toroidal field magnet, involved repositioning the joint away from the coaxial termination and eliminating the use of solder as a structural element. PF coils external to the toroidal field magnet are not so highly stressed and brazing is acceptable. The redesign easily accommodates repositioning the joint, but finding a substitute for solder, which was originally selected to avoid annealing the cold worked copper conductor, proved difficult. Localized annealing which occurs in welding and brazing processes eliminated the two most common methods of terminating copper coils. There is not enough space available in the vacuum vessel coil pockets to accommodate mechanical clamping devices. The use of fasteners such as screws and rivets was prohibited due to adverse effects on fatigue life. Electroforming, a process by which complex parts are formed by electroplating materials such as copper onto an electrically conductive mandrel, was selected to replace soldering the joint. Electroformed copper sheet exhibited superior material properties to those of the C-10700 coil conductor, which has yield strength of 290 MPa. Changes, development of an electroformed electromechanical joint, and coil manufacturing will be further described.
{"title":"Repair of poloidal field magnets on Alcator C-Mod","authors":"W. Beck","doi":"10.1109/FUSION.1993.518520","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518520","url":null,"abstract":"While bringing Alcator C-MOD on line, failure of a solder joint caused an open circuit in one of the PF coils located within the toroidal field magnet. A design review was conducted to analyze the failure and propose possible solutions. The exact reason for the failure was not determined, but the joint may have been weakened by high temperatures during bakeout of the vacuum vessel. Peeling forces also may have been induced by unforeseen temperature gradients and/or magnetic loads. Significant design changes, which are limited to highly stressed PF coils located within the toroidal field magnet, involved repositioning the joint away from the coaxial termination and eliminating the use of solder as a structural element. PF coils external to the toroidal field magnet are not so highly stressed and brazing is acceptable. The redesign easily accommodates repositioning the joint, but finding a substitute for solder, which was originally selected to avoid annealing the cold worked copper conductor, proved difficult. Localized annealing which occurs in welding and brazing processes eliminated the two most common methods of terminating copper coils. There is not enough space available in the vacuum vessel coil pockets to accommodate mechanical clamping devices. The use of fasteners such as screws and rivets was prohibited due to adverse effects on fatigue life. Electroforming, a process by which complex parts are formed by electroplating materials such as copper onto an electrically conductive mandrel, was selected to replace soldering the joint. Electroformed copper sheet exhibited superior material properties to those of the C-10700 coil conductor, which has yield strength of 290 MPa. Changes, development of an electroformed electromechanical joint, and coil manufacturing will be further described.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"114015859","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}
Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518413
D. Humphreys, M. Firestone, J. Morrow-Jones
The capabilities of the new DIII-D digital control system have motivated an effort to apply state-of-the-art multivariable techniques to control of the DIII-D tokamak. Tokamak plasma control is inherently multivariable in nature, since many closely coupled equilibrium parameters must be regulated simultaneously during a discharge. The present work describes the determination of dynamic models for plasma response and plasma interaction with conducting structures, necessary for calculation of accurate and robust multivariable control laws. Plasma response matrices and shape prediction matrices are calculated from analytic models and perturbed ideal MHD equilibria. Plasma resistive effects are described by a circuit equation which conserves poloidal flux on time scales shorter than the plasma L/R time. Shape estimation and plasma/conductor eigenmode spectrum results are presented along with experimental data and time-dependent simulations.
{"title":"Plasma response modeling for multivariable tokamak control design","authors":"D. Humphreys, M. Firestone, J. Morrow-Jones","doi":"10.1109/FUSION.1993.518413","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518413","url":null,"abstract":"The capabilities of the new DIII-D digital control system have motivated an effort to apply state-of-the-art multivariable techniques to control of the DIII-D tokamak. Tokamak plasma control is inherently multivariable in nature, since many closely coupled equilibrium parameters must be regulated simultaneously during a discharge. The present work describes the determination of dynamic models for plasma response and plasma interaction with conducting structures, necessary for calculation of accurate and robust multivariable control laws. Plasma response matrices and shape prediction matrices are calculated from analytic models and perturbed ideal MHD equilibria. Plasma resistive effects are described by a circuit equation which conserves poloidal flux on time scales shorter than the plasma L/R time. Shape estimation and plasma/conductor eigenmode spectrum results are presented along with experimental data and time-dependent simulations.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124001182","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}
Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518358
S. Fairfax, D. Montgomery
The Alcator C-MOD tokamak began operation in late March, 1992. The terminal of one of the poloidal field magnets, located inside the TF magnet, suffered a structural failure in routine service on April 10, 1992. The protective systems functioned as designed and there was virtually no associated damage. Post mortem analysis showed that the terminal might well have failed at or near the design current limit, but a satisfactory explanation for the failure at 17% of design stress has not been produced. The terminal details of the failed magnet were present in 9 other PF magnets. A new design was developed and applied to all affected magnets. The replacement of the magnet terminals and restart of the experimental facility took 13 months. The terminal assembly that failed was on a relatively simple magnet with only moderately high applied loads. The toroidal field magnet and ohmic heating solenoid are both much more complex and subjected to higher loads. The TF magnet, for example, utilizes an innovative sliding joint construction method to carry loads to a massive external superstructure. Yet the TF magnet continues to perform well and is now operating routinely at over 50% of design current. This presentation will examine the factors and decisions that led to the original PF terminal design and subsequent failure. The new design will be presented, followed by a discussion of the lessons learned from this experience.
{"title":"Anatomy of the PF magnet failure in Alcator C-MOD","authors":"S. Fairfax, D. Montgomery","doi":"10.1109/FUSION.1993.518358","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518358","url":null,"abstract":"The Alcator C-MOD tokamak began operation in late March, 1992. The terminal of one of the poloidal field magnets, located inside the TF magnet, suffered a structural failure in routine service on April 10, 1992. The protective systems functioned as designed and there was virtually no associated damage. Post mortem analysis showed that the terminal might well have failed at or near the design current limit, but a satisfactory explanation for the failure at 17% of design stress has not been produced. The terminal details of the failed magnet were present in 9 other PF magnets. A new design was developed and applied to all affected magnets. The replacement of the magnet terminals and restart of the experimental facility took 13 months. The terminal assembly that failed was on a relatively simple magnet with only moderately high applied loads. The toroidal field magnet and ohmic heating solenoid are both much more complex and subjected to higher loads. The TF magnet, for example, utilizes an innovative sliding joint construction method to carry loads to a massive external superstructure. Yet the TF magnet continues to perform well and is now operating routinely at over 50% of design current. This presentation will examine the factors and decisions that led to the original PF terminal design and subsequent failure. The new design will be presented, followed by a discussion of the lessons learned from this experience.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124409753","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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