Pub Date : 2002-11-07DOI: 10.1109/FUSION.2002.1027683
D. Mastrovito, W. Blanchard, J. Dong, R. Gernhardt, H. Kugel, G. Oliaro, T. Provost
The National Spherical Torus Experiment (NSTX) is the first step in an investigation of the physics principles of low-aspect-ratio spherical tori (ST) designed to study non-inductive start-up, current sustainability, current profile control, confinement, transport, pressure limits, stability and disruption resilience, as well as unique scrape-off layer (SOL) and divertor characteristics. NSTX started plasma operations in February 1999. During the first extended period of experiments, plasma discharge reproducibility and performance were strongly affected by impurity control and wall conditions. During this time, residual gas analyzer (RGA) data has been used during non-operating periods and between discharges to measure contributions to the vessel base-pressure (2-3E/sup -8/ Torr) that resulted from impurity gases evolving from internal surfaces. The RGA is a quadrupole mass spectrometer (QMS) with a Faraday Cup and electron multiplier. Recently, a second RGA system was installed to monitor fast changes in impurity gas production during and immediately after plasma discharges, which are indicative of changing conditions. In order to achieve the required response time, scan rates were increased by limiting the range of masses scanned. A data acquisition board was used to drive the new RGA system in concurrence with the NSTX shot cycle. Specialized interactive software to put acquired data into MDSplus, control data acquisition and assist in data analysis was written in Visual Basic and IDL.
{"title":"Residual gas analysis hardware and software data acquisition system at NSTX","authors":"D. Mastrovito, W. Blanchard, J. Dong, R. Gernhardt, H. Kugel, G. Oliaro, T. Provost","doi":"10.1109/FUSION.2002.1027683","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027683","url":null,"abstract":"The National Spherical Torus Experiment (NSTX) is the first step in an investigation of the physics principles of low-aspect-ratio spherical tori (ST) designed to study non-inductive start-up, current sustainability, current profile control, confinement, transport, pressure limits, stability and disruption resilience, as well as unique scrape-off layer (SOL) and divertor characteristics. NSTX started plasma operations in February 1999. During the first extended period of experiments, plasma discharge reproducibility and performance were strongly affected by impurity control and wall conditions. During this time, residual gas analyzer (RGA) data has been used during non-operating periods and between discharges to measure contributions to the vessel base-pressure (2-3E/sup -8/ Torr) that resulted from impurity gases evolving from internal surfaces. The RGA is a quadrupole mass spectrometer (QMS) with a Faraday Cup and electron multiplier. Recently, a second RGA system was installed to monitor fast changes in impurity gas production during and immediately after plasma discharges, which are indicative of changing conditions. In order to achieve the required response time, scan rates were increased by limiting the range of masses scanned. A data acquisition board was used to drive the new RGA system in concurrence with the NSTX shot cycle. Specialized interactive software to put acquired data into MDSplus, control data acquisition and assist in data analysis was written in Visual Basic and IDL.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":"37 1","pages":"234-237"},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"91022141","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"艺术学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2002-11-07DOI: 10.1109/FUSION.2002.1027724
G. Voss, E. Ciattaglia
The Mega Amp Spherical Tokamak (MAST) is an experimental fusion device currently operating at the Culham Science Centre. The low aspect ratio of the spherical tokamak combined with the need to develop a high flux swing from its central solenoid leads to high forces on the conductors that form the centre column. This column consists of a compact solenoid wound around the inner limbs of the toroidal field coils. In order to achieve the nominal plasma parameters, the dimensions of the centre column were optimised to make best possible use of the space available. The most critical region of the solenoid magnet design is the end turn and tail section where complex 3D magnetic fields interact with the current in the end turn conductors giving an asymmetric stress distribution. The hoop load in this end turn is particularly troublesome since it must be reacted across a bonded joint, which inevitably generates tensile and shear stresses within it. This paper describes how this problem has been solved for the new MAST centre column by the use of compact quadrupole tails formed close to the main coil and connected to it by tight bends formed in the water cooled copper conductor. A test coil employing these features and fitted with engineering diagnostics has been manufactured and tested at Culham to the maximum design current. This coil not only demonstrated the manufacturing processes involved but also provided test data, which compares well with engineering analysis.
Mega Amp球形托卡马克(MAST)是目前在Culham科学中心运行的实验性聚变装置。球形托卡马克的低长径比，加上需要从其中心螺线管发展高通量摆动，导致形成中心柱的导体受到高力。该柱由一个紧凑的螺线管组成，绕在环形场线圈的内肢上。为了达到标称等离子体参数，中心柱的尺寸进行了优化，以最大限度地利用可用空间。螺线管磁体设计中最关键的区域是端匝和尾部部分，复杂的三维磁场与端匝导体中的电流相互作用，产生不对称的应力分布。在这个末端转弯的环向载荷特别麻烦，因为它必须在一个粘合的关节上反应，这不可避免地在它内部产生拉伸和剪切应力。本文介绍了在新型桅杆中心柱上如何解决这一问题，即采用紧靠主线圈的紧凑型四极杆尾部，并通过水冷铜导体内形成的紧弯与主线圈相连。采用这些特性并配备工程诊断的测试线圈已在Culham制造并测试到最大设计电流。该线圈不仅演示了所涉及的制造过程，还提供了测试数据，与工程分析相比效果很好。
{"title":"Development of a high field solenoid magnet for the MAST Spherical Tokamak","authors":"G. Voss, E. Ciattaglia","doi":"10.1109/FUSION.2002.1027724","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027724","url":null,"abstract":"The Mega Amp Spherical Tokamak (MAST) is an experimental fusion device currently operating at the Culham Science Centre. The low aspect ratio of the spherical tokamak combined with the need to develop a high flux swing from its central solenoid leads to high forces on the conductors that form the centre column. This column consists of a compact solenoid wound around the inner limbs of the toroidal field coils. In order to achieve the nominal plasma parameters, the dimensions of the centre column were optimised to make best possible use of the space available. The most critical region of the solenoid magnet design is the end turn and tail section where complex 3D magnetic fields interact with the current in the end turn conductors giving an asymmetric stress distribution. The hoop load in this end turn is particularly troublesome since it must be reacted across a bonded joint, which inevitably generates tensile and shear stresses within it. This paper describes how this problem has been solved for the new MAST centre column by the use of compact quadrupole tails formed close to the main coil and connected to it by tight bends formed in the water cooled copper conductor. A test coil employing these features and fitted with engineering diagnostics has been manufactured and tested at Culham to the maximum design current. This coil not only demonstrated the manufacturing processes involved but also provided test data, which compares well with engineering analysis.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":"8 1","pages":"409-412"},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87648703","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"艺术学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2002-11-07DOI: 10.1109/FUSION.2002.1027687
B. Nelson, R. Benson, L. Berry, A. Brooks, M. Cole, P.J. Fogrty, P. Goranson, P. Heitzenroeder, S. Hirschman, G. Jones, J. Lyon, P. Mioduszewski, D. Monticello, D. Spong, D. Strickler, A. Ware, D. Williamson
The engineering design status of the Quasi-Poloidal Stellarator Experiment (QPS) is presented. The purpose, configuration, and possible manufacturing and assembly techniques of the various components of the core are described.
{"title":"Design of the Quasi-Poloidal Stellarator Experiment (QPS)","authors":"B. Nelson, R. Benson, L. Berry, A. Brooks, M. Cole, P.J. Fogrty, P. Goranson, P. Heitzenroeder, S. Hirschman, G. Jones, J. Lyon, P. Mioduszewski, D. Monticello, D. Spong, D. Strickler, A. Ware, D. Williamson","doi":"10.1109/FUSION.2002.1027687","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027687","url":null,"abstract":"The engineering design status of the Quasi-Poloidal Stellarator Experiment (QPS) is presented. The purpose, configuration, and possible manufacturing and assembly techniques of the various components of the core are described.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":"104 1","pages":"248-251"},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75979199","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"艺术学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2002-11-07DOI: 10.1109/FUSION.2002.1027735
Jiancheng Yan, C. Zhou, Yong Liu, Dequan Liu
This paper describes the outline of HL-2A project that is the first tokamak with a divertor in China. The recent achievements in the construction of the HL-2A tokamak is given. The development of major components and the installation of the machine are also described. The commissioning of HL-2A machine will start from the middle of 2002. The HL-2A experimental program and the future plan are sketched.
{"title":"Status and plan of the HL-2A project","authors":"Jiancheng Yan, C. Zhou, Yong Liu, Dequan Liu","doi":"10.1109/FUSION.2002.1027735","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027735","url":null,"abstract":"This paper describes the outline of HL-2A project that is the first tokamak with a divertor in China. The recent achievements in the construction of the HL-2A tokamak is given. The development of major components and the installation of the machine are also described. The commissioning of HL-2A machine will start from the middle of 2002. The HL-2A experimental program and the future plan are sketched.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":"1 1","pages":"461-464"},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77500906","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"艺术学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
An online trouble reporting system (TRS) has been introduced at NSTX. The TRS is used by NSTX operators to report problems that affect NSTX operations. The purpose of the TRS is to enhance NSTX reliability and maintainability by identifying components, occurrences, and trends that contribute to machine downtime. All NSTX personnel have access to the TRS. The user interface is via a Web browser, such as Netscape or Internet Explorer. This Web-based feature permits any X-terminal, PC, or MAC access to the TRS. The TRS is based upon a trouble reporting system developed at the DIII-D Tokamak, at General Atomics Technologies (Petersen and Miller, 1991). This paper provides a detailed description of the TRS software architecture, user interface, MS SQL server interface and operational experiences. In addition, sample data from the TRS database are summarized and presented.
{"title":"The NSTX trouble reporting system","authors":"S. Sengupta, G. Oliaro","doi":"10.2172/795714","DOIUrl":"https://doi.org/10.2172/795714","url":null,"abstract":"An online trouble reporting system (TRS) has been introduced at NSTX. The TRS is used by NSTX operators to report problems that affect NSTX operations. The purpose of the TRS is to enhance NSTX reliability and maintainability by identifying components, occurrences, and trends that contribute to machine downtime. All NSTX personnel have access to the TRS. The user interface is via a Web browser, such as Netscape or Internet Explorer. This Web-based feature permits any X-terminal, PC, or MAC access to the TRS. The TRS is based upon a trouble reporting system developed at the DIII-D Tokamak, at General Atomics Technologies (Petersen and Miller, 1991). This paper provides a detailed description of the TRS software architecture, user interface, MS SQL server interface and operational experiences. In addition, sample data from the TRS database are summarized and presented.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":"13 1","pages":"242-244"},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83808766","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"艺术学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2002-11-07DOI: 10.1109/FUSION.2002.1027705
P. Titus, J. Zaks, M. DeMaria, B. LaBombard, R. Granetz, J. Irby, B. Lipshultz, E. Fitzgerald, R. Childs, W. Beck, E. Marmar, D. Gwinn, I. Hutchinson, R. Boivin, W. Burke
In 1993 there were indications that the C-Mod inner divertor was being over-loaded. As a result, an inner divertor modification is being installed in C-Mod. The new design is intended to allow a change in profile of the divertor to accommodate a higher plasma triangularity, strengthen the inner vessel wall to survive 9.0 Tesla, 2.5 MA disruptions and, eliminate tile rotation. The thermal differentials between tiles and vessel wall, necessitated the introduction of compliance in the reinforcement hardware. Practical assembly within the vessel required manageable part size,and bolted interfaces which further reduced the strength and stiffness. A modest improvement of 22% was achieved while meeting the many design constraints.
{"title":"Alcator C-Mod inner divertor upgrade design and analysis","authors":"P. Titus, J. Zaks, M. DeMaria, B. LaBombard, R. Granetz, J. Irby, B. Lipshultz, E. Fitzgerald, R. Childs, W. Beck, E. Marmar, D. Gwinn, I. Hutchinson, R. Boivin, W. Burke","doi":"10.1109/FUSION.2002.1027705","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027705","url":null,"abstract":"In 1993 there were indications that the C-Mod inner divertor was being over-loaded. As a result, an inner divertor modification is being installed in C-Mod. The new design is intended to allow a change in profile of the divertor to accommodate a higher plasma triangularity, strengthen the inner vessel wall to survive 9.0 Tesla, 2.5 MA disruptions and, eliminate tile rotation. The thermal differentials between tiles and vessel wall, necessitated the introduction of compliance in the reinforcement hardware. Practical assembly within the vessel required manageable part size,and bolted interfaces which further reduced the strength and stiffness. A modest improvement of 22% was achieved while meeting the many design constraints.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":"2 1","pages":"333-336"},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86026951","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"艺术学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2002-11-07DOI: 10.1109/FUSION.2002.1027721
N. Her, S. Cho, J. Sa, K. Im, G. Hong, G.H. Kim, J.Y. Park, H. Kim, B.C. Kim, I. Yu, D.L. Kim, W. Kim, Y. Oh, C. Choi, J. Bak, M. Kwon, G.S. Lee, J.H. Kim, H. Ahn
The KSTAR cryostat is a 8.8 m diameter vacuum vessel that provides the necessary thermal barrier between the ambient temperature test cell and the supercritical helium cooled superconducting magnet providing the base pressure of 1/spl times/10/sup -5/ torr. The cryostat is a single walled vessel consisting of central cylindrical section and two end closures, a flat base structure with external reinforcements and a dome-shaped lid structure. The base structure has 8 equally spaced support legs anchored on the concrete base. The cryostat vessel design was executed to satisfy the performance and operation requirements. The mechanical penetration components with bellows were designed to restrict the displacements of all kinds of ports due to EM loads and thermal loads within the allowable limits. The major loads considered in this paper for the design of cryostat vessel are the vacuum pressure, the dead weight of vacuum vessel, PFC, and magnet which are total about 400 tons, the electromagnetic load driven by plasma disruption, and seismic loads. Based on these loads, structural analyses were performed. It was found that the maximum stress intensity was below the allowable limit, and that the cryostat vessel had buckling safety of over 5. Based on the results, structural robustness of the cryostat vessel has been proved.
{"title":"Structural design and analysis for the KSTAR cryostat","authors":"N. Her, S. Cho, J. Sa, K. Im, G. Hong, G.H. Kim, J.Y. Park, H. Kim, B.C. Kim, I. Yu, D.L. Kim, W. Kim, Y. Oh, C. Choi, J. Bak, M. Kwon, G.S. Lee, J.H. Kim, H. Ahn","doi":"10.1109/FUSION.2002.1027721","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027721","url":null,"abstract":"The KSTAR cryostat is a 8.8 m diameter vacuum vessel that provides the necessary thermal barrier between the ambient temperature test cell and the supercritical helium cooled superconducting magnet providing the base pressure of 1/spl times/10/sup -5/ torr. The cryostat is a single walled vessel consisting of central cylindrical section and two end closures, a flat base structure with external reinforcements and a dome-shaped lid structure. The base structure has 8 equally spaced support legs anchored on the concrete base. The cryostat vessel design was executed to satisfy the performance and operation requirements. The mechanical penetration components with bellows were designed to restrict the displacements of all kinds of ports due to EM loads and thermal loads within the allowable limits. The major loads considered in this paper for the design of cryostat vessel are the vacuum pressure, the dead weight of vacuum vessel, PFC, and magnet which are total about 400 tons, the electromagnetic load driven by plasma disruption, and seismic loads. Based on these loads, structural analyses were performed. It was found that the maximum stress intensity was below the allowable limit, and that the cryostat vessel had buckling safety of over 5. Based on the results, structural robustness of the cryostat vessel has been proved.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":"1 1","pages":"396-399"},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82914727","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"艺术学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2002-11-07DOI: 10.1109/FUSION.2002.1027631
M. Grimes, D. Gwinn, R. Parker, D. Terry, J. Alex
Alcator C-Mod is a high-field, high-density, diverted, compact tokamak, which, in its present form uses inductive current drive and is heated with 5 MW of ICRF auxiliary power. C-Mod is in the process of being upgraded with a 4.6 GHz Lower Hybrid heating and current drive system. The purpose of the experiment is to develop and explore the potential of "Advanced Tokamak Regimes", i.e., regimes with high bootstrap fraction (/spl sim/70%), high /spl beta//sub n/ (/spl sim/3) and high confinement (H/sub H//spl sim/1-2) under quasi-steady-state conditions. In this paper, an overview of the RF transmitter, high-voltage power supply and controls and protection systems for the Lower Hybrid Project is given. The transmitter will use twelve 250 kW klystrons operating simultaneously which will result in a total directed power of nearly 3 MW for a planned pulse width of 5 seconds. An expected upgrade of four additional klystrons will result in a total directed power of 4 MW. All klystrons will be powered in parallel by a single solid-state pulse-step-modulated (PSM) power supply with a rating of 50 kV and 208 amperes. Commissioning of the power supply is expected in February of 2002 with initial transmitter operation in late 2002.
{"title":"The Alcator C-Mod lower hybrid current drive experiment transmitter and power system","authors":"M. Grimes, D. Gwinn, R. Parker, D. Terry, J. Alex","doi":"10.1109/FUSION.2002.1027631","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027631","url":null,"abstract":"Alcator C-Mod is a high-field, high-density, diverted, compact tokamak, which, in its present form uses inductive current drive and is heated with 5 MW of ICRF auxiliary power. C-Mod is in the process of being upgraded with a 4.6 GHz Lower Hybrid heating and current drive system. The purpose of the experiment is to develop and explore the potential of \"Advanced Tokamak Regimes\", i.e., regimes with high bootstrap fraction (/spl sim/70%), high /spl beta//sub n/ (/spl sim/3) and high confinement (H/sub H//spl sim/1-2) under quasi-steady-state conditions. In this paper, an overview of the RF transmitter, high-voltage power supply and controls and protection systems for the Lower Hybrid Project is given. The transmitter will use twelve 250 kW klystrons operating simultaneously which will result in a total directed power of nearly 3 MW for a planned pulse width of 5 seconds. An expected upgrade of four additional klystrons will result in a total directed power of 4 MW. All klystrons will be powered in parallel by a single solid-state pulse-step-modulated (PSM) power supply with a rating of 50 kV and 208 amperes. Commissioning of the power supply is expected in February of 2002 with initial transmitter operation in late 2002.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":"57 1","pages":"16-19"},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89110997","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"艺术学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2002-11-07DOI: 10.1109/FUSION.2002.1027632
G. Loesser, J. Rushinski, S. Bernabei, J. Hosea, J. Wilson, B. Beck, R. Parker
The PSFC group of MIT and PPPL are jointly developing a Lower Hybrid Current Drive (LHCD) system for installation on the Alcator C-Mod tokamak, with the primary goal of driving plasma current. Twelve Klystrons will feed a coupler with an input power of 3 MW at 4.6 GHz and a pulse length of 5 seconds. The coupler is composed of 4 arrays, each with 24 wave-guides (96 total) which are stacked poloidally and are symmetric about both the vertical and horizontal planes. The four-stacked arrays are installed through a single equatorial port and can be axially adjusted to provide optimum plasma coupling. Material selections and spatial limitations required innovative design solutions that will be described.
{"title":"Design and engineering of the Alcator C-Mod lower hybrid current drive system","authors":"G. Loesser, J. Rushinski, S. Bernabei, J. Hosea, J. Wilson, B. Beck, R. Parker","doi":"10.1109/FUSION.2002.1027632","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027632","url":null,"abstract":"The PSFC group of MIT and PPPL are jointly developing a Lower Hybrid Current Drive (LHCD) system for installation on the Alcator C-Mod tokamak, with the primary goal of driving plasma current. Twelve Klystrons will feed a coupler with an input power of 3 MW at 4.6 GHz and a pulse length of 5 seconds. The coupler is composed of 4 arrays, each with 24 wave-guides (96 total) which are stacked poloidally and are symmetric about both the vertical and horizontal planes. The four-stacked arrays are installed through a single equatorial port and can be axially adjusted to provide optimum plasma coupling. Material selections and spatial limitations required innovative design solutions that will be described.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":"8 1","pages":"20-22"},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80693699","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"艺术学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2002-11-07DOI: 10.1109/FUSION.2002.1027630
F. Najmabadi
During the past ten years, the ARIES Team has studied a variety of tokamak power plants with different degrees of extrapolation in plasma physics and technology from present database. Continuation of research has allowed us to apply lessons learned from each ARIES design to the next. The results of ARIES tokamak power plant studies provide a large body of data that highlight the tradeoffs and relative leverage of advanced plasma physics and fusion technology directions. Our results indicate that for the same plasma physics (e.g., first-stability) and technology extrapolation, steady state operation is more attractive than pulsed-plasma operation. Dramatic improvement over first-stability operation can be obtained through either utilization of high-field magnets (e.g., high-temperature superconductors) or operation in advanced-tokamak modes (e.g., reversed-shear). In particular, if full benefits of reversed-shear operation are realized, as is assumed in ARIES-AT, tokamak power plants will have a cost of electricity competitive with other sources of electricity. In technology area, emerging technologies such as advanced Baryon cycle, high-temperature superconductor, and advanced manufacturing techniques can improve the cost and attractiveness of fusion plants. For blankets, liquid breeder/coolants are the most attractive because most of neutron power is directly deposited in the coolant. This property can be exploited to arrive at a blanket design with a coolant outlet temperature higher than the structure temperature in the radiation zone. The high coolant temperature leads to a high thermal conversion efficiency (as in ARIES-ST and ARIES-AT blankets). The dual-cooled (He and LiPb) ARIES-ST blanket using ferritic steel structural material represents a near-term option for fusion systems and achieves a thermal efficiency of 45%. Development of high-performance SIC composites leads to the high-performance ARIES-AT blanket (SiC composite/LiPb coolant) that achieves 59% thermal conversion efficiency as well as the full potential safety and environmental features of fusion power.
{"title":"Impact of advanced physics and technology on the attractiveness of tokamak fusion power plants","authors":"F. Najmabadi","doi":"10.1109/FUSION.2002.1027630","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027630","url":null,"abstract":"During the past ten years, the ARIES Team has studied a variety of tokamak power plants with different degrees of extrapolation in plasma physics and technology from present database. Continuation of research has allowed us to apply lessons learned from each ARIES design to the next. The results of ARIES tokamak power plant studies provide a large body of data that highlight the tradeoffs and relative leverage of advanced plasma physics and fusion technology directions. Our results indicate that for the same plasma physics (e.g., first-stability) and technology extrapolation, steady state operation is more attractive than pulsed-plasma operation. Dramatic improvement over first-stability operation can be obtained through either utilization of high-field magnets (e.g., high-temperature superconductors) or operation in advanced-tokamak modes (e.g., reversed-shear). In particular, if full benefits of reversed-shear operation are realized, as is assumed in ARIES-AT, tokamak power plants will have a cost of electricity competitive with other sources of electricity. In technology area, emerging technologies such as advanced Baryon cycle, high-temperature superconductor, and advanced manufacturing techniques can improve the cost and attractiveness of fusion plants. For blankets, liquid breeder/coolants are the most attractive because most of neutron power is directly deposited in the coolant. This property can be exploited to arrive at a blanket design with a coolant outlet temperature higher than the structure temperature in the radiation zone. The high coolant temperature leads to a high thermal conversion efficiency (as in ARIES-ST and ARIES-AT blankets). The dual-cooled (He and LiPb) ARIES-ST blanket using ferritic steel structural material represents a near-term option for fusion systems and achieves a thermal efficiency of 45%. Development of high-performance SIC composites leads to the high-performance ARIES-AT blanket (SiC composite/LiPb coolant) that achieves 59% thermal conversion efficiency as well as the full potential safety and environmental features of fusion power.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":"201 1","pages":"10-15"},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80140424","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"艺术学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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