Pub Date : 2002-11-07DOI: 10.1109/FUSION.2002.1027726
D. Williamson, A. Brooks, M. Cole, H. Fan, P. Fogarty, B. Nelson, D. Strickler, W. Reiersen
The National Compact Stellarator Experiment (NCSX) is proposed as a test of a low aspect ratio, quasi-axisymmetric plasma configuration that exhibits high beta and good confinement in a disruption-free environment. The experiment will be built at Princeton Plasma Physics Laboratory (PPPL) and utilize some ancillary equipment. The NCSX stellarator core is a complex assembly of four coil systems which provide the magnetic field for plasma shaping and position control, inductive current drive, and field error correction. The primary magnets are the modular coils, which provide up to 2-T at an average major radius of 1.4-m. Other magnets include toroidal field (TF) coils, poloidal field (PF) coils, and trim coils, which can be used to control resonant field errors. The magnets are supported by an integral shell structure, which also serves as the winding form for the modular coils. The coils and structure have been evaluated for thermal stress and electromagnetic loads during normal operating conditions. The results indicate that the performance of the modular coil system is acceptable.
{"title":"Design and analysis of the modular coils for the National Compact Stellarator Experiment (NCSX)","authors":"D. Williamson, A. Brooks, M. Cole, H. Fan, P. Fogarty, B. Nelson, D. Strickler, W. Reiersen","doi":"10.1109/FUSION.2002.1027726","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027726","url":null,"abstract":"The National Compact Stellarator Experiment (NCSX) is proposed as a test of a low aspect ratio, quasi-axisymmetric plasma configuration that exhibits high beta and good confinement in a disruption-free environment. The experiment will be built at Princeton Plasma Physics Laboratory (PPPL) and utilize some ancillary equipment. The NCSX stellarator core is a complex assembly of four coil systems which provide the magnetic field for plasma shaping and position control, inductive current drive, and field error correction. The primary magnets are the modular coils, which provide up to 2-T at an average major radius of 1.4-m. Other magnets include toroidal field (TF) coils, poloidal field (PF) coils, and trim coils, which can be used to control resonant field errors. The magnets are supported by an integral shell structure, which also serves as the winding form for the modular coils. The coils and structure have been evaluated for thermal stress and electromagnetic loads during normal operating conditions. The results indicate that the performance of the modular coil system is acceptable.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":"1 1","pages":"418-421"},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89489866","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.2002.1027646
S. Combs, C. Foust, L. Baylor, M. Cole, D. Fehling, P. W. Fisher, M. Gouge, R.A. Rasmussen, D. O. Sparks, J. Wilgen
A compact pellet injection system has been designed and constructed at the Oak Ridge National Laboratory to provide a flexible fueling system for studies in magnetic confinement fusion devices. The system, referred to as a "pellet injector in a suitcase," is a pipe gun device with a four-barrel capability (1- to 4-mm bore), and it uses a cryogenic refrigerator for in-situ hydrogen pellet formation. The injector design allows for simple change-out of barrel sizes and different pellet acceleration options, including mechanical punches and/or propellant valves to provide speeds in the range of /spl sim/100 to 1500 m/s. The stand-alone instrumentation and controls, as well as the data acquisition system, are personal-computer-based and housed in one standard instrument cabinet. The portable system has been developed to provide a flexible, low-cost fueling system that can be used on a number of plasma confinement experiments with minimal installation and operation costs. The prototype will be installed on the Madison Symmetric Torus (MST) at the University of Wisconsin. For the MST application, the pellet sizes will be in the range of. 1 to 1.8 mm.
{"title":"Compact flexible pellet injection system for plasma fueling experiments","authors":"S. Combs, C. Foust, L. Baylor, M. Cole, D. Fehling, P. W. Fisher, M. Gouge, R.A. Rasmussen, D. O. Sparks, J. Wilgen","doi":"10.1109/FUSION.2002.1027646","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027646","url":null,"abstract":"A compact pellet injection system has been designed and constructed at the Oak Ridge National Laboratory to provide a flexible fueling system for studies in magnetic confinement fusion devices. The system, referred to as a \"pellet injector in a suitcase,\" is a pipe gun device with a four-barrel capability (1- to 4-mm bore), and it uses a cryogenic refrigerator for in-situ hydrogen pellet formation. The injector design allows for simple change-out of barrel sizes and different pellet acceleration options, including mechanical punches and/or propellant valves to provide speeds in the range of /spl sim/100 to 1500 m/s. The stand-alone instrumentation and controls, as well as the data acquisition system, are personal-computer-based and housed in one standard instrument cabinet. The portable system has been developed to provide a flexible, low-cost fueling system that can be used on a number of plasma confinement experiments with minimal installation and operation costs. The prototype will be installed on the Madison Symmetric Torus (MST) at the University of Wisconsin. For the MST application, the pellet sizes will be in the range of. 1 to 1.8 mm.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":"22 1","pages":"76-79"},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87236104","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.2002.1027710
J.E. Pulsifer, A. Raffray
The use of structured porous media is a proposed technique to achieve higher heat transfer coefficients by increasing the specific surface area for heat transfer while aiming to maintain acceptable pressure drop and pumping power. The general design strategy is to minimize the coolant flow path through the porous medium while optimizing the porous medium characteristics to minimize the friction pressure drop for a given heat transfer performance. A comprehensive thermo-fluid model called MERLOT was used to assess the use of porous heat transfer media for fusion plasma facing component applications. A parametric study was performed to assess the relative importance on the heat transfer performance of key design parameters including the solid conductivity, the porosity magnitude and distribution, the microstructure characteristic dimension, and the local heat transfer coefficient. The analysis was carried out for different incident heat fluxes of up to 30 MW/m/sup 2/ with the goal of identifying particularly attractive sets of design parameters for plasma facing components.
{"title":"Structured porous media for high heat flux fusion applications","authors":"J.E. Pulsifer, A. Raffray","doi":"10.1109/FUSION.2002.1027710","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027710","url":null,"abstract":"The use of structured porous media is a proposed technique to achieve higher heat transfer coefficients by increasing the specific surface area for heat transfer while aiming to maintain acceptable pressure drop and pumping power. The general design strategy is to minimize the coolant flow path through the porous medium while optimizing the porous medium characteristics to minimize the friction pressure drop for a given heat transfer performance. A comprehensive thermo-fluid model called MERLOT was used to assess the use of porous heat transfer media for fusion plasma facing component applications. A parametric study was performed to assess the relative importance on the heat transfer performance of key design parameters including the solid conductivity, the porosity magnitude and distribution, the microstructure characteristic dimension, and the local heat transfer coefficient. The analysis was carried out for different incident heat fluxes of up to 30 MW/m/sup 2/ with the goal of identifying particularly attractive sets of design parameters for plasma facing components.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":"25 1","pages":"352-355"},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81415809","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.2002.1027636
K. Doan, J. Busath, D. Kellman
The DIII-D Neutral Beam Supervisory Control and Data Acquisition (NB SCADA) system is responsible for data and status communication between remote system devices. Some years ago, it was operated and controlled on a 486 PC with Microsoft Windows 3.1. A 16-bit software package called FIXDMACS was used to interface and communicate with Siemens programmable logic controllers (PLCs). Due to the ever-changing operation requirements, this system became antiquated and failed to adequately support new process conditions and meet the NB operational demands. It was, therefore, inevitable that a system upgrade would be needed to satisfy efficiency and performance. This project required a comprehensive survey of available hardware and software currently offered by the leading industries. The best solution was a complete replacement of the entire workstation. A new Dell Pentium III PC, equipped with Windows NT, was acquired to replace the old SCADA system. FIXDMACS was replaced by scalable iFIX, which was also developed by Intellution. In addition, data migration and conversion was performed to enable forward compatibility of all existing software and system configurations. Besides delivering an excellent solution to monitoring the neutral beam system operations, iFIX is able to accept Microsoft Visual Basic scripts and programs to automate routine or repetitive tasks, allowing system administrators to execute these tasks and controls quickly. This added feature provides flexibility and simplicity for maintaining and troubleshooting purposes. Although additional improvements are always possible as with all other software products, iFIX has proven to be a valuable tool in supporting the operations. Packaged with essential enhancements, new capabilities and powerful tools, Intellution has developed an application that certainly surpasses its predecessor. Today, the upgraded DIII-D Neutral Beam SCADA system is fully operational. Both hardware and software upgrades were a cost effective and necessary approach toward achieving the goals of maximizing system performance, improving efficiency and reliability, and providing better control of the neutral beam operational processes.
{"title":"The DIII-D Neutral Beam Supervisory Control and Data Acquisition workstation upgrade","authors":"K. Doan, J. Busath, D. Kellman","doi":"10.1109/FUSION.2002.1027636","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027636","url":null,"abstract":"The DIII-D Neutral Beam Supervisory Control and Data Acquisition (NB SCADA) system is responsible for data and status communication between remote system devices. Some years ago, it was operated and controlled on a 486 PC with Microsoft Windows 3.1. A 16-bit software package called FIXDMACS was used to interface and communicate with Siemens programmable logic controllers (PLCs). Due to the ever-changing operation requirements, this system became antiquated and failed to adequately support new process conditions and meet the NB operational demands. It was, therefore, inevitable that a system upgrade would be needed to satisfy efficiency and performance. This project required a comprehensive survey of available hardware and software currently offered by the leading industries. The best solution was a complete replacement of the entire workstation. A new Dell Pentium III PC, equipped with Windows NT, was acquired to replace the old SCADA system. FIXDMACS was replaced by scalable iFIX, which was also developed by Intellution. In addition, data migration and conversion was performed to enable forward compatibility of all existing software and system configurations. Besides delivering an excellent solution to monitoring the neutral beam system operations, iFIX is able to accept Microsoft Visual Basic scripts and programs to automate routine or repetitive tasks, allowing system administrators to execute these tasks and controls quickly. This added feature provides flexibility and simplicity for maintaining and troubleshooting purposes. Although additional improvements are always possible as with all other software products, iFIX has proven to be a valuable tool in supporting the operations. Packaged with essential enhancements, new capabilities and powerful tools, Intellution has developed an application that certainly surpasses its predecessor. Today, the upgraded DIII-D Neutral Beam SCADA system is fully operational. Both hardware and software upgrades were a cost effective and necessary approach toward achieving the goals of maximizing system performance, improving efficiency and reliability, and providing better control of the neutral beam operational processes.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":"36 1","pages":"36-39"},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83674087","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.2002.1027727
P. Chaudhuri, D. C. Reddy, S. Khirwadkar, N. R. Prakash, P. Santra, Y. Saxena
The plasma facing components (PFCs) are an important part of Steady State Superconducting Tokamak (SST-1) design. PFC of SST-1 consists of divertors, passive stabilizers, baffles and limiters are designed to be compatible for steady state operation. The main consideration in the design of the PFC is the steady state heat removal of up to 1 MW/m/sup 2/. During steady-state operation, the average heat loads on divertor and passive stabilizers are expected to be 0.6 and 0.25 MW/m/sup 2/ respectively. The PFC has been design to withstand the peak heat fluxes and also without significant erosion such that frequent replacement is not necessary. Design considerations included 2-D steady state and transient tile temperature distribution and resulting thermal loads in PFC during cooling. Coolant parameters necessary to maintain optimum thermal-hydraulic design, tile surface temperature, and tile fitting mechanism. A 2-D finite difference code has been developed to study of flow behavior and thermal response of PFC during cooling.
{"title":"Study and thermal-hydraulic design of water cooled PFC for SST-1 tokamak","authors":"P. Chaudhuri, D. C. Reddy, S. Khirwadkar, N. R. Prakash, P. Santra, Y. Saxena","doi":"10.1109/FUSION.2002.1027727","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027727","url":null,"abstract":"The plasma facing components (PFCs) are an important part of Steady State Superconducting Tokamak (SST-1) design. PFC of SST-1 consists of divertors, passive stabilizers, baffles and limiters are designed to be compatible for steady state operation. The main consideration in the design of the PFC is the steady state heat removal of up to 1 MW/m/sup 2/. During steady-state operation, the average heat loads on divertor and passive stabilizers are expected to be 0.6 and 0.25 MW/m/sup 2/ respectively. The PFC has been design to withstand the peak heat fluxes and also without significant erosion such that frequent replacement is not necessary. Design considerations included 2-D steady state and transient tile temperature distribution and resulting thermal loads in PFC during cooling. Coolant parameters necessary to maintain optimum thermal-hydraulic design, tile surface temperature, and tile fitting mechanism. A 2-D finite difference code has been developed to study of flow behavior and thermal response of PFC during cooling.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":"40 1","pages":"422-425"},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74938481","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.2002.1027656
P. Karditsas
Analysis of hypothetical loss-of-coolant accidents in fusion power stations, involves establishing low upper limits of temperature excursions in the structure due to decay heat. Previous studies and results have revealed several key areas where it is clear that more realistic modelling of thermal transient events in fusion power plants could remove excessive conservatism in accident analysis and improve results. Modelling of thermal transient events in fusion power plants following accidental loss of cooling has been improved by the use of a commercially available finite element analysis code which allows two and three dimensional geometry treatments as well as various heat loading conditions. The principal plant parameters were a set used in preliminary studies for the European Power Plant Conceptual Study, as follows: fusion power 3000 MW, major plasma radius 7.9 m, aspect ratio 3.0, elongation 1.7, and triangularity 0.3. The blanket is based on the water-cooled lithium lead concept, with low activation martensitic steel as the structural material.
{"title":"Temperature excursions in bounding accidents using finite element analysis for a water-cooled lithium lead blanket power plant","authors":"P. Karditsas","doi":"10.1109/FUSION.2002.1027656","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027656","url":null,"abstract":"Analysis of hypothetical loss-of-coolant accidents in fusion power stations, involves establishing low upper limits of temperature excursions in the structure due to decay heat. Previous studies and results have revealed several key areas where it is clear that more realistic modelling of thermal transient events in fusion power plants could remove excessive conservatism in accident analysis and improve results. Modelling of thermal transient events in fusion power plants following accidental loss of cooling has been improved by the use of a commercially available finite element analysis code which allows two and three dimensional geometry treatments as well as various heat loading conditions. The principal plant parameters were a set used in preliminary studies for the European Power Plant Conceptual Study, as follows: fusion power 3000 MW, major plasma radius 7.9 m, aspect ratio 3.0, elongation 1.7, and triangularity 0.3. The blanket is based on the water-cooled lithium lead concept, with low activation martensitic steel as the structural material.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":"25 1","pages":"114-117"},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74505969","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.2002.1027692
C. Kung, S. Bernabei, J. Gumbas, N. Greenough, E. Fredd
Due to confined space in a stack of reduced height waveguides, power detection of the incident and reflected wave in the reduced height waveguide is extremely difficult. A new compact probe to monitor the incident and reflected wave from the narrow side of the reduced height waveguide has been developed. This compact probe consists of 2 current loops, a directional coupler, and a small trimmer capacitor. The two current loops are placed on the narrow side of the waveguide with a quarter guide wavelength (/spl lambda//sub g//4) spacing. The outputs of the current loops are connected to the directional coupler. In order to optimize phase adjustment, a trimmer capacitor is attached in series with one of the current loops. Test results show that more than 30 dB directivity at 4.6 GHz is achieved by using this compact probe.
{"title":"Compact probe design for power monitoring from the narrow side of the reduced height waveguide","authors":"C. Kung, S. Bernabei, J. Gumbas, N. Greenough, E. Fredd","doi":"10.1109/FUSION.2002.1027692","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027692","url":null,"abstract":"Due to confined space in a stack of reduced height waveguides, power detection of the incident and reflected wave in the reduced height waveguide is extremely difficult. A new compact probe to monitor the incident and reflected wave from the narrow side of the reduced height waveguide has been developed. This compact probe consists of 2 current loops, a directional coupler, and a small trimmer capacitor. The two current loops are placed on the narrow side of the waveguide with a quarter guide wavelength (/spl lambda//sub g//4) spacing. The outputs of the current loops are connected to the directional coupler. In order to optimize phase adjustment, a trimmer capacitor is attached in series with one of the current loops. Test results show that more than 30 dB directivity at 4.6 GHz is achieved by using this compact probe.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":"4 1","pages":"268-271"},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"76260451","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.2002.1027700
J. Casazza
The author discusses his interest in fusion and his belief that one needs to think about US long-term national energy needs and develop the needed technology to supply them. He encourages people to understand the relationship between US electric energy policies and fusion research and emphasises that policies based on "profits now" harm the US. Topics covered are education needs, restructuring, lack of government competence, fusion research and national power and transmission surveys.
{"title":"Profits now versus long range needs [fusion research]","authors":"J. Casazza","doi":"10.1109/FUSION.2002.1027700","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027700","url":null,"abstract":"The author discusses his interest in fusion and his belief that one needs to think about US long-term national energy needs and develop the needed technology to supply them. He encourages people to understand the relationship between US electric energy policies and fusion research and emphasises that policies based on \"profits now\" harm the US. Topics covered are education needs, restructuring, lack of government competence, fusion research and national power and transmission surveys.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":"206 ","pages":"305-309"},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://sci-hub-pdf.com/10.1109/FUSION.2002.1027700","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"72435132","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.2002.1027707
R. Majeski, G. Antar, M. Boaz, D. Buchenauer, L. Cadwallader, R. Causey, R. Conn, R. Doerner, P. Efthimion, M. Finkenthal, D. Hoffman, B. Jones, R. Kaita, H. Kugel, S. Luckhardt, R. Maingi, M. Maiorano, T. Munsat, S. Raftopoulos, T. Rognlein, J. Spaleta, V. Soukhanovskii, D. Stutman, G. Taylor, J. Timberlake, M. Ulrickson, D. Whyte
Attention has focused recently on flowing liquid lithium as a first wall for a reactor because of its potentially attractive physics and engineering features. In order to test the suitability of liquid lithium as a plasma facing component, the Current Drive eXperiment - Upgrade (CDX-U) at the Princeton Plasma Physics Laboratory has recently installed a fully toroidal liquid lithium limiter. CDX-U is a compact (R = 34 cm, a = 22 cm, B/sub toroidal/ = 2 kG, I/sub p/ =100 kA, T/sub e/(O) /spl sim/ 100 eV, n/sub e/(0) /spl sim/ 5 /spl times/ 10/sup 19/ m/sup -3/ short-pulse (< 25 msec) spherical torus (ST) with extensive diagnostics. The limiter, which consists of a shallow circular stainless steel tray of radius 34 cm and width 10 cm, is filled with lithium to a depth of a few millimeters, and forms the lower limiting surface for the discharge. Heating elements beneath the tray are used to liquefy the lithium (melting point = 180.5/spl deg/C) prior to the experiment. The total area of liquid lithium exposed to the plasma is approximately 2000 cm/sup 2/. The design of the limiter, modifications to CDX-U to accommodate in-vessel inventories of approximately 1 liter of liquid lithium, techniques for loading lithium onto the limiter, and other preparations will be described. CDX-U has previously been successfully operated with a smaller area cm/sup 2/) liquid lithium rail limiter. Diagnostics specific to lithium operations include multichord spectrometry of the 135 /spl Aring/ LiIII line in the core plasma, monitors for neutral lithium light at the lithium limiter, and a fast (10,000 frame per second) camera which monitors motion of the liquid during the discharge. First results of plasma operations with the toroidal liquid lithium limiter will also be given.
{"title":"A toroidal liquid lithium limiter for CDX-U","authors":"R. Majeski, G. Antar, M. Boaz, D. Buchenauer, L. Cadwallader, R. Causey, R. Conn, R. Doerner, P. Efthimion, M. Finkenthal, D. Hoffman, B. Jones, R. Kaita, H. Kugel, S. Luckhardt, R. Maingi, M. Maiorano, T. Munsat, S. Raftopoulos, T. Rognlein, J. Spaleta, V. Soukhanovskii, D. Stutman, G. Taylor, J. Timberlake, M. Ulrickson, D. Whyte","doi":"10.1109/FUSION.2002.1027707","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027707","url":null,"abstract":"Attention has focused recently on flowing liquid lithium as a first wall for a reactor because of its potentially attractive physics and engineering features. In order to test the suitability of liquid lithium as a plasma facing component, the Current Drive eXperiment - Upgrade (CDX-U) at the Princeton Plasma Physics Laboratory has recently installed a fully toroidal liquid lithium limiter. CDX-U is a compact (R = 34 cm, a = 22 cm, B/sub toroidal/ = 2 kG, I/sub p/ =100 kA, T/sub e/(O) /spl sim/ 100 eV, n/sub e/(0) /spl sim/ 5 /spl times/ 10/sup 19/ m/sup -3/ short-pulse (< 25 msec) spherical torus (ST) with extensive diagnostics. The limiter, which consists of a shallow circular stainless steel tray of radius 34 cm and width 10 cm, is filled with lithium to a depth of a few millimeters, and forms the lower limiting surface for the discharge. Heating elements beneath the tray are used to liquefy the lithium (melting point = 180.5/spl deg/C) prior to the experiment. The total area of liquid lithium exposed to the plasma is approximately 2000 cm/sup 2/. The design of the limiter, modifications to CDX-U to accommodate in-vessel inventories of approximately 1 liter of liquid lithium, techniques for loading lithium onto the limiter, and other preparations will be described. CDX-U has previously been successfully operated with a smaller area cm/sup 2/) liquid lithium rail limiter. Diagnostics specific to lithium operations include multichord spectrometry of the 135 /spl Aring/ LiIII line in the core plasma, monitors for neutral lithium light at the lithium limiter, and a fast (10,000 frame per second) camera which monitors motion of the liquid during the discharge. First results of plasma operations with the toroidal liquid lithium limiter will also be given.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":"25 1","pages":"341-344"},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80975933","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.2002.1027659
J. Lohr, S. Delaware, R. Callis, W. Cary, J. Deboo, J. Doane, I. Gorelov, R. L. La Haye, H. Grunloh, C. Petty, R. Pinsker, D. Ponce, R. Prater, S. Pronko
Advances in gyrotron technology are resulting in new capabilities and scientific results on magnetic confinement devices for fusion research worldwide. Unit output power of 1 MW and higher, at frequencies greater than 100 GHz and quasi-cw operation have become possible. This has led to successful experiments on electron cyclotron heating, electron cyclotron current drive, non-inductive tokamak operation, tokamak energy transport, suppression of instabilities and advanced profile control leading to enhanced performance. The synthetic diamond gyrotron output window is being developed as the answer to the requirement for a low loss blocking window with excellent thermal and mechanical properties and the potential for cw operation at high power. Ancillary equipment for efficient microwave transmission over distances of hundreds of meters, polarization control, diagnostics and flexible launch geometry have all been developed and proven in regular service. There now is excellent convergence between the experimental measurements and theoretical understanding of the heating and current drive mechanisms. The reliability of high power gyrotron installations is at the level previously achieved by neutral beam systems.
{"title":"ECH comes of age for magnetic fusion research","authors":"J. Lohr, S. Delaware, R. Callis, W. Cary, J. Deboo, J. Doane, I. Gorelov, R. L. La Haye, H. Grunloh, C. Petty, R. Pinsker, D. Ponce, R. Prater, S. Pronko","doi":"10.1109/FUSION.2002.1027659","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027659","url":null,"abstract":"Advances in gyrotron technology are resulting in new capabilities and scientific results on magnetic confinement devices for fusion research worldwide. Unit output power of 1 MW and higher, at frequencies greater than 100 GHz and quasi-cw operation have become possible. This has led to successful experiments on electron cyclotron heating, electron cyclotron current drive, non-inductive tokamak operation, tokamak energy transport, suppression of instabilities and advanced profile control leading to enhanced performance. The synthetic diamond gyrotron output window is being developed as the answer to the requirement for a low loss blocking window with excellent thermal and mechanical properties and the potential for cw operation at high power. Ancillary equipment for efficient microwave transmission over distances of hundreds of meters, polarization control, diagnostics and flexible launch geometry have all been developed and proven in regular service. There now is excellent convergence between the experimental measurements and theoretical understanding of the heating and current drive mechanisms. The reliability of high power gyrotron installations is at the level previously achieved by neutral beam systems.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":"40 1","pages":"126-132"},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83581676","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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