Pub Date : 1991-09-30DOI: 10.1109/FUSION.1991.218867
D. Terry, F. Santucci, H. Huang, Xl Wang
The Texas Experimental Tokamak (TEXT) is being upgraded to include an inner poloidal divertor coil set and a new vacuum vessel. A 400 V DC, 25000 A pulsed-power supply is required for the divertor coil set and a similarly rated supply is required for the vertical field coil set to provide in-out positioning of the plasma during divertor experiments. Two-quadrant 12-pulse convertors with bypass switches have been designed and built to utilize surplus pulse-rated rectifier transformers provided by Oak Ridge National Laboratory. A description of these supplies is given.<>
{"title":"Divertor and vertical field power supplies for the TEXT upgrade","authors":"D. Terry, F. Santucci, H. Huang, Xl Wang","doi":"10.1109/FUSION.1991.218867","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218867","url":null,"abstract":"The Texas Experimental Tokamak (TEXT) is being upgraded to include an inner poloidal divertor coil set and a new vacuum vessel. A 400 V DC, 25000 A pulsed-power supply is required for the divertor coil set and a similarly rated supply is required for the vertical field coil set to provide in-out positioning of the plasma during divertor experiments. Two-quadrant 12-pulse convertors with bypass switches have been designed and built to utilize surplus pulse-rated rectifier transformers provided by Oak Ridge National Laboratory. A description of these supplies is given.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"18 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1991-09-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"125448116","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 : 1991-09-30DOI: 10.1109/FUSION.1991.218686
T. Simonen
Recent DIII-D tokamak experimental results are summarized, new hardware being implemented to carry out the DIII-D 1990s tokamak research program is described, and their implications for engineering designs for next-generation tokamaks, such as ITER (International Thermonuclear Experimental Reactor), are discussed. DIII-D is presently investigating several new and alternative methods to address key ITER design issues: divertor, disruption, and current profile control. DIII-D is also demonstrating improved tokamak performance regimes such as high beta second stability and high confinement such as VH-mode. On a more basic and broader research front, DIII-D is exploring innovative tokamak engineering and physics concepts, solidifying fundamental plasma physics understanding, and demonstrating essential engineering technologies with the goal of developing a safe and environmentally and commercially attractive fusion reactor concept.<>
{"title":"Recent results from DIII-D and future plans","authors":"T. Simonen","doi":"10.1109/FUSION.1991.218686","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218686","url":null,"abstract":"Recent DIII-D tokamak experimental results are summarized, new hardware being implemented to carry out the DIII-D 1990s tokamak research program is described, and their implications for engineering designs for next-generation tokamaks, such as ITER (International Thermonuclear Experimental Reactor), are discussed. DIII-D is presently investigating several new and alternative methods to address key ITER design issues: divertor, disruption, and current profile control. DIII-D is also demonstrating improved tokamak performance regimes such as high beta second stability and high confinement such as VH-mode. On a more basic and broader research front, DIII-D is exploring innovative tokamak engineering and physics concepts, solidifying fundamental plasma physics understanding, and demonstrating essential engineering technologies with the goal of developing a safe and environmentally and commercially attractive fusion reactor concept.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"72 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1991-09-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"115958690","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 : 1991-09-30DOI: 10.1109/FUSION.1991.218684
D. Lousteau, F. Davis, B. Nelson
During the conceptual design activity (CDA) for the International Thermonuclear Experimental Reactor (ITER), a reactor design was established that emphasized performance of the individual systems in a minimum overall reactor size. The resulting high component density arrangement dictates including assembly and maintenance (A&M) considerations in the development of the configuration. The A&M task is complicated further since remote handling equipment will be required after the start of deuterium-tritium operations. During the CDA, the Assembly and Maintenance Design Unit addressed many aspects of an overall A&M system. The authors discuss the ITER A&M philosophy that evolved, describe the ITER configuration as it relates to maintenance, and describe the procedures and equipment required for specific maintenance operations. Change-out of the in-vessel divertors and blanket/shields modules is discussed in detail.<>
{"title":"ITER remote maintenance","authors":"D. Lousteau, F. Davis, B. Nelson","doi":"10.1109/FUSION.1991.218684","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218684","url":null,"abstract":"During the conceptual design activity (CDA) for the International Thermonuclear Experimental Reactor (ITER), a reactor design was established that emphasized performance of the individual systems in a minimum overall reactor size. The resulting high component density arrangement dictates including assembly and maintenance (A&M) considerations in the development of the configuration. The A&M task is complicated further since remote handling equipment will be required after the start of deuterium-tritium operations. During the CDA, the Assembly and Maintenance Design Unit addressed many aspects of an overall A&M system. The authors discuss the ITER A&M philosophy that evolved, describe the ITER configuration as it relates to maintenance, and describe the procedures and equipment required for specific maintenance operations. Change-out of the in-vessel divertors and blanket/shields modules is discussed in detail.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"23 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1991-09-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"115961214","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 : 1991-09-30DOI: 10.1109/FUSION.1991.218765
L. Liu
The author presents the conceptual design of the plasma vertical instability control system in the EHR (Experimental Hybrid Reactor). It has been demonstrated that the plasma-vacuum vessel-shell-active coil control system can stabilize the EHR plasma displacement in the vertical direction successfully. The estimated active control system power supply capacity is >
{"title":"Control of the EHR plasma vertical instabilities","authors":"L. Liu","doi":"10.1109/FUSION.1991.218765","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218765","url":null,"abstract":"The author presents the conceptual design of the plasma vertical instability control system in the EHR (Experimental Hybrid Reactor). It has been demonstrated that the plasma-vacuum vessel-shell-active coil control system can stabilize the EHR plasma displacement in the vertical direction successfully. The estimated active control system power supply capacity is <or=1 MW. The plasma position is regulated for approximately 150 ms.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1991-09-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"130230264","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 : 1991-09-30DOI: 10.1109/FUSION.1991.218680
D. Kuban, N. Busko
The goal of the TeleMate manipulator system is to provide a single system that works effectively in all three of the major maintenance areas of the BPX (Burning Plasma Experiment): in-vessel, ex-vessel, and under-vessel. The system, the development status, initial test data, and control system performance are described. TeleMate combines enhancements to a proven mechanical design with state-of-the-art control technology to produce a flexible system that can be configured to address the numerous remote fusion applications. The mechanical portion of the system has many years of operation in radioactive facilities. The subject development includes upgrading the mechanical system, correcting deficiencies identified by users, and incorporating a custom distributed digital controller. A commercial high-speed transformation processor will mate master controller to slave arm while maintaining the force reflecting properties which are fundamental to efficient remote operations. This transformer and a innovative high-speed communications board, Telelink, will accommodate dissimilar master-slave combinations and provide a centralized control platform for multiple arms and support system configurations. The operator interface utilizes a proprietary visual programming language capable of controlling multiple systems from a single terminal. Advanced control features will also be implemented.<>
{"title":"A remote in-vessel and ex-vessel force-reflecting telerobotic system for the Burning Plasma Experiment","authors":"D. Kuban, N. Busko","doi":"10.1109/FUSION.1991.218680","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218680","url":null,"abstract":"The goal of the TeleMate manipulator system is to provide a single system that works effectively in all three of the major maintenance areas of the BPX (Burning Plasma Experiment): in-vessel, ex-vessel, and under-vessel. The system, the development status, initial test data, and control system performance are described. TeleMate combines enhancements to a proven mechanical design with state-of-the-art control technology to produce a flexible system that can be configured to address the numerous remote fusion applications. The mechanical portion of the system has many years of operation in radioactive facilities. The subject development includes upgrading the mechanical system, correcting deficiencies identified by users, and incorporating a custom distributed digital controller. A commercial high-speed transformation processor will mate master controller to slave arm while maintaining the force reflecting properties which are fundamental to efficient remote operations. This transformer and a innovative high-speed communications board, Telelink, will accommodate dissimilar master-slave combinations and provide a centralized control platform for multiple arms and support system configurations. The operator interface utilizes a proprietary visual programming language capable of controlling multiple systems from a single terminal. Advanced control features will also be implemented.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"63 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1991-09-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"134327986","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 : 1991-09-30DOI: 10.1109/FUSION.1991.218820
R. Wells, T. Stevens, C. V. van Os
The details of a prototype surface conversion ion source, designed for continuous operation are described. The source consists of an annular, convectively water-cooled, magnetic multi-cusp driver chamber separated by a magnetic field from an inner chamber. A barium-coated spherically contoured disc-shaped converter in the inner chamber produces a stream of H/sup -/ ions focused at the exit aperture. The converter is electrically biased up to 300 V negative with respect to the source potential. Jacketed, water-cooled permanent magnets near the source exit deflect electrons out of the beam path. Techniques for fabrication and maintenance of a clean barium metal surface are also discussed.<>
{"title":"Mechanical design and fabrication of a barium surface conversion H/sup -/ ion source","authors":"R. Wells, T. Stevens, C. V. van Os","doi":"10.1109/FUSION.1991.218820","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218820","url":null,"abstract":"The details of a prototype surface conversion ion source, designed for continuous operation are described. The source consists of an annular, convectively water-cooled, magnetic multi-cusp driver chamber separated by a magnetic field from an inner chamber. A barium-coated spherically contoured disc-shaped converter in the inner chamber produces a stream of H/sup -/ ions focused at the exit aperture. The converter is electrically biased up to 300 V negative with respect to the source potential. Jacketed, water-cooled permanent magnets near the source exit deflect electrons out of the beam path. Techniques for fabrication and maintenance of a clean barium metal surface are also discussed.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"47 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1991-09-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"134380543","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 : 1991-09-30DOI: 10.1109/FUSION.1991.218741
S. Saka, M. Kanno, S. Sudo, T. Baba
Hydrogen pellets without sabots were accelerated to high speeds using a two-stage light gas gun for Heliotron-E. The primary application of this technology is plasma fueling of fusion devices. Conventional pellet injectors have limited pellet speeds in the range of 1-2 km/s. Higher velocities are desirable for more flexible density profile control and for deep penetration into a high-temperature plasma. By developing a new fast valve with high conductance to accelerate the piston and other components, a 2-mm-diam hydrogen pellet with a velocity of 3.2 km/s has been successfully accelerated without a sabot. The pipe gun technique for freezing hydrogen operation is simulated with the code MYKE. Development of new pistons instead of the plastic piston for the repetitive two-stage pellet injector is being carried out using metal and ceramic, as the surface of the plastic piston tends to wear out through repetitive operation and carbon powder produced from the plastic piston may cause trouble with the fusion device.<>
{"title":"Development of hydrogen pellet acceleration system using a two-stage light gas gun","authors":"S. Saka, M. Kanno, S. Sudo, T. Baba","doi":"10.1109/FUSION.1991.218741","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218741","url":null,"abstract":"Hydrogen pellets without sabots were accelerated to high speeds using a two-stage light gas gun for Heliotron-E. The primary application of this technology is plasma fueling of fusion devices. Conventional pellet injectors have limited pellet speeds in the range of 1-2 km/s. Higher velocities are desirable for more flexible density profile control and for deep penetration into a high-temperature plasma. By developing a new fast valve with high conductance to accelerate the piston and other components, a 2-mm-diam hydrogen pellet with a velocity of 3.2 km/s has been successfully accelerated without a sabot. The pipe gun technique for freezing hydrogen operation is simulated with the code MYKE. Development of new pistons instead of the plastic piston for the repetitive two-stage pellet injector is being carried out using metal and ceramic, as the surface of the plastic piston tends to wear out through repetitive operation and carbon powder produced from the plastic piston may cause trouble with the fusion device.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1991-09-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129377159","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 : 1991-09-30DOI: 10.1109/FUSION.1991.218830
J. Citrolo
The Burning Plasma Experiment (BPX) toroidal field (TF) coil design is based on a wedged inner leg support concept with the rear leg of each coil contained in a stainless-steel case. The cases are welded together to form a continuous structure that provides support for both in-plane and overturning forces. The 18 coils and cases are grouped into three coil modules to facilitate remote replacement in the event of a coil failure. The coils are a modified Bitter design with radially oriented flat plate conductors of high-performance, beryllium-copper. The coils are precooled with liquid nitrogen. The inner leg temperature rises to room temperature at the end of the pulse and approximately 1 h is required to cool down the coil after a full power pulse. An important consideration in the design is that the structure should be robust and use conventional materials. In addition to describing the structural design, the author discusses the applied loads, the design approach, and the fabrication sequence.<>
{"title":"Description of the BPX toroidal field coil support structure","authors":"J. Citrolo","doi":"10.1109/FUSION.1991.218830","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218830","url":null,"abstract":"The Burning Plasma Experiment (BPX) toroidal field (TF) coil design is based on a wedged inner leg support concept with the rear leg of each coil contained in a stainless-steel case. The cases are welded together to form a continuous structure that provides support for both in-plane and overturning forces. The 18 coils and cases are grouped into three coil modules to facilitate remote replacement in the event of a coil failure. The coils are a modified Bitter design with radially oriented flat plate conductors of high-performance, beryllium-copper. The coils are precooled with liquid nitrogen. The inner leg temperature rises to room temperature at the end of the pulse and approximately 1 h is required to cool down the coil after a full power pulse. An important consideration in the design is that the structure should be robust and use conventional materials. In addition to describing the structural design, the author discusses the applied loads, the design approach, and the fabrication sequence.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"23 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1991-09-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"130802485","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 : 1991-09-30DOI: 10.1109/FUSION.1991.218848
G. Cambi, G. Cavallone, A. Boschi, S. Sarto
The authors describe a safety study performed on some design solutions for the divertor of ITER (International Thermonuclear Experimental Reactor). The event tree technique is used to delineate the accident scenarios and to outline reference (or critical) accident sequences. Common features of the solutions considered are double null magnetic configuration, with upper and lower divertor plates, water as primary coolant, separate and independent cooling loops for the divertor and the first wall, carbon-based protective tiles for the physics phase, and tungsten tiles for the technology phase. Fault trees are used to evaluate the occurrence rate of the initiating events and the unavailability of the different event tree headings (i.e., the protective systems, both passive and active). Deterministic safety studies have been considered, where applicable, to describe the phenomenology of the accident sequences and then to estimate the environmental radioactive releases. Consequences are evaluated in terms of doses to the maximum exposed individual of the public for instantaneous releases.<>
{"title":"Risk analysis for ITER divertor system","authors":"G. Cambi, G. Cavallone, A. Boschi, S. Sarto","doi":"10.1109/FUSION.1991.218848","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218848","url":null,"abstract":"The authors describe a safety study performed on some design solutions for the divertor of ITER (International Thermonuclear Experimental Reactor). The event tree technique is used to delineate the accident scenarios and to outline reference (or critical) accident sequences. Common features of the solutions considered are double null magnetic configuration, with upper and lower divertor plates, water as primary coolant, separate and independent cooling loops for the divertor and the first wall, carbon-based protective tiles for the physics phase, and tungsten tiles for the technology phase. Fault trees are used to evaluate the occurrence rate of the initiating events and the unavailability of the different event tree headings (i.e., the protective systems, both passive and active). Deterministic safety studies have been considered, where applicable, to describe the phenomenology of the accident sequences and then to estimate the environmental radioactive releases. Consequences are evaluated in terms of doses to the maximum exposed individual of the public for instantaneous releases.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1991-09-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"130836156","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 : 1991-09-30DOI: 10.1109/FUSION.1991.218788
G.D. Loesser, D. Owens, G. Barnes
To optimize power loading on the first wall components in the TFTR (Tokamak Fusion Test Reactor), parts must be carefully aligned with the toroidal magnetic field (TF) surfaces. First, the location of a template with six fixed positions was established with respect to the TF using measurements of magnetic field strength at several positions on the template. Next, a measuring arm was installed at each of these locations and manually positioned in order to resolve relative and absolute coordinates of the in-vessel components with respect to the TF. The measuring arm is a flexible linkage consisting of six links and five joints instrumented with absolute optical encoders. Prior to use in the vacuum vessel, the arm was calibrated to determine the best fit for the 23 unknowns involving link lengths, encoder zero positions, and shaft-to-link angles. The position of the indicator point with respect to the measuring arm base is known to +or-0.75 mm after this calibration process. During operation, the shaft encoder positions are processed by an algorithm to determine the relative and global coordinates of the indicator point for real-time use and stored for later detailed analysis. Data obtained with this device are shown.<>
{"title":"Five degree of freedom measuring arm for resolving spatial relationships within TFTR vacuum vessel","authors":"G.D. Loesser, D. Owens, G. Barnes","doi":"10.1109/FUSION.1991.218788","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218788","url":null,"abstract":"To optimize power loading on the first wall components in the TFTR (Tokamak Fusion Test Reactor), parts must be carefully aligned with the toroidal magnetic field (TF) surfaces. First, the location of a template with six fixed positions was established with respect to the TF using measurements of magnetic field strength at several positions on the template. Next, a measuring arm was installed at each of these locations and manually positioned in order to resolve relative and absolute coordinates of the in-vessel components with respect to the TF. The measuring arm is a flexible linkage consisting of six links and five joints instrumented with absolute optical encoders. Prior to use in the vacuum vessel, the arm was calibrated to determine the best fit for the 23 unknowns involving link lengths, encoder zero positions, and shaft-to-link angles. The position of the indicator point with respect to the measuring arm base is known to +or-0.75 mm after this calibration process. During operation, the shaft encoder positions are processed by an algorithm to determine the relative and global coordinates of the indicator point for real-time use and stored for later detailed analysis. Data obtained with this device are shown.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1991-09-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"130963224","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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