Pub Date : 1991-09-30DOI: 10.1109/FUSION.1991.218810
J. Alonso, M. Blaumoser
The TJ-II is a flexible heliac under construction, to be mounted at the Euratom/Ciemat Association Laboratory in Madrid, Spain. The machine can explore different magnetic configurations (mainly with different values of the rotational transform) by means of the adjustment of the currents in the coils. The hard core (HC) is one of the main components of the device and it constitutes what can be called the most critical part of the machine, since its close proximity to the plasma places on it the requirement of strict tolerances. The authors describe the engineering design features of the HC, the main characteristics, and the design details An overview of the manufacturing methods is also presented.<>
{"title":"Design and feasibility of the TJ-II hard core","authors":"J. Alonso, M. Blaumoser","doi":"10.1109/FUSION.1991.218810","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218810","url":null,"abstract":"The TJ-II is a flexible heliac under construction, to be mounted at the Euratom/Ciemat Association Laboratory in Madrid, Spain. The machine can explore different magnetic configurations (mainly with different values of the rotational transform) by means of the adjustment of the currents in the coils. The hard core (HC) is one of the main components of the device and it constitutes what can be called the most critical part of the machine, since its close proximity to the plasma places on it the requirement of strict tolerances. The authors describe the engineering design features of the HC, the main characteristics, and the design details An overview of the manufacturing methods is also presented.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"25 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":"116562947","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.1991.218916
S. K. Ho
An alternative fueling scheme for ARIES-III using accelerated compact-toroids (CT) is studied. The methodology of L.J. Perkins et al. (1988) for modeling CT penetration and deposition is followed. The calculations are performed using a self-consistent, radial-zoning scheme which includes several interrelated constraints such as CT ring decay, tilting, field-line reconnection, deceleration in the external field gradient, and ring expansion/contraction. The CT injection parameters optimized for fueling of ARIES-III are presented. Advantages of CT fueling with respect to other aspects of tokamak operations and uncertainties in the CT injection modeling are also discussed. From the conservative upper-limit estimation, CTs of 40 mg with 40 cm diameter can deposit fuel directly into the center and half-way of the plasma, requiring 49 MW and 14 MW, respectively.<>
{"title":"Controlled central fueling for ARIES-III by compact-toroid injection","authors":"S. K. Ho","doi":"10.1109/FUSION.1991.218916","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218916","url":null,"abstract":"An alternative fueling scheme for ARIES-III using accelerated compact-toroids (CT) is studied. The methodology of L.J. Perkins et al. (1988) for modeling CT penetration and deposition is followed. The calculations are performed using a self-consistent, radial-zoning scheme which includes several interrelated constraints such as CT ring decay, tilting, field-line reconnection, deceleration in the external field gradient, and ring expansion/contraction. The CT injection parameters optimized for fueling of ARIES-III are presented. Advantages of CT fueling with respect to other aspects of tokamak operations and uncertainties in the CT injection modeling are also discussed. From the conservative upper-limit estimation, CTs of 40 mg with 40 cm diameter can deposit fuel directly into the center and half-way of the plasma, requiring 49 MW and 14 MW, respectively.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"142 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":"116570817","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.1991.218743
Y. Oda, M. Onozuka, S. Tsujimura, S. Kuribayashi, K. Shimizu, A. Sawaoka, H. Tamura
Railgun systems for the application of pellet injectors have been investigated and developed in the experimental stage. One of the main features of these railgun systems is the use of a pulse laser beam to induce the initial plasma armature between rails to be accelerated. This unique feature provides a reduction of the supplied voltage to the breakdown between the rails in order to avoid any unnecessary breakdown between the rails and to reduce the erosion of the rails. The authors present results of experimental and theoretical research and introduce the design study for a repetitive pellet injection systems with an electromagnetic railgun based on the research progress.<>
{"title":"Development and design of railgun system to pellet injector","authors":"Y. Oda, M. Onozuka, S. Tsujimura, S. Kuribayashi, K. Shimizu, A. Sawaoka, H. Tamura","doi":"10.1109/FUSION.1991.218743","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218743","url":null,"abstract":"Railgun systems for the application of pellet injectors have been investigated and developed in the experimental stage. One of the main features of these railgun systems is the use of a pulse laser beam to induce the initial plasma armature between rails to be accelerated. This unique feature provides a reduction of the supplied voltage to the breakdown between the rails in order to avoid any unnecessary breakdown between the rails and to reduce the erosion of the rails. The authors present results of experimental and theoretical research and introduce the design study for a repetitive pellet injection systems with an electromagnetic railgun based on the research progress.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"48 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":"130986812","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.1991.218732
W. W. Mixon, D. Terry, D. Patterson, J. Dibble
The upgrade to the Texas Experimental Tokamak, (TEXT) includes two new motor-generator sets, new vertical-field and divertor power supplies, and a new 500-kW ECH (electron cyclotron heating) gyrotron and its power supplies. Much of the monitoring and control of this hardware are done by a network of UNIX workstations running TACL control software. Additional capabilities needed during actual pulses, such as fast data-logging and the generation of waveforms and fast timing sequences, is provided by CAMAC modules controlled by VAX computers and X Windows terminals. These new systems are integrated with the old TEXT control and data-acquisition systems.<>
{"title":"Overview of TEXT control system upgrade","authors":"W. W. Mixon, D. Terry, D. Patterson, J. Dibble","doi":"10.1109/FUSION.1991.218732","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218732","url":null,"abstract":"The upgrade to the Texas Experimental Tokamak, (TEXT) includes two new motor-generator sets, new vertical-field and divertor power supplies, and a new 500-kW ECH (electron cyclotron heating) gyrotron and its power supplies. Much of the monitoring and control of this hardware are done by a network of UNIX workstations running TACL control software. Additional capabilities needed during actual pulses, such as fast data-logging and the generation of waveforms and fast timing sequences, is provided by CAMAC modules controlled by VAX computers and X Windows terminals. These new systems are integrated with the old TEXT control and data-acquisition systems.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"35 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":"131073239","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.1991.218836
D. Huttar, G. Bronner, N. Fromm
The design of the BPX (Burning Plasma Experiment) power system has evolved over a period of several years and has included studies of several alternative approaches. The reapplication of the existing TFTR (Tokamak Fusion Test Reactor) power and energy facilities has been basic to all approaches. The dynamics of the power requirements for the BPX poloidal coil system suggest that the TFTR facilities would be most suitably applied to that requirement. The chief concern related to that match has been the adequacy of the 4.5-GJ energy rating of the TFTR flywheel units. The toroidal field power requirements are the greatest of the BPX subsystems and, fortunately, are sufficiently free of dynamics to allow the consideration of different approaches to providing pulse power and energy. Additional design challenges were presented by the multiplicity of plasma control scenarios incorporated in the BPX physics planning and the power response demanded of the plasma position control system. The plasma control scenarios include upper, lower, and symmetrical poloidal diverter operation as well as limiter operation. The plasma position control coils (internal to the TF bore) have a collective peak power demand of 640 MVA, require four quadrant drive, and require 1 ms voltage response.<>
{"title":"The BPX electrical power system","authors":"D. Huttar, G. Bronner, N. Fromm","doi":"10.1109/FUSION.1991.218836","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218836","url":null,"abstract":"The design of the BPX (Burning Plasma Experiment) power system has evolved over a period of several years and has included studies of several alternative approaches. The reapplication of the existing TFTR (Tokamak Fusion Test Reactor) power and energy facilities has been basic to all approaches. The dynamics of the power requirements for the BPX poloidal coil system suggest that the TFTR facilities would be most suitably applied to that requirement. The chief concern related to that match has been the adequacy of the 4.5-GJ energy rating of the TFTR flywheel units. The toroidal field power requirements are the greatest of the BPX subsystems and, fortunately, are sufficiently free of dynamics to allow the consideration of different approaches to providing pulse power and energy. Additional design challenges were presented by the multiplicity of plasma control scenarios incorporated in the BPX physics planning and the power response demanded of the plasma position control system. The plasma control scenarios include upper, lower, and symmetrical poloidal diverter operation as well as limiter operation. The plasma position control coils (internal to the TF bore) have a collective peak power demand of 640 MVA, require four quadrant drive, and require 1 ms voltage response.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"58 10","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1991-09-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"132870926","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.1991.218859
S. Mills, A. Loving, M. Irving
The authors report on remote handling tools which have been specifically designed to meet requirements for pipe cutting at JET (Joint European Torus). The principal requirements were the quality of cut necessary for rewelding, effective swarf removal, and compactness for remote handling. The designs of tools had to be compatible with the severe access restrictions imposed by the JET machine. The processes used by the tools are sawing from the inside and outside of pipes, and orbital lathe for larger pipes. Special features were created on the pipes to facilitate tool location. The blade and toolbit designs have evolved to optimize cutting forces and tool durability. Satisfactory reliability has been achieved by performing 200 h of cutting during the two year period of development. Subsequently, over 100 'hands-on' cutting operations have been made on the JET machine since 1988 and a further 150 cuts are planned for 1992. Using a programmable controller the feed rate can be changed throughout the cutting operation into a predetermined way, thereby optimizing the tools' efficiency.<>
{"title":"The design, development and use of pipe cutting tools for remote handling in JET","authors":"S. Mills, A. Loving, M. Irving","doi":"10.1109/FUSION.1991.218859","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218859","url":null,"abstract":"The authors report on remote handling tools which have been specifically designed to meet requirements for pipe cutting at JET (Joint European Torus). The principal requirements were the quality of cut necessary for rewelding, effective swarf removal, and compactness for remote handling. The designs of tools had to be compatible with the severe access restrictions imposed by the JET machine. The processes used by the tools are sawing from the inside and outside of pipes, and orbital lathe for larger pipes. Special features were created on the pipes to facilitate tool location. The blade and toolbit designs have evolved to optimize cutting forces and tool durability. Satisfactory reliability has been achieved by performing 200 h of cutting during the two year period of development. Subsequently, over 100 'hands-on' cutting operations have been made on the JET machine since 1988 and a further 150 cuts are planned for 1992. Using a programmable controller the feed rate can be changed throughout the cutting operation into a predetermined way, thereby optimizing the tools' efficiency.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"26 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":"131836781","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.1991.218914
F. Najmabadi, R. Conn, C. Bathke, J. Blanchard, L. Bromberg, J. Brooks, E. Cheng, D. Cohn, D. Ehst, L. El-Guebaly, G. Emmert, T. Dolan, P. Gierszewski, S. Grotz, M.S. Hasan, J. Herring, S. K. Ho, A. Hollies, J. Holmes, E. Ibrahim, S. Jardin, C. Kessel, H. Khater, R. Krakowski, G.L. Kuleinski, J. Mandrekas, T. Mau, G. Miley, R.L. Miller, E. Mogahed, E. Reis, J. Santarius, M. Sawan, J. Schultz, K. Schultz, S. Sharafat, D. Steiner, D. Strickler, I. Sviatoslavsky, D. Sze, P. Titus, M. Valenti, K. Werley, J. H. Whealton, J.E.C. Williams, L. Wittenberg, C. Wong
A description of the ARIES-III research effort is presented, and the general features of the ARIES-III reactor are described. The plasma engineering and fusion-power-core design are summarized, including the major results, the key technical issues, and the central conclusions. Analyses have shown that the plasma power-balance window for D-/sup 3/He tokamak reactors is small and requires a first wall (or coating) that is highly reflective to synchrotron radiation and small values of tau /sub ash// epsilon /sub e/ (the ratio of ash-particle to energy confinement times in the core plasma). Both first and second stability regimes of operation have been considered. The second stability regime is chosen for the ARIES-III design point because the reactor can operate at a higher value of tau /sub ash// tau /sub E// tau /sub E/ approximately=2 (twice that of a first stability version), and because it has a reduced plasma current (30 MA), magnetic field at the coil (14 T), mass, and cost (also compared to a first-stability D-/sup 3/He reactor). The major and minor radii are, respectively 7.5 and 2.5 m.<>
{"title":"The ARIES-III D-3He tokamak-reactor study","authors":"F. Najmabadi, R. Conn, C. Bathke, J. Blanchard, L. Bromberg, J. Brooks, E. Cheng, D. Cohn, D. Ehst, L. El-Guebaly, G. Emmert, T. Dolan, P. Gierszewski, S. Grotz, M.S. Hasan, J. Herring, S. K. Ho, A. Hollies, J. Holmes, E. Ibrahim, S. Jardin, C. Kessel, H. Khater, R. Krakowski, G.L. Kuleinski, J. Mandrekas, T. Mau, G. Miley, R.L. Miller, E. Mogahed, E. Reis, J. Santarius, M. Sawan, J. Schultz, K. Schultz, S. Sharafat, D. Steiner, D. Strickler, I. Sviatoslavsky, D. Sze, P. Titus, M. Valenti, K. Werley, J. H. Whealton, J.E.C. Williams, L. Wittenberg, C. Wong","doi":"10.1109/FUSION.1991.218914","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218914","url":null,"abstract":"A description of the ARIES-III research effort is presented, and the general features of the ARIES-III reactor are described. The plasma engineering and fusion-power-core design are summarized, including the major results, the key technical issues, and the central conclusions. Analyses have shown that the plasma power-balance window for D-/sup 3/He tokamak reactors is small and requires a first wall (or coating) that is highly reflective to synchrotron radiation and small values of tau /sub ash// epsilon /sub e/ (the ratio of ash-particle to energy confinement times in the core plasma). Both first and second stability regimes of operation have been considered. The second stability regime is chosen for the ARIES-III design point because the reactor can operate at a higher value of tau /sub ash// tau /sub E// tau /sub E/ approximately=2 (twice that of a first stability version), and because it has a reduced plasma current (30 MA), magnetic field at the coil (14 T), mass, and cost (also compared to a first-stability D-/sup 3/He reactor). The major and minor radii are, respectively 7.5 and 2.5 m.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"11 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":"133838970","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.1991.218779
S. Pelloni, E. Cheng
The effect of ENDF/B-VI beryllium data on the neutronic characteristics of the reference ARIES-I fusion blanket is investigated. It is found that the total initial tritium breeding ratio (1.2154) calculated with ENDF/B-V is significantly higher, by about 5.7%, than that calculated with ENDF/B-VI beryllium cross sections (1.1504). When using beryllium data from ENDF/B-VI instead of ENDF/B-V, the maximum fast neutron flux above 0.111 MeV in the superconducting magnet calculated assuming a wall loading of 1 MW/m/sup 2/ increases by 4.6% (1.915*10/sup 9/ against 1.830*10/sup 9/ neutrons/cm/sup 2//s), whereas the total blanket energy multiplication decreases by about 3.3% (1.2518 against 1.2937), the average volumetric nuclear heating in the first wall by about 3% (4.8012 against 4.9484 W/cm/sup 3/), the maximum helium production rate in the neutron multiplier after one year irradiation significantly by 11.2% (1535 against 1720 parts per million), and the maximum hydrogen production rate in the first wall by 2.5% (385 against 395 parts per million).<>
{"title":"Implications of ENDF/B-VI beryllium data on the performance of the reference ARIES-I blanket","authors":"S. Pelloni, E. Cheng","doi":"10.1109/FUSION.1991.218779","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218779","url":null,"abstract":"The effect of ENDF/B-VI beryllium data on the neutronic characteristics of the reference ARIES-I fusion blanket is investigated. It is found that the total initial tritium breeding ratio (1.2154) calculated with ENDF/B-V is significantly higher, by about 5.7%, than that calculated with ENDF/B-VI beryllium cross sections (1.1504). When using beryllium data from ENDF/B-VI instead of ENDF/B-V, the maximum fast neutron flux above 0.111 MeV in the superconducting magnet calculated assuming a wall loading of 1 MW/m/sup 2/ increases by 4.6% (1.915*10/sup 9/ against 1.830*10/sup 9/ neutrons/cm/sup 2//s), whereas the total blanket energy multiplication decreases by about 3.3% (1.2518 against 1.2937), the average volumetric nuclear heating in the first wall by about 3% (4.8012 against 4.9484 W/cm/sup 3/), the maximum helium production rate in the neutron multiplier after one year irradiation significantly by 11.2% (1535 against 1720 parts per million), and the maximum hydrogen production rate in the first wall by 2.5% (385 against 395 parts per million).<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"54 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":"127727437","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.1991.218879
R. Thome, B.A. Smith, R. Pillsbury, P. Titus, R. Myatt
The design for the Burning Plasma Experiment (BPX) poloidal field (PF) coil system has evolved through several stages of machine size and physics requirements. The result has been a firm basis for a conceptual design with a significant R&D supporting activity on critical components. The authors review the characteristics of the latest PF system design and various facets of the R&D activity in the program as the machine has progressed. The BPX PF system design activity has satisfied machine physics requirements and dimensional constraints. Concepts for critical components have been developed to the point where detailed dimensions could begin to be solidified. Mechanical and electrical evaluation of materials, testing of selected components, and development of design criteria have spanned several iterations on machine requirements. This provides a strong basis even for initiating design of a new machine.<>
{"title":"Poloidal field (PF) coil system design and R&D for the Burning Plasma Experiment (BPX)","authors":"R. Thome, B.A. Smith, R. Pillsbury, P. Titus, R. Myatt","doi":"10.1109/FUSION.1991.218879","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218879","url":null,"abstract":"The design for the Burning Plasma Experiment (BPX) poloidal field (PF) coil system has evolved through several stages of machine size and physics requirements. The result has been a firm basis for a conceptual design with a significant R&D supporting activity on critical components. The authors review the characteristics of the latest PF system design and various facets of the R&D activity in the program as the machine has progressed. The BPX PF system design activity has satisfied machine physics requirements and dimensional constraints. Concepts for critical components have been developed to the point where detailed dimensions could begin to be solidified. Mechanical and electrical evaluation of materials, testing of selected components, and development of design criteria have spanned several iterations on machine requirements. This provides a strong basis even for initiating design of a new machine.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"33 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":"128605659","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.1991.218814
T. Fujii, M. Saigusa, S. Moriyama, K. Annoh, S. Shinozaki, M. Terakado, H. Kimura, M. Ohta, M. Texuka, K. Wakabayashi, J. Ohmori, N. Miki, N. Kobayashi, K. Itoh
An upgrading of the ICRF (ion cyclotron range of frequencies) heating system for JT-60 (JAERI Tokamak-60) was performed during the modification of the JT-60 tokamak, which, after the upgrade, allowed 6 MA of plasma current and 100 m/sup 3/ of plasma volume. In the upgrade, the old ICRF launcher was replaced by two new ones in order to inject more power ( approximately 4.5 MW). The new launcher has severe design conditions of high heat flux from the plasma (max. 0.4 MW/m/sup 2/) and large electromagnetic force induced by plasma disruption (6 MA/5 ms). The total torque acting on the launcher by the electromagnetic force is about 35 t-m. Structural analysis was carried out to evaluate the integrity of the launcher, particularly of the feedthrough and the Faraday shield, under these severe conditions. The launchers are now being installed on the horizontal ports of the JT-60U vacuum vessel, which are movable by 40 mm in a radial direction.<>
{"title":"Mechanical design and analysis of JT-60U ICRF launcher","authors":"T. Fujii, M. Saigusa, S. Moriyama, K. Annoh, S. Shinozaki, M. Terakado, H. Kimura, M. Ohta, M. Texuka, K. Wakabayashi, J. Ohmori, N. Miki, N. Kobayashi, K. Itoh","doi":"10.1109/FUSION.1991.218814","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218814","url":null,"abstract":"An upgrading of the ICRF (ion cyclotron range of frequencies) heating system for JT-60 (JAERI Tokamak-60) was performed during the modification of the JT-60 tokamak, which, after the upgrade, allowed 6 MA of plasma current and 100 m/sup 3/ of plasma volume. In the upgrade, the old ICRF launcher was replaced by two new ones in order to inject more power ( approximately 4.5 MW). The new launcher has severe design conditions of high heat flux from the plasma (max. 0.4 MW/m/sup 2/) and large electromagnetic force induced by plasma disruption (6 MA/5 ms). The total torque acting on the launcher by the electromagnetic force is about 35 t-m. Structural analysis was carried out to evaluate the integrity of the launcher, particularly of the feedthrough and the Faraday shield, under these severe conditions. The launchers are now being installed on the horizontal ports of the JT-60U vacuum vessel, which are movable by 40 mm in a radial direction.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"328 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":"115958583","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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