Pub Date : 1991-09-30DOI: 10.1109/FUSION.1991.218717
D. Petersen
A new timing system has been designed to meet the requirements for the one megawatt test set (MWTS). This test set is a system of power supplies, support systems, controls, and monitors that will be used for the development of 1-MW gyrotrons. The purpose of the timing system is to provide timing signals for control of the gyrotron's cathode, anode, and heater power supplies as well as oscilloscope, digitizer, and other control or diagnostic triggers. Some unique timing requirements include four main operating modes: single pulse, repetitive pulse, continuous wave (CW), and CW cathode with repetitively pulsed anode. Timing adjustments can be made while the tube is running with no adverse effects. This modular, distributed timing system is implemented using two CAMAC modules: the master timing controller and the two channel timing generator. A single master timing controller provides real-time control and synchronization for all timing generators in the system. The timing generators provide triggers having adjustable delay and duration to the various system components.<>
{"title":"A modular timing system for megawatt gyrotrons","authors":"D. Petersen","doi":"10.1109/FUSION.1991.218717","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218717","url":null,"abstract":"A new timing system has been designed to meet the requirements for the one megawatt test set (MWTS). This test set is a system of power supplies, support systems, controls, and monitors that will be used for the development of 1-MW gyrotrons. The purpose of the timing system is to provide timing signals for control of the gyrotron's cathode, anode, and heater power supplies as well as oscilloscope, digitizer, and other control or diagnostic triggers. Some unique timing requirements include four main operating modes: single pulse, repetitive pulse, continuous wave (CW), and CW cathode with repetitively pulsed anode. Timing adjustments can be made while the tube is running with no adverse effects. This modular, distributed timing system is implemented using two CAMAC modules: the master timing controller and the two channel timing generator. A single master timing controller provides real-time control and synchronization for all timing generators in the system. The timing generators provide triggers having adjustable delay and duration to the various system components.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"339 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":"124770234","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.1991.218850
R. Russ, A. Haigh, S. J. Booth
The measures detailed in the 'Code of Practice for the Safe use of Beryllium at JET' are summarized. JETs (Joint European Torus') policy of segregating beryllium related work into BeCAs (beryllium controlled areas) has ensured that the general workplace (torus hall, assembly hall, etc.) remain free of airborne and surface contamination, and hence no safety precautions have had to be adopted in these areas. There have been no beryllium incidents leading to airborne concentrations in excess of 0.2 mu gm/sup -3/ outside of any BeCA. Aerial discharges of beryllium have generally been undetectable and have never approached the limit of 0.01 mu gm/sup -3/. No cases of acute or chronic beryllium intoxication have been diagnosed; some minor cuts have been sustained in BeCAs (no contamination of the wound was found on samples) and no treatment has been required beyond conventional first-aid procedures. In-vessel work has continued to be possible despite high levels of surface contamination and variable air concentrations, but the logistics support for in-vessel operation has increased very noticeably as a result of beryllium.<>
{"title":"Beryllium safety at JET","authors":"R. Russ, A. Haigh, S. J. Booth","doi":"10.1109/FUSION.1991.218850","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218850","url":null,"abstract":"The measures detailed in the 'Code of Practice for the Safe use of Beryllium at JET' are summarized. JETs (Joint European Torus') policy of segregating beryllium related work into BeCAs (beryllium controlled areas) has ensured that the general workplace (torus hall, assembly hall, etc.) remain free of airborne and surface contamination, and hence no safety precautions have had to be adopted in these areas. There have been no beryllium incidents leading to airborne concentrations in excess of 0.2 mu gm/sup -3/ outside of any BeCA. Aerial discharges of beryllium have generally been undetectable and have never approached the limit of 0.01 mu gm/sup -3/. No cases of acute or chronic beryllium intoxication have been diagnosed; some minor cuts have been sustained in BeCAs (no contamination of the wound was found on samples) and no treatment has been required beyond conventional first-aid procedures. In-vessel work has continued to be possible despite high levels of surface contamination and variable air concentrations, but the logistics support for in-vessel operation has increased very noticeably as a result of beryllium.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"135 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":"127297751","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.1991.218729
W. Parkin
A design utilizing industry-standard EURO card packaging, connected via ribbon cable to a custom I/O backplane, accommodates the many different circuits to be used on Alcator C-MOD. The chassis approach eliminates packaging, powering, and mounting problems associated with one-of-a-kind circuits and allows dense packaging of multiple circuits. The separate I/O backplane allows easy removal and replacement of circuit cards, while leaving front panel space available for circuit-specific indicators. The DIN pin-and-socket connectors are considerably more robust and reliable than edge connectors. The pin-out allows up to four circuits per card, allowing up to eighty channels in one 5-1/2-in-tall, rack mountable chassis. A variety of general-purpose signal conditioning circuit cards was designed and built for engineering and diagnostic systems. Four channel circuits include low-drift integrators, analog multiplexers, temperature controllers, instrumentation amplifiers, transimpedance amplifiers, and an eight-pole Bessel filter for anti-aliasing applications. A two-channel card provides thermal electric cooler control. Fiber-optic interface is provided by an 8-b digital transceiver and an analog fiber-optic link.<>
{"title":"Signal conditioning electronics and packaging for the Alcator C-MOD tokamak","authors":"W. Parkin","doi":"10.1109/FUSION.1991.218729","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218729","url":null,"abstract":"A design utilizing industry-standard EURO card packaging, connected via ribbon cable to a custom I/O backplane, accommodates the many different circuits to be used on Alcator C-MOD. The chassis approach eliminates packaging, powering, and mounting problems associated with one-of-a-kind circuits and allows dense packaging of multiple circuits. The separate I/O backplane allows easy removal and replacement of circuit cards, while leaving front panel space available for circuit-specific indicators. The DIN pin-and-socket connectors are considerably more robust and reliable than edge connectors. The pin-out allows up to four circuits per card, allowing up to eighty channels in one 5-1/2-in-tall, rack mountable chassis. A variety of general-purpose signal conditioning circuit cards was designed and built for engineering and diagnostic systems. Four channel circuits include low-drift integrators, analog multiplexers, temperature controllers, instrumentation amplifiers, transimpedance amplifiers, and an eight-pole Bessel filter for anti-aliasing applications. A two-channel card provides thermal electric cooler control. Fiber-optic interface is provided by an 8-b digital transceiver and an analog fiber-optic link.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"176 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":"121927994","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.1991.218880
P. Heitzenroeder
The toroidal field coil system of the Burning Plasma Experiment (BPX) consists of 18 beryllium copper magnets arrayed in a wedged configuration with a major radius of 2.6 m and a field strength capability on axis of 9.0 T. The toroidal array is constructed from six 3-coil modules to facilitate remote recovery in the event of a magnet failure after nuclear activation precludes hands-on servicing. The magnets are of a modified Bitter plate design with partial cases of type 316-LN stainless steel welded with Inconel 182 weld wire. The coil turn plates are fabricated from CDA C17510 beryllium copper with optimized mechanical, thermal, and electrical characteristics. Cryogenic cooling is used to reduce power dissipation and to enhance performance. The magnets are cooled between experimental pulses by pressurized liquid nitrogen flowing through channels in the edges of the coil turns, allowing one full-power pulse per hour. Overturning forces are reacted by friction forces generated by the wedging pressure in the central leg regions and by case and structure in the other regions of the coil. The magnets are designed to structural design criteria modeled after the ASME Boiler and Pressure Vessel Code, Section III, but with appropriate changes.<>
{"title":"BPX toroidal field coil design","authors":"P. Heitzenroeder","doi":"10.1109/FUSION.1991.218880","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218880","url":null,"abstract":"The toroidal field coil system of the Burning Plasma Experiment (BPX) consists of 18 beryllium copper magnets arrayed in a wedged configuration with a major radius of 2.6 m and a field strength capability on axis of 9.0 T. The toroidal array is constructed from six 3-coil modules to facilitate remote recovery in the event of a magnet failure after nuclear activation precludes hands-on servicing. The magnets are of a modified Bitter plate design with partial cases of type 316-LN stainless steel welded with Inconel 182 weld wire. The coil turn plates are fabricated from CDA C17510 beryllium copper with optimized mechanical, thermal, and electrical characteristics. Cryogenic cooling is used to reduce power dissipation and to enhance performance. The magnets are cooled between experimental pulses by pressurized liquid nitrogen flowing through channels in the edges of the coil turns, allowing one full-power pulse per hour. Overturning forces are reacted by friction forces generated by the wedging pressure in the central leg regions and by case and structure in the other regions of the coil. The magnets are designed to structural design criteria modeled after the ASME Boiler and Pressure Vessel Code, Section III, but with appropriate changes.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"30 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":"123053524","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.1991.218876
G. Clerc, J.P. Ichac, M. Tardy
Thomson Tubes Electroniques tetrodes of the 2 MW class, the TH 525 and the TH 526, have now entered the industrial production line, and their amplifiers are fully operational. They have already made it possible to reach outstanding results on plasmas. At JET (Joint European Torus), 16 amplifiers (25-27 MHz) are fitted with TH 525 tetrodes, which each deliver 2 MW of RF power in 20 s pulses to a plasma load (voltage standing wave ratio up to 1.5:1). The JET teams have tested these tubes up to 2.7 MW peak. A record peak of 22.5 MW of ICRH (ion cyclotron resonance heating) power has been coupled to the plasma (26.3 MW at the amplifier output). Two TH 525 tetrodes have also delivered their RF power to the TEXTOR plasma for more than two years. Six generators with TH 526 tetrodes are also delivering 2 MW each during 30 s, between 35 and 80 MHz, at TORE SUPRA. They have also been tested with the same pulse length at 1.4 MW/120 MHz, and are operational at 1 MW at this frequency, with high figures of gain, bandwidth, and efficiency. The next generation of Thomson tetrodes will make it possible to reach the same levels of power at these frequencies in CW (continuous wave) operation, with an even higher efficiency.<>
{"title":"ICRH Thomson tetrodes: from long pulses to CW","authors":"G. Clerc, J.P. Ichac, M. Tardy","doi":"10.1109/FUSION.1991.218876","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218876","url":null,"abstract":"Thomson Tubes Electroniques tetrodes of the 2 MW class, the TH 525 and the TH 526, have now entered the industrial production line, and their amplifiers are fully operational. They have already made it possible to reach outstanding results on plasmas. At JET (Joint European Torus), 16 amplifiers (25-27 MHz) are fitted with TH 525 tetrodes, which each deliver 2 MW of RF power in 20 s pulses to a plasma load (voltage standing wave ratio up to 1.5:1). The JET teams have tested these tubes up to 2.7 MW peak. A record peak of 22.5 MW of ICRH (ion cyclotron resonance heating) power has been coupled to the plasma (26.3 MW at the amplifier output). Two TH 525 tetrodes have also delivered their RF power to the TEXTOR plasma for more than two years. Six generators with TH 526 tetrodes are also delivering 2 MW each during 30 s, between 35 and 80 MHz, at TORE SUPRA. They have also been tested with the same pulse length at 1.4 MW/120 MHz, and are operational at 1 MW at this frequency, with high figures of gain, bandwidth, and efficiency. The next generation of Thomson tetrodes will make it possible to reach the same levels of power at these frequencies in CW (continuous wave) operation, with an even higher efficiency.<<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":"126725057","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.1991.218777
M. Viola, L. Dudek, V. Garzotto, J. Kamperschroer, G. Martin, R. Rossmassler, R. Sissingh
The Tokamak Fusion Test Reactor (TFTR) vacuum system is being upgraded to trap SF/sub 6/ gas which may leak from the neutral beam injectors during the planned deuterium-tritium experiments. This is necessary since SF/sub 6/ can, in 25% of a gram-mole, completely inactivate the platinum catalyst in the tritium cleanup system (TCS). Any SF/sub 6/ leaking into the injectors will initially be held on the liquid helium or liquid nitrogen cryopanels. During regens of the liquid helium panel, the deuterium and tritium will pass to the TCS, with the SF/sub 6/ being retained on the cryopanels. In the final phase of the accident recovery scenario, the cryopanels will be warmed to room temperature and the SF/sub 6/ collected by a liquid-nitrogen cooled cold trap. Design of the cold trap and measurements of the vapor pressure of SF/sub 6/ at 77 K as a function of surface coverage are presented. The SF/sub 6/ vapor pressure at 77 K, with only a few monolayers absorbed, was found to be <10/sup -7/ torr. Modest pressures of approximately 10/sup -5/ torr were measured after nearly 3*10/sup 5/ torr-1 were deposited on a cold trap with an area of 0.1 m/sup 2/. In the final configuration, the cold trap will also be used to collect water which may have accidentally leaked into the vacuum system.<>
{"title":"SF/sub 6/ and water cold trap for TFTR deuterium-tritium operations","authors":"M. Viola, L. Dudek, V. Garzotto, J. Kamperschroer, G. Martin, R. Rossmassler, R. Sissingh","doi":"10.1109/FUSION.1991.218777","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218777","url":null,"abstract":"The Tokamak Fusion Test Reactor (TFTR) vacuum system is being upgraded to trap SF/sub 6/ gas which may leak from the neutral beam injectors during the planned deuterium-tritium experiments. This is necessary since SF/sub 6/ can, in 25% of a gram-mole, completely inactivate the platinum catalyst in the tritium cleanup system (TCS). Any SF/sub 6/ leaking into the injectors will initially be held on the liquid helium or liquid nitrogen cryopanels. During regens of the liquid helium panel, the deuterium and tritium will pass to the TCS, with the SF/sub 6/ being retained on the cryopanels. In the final phase of the accident recovery scenario, the cryopanels will be warmed to room temperature and the SF/sub 6/ collected by a liquid-nitrogen cooled cold trap. Design of the cold trap and measurements of the vapor pressure of SF/sub 6/ at 77 K as a function of surface coverage are presented. The SF/sub 6/ vapor pressure at 77 K, with only a few monolayers absorbed, was found to be <10/sup -7/ torr. Modest pressures of approximately 10/sup -5/ torr were measured after nearly 3*10/sup 5/ torr-1 were deposited on a cold trap with an area of 0.1 m/sup 2/. In the final configuration, the cold trap will also be used to collect water which may have accidentally leaked into the vacuum system.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"04 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":"127184965","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.1991.218779
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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