Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518350
R. Gentzlinger, S. Mendelsohn, B. Abel, I. Birnbaum, U. Christensen, S. Kalsi, J. Mueller, M. Phillips, J. Swinton, D. Weissenburger, S. Cohen, E. Fredd, R. Majeski, R. Motley, R. Walls
A pre-conceptual design has been produced for a plasma device to further divertor concepts and validate technology in support of the International Thermonuclear Experimental Reactor program. The ITER Divertor Experiment and Laboratory (IDEAL) design effort is to develop a reliable, maintainable and robust facility for steady-state divertor simulation experiments. The configuration includes a 30 meter vacuum vessel, enclosed within a set of 30 high field superconducting solenoid modules, a resistive quadrupole coil set, a radio-frequency heating system and a complement of diagnostics. It is planned to utilize existing facilities, and off-the-shelf hardware, wherever possible to maximize technological return on investment.
{"title":"IDEAL preconceptual design development","authors":"R. Gentzlinger, S. Mendelsohn, B. Abel, I. Birnbaum, U. Christensen, S. Kalsi, J. Mueller, M. Phillips, J. Swinton, D. Weissenburger, S. Cohen, E. Fredd, R. Majeski, R. Motley, R. Walls","doi":"10.1109/FUSION.1993.518350","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518350","url":null,"abstract":"A pre-conceptual design has been produced for a plasma device to further divertor concepts and validate technology in support of the International Thermonuclear Experimental Reactor program. The ITER Divertor Experiment and Laboratory (IDEAL) design effort is to develop a reliable, maintainable and robust facility for steady-state divertor simulation experiments. The configuration includes a 30 meter vacuum vessel, enclosed within a set of 30 high field superconducting solenoid modules, a resistive quadrupole coil set, a radio-frequency heating system and a complement of diagnostics. It is planned to utilize existing facilities, and off-the-shelf hardware, wherever possible to maximize technological return on investment.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"42 10 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"128684574","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 : 1993-10-11DOI: 10.1109/FUSION.1993.518539
K. Takase, M. Z. Hasan
Convective heat transfer in MHD laminar flow through a square duct in the plasma facing components (PFCs) of fusion reactors is analyzed numerically to investigate the effects of a transverse magnetic field and the nonuniformity of surface heat flux. As in the case of non-MHD laminar flow, analyzed earlier, the corners of the plasma facing (PF) side are possible hot-spot areas; the presence of a transverse magnetic field does not alleviate this situation to any significant degree. The nonuniformity of surface heat flux nearly cancels this increase of Nu at the PF side. At Hartmann number (Ha) of 40, Nu at the center increases from 6.9 to 23, but at the corner from 2.2 to 3.2 only with uniform heat flux. But, as the extent of nonuniformity of surface heat flux increases, Nu at the center decreases rapidly. This effect, however, saturates rapidly with the increase of the nonuniformity of heat flux. The increase of Nu with Ha is very small for large nonuniformity of the heat flux. Under this condition, Nu at the center of the plasma facing side, is 2.8 at Ha=O, 3.02 at Ha=16 and 3.4 at Ha=400. At the corner of the PF side, the corresponding values of Nu are 2.65, 2.88, and 3.0, respectively. The effect of Ha on entry length is small for highly nonuniform heat flux.
{"title":"Heat transfer in MHD laminar flow through a rectangular channel in the plasma facing components of fusion reactors","authors":"K. Takase, M. Z. Hasan","doi":"10.1109/FUSION.1993.518539","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518539","url":null,"abstract":"Convective heat transfer in MHD laminar flow through a square duct in the plasma facing components (PFCs) of fusion reactors is analyzed numerically to investigate the effects of a transverse magnetic field and the nonuniformity of surface heat flux. As in the case of non-MHD laminar flow, analyzed earlier, the corners of the plasma facing (PF) side are possible hot-spot areas; the presence of a transverse magnetic field does not alleviate this situation to any significant degree. The nonuniformity of surface heat flux nearly cancels this increase of Nu at the PF side. At Hartmann number (Ha) of 40, Nu at the center increases from 6.9 to 23, but at the corner from 2.2 to 3.2 only with uniform heat flux. But, as the extent of nonuniformity of surface heat flux increases, Nu at the center decreases rapidly. This effect, however, saturates rapidly with the increase of the nonuniformity of heat flux. The increase of Nu with Ha is very small for large nonuniformity of the heat flux. Under this condition, Nu at the center of the plasma facing side, is 2.8 at Ha=O, 3.02 at Ha=16 and 3.4 at Ha=400. At the corner of the PF side, the corresponding values of Nu are 2.65, 2.88, and 3.0, respectively. The effect of Ha on entry length is small for highly nonuniform heat flux.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"2010 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127350854","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 : 1993-10-11DOI: 10.1109/FUSION.1993.518474
G. Sager, C. Wong, D.D. Kapich, C. Mcdonald, R. Schleicher
The impact of cyclic loading of the power conversion system of a helium-cooled, pulsed tokamak power plant is assessed. Design limits of key components of heat transport systems employing Rankine and Brayton thermodynamic cycles are quantified based on experience in gas-cooled fission reactor design and operation. Cyclic loads due to pulsed tokamak operation are estimated. Expected performance of the steam generator is shown to be incompatible with pulsed tokamak operation without load leveling thermal energy storage. The closed cycle gas turbine is evaluated qualitatively based on performance of existing industrial and aeroderivative gas turbines. Advances in key technologies which significantly improve prospects for operation with tokamak fusion plants are reviewed.
{"title":"Limitations of power conversion systems under transient loads and impact on the pulsed tokamak power reactor","authors":"G. Sager, C. Wong, D.D. Kapich, C. Mcdonald, R. Schleicher","doi":"10.1109/FUSION.1993.518474","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518474","url":null,"abstract":"The impact of cyclic loading of the power conversion system of a helium-cooled, pulsed tokamak power plant is assessed. Design limits of key components of heat transport systems employing Rankine and Brayton thermodynamic cycles are quantified based on experience in gas-cooled fission reactor design and operation. Cyclic loads due to pulsed tokamak operation are estimated. Expected performance of the steam generator is shown to be incompatible with pulsed tokamak operation without load leveling thermal energy storage. The closed cycle gas turbine is evaluated qualitatively based on performance of existing industrial and aeroderivative gas turbines. Advances in key technologies which significantly improve prospects for operation with tokamak fusion plants are reviewed.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"126774295","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 : 1993-10-11DOI: 10.1109/FUSION.1993.518543
B. Felker, S. Allen, H. Bell, J. Bowman, M. Delong, M. Fenstermacher, S. W. Ferguson, W. F. Fields, D. Hathaway, E. Hooper, S. Hulsey, M. Jackson, D. Lang, C. Lasnier, M. Makowski, J. Moller, W. Meyer, D. Nilson, D. Peterson, D. Seilhymer, B. Stallard
The MTX explored the plasma heating effects of 140 GHz microwaves from both Gyrotrons and from the IMP FEL wiggler. The Gyrotron was long pulse length (0.5 seconds maximum) and the FEL produced short-pulse length, high-peak power, single and burst modes of 140 GHz microwaves. Full-power operations of the IMP FEL wiggler were commenced in April of 1992 and continued into October of 1992. The Experimental Test Accelerator II (ETA-II) provided a 50-nanosecond, 6-MeV, 2-3 kAmp electron beam that was introduced co-linear into the IMP FEL with a 140 GHz Gyrotron master oscillator (MO). The FEL was able to amplify the MO signal from approximately 7 kW to peaks consistently in the range of 1-2 GW. This microwave pulse was transmitted into the MTX and allowed the exploration of the linear and non-linear effects of short pulse, intense power in the MTX plasma. Single pulses were used to explore and gain operating experience in the parameter space of the IMP FEL, and finally evaluate transmission and absorption in the MTX. Single-pulse operations were repeatable. After the MTX was shut down burst-mode operations were successful at 2 kHz. This paper will describe the IMP FEL, Microwave Transmission System to MTX, the diagnostics used for measurements, and tile operations of the entire Microwave system. A discussion of correlated and uncorrelated errors that affect FEL performance will be made. Linear and nonlinear absorption data of the microwaves in the MTX plasma will be presented.
{"title":"Two-gigawatt burst-mode operation of the intense microwave prototype (IMP) free-electron laser (FEL) for the microwave tokamak experiment (MTX)","authors":"B. Felker, S. Allen, H. Bell, J. Bowman, M. Delong, M. Fenstermacher, S. W. Ferguson, W. F. Fields, D. Hathaway, E. Hooper, S. Hulsey, M. Jackson, D. Lang, C. Lasnier, M. Makowski, J. Moller, W. Meyer, D. Nilson, D. Peterson, D. Seilhymer, B. Stallard","doi":"10.1109/FUSION.1993.518543","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518543","url":null,"abstract":"The MTX explored the plasma heating effects of 140 GHz microwaves from both Gyrotrons and from the IMP FEL wiggler. The Gyrotron was long pulse length (0.5 seconds maximum) and the FEL produced short-pulse length, high-peak power, single and burst modes of 140 GHz microwaves. Full-power operations of the IMP FEL wiggler were commenced in April of 1992 and continued into October of 1992. The Experimental Test Accelerator II (ETA-II) provided a 50-nanosecond, 6-MeV, 2-3 kAmp electron beam that was introduced co-linear into the IMP FEL with a 140 GHz Gyrotron master oscillator (MO). The FEL was able to amplify the MO signal from approximately 7 kW to peaks consistently in the range of 1-2 GW. This microwave pulse was transmitted into the MTX and allowed the exploration of the linear and non-linear effects of short pulse, intense power in the MTX plasma. Single pulses were used to explore and gain operating experience in the parameter space of the IMP FEL, and finally evaluate transmission and absorption in the MTX. Single-pulse operations were repeatable. After the MTX was shut down burst-mode operations were successful at 2 kHz. This paper will describe the IMP FEL, Microwave Transmission System to MTX, the diagnostics used for measurements, and tile operations of the entire Microwave system. A discussion of correlated and uncorrelated errors that affect FEL performance will be made. Linear and nonlinear absorption data of the microwaves in the MTX plasma will be presented.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"21 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127174234","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 : 1993-10-11DOI: 10.1109/FUSION.1993.518284
A.G. Heics, W. Shmayda
A prototypical metal hydride based recirculating glovebox cleanup system was commissioned and tested with tritium. Getter material SAES St 198 was selected for its ability to effectively remove tritium and trace impurities from inert or nitrogen glovebox cover gases and its ease of tritium recovery by heating to elevated temperatures. The Secondary Enclosure Clean-up (SEC) system utilizes a programmable controller for process control and system isolation and alarm in the event of an abnormal condition. The system was used to detect glovebox air in leakage by tracking the moisture level within the glovebox when the bed is bypassed. An aliquot of 0.5 Ci of tritium, intentionally released into a glovebox to demonstrate the system performance, was effectively removed by the getter bed in about 20 minutes or about 7 system time constants.
一个基于金属氢化物的循环手套箱清理系统原型进行了调试，并对其进行了氚测试。吸气材料SAES St 198之所以被选中，是因为它能够有效地从惰性气体或氮气手套箱盖气体中去除氚和微量杂质，并且易于通过加热到高温回收氚。次级外壳清理(SEC)系统利用可编程控制器进行过程控制和系统隔离，并在异常情况下发出警报。当床层被绕过时，该系统通过跟踪手套箱内的湿度水平来检测手套箱中的泄漏空气。将0.5 Ci的氚故意释放到手套箱中以演示系统性能，在大约20分钟或大约7个系统时间常数内被吸气床有效地去除。
{"title":"Development of a secondary enclosure clean-up system for tritium systems","authors":"A.G. Heics, W. Shmayda","doi":"10.1109/FUSION.1993.518284","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518284","url":null,"abstract":"A prototypical metal hydride based recirculating glovebox cleanup system was commissioned and tested with tritium. Getter material SAES St 198 was selected for its ability to effectively remove tritium and trace impurities from inert or nitrogen glovebox cover gases and its ease of tritium recovery by heating to elevated temperatures. The Secondary Enclosure Clean-up (SEC) system utilizes a programmable controller for process control and system isolation and alarm in the event of an abnormal condition. The system was used to detect glovebox air in leakage by tracking the moisture level within the glovebox when the bed is bypassed. An aliquot of 0.5 Ci of tritium, intentionally released into a glovebox to demonstrate the system performance, was effectively removed by the getter bed in about 20 minutes or about 7 system time constants.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"7 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"114241527","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 : 1993-10-11DOI: 10.1109/FUSION.1993.518346
F. Bellina, G. Chitarin, M. Guarnieri, A. Stella, F. Trevisan
RFX, the largest RFP machine, has air-core poloidal windings. A tree-shaped earthing geometry has been adopted for all the machine components, to avoid electrical loops. Nevertheless, during the first operation phase a number of accidental contacts occurred, which caused loops currents high enough to distort plasma equilibrium. These loops could be detected by means of RGM, a system designed to perform fast winding protection, but able to detect accidental earth currents as well. After careful analyses of the signals, these earth faults were always located and removed. The use of a compass resulted particularly useful in the occasion of a number of these faults, the others were detected by means of rogowski coil probes.
{"title":"Earth faults during RFX initial operations","authors":"F. Bellina, G. Chitarin, M. Guarnieri, A. Stella, F. Trevisan","doi":"10.1109/FUSION.1993.518346","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518346","url":null,"abstract":"RFX, the largest RFP machine, has air-core poloidal windings. A tree-shaped earthing geometry has been adopted for all the machine components, to avoid electrical loops. Nevertheless, during the first operation phase a number of accidental contacts occurred, which caused loops currents high enough to distort plasma equilibrium. These loops could be detected by means of RGM, a system designed to perform fast winding protection, but able to detect accidental earth currents as well. After careful analyses of the signals, these earth faults were always located and removed. The use of a compass resulted particularly useful in the occasion of a number of these faults, the others were detected by means of rogowski coil probes.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"64 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127018120","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 : 1993-10-11DOI: 10.1109/FUSION.1993.518276
S. Sudo, M. Kanno, H. Kaneko, S. Saka, T. Shirai, T. Baba
From the results of the simulation study including pellet ablation and 1-D transport code, it is found that a high speed pellet injector with pellet velocity of more than 3 km/s is necessary for the penetration of the pellet with diameter of 3 mm into the core region under the expected plasma condition of Large Helical Device (LHD) of heliotron/stellarator type with superconducting coils at NIFS in Japan. Therefore, a two stage pellet injector was constructed and tested successfully in order to obtain the pellet velocity range of 3 km/s. Based upon the above results, a high speed flexible multiple-pellet injection system "HIPEL" for LHD is proposed. HIPEL consists of independent (1) 10 two-stage gun barrels and (2) 10 single-stage gun barrels. It has multi purposes such as refueling and flexible density profile control, diagnostics and the other functions.
{"title":"Proposed high speed pellet injection system \"HIPEL\" for Large Helical Device","authors":"S. Sudo, M. Kanno, H. Kaneko, S. Saka, T. Shirai, T. Baba","doi":"10.1109/FUSION.1993.518276","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518276","url":null,"abstract":"From the results of the simulation study including pellet ablation and 1-D transport code, it is found that a high speed pellet injector with pellet velocity of more than 3 km/s is necessary for the penetration of the pellet with diameter of 3 mm into the core region under the expected plasma condition of Large Helical Device (LHD) of heliotron/stellarator type with superconducting coils at NIFS in Japan. Therefore, a two stage pellet injector was constructed and tested successfully in order to obtain the pellet velocity range of 3 km/s. Based upon the above results, a high speed flexible multiple-pellet injection system \"HIPEL\" for LHD is proposed. HIPEL consists of independent (1) 10 two-stage gun barrels and (2) 10 single-stage gun barrels. It has multi purposes such as refueling and flexible density profile control, diagnostics and the other functions.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"24 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"131297958","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 : 1993-10-11DOI: 10.1109/FUSION.1993.518520
W. Beck
While bringing Alcator C-MOD on line, failure of a solder joint caused an open circuit in one of the PF coils located within the toroidal field magnet. A design review was conducted to analyze the failure and propose possible solutions. The exact reason for the failure was not determined, but the joint may have been weakened by high temperatures during bakeout of the vacuum vessel. Peeling forces also may have been induced by unforeseen temperature gradients and/or magnetic loads. Significant design changes, which are limited to highly stressed PF coils located within the toroidal field magnet, involved repositioning the joint away from the coaxial termination and eliminating the use of solder as a structural element. PF coils external to the toroidal field magnet are not so highly stressed and brazing is acceptable. The redesign easily accommodates repositioning the joint, but finding a substitute for solder, which was originally selected to avoid annealing the cold worked copper conductor, proved difficult. Localized annealing which occurs in welding and brazing processes eliminated the two most common methods of terminating copper coils. There is not enough space available in the vacuum vessel coil pockets to accommodate mechanical clamping devices. The use of fasteners such as screws and rivets was prohibited due to adverse effects on fatigue life. Electroforming, a process by which complex parts are formed by electroplating materials such as copper onto an electrically conductive mandrel, was selected to replace soldering the joint. Electroformed copper sheet exhibited superior material properties to those of the C-10700 coil conductor, which has yield strength of 290 MPa. Changes, development of an electroformed electromechanical joint, and coil manufacturing will be further described.
{"title":"Repair of poloidal field magnets on Alcator C-Mod","authors":"W. Beck","doi":"10.1109/FUSION.1993.518520","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518520","url":null,"abstract":"While bringing Alcator C-MOD on line, failure of a solder joint caused an open circuit in one of the PF coils located within the toroidal field magnet. A design review was conducted to analyze the failure and propose possible solutions. The exact reason for the failure was not determined, but the joint may have been weakened by high temperatures during bakeout of the vacuum vessel. Peeling forces also may have been induced by unforeseen temperature gradients and/or magnetic loads. Significant design changes, which are limited to highly stressed PF coils located within the toroidal field magnet, involved repositioning the joint away from the coaxial termination and eliminating the use of solder as a structural element. PF coils external to the toroidal field magnet are not so highly stressed and brazing is acceptable. The redesign easily accommodates repositioning the joint, but finding a substitute for solder, which was originally selected to avoid annealing the cold worked copper conductor, proved difficult. Localized annealing which occurs in welding and brazing processes eliminated the two most common methods of terminating copper coils. There is not enough space available in the vacuum vessel coil pockets to accommodate mechanical clamping devices. The use of fasteners such as screws and rivets was prohibited due to adverse effects on fatigue life. Electroforming, a process by which complex parts are formed by electroplating materials such as copper onto an electrically conductive mandrel, was selected to replace soldering the joint. Electroformed copper sheet exhibited superior material properties to those of the C-10700 coil conductor, which has yield strength of 290 MPa. Changes, development of an electroformed electromechanical joint, and coil manufacturing will be further described.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"64 2","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"114015859","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 : 1993-10-11DOI: 10.1109/FUSION.1993.518413
D. Humphreys, M. Firestone, J. Morrow-Jones
The capabilities of the new DIII-D digital control system have motivated an effort to apply state-of-the-art multivariable techniques to control of the DIII-D tokamak. Tokamak plasma control is inherently multivariable in nature, since many closely coupled equilibrium parameters must be regulated simultaneously during a discharge. The present work describes the determination of dynamic models for plasma response and plasma interaction with conducting structures, necessary for calculation of accurate and robust multivariable control laws. Plasma response matrices and shape prediction matrices are calculated from analytic models and perturbed ideal MHD equilibria. Plasma resistive effects are described by a circuit equation which conserves poloidal flux on time scales shorter than the plasma L/R time. Shape estimation and plasma/conductor eigenmode spectrum results are presented along with experimental data and time-dependent simulations.
{"title":"Plasma response modeling for multivariable tokamak control design","authors":"D. Humphreys, M. Firestone, J. Morrow-Jones","doi":"10.1109/FUSION.1993.518413","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518413","url":null,"abstract":"The capabilities of the new DIII-D digital control system have motivated an effort to apply state-of-the-art multivariable techniques to control of the DIII-D tokamak. Tokamak plasma control is inherently multivariable in nature, since many closely coupled equilibrium parameters must be regulated simultaneously during a discharge. The present work describes the determination of dynamic models for plasma response and plasma interaction with conducting structures, necessary for calculation of accurate and robust multivariable control laws. Plasma response matrices and shape prediction matrices are calculated from analytic models and perturbed ideal MHD equilibria. Plasma resistive effects are described by a circuit equation which conserves poloidal flux on time scales shorter than the plasma L/R time. Shape estimation and plasma/conductor eigenmode spectrum results are presented along with experimental data and time-dependent simulations.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"4 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124001182","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 : 1993-10-11DOI: 10.1109/FUSION.1993.518358
S. Fairfax, D. Montgomery
The Alcator C-MOD tokamak began operation in late March, 1992. The terminal of one of the poloidal field magnets, located inside the TF magnet, suffered a structural failure in routine service on April 10, 1992. The protective systems functioned as designed and there was virtually no associated damage. Post mortem analysis showed that the terminal might well have failed at or near the design current limit, but a satisfactory explanation for the failure at 17% of design stress has not been produced. The terminal details of the failed magnet were present in 9 other PF magnets. A new design was developed and applied to all affected magnets. The replacement of the magnet terminals and restart of the experimental facility took 13 months. The terminal assembly that failed was on a relatively simple magnet with only moderately high applied loads. The toroidal field magnet and ohmic heating solenoid are both much more complex and subjected to higher loads. The TF magnet, for example, utilizes an innovative sliding joint construction method to carry loads to a massive external superstructure. Yet the TF magnet continues to perform well and is now operating routinely at over 50% of design current. This presentation will examine the factors and decisions that led to the original PF terminal design and subsequent failure. The new design will be presented, followed by a discussion of the lessons learned from this experience.
{"title":"Anatomy of the PF magnet failure in Alcator C-MOD","authors":"S. Fairfax, D. Montgomery","doi":"10.1109/FUSION.1993.518358","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518358","url":null,"abstract":"The Alcator C-MOD tokamak began operation in late March, 1992. The terminal of one of the poloidal field magnets, located inside the TF magnet, suffered a structural failure in routine service on April 10, 1992. The protective systems functioned as designed and there was virtually no associated damage. Post mortem analysis showed that the terminal might well have failed at or near the design current limit, but a satisfactory explanation for the failure at 17% of design stress has not been produced. The terminal details of the failed magnet were present in 9 other PF magnets. A new design was developed and applied to all affected magnets. The replacement of the magnet terminals and restart of the experimental facility took 13 months. The terminal assembly that failed was on a relatively simple magnet with only moderately high applied loads. The toroidal field magnet and ohmic heating solenoid are both much more complex and subjected to higher loads. The TF magnet, for example, utilizes an innovative sliding joint construction method to carry loads to a massive external superstructure. Yet the TF magnet continues to perform well and is now operating routinely at over 50% of design current. This presentation will examine the factors and decisions that led to the original PF terminal design and subsequent failure. The new design will be presented, followed by a discussion of the lessons learned from this experience.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"107 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124409753","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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