Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518527
I. Anashkin, V.V. Asmalovksy, A. N. Vertiporokh, D. Ivanov, I. A. Posadsky, P. Khvostenko
The paper is devoted to the experimental results obtained in 1992-1993 on the T-15 superconducting toroidal field coil with a stored energy of 380 MJ using Nb/sub 3/Sn. The critical current dependence on the temperature in the range T/sub in/=4.5-12 K was obtained. The current 4.0 kA was achieved, which corresponds to a magnetic field at the axis of 3.6 T and a peak field at the conductor of 6.5 T. The level of the resistive heat load versus transport current was studied-Q/sub res/ achieves 300 W at 3.6 kA. It was shown that plasma current disruptions with a rate up to 80 MA/s did not affect the coil stability.
{"title":"Status and test results of the Tokamak-15 superconducting toroidal field coil","authors":"I. Anashkin, V.V. Asmalovksy, A. N. Vertiporokh, D. Ivanov, I. A. Posadsky, P. Khvostenko","doi":"10.1109/FUSION.1993.518527","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518527","url":null,"abstract":"The paper is devoted to the experimental results obtained in 1992-1993 on the T-15 superconducting toroidal field coil with a stored energy of 380 MJ using Nb/sub 3/Sn. The critical current dependence on the temperature in the range T/sub in/=4.5-12 K was obtained. The current 4.0 kA was achieved, which corresponds to a magnetic field at the axis of 3.6 T and a peak field at the conductor of 6.5 T. The level of the resistive heat load versus transport current was studied-Q/sub res/ achieves 300 W at 3.6 kA. It was shown that plasma current disruptions with a rate up to 80 MA/s did not affect the coil stability.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129356391","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":null,"pages":null},"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.1993.518350
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":null,"pages":null},"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.1993.518300
W. Stark, V.J. Abraitis, S. Davis, R. Simmons, S.M. Young, C. Flanagan
The Tokamak Physics Experiment (TPX) and the International Thermonuclear Engineering Reactor (ITER) are the major national design efforts currently underway by the US fusion program. These design efforts are being performed at many different sites consisting of national laboratories, universities, and industrial partners. This paper presents the current status of work in support of expediting communication and enhancing the collaborative process by providing a framework of common tools and resources for all collaborators. The key components of this framework are common applications which support desktop mail, directory services, and convenient desktop access to project design documents. A discussion of the issues associated with the deployment, support, and future evolution of this framework will be presented.
{"title":"Tools and frameworks supporting TPX and ITER design collaborations","authors":"W. Stark, V.J. Abraitis, S. Davis, R. Simmons, S.M. Young, C. Flanagan","doi":"10.1109/FUSION.1993.518300","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518300","url":null,"abstract":"The Tokamak Physics Experiment (TPX) and the International Thermonuclear Engineering Reactor (ITER) are the major national design efforts currently underway by the US fusion program. These design efforts are being performed at many different sites consisting of national laboratories, universities, and industrial partners. This paper presents the current status of work in support of expediting communication and enhancing the collaborative process by providing a framework of common tools and resources for all collaborators. The key components of this framework are common applications which support desktop mail, directory services, and convenient desktop access to project design documents. A discussion of the issues associated with the deployment, support, and future evolution of this framework will be presented.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"121237662","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":null,"pages":null},"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":null,"pages":null},"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.1993.518539
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":null,"pages":null},"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.1993.518346
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":null,"pages":null},"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.1993.518456
H.C. Mantz, D. Bowers, J. Haines, F. Williams
Poloidal limiters are required on the Tokamak Physics Experiment (TPX) for startup and to provide protection to the RF antennas. This paper discusses the conceptual design of these limiters and the analyses associated with the design. The configuration of the limiters is described and includes the carbon-carbon (C-C) tile design and their attachment concept. The thermal response of the limiters to startup and steady state plasma conditions is discussed. The required water coolant flow rates and resulting pressure drops are presented. The structural response of the limiter to both halo and eddy current disruption conditions is also included.
{"title":"TPX poloidal limiter design","authors":"H.C. Mantz, D. Bowers, J. Haines, F. Williams","doi":"10.1109/FUSION.1993.518456","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518456","url":null,"abstract":"Poloidal limiters are required on the Tokamak Physics Experiment (TPX) for startup and to provide protection to the RF antennas. This paper discusses the conceptual design of these limiters and the analyses associated with the design. The configuration of the limiters is described and includes the carbon-carbon (C-C) tile design and their attachment concept. The thermal response of the limiters to startup and steady state plasma conditions is discussed. The required water coolant flow rates and resulting pressure drops are presented. The structural response of the limiter to both halo and eddy current disruption conditions is also included.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"128173098","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.1993.518377
R. Uhlemann, H. Reimer, M. Lochter
Neutral particle beams of the /sup 3/He isotope in the megawatt range were successfully produced by the neutral beam injectors of TEXTOR. Measurements of beam and ion source characterizing data, i.e. perveance, divergence, neutralization efficiency, beamline transmission and gas flow rates for 50 kV, 49 A, 2.5 MW, 3 sec /sup 3/He beams are compared with relevant data of hydrogen, deuterium and /sup 4/He. The LHe cryo condensation pumps are capable of pumping the resulting /sup 3/He gas flow rates of 50 mbl/s without technical modification at 4.2 K, if covered prior to each beam pulse by an amount of argon frost with proper ratio of Ar/He. Beamline pressure, reionization losses and preliminary results of /sup 3/He-injection into TEXTOR are given.
{"title":"/sup 3/He high power neutral beams for TEXTOR comparison of 55 KeV particle beam and vacuum system data with /sup 4/He, deuterium and hydrogen","authors":"R. Uhlemann, H. Reimer, M. Lochter","doi":"10.1109/FUSION.1993.518377","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518377","url":null,"abstract":"Neutral particle beams of the /sup 3/He isotope in the megawatt range were successfully produced by the neutral beam injectors of TEXTOR. Measurements of beam and ion source characterizing data, i.e. perveance, divergence, neutralization efficiency, beamline transmission and gas flow rates for 50 kV, 49 A, 2.5 MW, 3 sec /sup 3/He beams are compared with relevant data of hydrogen, deuterium and /sup 4/He. The LHe cryo condensation pumps are capable of pumping the resulting /sup 3/He gas flow rates of 50 mbl/s without technical modification at 4.2 K, if covered prior to each beam pulse by an amount of argon frost with proper ratio of Ar/He. Beamline pressure, reionization losses and preliminary results of /sup 3/He-injection into TEXTOR are given.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"131873185","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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