Pub Date : 1991-09-30DOI: 10.1109/FUSION.1991.218698
K. Yoshikawa, Y. Yamamoto, H. Toku, S. Hashimoto, N. Komoda
A novel plasma neutralizer concept for a very simple structure is presented by making use of magnetron discharge for efficient plasma production with higher electron temperatures. Preliminary experimental results show electron temperatures as high as 20 eV, and 8*10/sup 11/ cm/sup -3/ electron density in an argon plasma, which are both promising for the plasma neutralizer. The effects of magnetic fields on the beam deflection are found negligible for extremely energetic beams like the 1.3-MeV D/sup -/ beams for ITER (International Thermonuclear Experimental Reactor). Issues for further improvements of plasma parameters are discussed.<>
{"title":"A plasma neutralizer by use of magnetron discharge","authors":"K. Yoshikawa, Y. Yamamoto, H. Toku, S. Hashimoto, N. Komoda","doi":"10.1109/FUSION.1991.218698","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218698","url":null,"abstract":"A novel plasma neutralizer concept for a very simple structure is presented by making use of magnetron discharge for efficient plasma production with higher electron temperatures. Preliminary experimental results show electron temperatures as high as 20 eV, and 8*10/sup 11/ cm/sup -3/ electron density in an argon plasma, which are both promising for the plasma neutralizer. The effects of magnetic fields on the beam deflection are found negligible for extremely energetic beams like the 1.3-MeV D/sup -/ beams for ITER (International Thermonuclear Experimental Reactor). Issues for further improvements of plasma parameters are discussed.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"19 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":"125032977","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 1991-09-30DOI: 10.1109/FUSION.1991.218859
S. Mills, A. Loving, M. Irving
The authors report on remote handling tools which have been specifically designed to meet requirements for pipe cutting at JET (Joint European Torus). The principal requirements were the quality of cut necessary for rewelding, effective swarf removal, and compactness for remote handling. The designs of tools had to be compatible with the severe access restrictions imposed by the JET machine. The processes used by the tools are sawing from the inside and outside of pipes, and orbital lathe for larger pipes. Special features were created on the pipes to facilitate tool location. The blade and toolbit designs have evolved to optimize cutting forces and tool durability. Satisfactory reliability has been achieved by performing 200 h of cutting during the two year period of development. Subsequently, over 100 'hands-on' cutting operations have been made on the JET machine since 1988 and a further 150 cuts are planned for 1992. Using a programmable controller the feed rate can be changed throughout the cutting operation into a predetermined way, thereby optimizing the tools' efficiency.<>
{"title":"The design, development and use of pipe cutting tools for remote handling in JET","authors":"S. Mills, A. Loving, M. Irving","doi":"10.1109/FUSION.1991.218859","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218859","url":null,"abstract":"The authors report on remote handling tools which have been specifically designed to meet requirements for pipe cutting at JET (Joint European Torus). The principal requirements were the quality of cut necessary for rewelding, effective swarf removal, and compactness for remote handling. The designs of tools had to be compatible with the severe access restrictions imposed by the JET machine. The processes used by the tools are sawing from the inside and outside of pipes, and orbital lathe for larger pipes. Special features were created on the pipes to facilitate tool location. The blade and toolbit designs have evolved to optimize cutting forces and tool durability. Satisfactory reliability has been achieved by performing 200 h of cutting during the two year period of development. Subsequently, over 100 'hands-on' cutting operations have been made on the JET machine since 1988 and a further 150 cuts are planned for 1992. Using a programmable controller the feed rate can be changed throughout the cutting operation into a predetermined way, thereby optimizing the tools' efficiency.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"26 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1991-09-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"131836781","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 1991-09-30DOI: 10.1109/FUSION.1991.218914
F. Najmabadi, R. Conn, C. Bathke, J. Blanchard, L. Bromberg, J. Brooks, E. Cheng, D. Cohn, D. Ehst, L. El-Guebaly, G. Emmert, T. Dolan, P. Gierszewski, S. Grotz, M.S. Hasan, J. Herring, S. K. Ho, A. Hollies, J. Holmes, E. Ibrahim, S. Jardin, C. Kessel, H. Khater, R. Krakowski, G.L. Kuleinski, J. Mandrekas, T. Mau, G. Miley, R.L. Miller, E. Mogahed, E. Reis, J. Santarius, M. Sawan, J. Schultz, K. Schultz, S. Sharafat, D. Steiner, D. Strickler, I. Sviatoslavsky, D. Sze, P. Titus, M. Valenti, K. Werley, J. H. Whealton, J.E.C. Williams, L. Wittenberg, C. Wong
A description of the ARIES-III research effort is presented, and the general features of the ARIES-III reactor are described. The plasma engineering and fusion-power-core design are summarized, including the major results, the key technical issues, and the central conclusions. Analyses have shown that the plasma power-balance window for D-/sup 3/He tokamak reactors is small and requires a first wall (or coating) that is highly reflective to synchrotron radiation and small values of tau /sub ash// epsilon /sub e/ (the ratio of ash-particle to energy confinement times in the core plasma). Both first and second stability regimes of operation have been considered. The second stability regime is chosen for the ARIES-III design point because the reactor can operate at a higher value of tau /sub ash// tau /sub E// tau /sub E/ approximately=2 (twice that of a first stability version), and because it has a reduced plasma current (30 MA), magnetic field at the coil (14 T), mass, and cost (also compared to a first-stability D-/sup 3/He reactor). The major and minor radii are, respectively 7.5 and 2.5 m.<>
{"title":"The ARIES-III D-3He tokamak-reactor study","authors":"F. Najmabadi, R. Conn, C. Bathke, J. Blanchard, L. Bromberg, J. Brooks, E. Cheng, D. Cohn, D. Ehst, L. El-Guebaly, G. Emmert, T. Dolan, P. Gierszewski, S. Grotz, M.S. Hasan, J. Herring, S. K. Ho, A. Hollies, J. Holmes, E. Ibrahim, S. Jardin, C. Kessel, H. Khater, R. Krakowski, G.L. Kuleinski, J. Mandrekas, T. Mau, G. Miley, R.L. Miller, E. Mogahed, E. Reis, J. Santarius, M. Sawan, J. Schultz, K. Schultz, S. Sharafat, D. Steiner, D. Strickler, I. Sviatoslavsky, D. Sze, P. Titus, M. Valenti, K. Werley, J. H. Whealton, J.E.C. Williams, L. Wittenberg, C. Wong","doi":"10.1109/FUSION.1991.218914","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218914","url":null,"abstract":"A description of the ARIES-III research effort is presented, and the general features of the ARIES-III reactor are described. The plasma engineering and fusion-power-core design are summarized, including the major results, the key technical issues, and the central conclusions. Analyses have shown that the plasma power-balance window for D-/sup 3/He tokamak reactors is small and requires a first wall (or coating) that is highly reflective to synchrotron radiation and small values of tau /sub ash// epsilon /sub e/ (the ratio of ash-particle to energy confinement times in the core plasma). Both first and second stability regimes of operation have been considered. The second stability regime is chosen for the ARIES-III design point because the reactor can operate at a higher value of tau /sub ash// tau /sub E// tau /sub E/ approximately=2 (twice that of a first stability version), and because it has a reduced plasma current (30 MA), magnetic field at the coil (14 T), mass, and cost (also compared to a first-stability D-/sup 3/He reactor). The major and minor radii are, respectively 7.5 and 2.5 m.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"11 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1991-09-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"133838970","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 1991-09-30DOI: 10.1109/FUSION.1991.218810
J. Alonso, M. Blaumoser
The TJ-II is a flexible heliac under construction, to be mounted at the Euratom/Ciemat Association Laboratory in Madrid, Spain. The machine can explore different magnetic configurations (mainly with different values of the rotational transform) by means of the adjustment of the currents in the coils. The hard core (HC) is one of the main components of the device and it constitutes what can be called the most critical part of the machine, since its close proximity to the plasma places on it the requirement of strict tolerances. The authors describe the engineering design features of the HC, the main characteristics, and the design details An overview of the manufacturing methods is also presented.<>
{"title":"Design and feasibility of the TJ-II hard core","authors":"J. Alonso, M. Blaumoser","doi":"10.1109/FUSION.1991.218810","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218810","url":null,"abstract":"The TJ-II is a flexible heliac under construction, to be mounted at the Euratom/Ciemat Association Laboratory in Madrid, Spain. The machine can explore different magnetic configurations (mainly with different values of the rotational transform) by means of the adjustment of the currents in the coils. The hard core (HC) is one of the main components of the device and it constitutes what can be called the most critical part of the machine, since its close proximity to the plasma places on it the requirement of strict tolerances. The authors describe the engineering design features of the HC, the main characteristics, and the design details An overview of the manufacturing methods is also presented.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"25 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1991-09-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"116562947","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 1991-09-30DOI: 10.1109/FUSION.1991.218766
R. Peterson, R. Englestad, G. Kulcinski, E. Lovell, J. Macfarlane, E. Mogahed, G. Moses, S. Rutledge, M. Sawan, I. Sviatoslavsky, G. Sviatoslavsky, L. Wittenberg
LIBRA-LiTE is a conceptual design for a 1300-MWe power plant using light ion inertial fusion. LIBRA-LiTE differs from the LIBRA design in its use of ballistically focused light ions to drive the target. Focusing magnets are positioned 2.05 m from the target, which, to mitigate neutron damage effects, has required a novel magnet design using liquid lithium as a conductor. A sacrificial film of liquid lithium protects the magnets, the target chamber side walls and bottom from the X-rays and debris released by the target microexplosion. The target neutrons deposit in a tritrium breeding blanket of liquid lithium confined to woven metal tubes on the sides and in a pool on the bottom. The top of the target chamber is a metallic dome removed far enough (16 m) from the target to be a lifetime component.<>
{"title":"LIBRA-LiTE, a light ion inertial confinement fusion reactor with ballistic ion propagation","authors":"R. Peterson, R. Englestad, G. Kulcinski, E. Lovell, J. Macfarlane, E. Mogahed, G. Moses, S. Rutledge, M. Sawan, I. Sviatoslavsky, G. Sviatoslavsky, L. Wittenberg","doi":"10.1109/FUSION.1991.218766","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218766","url":null,"abstract":"LIBRA-LiTE is a conceptual design for a 1300-MWe power plant using light ion inertial fusion. LIBRA-LiTE differs from the LIBRA design in its use of ballistically focused light ions to drive the target. Focusing magnets are positioned 2.05 m from the target, which, to mitigate neutron damage effects, has required a novel magnet design using liquid lithium as a conductor. A sacrificial film of liquid lithium protects the magnets, the target chamber side walls and bottom from the X-rays and debris released by the target microexplosion. The target neutrons deposit in a tritrium breeding blanket of liquid lithium confined to woven metal tubes on the sides and in a pool on the bottom. The top of the target chamber is a metallic dome removed far enough (16 m) from the target to be a lifetime component.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"355 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":"115168777","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 1991-09-30DOI: 10.1109/FUSION.1991.218916
S. K. Ho
An alternative fueling scheme for ARIES-III using accelerated compact-toroids (CT) is studied. The methodology of L.J. Perkins et al. (1988) for modeling CT penetration and deposition is followed. The calculations are performed using a self-consistent, radial-zoning scheme which includes several interrelated constraints such as CT ring decay, tilting, field-line reconnection, deceleration in the external field gradient, and ring expansion/contraction. The CT injection parameters optimized for fueling of ARIES-III are presented. Advantages of CT fueling with respect to other aspects of tokamak operations and uncertainties in the CT injection modeling are also discussed. From the conservative upper-limit estimation, CTs of 40 mg with 40 cm diameter can deposit fuel directly into the center and half-way of the plasma, requiring 49 MW and 14 MW, respectively.<>
{"title":"Controlled central fueling for ARIES-III by compact-toroid injection","authors":"S. K. Ho","doi":"10.1109/FUSION.1991.218916","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218916","url":null,"abstract":"An alternative fueling scheme for ARIES-III using accelerated compact-toroids (CT) is studied. The methodology of L.J. Perkins et al. (1988) for modeling CT penetration and deposition is followed. The calculations are performed using a self-consistent, radial-zoning scheme which includes several interrelated constraints such as CT ring decay, tilting, field-line reconnection, deceleration in the external field gradient, and ring expansion/contraction. The CT injection parameters optimized for fueling of ARIES-III are presented. Advantages of CT fueling with respect to other aspects of tokamak operations and uncertainties in the CT injection modeling are also discussed. From the conservative upper-limit estimation, CTs of 40 mg with 40 cm diameter can deposit fuel directly into the center and half-way of the plasma, requiring 49 MW and 14 MW, respectively.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"142 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1991-09-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"116570817","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 1991-09-30DOI: 10.1109/FUSION.1991.218709
F. McClintock, J. Feng, R. Vieira
An uncertainty analysis combines the resulting prediction limits with estimated effects of a number of factors, of which the most important are: the statistical variability and limited number of specimens, the largest undetected crack length, the plate-to-plate variability, and the allowance for oversights. The total effect of these (and many more minor) effects is to reduce the allowable stress, for chosen odds against fracture of 10000 to 1, to about 2/3 of the central statistical value. The analysis highlights the factors needing further study and illustrates the value of statistical and uncertainty analyses, which should be combined with cost-benefit and fault-free analyses to complement code-based design.<>
{"title":"Using statistical and uncertainty analyses in design, applied to a tokamak central solenoid","authors":"F. McClintock, J. Feng, R. Vieira","doi":"10.1109/FUSION.1991.218709","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218709","url":null,"abstract":"An uncertainty analysis combines the resulting prediction limits with estimated effects of a number of factors, of which the most important are: the statistical variability and limited number of specimens, the largest undetected crack length, the plate-to-plate variability, and the allowance for oversights. The total effect of these (and many more minor) effects is to reduce the allowable stress, for chosen odds against fracture of 10000 to 1, to about 2/3 of the central statistical value. The analysis highlights the factors needing further study and illustrates the value of statistical and uncertainty analyses, which should be combined with cost-benefit and fault-free analyses to complement code-based design.<<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":"115541164","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.1991.218703
T. Wade, J. Jacquinot, G. Bosia, A. Sibley, M. Schmid
In excess of 22 MW of ion cyclotron resonance heating at the JET (Joint European Torus) plasma center has been achieved with increased plant availability by changes to the plant and its control and protection philosophy. Nested feedback systems, including RF control of the plasma position, enable power to be sustained throughout L mode to H mode fast plasma load variations. Modified plasma sawtooth behavior has been obtained by pi /2 phasing of the antenna currents. These control developments and the proposed fast wave current drive experiments with the new four-strap, JET ICRH antennas will be relevant to ITER (International Thermonuclear Experimental Reactor) and next step devices.<>
{"title":"High power (22 MW) ICRH at JET and developments for next step devices","authors":"T. Wade, J. Jacquinot, G. Bosia, A. Sibley, M. Schmid","doi":"10.1109/FUSION.1991.218703","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218703","url":null,"abstract":"In excess of 22 MW of ion cyclotron resonance heating at the JET (Joint European Torus) plasma center has been achieved with increased plant availability by changes to the plant and its control and protection philosophy. Nested feedback systems, including RF control of the plasma position, enable power to be sustained throughout L mode to H mode fast plasma load variations. Modified plasma sawtooth behavior has been obtained by pi /2 phasing of the antenna currents. These control developments and the proposed fast wave current drive experiments with the new four-strap, JET ICRH antennas will be relevant to ITER (International Thermonuclear Experimental Reactor) and next step devices.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"116 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":"121352815","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.1991.218794
R. Peterson, J. Macfarlane, P. Wang
The authors have investigated the target chamber designs for two IFE (inertial-confinement fusion energy) reactors (SOMBRERO and OSIRIS). The CONRAD computer code has been used to analyze certain critical aspects of these designs. Auto-neutralized transport is considered and a gas density is used that precludes protection of the first surface of the target chamber from X-rays and ions. The dominant issue in the design of the SOMBRERO laser fusion target chamber is the reradiation of absorbed target energy from the gas to the wall of the target chamber. In the OSIRIS heavy ion fusion target chamber, vaporization of material from the wall is the most important consideration. In SOMBRERO, 0.5 torr of xenon gas should allow beam transport and will protect the graphite wall vaporization by target energy. In OSIRIS, it was found that the FLIBE is vaporized and that a high peak pressure but moderate impulse shock reaches the vapor/liquid interface in the FLIBE.<>
{"title":"Target chamber gas response and vaporization in a laser and a heavy ion beam IFE reactor","authors":"R. Peterson, J. Macfarlane, P. Wang","doi":"10.1109/FUSION.1991.218794","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218794","url":null,"abstract":"The authors have investigated the target chamber designs for two IFE (inertial-confinement fusion energy) reactors (SOMBRERO and OSIRIS). The CONRAD computer code has been used to analyze certain critical aspects of these designs. Auto-neutralized transport is considered and a gas density is used that precludes protection of the first surface of the target chamber from X-rays and ions. The dominant issue in the design of the SOMBRERO laser fusion target chamber is the reradiation of absorbed target energy from the gas to the wall of the target chamber. In the OSIRIS heavy ion fusion target chamber, vaporization of material from the wall is the most important consideration. In SOMBRERO, 0.5 torr of xenon gas should allow beam transport and will protect the graphite wall vaporization by target energy. In OSIRIS, it was found that the FLIBE is vaporized and that a high peak pressure but moderate impulse shock reaches the vapor/liquid interface in the FLIBE.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"13 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":"124129561","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.1991.218653
E. Frattolillo, S. Migliori, F. Mirizzi, F. Scaramuzzi
A diagnostic apparatus has been designed and tested, giving mass and speed measurements and in-flight pictures of the solid D/sub 2/ pellets for the high-speed (up to 3.3 km/s) single-pellet injector (SPIN) to be installed on the Frascati Tokamak Upgrade, (FTU), in order to completely characterize each of the launched pellets. A detailed description of the apparatus and of experimental results obtained with a prototype injector is given. From the point of view of speed determination, the system gives reliable results. Some problems have been detected on the volume (mass) determination, because the inner surfaces of the cavities have not been accurately treated for vacuum since they are prototypes. A new set of cavities is under construction and the whole setup will be improved according to what the experience with the prototypes suggests. Good-quality in-flight pictures of the pellets are reliably obtained by means of a video recording apparatus and a 20-ns pulsed flash lamp.<>
{"title":"Diagnostics of the high-speed single-pellet injector for the Frascati Tokamak Upgrade","authors":"E. Frattolillo, S. Migliori, F. Mirizzi, F. Scaramuzzi","doi":"10.1109/FUSION.1991.218653","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218653","url":null,"abstract":"A diagnostic apparatus has been designed and tested, giving mass and speed measurements and in-flight pictures of the solid D/sub 2/ pellets for the high-speed (up to 3.3 km/s) single-pellet injector (SPIN) to be installed on the Frascati Tokamak Upgrade, (FTU), in order to completely characterize each of the launched pellets. A detailed description of the apparatus and of experimental results obtained with a prototype injector is given. From the point of view of speed determination, the system gives reliable results. Some problems have been detected on the volume (mass) determination, because the inner surfaces of the cavities have not been accurately treated for vacuum since they are prototypes. A new set of cavities is under construction and the whole setup will be improved according to what the experience with the prototypes suggests. Good-quality in-flight pictures of the pellets are reliably obtained by means of a video recording apparatus and a 20-ns pulsed flash lamp.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"13 3 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":"124263482","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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