Pub Date : 1991-09-30DOI: 10.1109/FUSION.1991.218704
D. Beller
Antiproton-induced fission of uranium or plutonium has been proposed as a method to assist the driver of an inertial confinement fusion (ICF) pellet or as a spark initiator. In past studies with one-dimensional radiation-hydrodynamics codes, factors that have reduced predicted yields in past ICF experiments were neglected or not discussed. In order to validate the feasibility of this concept with higher confidence, a three-phase program has been initiated. The first phase is an investigation of the theoretical aspects of antiproton-initiated fission/ICF by using more competent 2D and/or 3D codes and extensive data libraries that were not available for past studies. Next, a technology development project will include the design and construction of systems for accumulating, storing, and transporting antiprotons. Finally, three proof-of-principle implosion experiments will be conducted at the Shiva Star facility. An overview of this program is given, including a discussion of the justification.<>
{"title":"Feasibility of antiproton-boosted fission and inertial confinement fusion","authors":"D. Beller","doi":"10.1109/FUSION.1991.218704","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218704","url":null,"abstract":"Antiproton-induced fission of uranium or plutonium has been proposed as a method to assist the driver of an inertial confinement fusion (ICF) pellet or as a spark initiator. In past studies with one-dimensional radiation-hydrodynamics codes, factors that have reduced predicted yields in past ICF experiments were neglected or not discussed. In order to validate the feasibility of this concept with higher confidence, a three-phase program has been initiated. The first phase is an investigation of the theoretical aspects of antiproton-initiated fission/ICF by using more competent 2D and/or 3D codes and extensive data libraries that were not available for past studies. Next, a technology development project will include the design and construction of systems for accumulating, storing, and transporting antiprotons. Finally, three proof-of-principle implosion experiments will be conducted at the Shiva Star facility. An overview of this program is given, including a discussion of the justification.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"7 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":"127877856","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.1991.218840
R. Bourque
It is shown that a rotating reactor chamber with a liquid blanket can be designed with sufficient turbulence in the free surface to remove the heat generated by target explosions. Because viscosity, not thermal conductivity, limits heat removal, both lithium-lead and Flibe can be used. Although LiPb has a fairly high rotational stored energy, it translates into a moderate dynamic pressure and thermalized temperature rise. The stored energy in Flibe is quite low and poses no problem. A rough cost estimate shows that the primary reactor components are not very expensive. Flibe has the advantages of low activation and lower atomic number constituents than LiPb, increasing the allowable chamber pressure for heavy ion beam propagation. However, it is a poor heat transfer medium and must rely on higher levels of turbulence than LiPb to avoid overheating of the liquid surface. The LIFE (liner inertial fusion energy) reactor concept appears to be an improvement over the Cascade concept in that granule refurbishing is eliminated, the blanket is self-healing, and heat transfer to the power conversion system is by convection rather than radiation. One disadvantage is the lower operating temperature compared to the ceramic granules in Cascade. Nevertheless, power conversion efficiencies in the 45% range can be expected.<>
{"title":"ICF reactor chambers with rotating liquid blankets","authors":"R. Bourque","doi":"10.1109/FUSION.1991.218840","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218840","url":null,"abstract":"It is shown that a rotating reactor chamber with a liquid blanket can be designed with sufficient turbulence in the free surface to remove the heat generated by target explosions. Because viscosity, not thermal conductivity, limits heat removal, both lithium-lead and Flibe can be used. Although LiPb has a fairly high rotational stored energy, it translates into a moderate dynamic pressure and thermalized temperature rise. The stored energy in Flibe is quite low and poses no problem. A rough cost estimate shows that the primary reactor components are not very expensive. Flibe has the advantages of low activation and lower atomic number constituents than LiPb, increasing the allowable chamber pressure for heavy ion beam propagation. However, it is a poor heat transfer medium and must rely on higher levels of turbulence than LiPb to avoid overheating of the liquid surface. The LIFE (liner inertial fusion energy) reactor concept appears to be an improvement over the Cascade concept in that granule refurbishing is eliminated, the blanket is self-healing, and heat transfer to the power conversion system is by convection rather than radiation. One disadvantage is the lower operating temperature compared to the ceramic granules in Cascade. Nevertheless, power conversion efficiencies in the 45% range can be expected.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"18 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":"123347230","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.1991.218889
F. Alladio, E. Barbato, Rosario Bartiromo, A. Cardinali, F. D. De Marco, C. Ferro, M. Gasparotto, L. Lovisetto, A. Mancuso, P. Micozzi, N. Mitchell, F. Orsitto, L. Pieroni, A. Pizzuto, M. Roccella, F. Romanelli, E. Salpietro, A. Tanga
The DIOSCUR tokamak is a superconducting machine using Nb/sub 3/Sn CIC coils and having a size similar to JET (Joint European Torus), although with a toroidal magnetic field of 7 T and a sustainable plasma current of 10 MA for long pulses (1000 s). The machine will have a divertor and great configurational flexibility in plasma aspect ratio up to A=5.2, current, up to 18 MA, and attainable poloidal beta , to allow the study of stationary regimes and also the influence of bootstrap current. For CD and profile control, 50 MW of LH and negative ion beam power are foreseen. ICRF (ion cyclotron resonant frequency) heating and fast wave current drive have also been considered.<>
{"title":"DIOSCUR-divertor optimization and steady current study of a tokamak aimed at steady state operation with reactor relevant plasma parameters","authors":"F. Alladio, E. Barbato, Rosario Bartiromo, A. Cardinali, F. D. De Marco, C. Ferro, M. Gasparotto, L. Lovisetto, A. Mancuso, P. Micozzi, N. Mitchell, F. Orsitto, L. Pieroni, A. Pizzuto, M. Roccella, F. Romanelli, E. Salpietro, A. Tanga","doi":"10.1109/FUSION.1991.218889","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218889","url":null,"abstract":"The DIOSCUR tokamak is a superconducting machine using Nb/sub 3/Sn CIC coils and having a size similar to JET (Joint European Torus), although with a toroidal magnetic field of 7 T and a sustainable plasma current of 10 MA for long pulses (1000 s). The machine will have a divertor and great configurational flexibility in plasma aspect ratio up to A=5.2, current, up to 18 MA, and attainable poloidal beta , to allow the study of stationary regimes and also the influence of bootstrap current. For CD and profile control, 50 MW of LH and negative ion beam power are foreseen. ICRF (ion cyclotron resonant frequency) heating and fast wave current drive have also been considered.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"57 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":"125712016","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.1991.218858
L. Galbiati, T. Raimondi, P. Garetti, G. Costi
The telemanipulator developed for maintenance of the JET (Joint European Torus) tokamak (Mascot 4) is a microprocessor-controlled unit based on bilateral position servosystems. The main objective of this type of force-feedback servomanipulator is to give the operator, as nearly as possible, the tactile sensations of actually doing the job. Servomanipulator sensitivity, stiffness, time response and low reflected inertia characteristics are achieved by using low-friction AC motors, high-resolution R/D converters, fast sampling time, and short delays in the signal transmissions. There is a 16-bit Z8001 CPU and a servo amplifier for each servo system, and a global CPU with 16+16 I/O for general supervision. For the advanced function an auxiliary CPU with 256 Kb of RAM is used. Communication between master and slave is via a high-speed full duplex serial link at 1 MHz. The digital system has permitted a reduction of the cabling required and the introduction of advanced aids to the operator: teach and repeat, tool weight compensation, and constraint of the trajectory on given planes or lines.<>
{"title":"Control and operational aspects of the Mascot 4 force feedback servomanipulator of JET","authors":"L. Galbiati, T. Raimondi, P. Garetti, G. Costi","doi":"10.1109/FUSION.1991.218858","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218858","url":null,"abstract":"The telemanipulator developed for maintenance of the JET (Joint European Torus) tokamak (Mascot 4) is a microprocessor-controlled unit based on bilateral position servosystems. The main objective of this type of force-feedback servomanipulator is to give the operator, as nearly as possible, the tactile sensations of actually doing the job. Servomanipulator sensitivity, stiffness, time response and low reflected inertia characteristics are achieved by using low-friction AC motors, high-resolution R/D converters, fast sampling time, and short delays in the signal transmissions. There is a 16-bit Z8001 CPU and a servo amplifier for each servo system, and a global CPU with 16+16 I/O for general supervision. For the advanced function an auxiliary CPU with 256 Kb of RAM is used. Communication between master and slave is via a high-speed full duplex serial link at 1 MHz. The digital system has permitted a reduction of the cabling required and the introduction of advanced aids to the operator: teach and repeat, tool weight compensation, and constraint of the trajectory on given planes or lines.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"24 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":"126085891","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.1991.218691
K. Ioki, K. Namiki, Y. Suzuki, T. Shimomura, H. Kaguchi, K. Ue, T. Ando, M. Yamamoto, Y. Neyatani, H. Ninomiya, M. Matsukawa, H. Horiike
The divertor plates have been designed and fabricated for JT-60U (JAERI Tokamak-60 upgrade). The bottom part of the vacuum vessel is fully covered with the divertor plates. The divertor plates are subject to a heat load of 20 MW/m/sup 2/ for 5 s in a 30-mm-wide band. The divertor plates consist of 125 base plates with cooling tubes. Each divertor base plate is held on the vessel wall by a fixed support and sliding supports in order to accommodate thermal expansion differences and withstand magnetic forces. The divertor base plates have been designed to take up as little space as possible in order to gain larger plasma volume. High dimensional and positional accuracy was obtained successfully in the divertor base plates fabricated and installed into the JT-60U vacuum vessel. The achieved accuracy of the base-plate height was within 0.6 mm for neighboring base plates.<>
{"title":"Design and fabrication of divertor base-plate for JT-60U","authors":"K. Ioki, K. Namiki, Y. Suzuki, T. Shimomura, H. Kaguchi, K. Ue, T. Ando, M. Yamamoto, Y. Neyatani, H. Ninomiya, M. Matsukawa, H. Horiike","doi":"10.1109/FUSION.1991.218691","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218691","url":null,"abstract":"The divertor plates have been designed and fabricated for JT-60U (JAERI Tokamak-60 upgrade). The bottom part of the vacuum vessel is fully covered with the divertor plates. The divertor plates are subject to a heat load of 20 MW/m/sup 2/ for 5 s in a 30-mm-wide band. The divertor plates consist of 125 base plates with cooling tubes. Each divertor base plate is held on the vessel wall by a fixed support and sliding supports in order to accommodate thermal expansion differences and withstand magnetic forces. The divertor base plates have been designed to take up as little space as possible in order to gain larger plasma volume. High dimensional and positional accuracy was obtained successfully in the divertor base plates fabricated and installed into the JT-60U vacuum vessel. The achieved accuracy of the base-plate height was within 0.6 mm for neighboring base plates.<<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":"126691875","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.1991.218813
Zengsi Guo, Wei-Hau Wu, W. Xu, Yue Chen
The results of the EHR (Experimental Hybrid Reactor) magnet conceptual design are presented. The EHR magnet systems include the toroidal field (TF) magnet system, the poloidal field (PF) magnet system, and the in-vessel plasma active control coils. The TF and PF magnet systems are superconductive to satisfy the steady state operation of the EHR. The plasma active control coils are water-cooled normal copper coils. The TF and PF magnet systems are designed to be semipermanent, i.e. they are not expected to be repaired or replaced during the life time of the machine, because they are the core of the basic machine. Thus, the problem of achieving high reliability is a key problem of the magnet design. It is concluded that the magnet system concept design meets the requirements specified in the EHR terms of reference.<>
{"title":"The EHR magnet system conceptual design","authors":"Zengsi Guo, Wei-Hau Wu, W. Xu, Yue Chen","doi":"10.1109/FUSION.1991.218813","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218813","url":null,"abstract":"The results of the EHR (Experimental Hybrid Reactor) magnet conceptual design are presented. The EHR magnet systems include the toroidal field (TF) magnet system, the poloidal field (PF) magnet system, and the in-vessel plasma active control coils. The TF and PF magnet systems are superconductive to satisfy the steady state operation of the EHR. The plasma active control coils are water-cooled normal copper coils. The TF and PF magnet systems are designed to be semipermanent, i.e. they are not expected to be repaired or replaced during the life time of the machine, because they are the core of the basic machine. Thus, the problem of achieving high reliability is a key problem of the magnet design. It is concluded that the magnet system concept design meets the requirements specified in the EHR terms of reference.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"76 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":"121108630","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.1991.218752
T. Fuller, C. Fiore
The challenge of providing an integrated health and safety program for an advanced fusion facility within the university atmosphere is addressed in the environment, health, and safety program design for the Alcator C-MOD project at MIT. The hazards common to all fusion experiments, such as radiation, cryogenics, high power and voltage, confined spaces, hazardous chemicals, and heavy mechanical and machine equipment, must be controlled in an acceptable manner with the coordination of university and facility programs. The development of this program is detailed. The combination of programs, procedures, equipment, and training described includes effective control and coordination of activities when different types of occupational hazards occur on one job such as a confined entry into a radiologically contaminated area or during machining of potentially contaminated materials. Procedures provide a means for documentation of required activities and a means to ensure that checks and controls are in place before work begins at a hazardous job site.<>
{"title":"Alcator C-MOD environmental health and radiation program development","authors":"T. Fuller, C. Fiore","doi":"10.1109/FUSION.1991.218752","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218752","url":null,"abstract":"The challenge of providing an integrated health and safety program for an advanced fusion facility within the university atmosphere is addressed in the environment, health, and safety program design for the Alcator C-MOD project at MIT. The hazards common to all fusion experiments, such as radiation, cryogenics, high power and voltage, confined spaces, hazardous chemicals, and heavy mechanical and machine equipment, must be controlled in an acceptable manner with the coordination of university and facility programs. The development of this program is detailed. The combination of programs, procedures, equipment, and training described includes effective control and coordination of activities when different types of occupational hazards occur on one job such as a confined entry into a radiologically contaminated area or during machining of potentially contaminated materials. Procedures provide a means for documentation of required activities and a means to ensure that checks and controls are in place before work begins at a hazardous job site.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"16 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":"121653214","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.1991.218929
R. M. Mayo, F. Wysocki, J. Fernández
The authors apply the results of the edge helicity dissipation model (see T.R. Jarboe and B. Alper, 1987) to determine the relative magnitudes of electric fields during relaxation in spheromaks. This is achieved by quantifying the average electric field in the plasma edge region generated by flux decay and relaxation mechanism(s). It is shown that relaxation electric fields can be as much as three times the flux decay field in the edge. The model also correctly predicts no relaxation electric field when the spheromak is a cold, purely resistively decaying object. In addition, the model provides an estimate for the quantity of magnetic decay power from relaxation, which can be as much as 75% of the total decay power.<>
{"title":"Relaxation electric fields in spheromaks","authors":"R. M. Mayo, F. Wysocki, J. Fernández","doi":"10.1109/FUSION.1991.218929","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218929","url":null,"abstract":"The authors apply the results of the edge helicity dissipation model (see T.R. Jarboe and B. Alper, 1987) to determine the relative magnitudes of electric fields during relaxation in spheromaks. This is achieved by quantifying the average electric field in the plasma edge region generated by flux decay and relaxation mechanism(s). It is shown that relaxation electric fields can be as much as three times the flux decay field in the edge. The model also correctly predicts no relaxation electric field when the spheromak is a cold, purely resistively decaying object. In addition, the model provides an estimate for the quantity of magnetic decay power from relaxation, which can be as much as 75% of the total decay power.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"45 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":"125136219","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.1991.218825
C. Challis, A. Bickley, A. Browne, H. D. de Esch, M. Fogg, T. Jones, D. Stork, L. Svensson
The JET (Joint European Torus) neutral beam injection (NBI) system, has demonstrated a very high degree of reliability and availability. Data on the reliability and availability of the beamlines and the associated subsystems (e.g. power supplies, cooling and cryogenic supplies, computerized control, etc.) have been recorded over the past three years. These data are used to quantify the overall reliability of the NBI systems and to identify problem areas. The overall reliability (energy injected into JET/energy requested) and availability for the 1990 experimental campaign are both better than 80% and improve significantly during routine operation to nearly 90% after the system has been fully commissioned.<>
JET (Joint European Torus)中性束注入(NBI)系统已经证明了非常高的可靠性和可用性。过去三年记录了有关光束线和相关子系统(例如电源、冷却和低温电源、计算机控制等)的可靠性和可用性的数据。这些数据用于量化NBI系统的整体可靠性，并确定问题区域。总体可靠性(注入JET的能量/要求的能量)和1990年试验活动的可用性都优于80%，并且在系统完全投入使用后，在常规操作期间显着提高到接近90%。
{"title":"Reliability study of the JET neutral injection system","authors":"C. Challis, A. Bickley, A. Browne, H. D. de Esch, M. Fogg, T. Jones, D. Stork, L. Svensson","doi":"10.1109/FUSION.1991.218825","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218825","url":null,"abstract":"The JET (Joint European Torus) neutral beam injection (NBI) system, has demonstrated a very high degree of reliability and availability. Data on the reliability and availability of the beamlines and the associated subsystems (e.g. power supplies, cooling and cryogenic supplies, computerized control, etc.) have been recorded over the past three years. These data are used to quantify the overall reliability of the NBI systems and to identify problem areas. The overall reliability (energy injected into JET/energy requested) and availability for the 1990 experimental campaign are both better than 80% and improve significantly during routine operation to nearly 90% after the system has been fully commissioned.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"23 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":"122816288","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.1991.218737
B. Yuan, B. Jiao, K. Yang, Q. Jin, Y. Jiang, M. Tan, S. Wang
A feedback control system is installed in the HL-1 tokamak with a thick copper shell (d=5 cm). The control field coils outside the copper shell are fed by a six-pulse thyristor rectifier. The power supply is operated by a pulse-width-modulated regulator which consists of a dual-mode circuit without dead zone and a PID circuit. The operation on the HL-1 shows that the horizontal and vertical plasma positions can be controlled to +or-2 mm.<>
{"title":"Feedback control of the plasma position in HL-1 tokamak","authors":"B. Yuan, B. Jiao, K. Yang, Q. Jin, Y. Jiang, M. Tan, S. Wang","doi":"10.1109/FUSION.1991.218737","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218737","url":null,"abstract":"A feedback control system is installed in the HL-1 tokamak with a thick copper shell (d=5 cm). The control field coils outside the copper shell are fed by a six-pulse thyristor rectifier. The power supply is operated by a pulse-width-modulated regulator which consists of a dual-mode circuit without dead zone and a PID circuit. The operation on the HL-1 shows that the horizontal and vertical plasma positions can be controlled to +or-2 mm.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"86 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":"121274296","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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