Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518492
L. Waganer
An industry team, led by McDonnell Douglas Aerospace, has developed two inertial fusion energy reactor plant designs that are predicted to compete favorably with other energy sources. Symmetrically illuminated, direct drive targets were chosen for the KrF laser driver for better gain at a lower cost. The 4 MJ pulse energy to the target is provided by 960 medium-sized (6 kJ) electric discharge lasers. This provides redundancy, system reliability, and stringent illumination requirements. Innovative non-linear optics (NLO) design for beam combination and pulse compression converts the beams into temporally-shaped, high quality beams. The 60 beamlines are shielded to minimize the radiation to the reactor and driver buildings. Long-lived, aluminum-coated, grazing incidence metal mirrors (GIMM), supported by SiC structure, are located within 20 meters of the center of the reactor cavity. These GIMMs and the final focusing mirrors are high-speed, adaptable structures that help steer the beams to the position of the tracked target. The second reactor design uses an innovative, lower-cost, pulsed, single beamline LINAC to deliver heavy ion beams into 14 storage rings. Tailoring of the beam storage, compression, and extraction from the storage rings provides the required temporal and energy shaping. The sets of beams are divided to illuminate the indirectly-driven target from two sides with 7 MJ of energy. The beams are focused to a point on the backside of the blanket where the beams are stripped to a high charge state. This enables formation of a small diameter transport channel through the blanket and across the cavity to the target. The first wall system for both designs uses a SiC structure cooled internally and on the surface with liquid lead. The lead is evaporated and recondensed on the surface to protect and cool the first wall from the 3 to 5 Hz target explosions. The blanket is a Li/sub 2/O solid breeder, cooled with low pressure helium. The two Prometheus IFE reactor designs show that IFE has the potential to be economically competitive with very attractive safety and environmental features while maintaining a high degree of technical credibility.
{"title":"Industry cooperates to advance IFE reactor design-the results of Prometheus","authors":"L. Waganer","doi":"10.1109/FUSION.1993.518492","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518492","url":null,"abstract":"An industry team, led by McDonnell Douglas Aerospace, has developed two inertial fusion energy reactor plant designs that are predicted to compete favorably with other energy sources. Symmetrically illuminated, direct drive targets were chosen for the KrF laser driver for better gain at a lower cost. The 4 MJ pulse energy to the target is provided by 960 medium-sized (6 kJ) electric discharge lasers. This provides redundancy, system reliability, and stringent illumination requirements. Innovative non-linear optics (NLO) design for beam combination and pulse compression converts the beams into temporally-shaped, high quality beams. The 60 beamlines are shielded to minimize the radiation to the reactor and driver buildings. Long-lived, aluminum-coated, grazing incidence metal mirrors (GIMM), supported by SiC structure, are located within 20 meters of the center of the reactor cavity. These GIMMs and the final focusing mirrors are high-speed, adaptable structures that help steer the beams to the position of the tracked target. The second reactor design uses an innovative, lower-cost, pulsed, single beamline LINAC to deliver heavy ion beams into 14 storage rings. Tailoring of the beam storage, compression, and extraction from the storage rings provides the required temporal and energy shaping. The sets of beams are divided to illuminate the indirectly-driven target from two sides with 7 MJ of energy. The beams are focused to a point on the backside of the blanket where the beams are stripped to a high charge state. This enables formation of a small diameter transport channel through the blanket and across the cavity to the target. The first wall system for both designs uses a SiC structure cooled internally and on the surface with liquid lead. The lead is evaporated and recondensed on the surface to protect and cool the first wall from the 3 to 5 Hz target explosions. The blanket is a Li/sub 2/O solid breeder, cooled with low pressure helium. The two Prometheus IFE reactor designs show that IFE has the potential to be economically competitive with very attractive safety and environmental features while maintaining a high degree of technical credibility.","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":"115005503","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.1993.518321
F. Bellina, P. Campostrini, G. Chitarin, M. Guarnieri, A. Stella, F. Trevisan
The paper presents the application and the first results of a plasma shape identification code used in the RFX experiment. This code has been specifically developed for machines provided with a stabilizing shell and with magnetic probes placed outside the vacuum vessel. In RFX, as in all the present-generation RFP machines, no steady-state magnetic configuration is reached, being the eddy currents in the vessel and in the stabilizing shell not negligible during the whole pulse duration. The new code, known as PLACID (Plasma Contour Identification) makes use of the information from the electromagnetic probes to estimate, by means of a filament current model, the currents in the vacuum vessel and in the stabilizing shell. In this way the solution to the magnetostatic problem of the determination of the magnetic flux surfaces outside the plasma is provided.
{"title":"Identification of RFX plasma shape from the electromagnetic probe signals","authors":"F. Bellina, P. Campostrini, G. Chitarin, M. Guarnieri, A. Stella, F. Trevisan","doi":"10.1109/FUSION.1993.518321","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518321","url":null,"abstract":"The paper presents the application and the first results of a plasma shape identification code used in the RFX experiment. This code has been specifically developed for machines provided with a stabilizing shell and with magnetic probes placed outside the vacuum vessel. In RFX, as in all the present-generation RFP machines, no steady-state magnetic configuration is reached, being the eddy currents in the vessel and in the stabilizing shell not negligible during the whole pulse duration. The new code, known as PLACID (Plasma Contour Identification) makes use of the information from the electromagnetic probes to estimate, by means of a filament current model, the currents in the vacuum vessel and in the stabilizing shell. In this way the solution to the magnetostatic problem of the determination of the magnetic flux surfaces outside the plasma is provided.","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":"115153454","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.1993.518437
R. Stephens
Advanced ICF targets will have an inner layer of solid or liquid fuel. Their inner surfaces must be smooth and contamination free. All of the current means to produce such a surface have problems: Liquid surfaces sag, solid surfaces tend to facet, and polymer-foam-stabilized surfaces are contaminated by carbon from the foam. An alternative may be to generate a liquid surface immediately before a shot by rapid thermal expansion of a fuel-saturated foam-walled capsule. This approach makes use of the large coefficient of expansion of liquid hydrogen relative to its foam matrix. The shell is filled by exposure to hydrogen vapor during cooling; liquid in the foam has a lower vapor pressure than free liquid, so the shell will fill to exactly 100%. It will stay at that fill fraction as the shell cools and the density of the liquid it contains increases. The shell may be frozen and cooled to 4 K so that it can be stored and handled in vacuum. When the shell is warmed, the liquid expands; the elastic modulus of the foam will force some liquid out of the surface. A simple analysis suggests that a 1 /spl mu/m thick liquid film might be generated in 1 /spl mu/s; that depends on the compressibility of the foam and the flow resistance of its cell structure. Surface tension would smooth this surface layer very rapidly. It would not begin to sag for 1000 /spl mu/s, so there would be sufficient time during which the layer would be satisfactory. An analysis will be presented showing the feasibility of this approach, and the constraints it puts on shell wall structure and insertion-and-shot procedures.
{"title":"Fast-formed liquid surfaces for inertial confinement fusion target shells","authors":"R. Stephens","doi":"10.1109/FUSION.1993.518437","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518437","url":null,"abstract":"Advanced ICF targets will have an inner layer of solid or liquid fuel. Their inner surfaces must be smooth and contamination free. All of the current means to produce such a surface have problems: Liquid surfaces sag, solid surfaces tend to facet, and polymer-foam-stabilized surfaces are contaminated by carbon from the foam. An alternative may be to generate a liquid surface immediately before a shot by rapid thermal expansion of a fuel-saturated foam-walled capsule. This approach makes use of the large coefficient of expansion of liquid hydrogen relative to its foam matrix. The shell is filled by exposure to hydrogen vapor during cooling; liquid in the foam has a lower vapor pressure than free liquid, so the shell will fill to exactly 100%. It will stay at that fill fraction as the shell cools and the density of the liquid it contains increases. The shell may be frozen and cooled to 4 K so that it can be stored and handled in vacuum. When the shell is warmed, the liquid expands; the elastic modulus of the foam will force some liquid out of the surface. A simple analysis suggests that a 1 /spl mu/m thick liquid film might be generated in 1 /spl mu/s; that depends on the compressibility of the foam and the flow resistance of its cell structure. Surface tension would smooth this surface layer very rapidly. It would not begin to sag for 1000 /spl mu/s, so there would be sufficient time during which the layer would be satisfactory. An analysis will be presented showing the feasibility of this approach, and the constraints it puts on shell wall structure and insertion-and-shot procedures.","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":"115475466","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.1993.518329
R. Bourque, C. Wong
A helium-cooled blanket/shield for ITER is presented that can provide high-grade heat and tritium self-sufficiency. It consists of narrow and relatively simple canisters filled with static liquid metal which is cooled by high pressure helium flowing through small double-walled tubes immersed in the liquid metal. The gaps between the tubes are also filled with static liquid metal. There are therefore three barriers between high-pressure helium and vacuum. Thermal-hydraulic analyses are presented that show the concept to be viable with both ferritic steel and vanadium alloy and with lithium and NaK liquid metal.
{"title":"Thermal-hydraulic analysis of a high-pressure helium-cooled shield/blanket for ITER","authors":"R. Bourque, C. Wong","doi":"10.1109/FUSION.1993.518329","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518329","url":null,"abstract":"A helium-cooled blanket/shield for ITER is presented that can provide high-grade heat and tritium self-sufficiency. It consists of narrow and relatively simple canisters filled with static liquid metal which is cooled by high pressure helium flowing through small double-walled tubes immersed in the liquid metal. The gaps between the tubes are also filled with static liquid metal. There are therefore three barriers between high-pressure helium and vacuum. Thermal-hydraulic analyses are presented that show the concept to be viable with both ferritic steel and vanadium alloy and with lithium and NaK liquid metal.","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":"115650182","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.1993.518371
F. Dahlgren, K. Wright, J. Kamperschroer, L. Grisham, L. Lontai, C. Peters, A. VonHalle
The present beam energy absorbing surfaces on the TFTR neutral beams such as ion dumps, calorimeters, beam defining apertures, and scrapers, are simple water cooled copper plates which were designed to absorb (via their thermal inertia) the incident beam power for two seconds with a five minute cool down interval between pulses. These components are not capable of absorbing the anticipated beam power loading for 1000 second TPX pulses and will have to be replaced with an actively cooled design. While several actively cooled energy absorbing designs were considered, the hypervapotron elements currently being used on the JET beamlines were chosen due to their lower cooling water demands and reliable performance on JET. The authors summarize the size, location (relative to the source) and the peak power requirements of the various beam components.
{"title":"TPX/TFTR neutral beam energy absorbers","authors":"F. Dahlgren, K. Wright, J. Kamperschroer, L. Grisham, L. Lontai, C. Peters, A. VonHalle","doi":"10.1109/FUSION.1993.518371","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518371","url":null,"abstract":"The present beam energy absorbing surfaces on the TFTR neutral beams such as ion dumps, calorimeters, beam defining apertures, and scrapers, are simple water cooled copper plates which were designed to absorb (via their thermal inertia) the incident beam power for two seconds with a five minute cool down interval between pulses. These components are not capable of absorbing the anticipated beam power loading for 1000 second TPX pulses and will have to be replaced with an actively cooled design. While several actively cooled energy absorbing designs were considered, the hypervapotron elements currently being used on the JET beamlines were chosen due to their lower cooling water demands and reliable performance on JET. The authors summarize the size, location (relative to the source) and the peak power requirements of the various beam components.","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":"116903249","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.1993.518338
L. Ku, S. Liew
This paper presents our recent analysis on the global radioactivation dose rate in TFTR for the upcoming DT operation. We show the profiles for the dose rate decay at four representative locations of interest in the test cell area. The dose rate data, together with the extensive radioactivity data base that we have developed, provide us with the essential information needed for the planning of the TFTR operation, decommissioning and waste disposal.
{"title":"Global dose rate in TFTR due to neutron induced residual radioactivities during DT operation","authors":"L. Ku, S. Liew","doi":"10.1109/FUSION.1993.518338","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518338","url":null,"abstract":"This paper presents our recent analysis on the global radioactivation dose rate in TFTR for the upcoming DT operation. We show the profiles for the dose rate decay at four representative locations of interest in the test cell area. The dose rate data, together with the extensive radioactivity data base that we have developed, provide us with the essential information needed for the planning of the TFTR operation, decommissioning and waste disposal.","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":"116922142","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.1993.518472
E. Lu, G. Bronner, C. Neumayer
This paper describes the modeling and simulation of the PF power system for the Tokamak Physics Experiment (TPX), which is required to supply pulsed DC current to the poloidal field (PF) superconducting coil system. An analytical model was developed to simulate the dynamics of the PF power system for any PF current scenario and thereby provide the basis for selection of PF circuit topology, in support of the major design goal of optimizing the use of the existing Tokamak Fusion Test Reactor (TFTR) facilities at the Princeton Plasma Physics Lab (PPPL).
{"title":"TPX poloidal field (PF) power systems simulation","authors":"E. Lu, G. Bronner, C. Neumayer","doi":"10.1109/FUSION.1993.518472","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518472","url":null,"abstract":"This paper describes the modeling and simulation of the PF power system for the Tokamak Physics Experiment (TPX), which is required to supply pulsed DC current to the poloidal field (PF) superconducting coil system. An analytical model was developed to simulate the dynamics of the PF power system for any PF current scenario and thereby provide the basis for selection of PF circuit topology, in support of the major design goal of optimizing the use of the existing Tokamak Fusion Test Reactor (TFTR) facilities at the Princeton Plasma Physics Lab (PPPL).","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":"124883873","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.1993.518362
S. Sharafat, R. Junge, F. Najmabadi, I. Sviatoslavsky, C. Wong
The ARIES-IV fusion power plant is a conceptual, steady-state, D-T burning, 1000-MWe net-power tokamak, operating in the second-stability plasma regime. Design simplicity, maintainability, and reasonable maintenance downtimes are crucial for viable and competitive fusion economics. The ARIES-IV design team developed a maintenance scheme that allows rapid and simultaneous replacement of all of the components in a large, self-contained fusion-power-core (FPC) section. The FPC is divided into 16 self-contained "pie-shaped" sections that can be removed horizontally through large vacuum-vessel access ports located between the outer legs of the toroidal-field (TF) coils. Prior to commitment to service, the entire FPC section assembly is pretested extensively to maximize operating reliability. To facilitate this radial section-removal scheme, the TF coils and the poloidal-field (PF) coils of the ARIES-IV tokamak had to be enlarged. The advantages of the ARIES-IV section-replacement scheme over more traditional approaches, where individual FPC components are removed sequentially from the vacuum vessel, outweigh the cost increases associated with larger TF- and PF-coil systems.
{"title":"Design layout and maintenance of the ARIES-IV tokamak fusion power plant","authors":"S. Sharafat, R. Junge, F. Najmabadi, I. Sviatoslavsky, C. Wong","doi":"10.1109/FUSION.1993.518362","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518362","url":null,"abstract":"The ARIES-IV fusion power plant is a conceptual, steady-state, D-T burning, 1000-MWe net-power tokamak, operating in the second-stability plasma regime. Design simplicity, maintainability, and reasonable maintenance downtimes are crucial for viable and competitive fusion economics. The ARIES-IV design team developed a maintenance scheme that allows rapid and simultaneous replacement of all of the components in a large, self-contained fusion-power-core (FPC) section. The FPC is divided into 16 self-contained \"pie-shaped\" sections that can be removed horizontally through large vacuum-vessel access ports located between the outer legs of the toroidal-field (TF) coils. Prior to commitment to service, the entire FPC section assembly is pretested extensively to maximize operating reliability. To facilitate this radial section-removal scheme, the TF coils and the poloidal-field (PF) coils of the ARIES-IV tokamak had to be enlarged. The advantages of the ARIES-IV section-replacement scheme over more traditional approaches, where individual FPC components are removed sequentially from the vacuum vessel, outweigh the cost increases associated with larger TF- and PF-coil systems.","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":"124894135","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.1993.518523
M. Tekula, L. Bromberg
In this paper a system evaluation of MHD coils for divertor coil and as passive stabilizer against plasma vertical displacements is presented. The advantages of placing the poloidal field coils are quantified. Large decreases in the conventional divertor coil current and associated torques in the toroidal field system are obtained when the current in the MHD coil is as little as 1 MA. To generate this current, flows of /spl sim/1 m/s are needed. The sensitivity of the separatrix to changing plasma or external coil currents is evaluated. Finally, the use of low activation metals in in the coil segments is evaluated.
{"title":"MHD-driven internal coils for tokamak divertor operation","authors":"M. Tekula, L. Bromberg","doi":"10.1109/FUSION.1993.518523","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518523","url":null,"abstract":"In this paper a system evaluation of MHD coils for divertor coil and as passive stabilizer against plasma vertical displacements is presented. The advantages of placing the poloidal field coils are quantified. Large decreases in the conventional divertor coil current and associated torques in the toroidal field system are obtained when the current in the MHD coil is as little as 1 MA. To generate this current, flows of /spl sim/1 m/s are needed. The sensitivity of the separatrix to changing plasma or external coil currents is evaluated. Finally, the use of low activation metals in in the coil segments is evaluated.","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":"123702325","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.1993.518447
M. Chaplin, W. Hassenzahl, H. Schultz
The design of the Local Instrumentation & Control (I&C) System for the Tokamak Physics Experiment (TPX) superconducting PF & TF magnets is presented. The local I&C system monitors the status of the magnet systems and initiates the proper control sequences to protect the magnets from any foreseeable fault. Local I&C also stores magnet-system data for analysis and archiving. Quench Detection for the TPX magnets must use a minimum of two independent sensing methods and is allowed a detection time of one second. Proposed detection methods include the measurement of; (1) normal-zone resistive voltage, (2) cooling-path helium flow, (3) local temperature in the winding pack, (4) local pressure in the winding pack. Fiber-optic based isolation systems are used to remove high common-mode magnet voltages and eliminate ground loops. The data acquisition and fault-detection systems are computer based. The design of the local I&C system incorporates redundant, fault-tolerant, and/or fail-safe features at all component levels. As part of a quench detection R&D plan, a Quench Detection Model Coil has been proposed to test all detection methods. Initial cost estimates and schedule for the local I&C system are presented.
{"title":"Quench detection & instrumentation for the Tokamak Physics Experiment magnets","authors":"M. Chaplin, W. Hassenzahl, H. Schultz","doi":"10.1109/FUSION.1993.518447","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518447","url":null,"abstract":"The design of the Local Instrumentation & Control (I&C) System for the Tokamak Physics Experiment (TPX) superconducting PF & TF magnets is presented. The local I&C system monitors the status of the magnet systems and initiates the proper control sequences to protect the magnets from any foreseeable fault. Local I&C also stores magnet-system data for analysis and archiving. Quench Detection for the TPX magnets must use a minimum of two independent sensing methods and is allowed a detection time of one second. Proposed detection methods include the measurement of; (1) normal-zone resistive voltage, (2) cooling-path helium flow, (3) local temperature in the winding pack, (4) local pressure in the winding pack. Fiber-optic based isolation systems are used to remove high common-mode magnet voltages and eliminate ground loops. The data acquisition and fault-detection systems are computer based. The design of the local I&C system incorporates redundant, fault-tolerant, and/or fail-safe features at all component levels. As part of a quench detection R&D plan, a Quench Detection Model Coil has been proposed to test all detection methods. Initial cost estimates and schedule for the local I&C system are 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":"125278615","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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