Pub Date : 1991-09-30DOI: 10.1109/FUSION.1991.218901
H. Murray, I. Harris, J.A. Ratka
The BPX (Burning Plasma Experiment) toroidal field coil conductor is fabricated from plates of C17510, a copper-beryllium alloy. A development program was initiated by Princeton Plasma Physics Laboratory (PPL) to establish welding techniques and investigate the benefits of post-weld heat treatment for this alloy. Several process options were investigated. Since both the plate material and weld filler contain beryllium, the establishment of health/safety procedures was prerequisite to the start of the development program. Significant progress in elevating the physical integrity of the welds was achieved by simultaneously optimizing the weld procedure and the heat-treatment processing. The initial program phase resulted in reproducible welds which exhibited 0.2% offset yield strengths in excess of 80 ksi. The welding techniques were developed on commercially available automated welding equipment and the heat-treatment procedures were established to be compatible with large welded assemblies. In-process quality assurance techniques were investigated.<>
{"title":"Development of a welding procedure for high conductivity, copper-beryllium alloy C17510","authors":"H. Murray, I. Harris, J.A. Ratka","doi":"10.1109/FUSION.1991.218901","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218901","url":null,"abstract":"The BPX (Burning Plasma Experiment) toroidal field coil conductor is fabricated from plates of C17510, a copper-beryllium alloy. A development program was initiated by Princeton Plasma Physics Laboratory (PPL) to establish welding techniques and investigate the benefits of post-weld heat treatment for this alloy. Several process options were investigated. Since both the plate material and weld filler contain beryllium, the establishment of health/safety procedures was prerequisite to the start of the development program. Significant progress in elevating the physical integrity of the welds was achieved by simultaneously optimizing the weld procedure and the heat-treatment processing. The initial program phase resulted in reproducible welds which exhibited 0.2% offset yield strengths in excess of 80 ksi. The welding techniques were developed on commercially available automated welding equipment and the heat-treatment procedures were established to be compatible with large welded assemblies. In-process quality assurance techniques were investigated.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"120 12","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1991-09-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"131913593","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.1991.218720
L. Lagin
The author describes the application of neural networks to the control of the neural beam long-pulse positive ion source accelerators on the Tokamak Fusion Test Reactor (TFTR) at Princeton University. Neural networks were used to learn how the operators adjust the control setpoints when running these sources. The data sets used to train these networks were derived from a large database containing actual setpoints and power supply waveform calculations for the 1990 run period. The networks learned what the optimum control setpoints should initially be set based upon desired accel voltage and perveance levels. Neural networks were also used to predict the divergence of the ion beam.<>
{"title":"Applying neural networks to control the TFTR neural beam ion sources","authors":"L. Lagin","doi":"10.1109/FUSION.1991.218720","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218720","url":null,"abstract":"The author describes the application of neural networks to the control of the neural beam long-pulse positive ion source accelerators on the Tokamak Fusion Test Reactor (TFTR) at Princeton University. Neural networks were used to learn how the operators adjust the control setpoints when running these sources. The data sets used to train these networks were derived from a large database containing actual setpoints and power supply waveform calculations for the 1990 run period. The networks learned what the optimum control setpoints should initially be set based upon desired accel voltage and perveance levels. Neural networks were also used to predict the divergence of the ion beam.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"33 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":"127398671","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.1991.218783
S. Dinkevich, J. Swanson, T. Feng, M. Z. Khan, P. Hsueh
The authors provide a summary of the vacuum vessel structural analyses performed in support of the BPX (Burning Plasma Experiment) Conceptual Design Review held in March 1991. They present descriptions of the vessel finite element model and analysis results for the principal loadings associated with electromagnetic forces induced in the vessel walls by plasma disruptions. The structural analysis performed with the U6M model indicates that the BPX vessel is adequate, except for the need to provide minor reinforcement in various local areas.<>
{"title":"Structural analysis for the conceptual design of the BPX vacuum vessel","authors":"S. Dinkevich, J. Swanson, T. Feng, M. Z. Khan, P. Hsueh","doi":"10.1109/FUSION.1991.218783","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218783","url":null,"abstract":"The authors provide a summary of the vacuum vessel structural analyses performed in support of the BPX (Burning Plasma Experiment) Conceptual Design Review held in March 1991. They present descriptions of the vessel finite element model and analysis results for the principal loadings associated with electromagnetic forces induced in the vessel walls by plasma disruptions. The structural analysis performed with the U6M model indicates that the BPX vessel is adequate, except for the need to provide minor reinforcement in various local areas.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"91 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":"124213005","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.1991.218910
V. Lee
The author presents the status of design integration efforts on a conceptual design study for a power producing reactor utilizing inertial fusion energy (IFE) technology. PROMETHEUS-L is a 1000 MWe power plant utilizing a KrF excimer laser driver to deliver 5 MJ of energy through 60 beamlines to injected, direct driver targets in the center of a cylindrical reactor cavity at a rate of 4 times/s. The first wall (FW) system of the cavity uses liquid lead which both cools the FW structure of silicon carbide composite and bleeds through the porous, inner surface of the cavity, forming a protective liquid film on the FW surface. The blanket system features SiC composite structure for a Li/sub 2/O packed bed breeder with separate helium purge and helium coolant. The tritium breeding ratio is 1.30. The thermal conversion system features an advanced supercritical Rankine steam cycle with double reheat. The gross thermal efficiency is 43%.<>
{"title":"Design integration of an inertial fusion energy reactor power plant","authors":"V. Lee","doi":"10.1109/FUSION.1991.218910","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218910","url":null,"abstract":"The author presents the status of design integration efforts on a conceptual design study for a power producing reactor utilizing inertial fusion energy (IFE) technology. PROMETHEUS-L is a 1000 MWe power plant utilizing a KrF excimer laser driver to deliver 5 MJ of energy through 60 beamlines to injected, direct driver targets in the center of a cylindrical reactor cavity at a rate of 4 times/s. The first wall (FW) system of the cavity uses liquid lead which both cools the FW structure of silicon carbide composite and bleeds through the porous, inner surface of the cavity, forming a protective liquid film on the FW surface. The blanket system features SiC composite structure for a Li/sub 2/O packed bed breeder with separate helium purge and helium coolant. The tritium breeding ratio is 1.30. The thermal conversion system features an advanced supercritical Rankine steam cycle with double reheat. The gross thermal efficiency is 43%.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"12 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":"114563638","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.1991.218885
M. Huguet, H. Altmann, P. Barabaschi, E. Bertolini, K. Dietz, E. Deksnis, H. Falter, C. Froger, M. Garribba, A. Kaye, J. Last, R. Lobel, E. Martin, P. Massmann, P. Noll, W. Obert, S. Papastergiou, A. Peacock, M. Pick, P. Rebut, L. Rossi, C. Sborchia, G. Sannazzaro, A. Tesini, R. Tivey
The JET (Joint European Torus) pumped divertor aims at demonstrating an effective method of impurity control with quasi-stationary plasmas of thermonuclear grade in a next step relevant, axisymmetric configuration. The magnetic configuration is produced by a set of four coils internal to the JET vacuum vessel. These coils can produce a range of configurations and also sweep the magnetic field lines along the target plates. The target plates will initially use radiation-cooled beryllium tiles, but actively cooled target plates able to operate in steady state at up to 40 MW are planned in a second phase. The design also features a cryopump which will remove a fraction of the particles recycled in the vicinity of the target plates. The configuration of the ICRH (ion cyclotron resonance heating) antennae and wall protections has been modified to match the new plasma shape. All components have been designed to resist the large forces generated by halo currents.<>
{"title":"Design of the JET pumped divertor","authors":"M. Huguet, H. Altmann, P. Barabaschi, E. Bertolini, K. Dietz, E. Deksnis, H. Falter, C. Froger, M. Garribba, A. Kaye, J. Last, R. Lobel, E. Martin, P. Massmann, P. Noll, W. Obert, S. Papastergiou, A. Peacock, M. Pick, P. Rebut, L. Rossi, C. Sborchia, G. Sannazzaro, A. Tesini, R. Tivey","doi":"10.1109/FUSION.1991.218885","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218885","url":null,"abstract":"The JET (Joint European Torus) pumped divertor aims at demonstrating an effective method of impurity control with quasi-stationary plasmas of thermonuclear grade in a next step relevant, axisymmetric configuration. The magnetic configuration is produced by a set of four coils internal to the JET vacuum vessel. These coils can produce a range of configurations and also sweep the magnetic field lines along the target plates. The target plates will initially use radiation-cooled beryllium tiles, but actively cooled target plates able to operate in steady state at up to 40 MW are planned in a second phase. The design also features a cryopump which will remove a fraction of the particles recycled in the vicinity of the target plates. The configuration of the ICRH (ion cyclotron resonance heating) antennae and wall protections has been modified to match the new plasma shape. All components have been designed to resist the large forces generated by halo currents.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"40 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":"117050831","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.1991.218873
R. Sledge, K. Hsieh, W. Weldon, M. Werst
The Ignition Technology Demonstration (ITD) is a 0.06 scale prototype toroidal field magnet of the proposed full-scale IGNITEX (Ignition Experiment) tokamak. The goal of ITD is to achieve an on-axis magnetic confinement field of 20 T while demonstrating the magnet's ability to withstand high magnetic and thermal stresses. To accomplish this task, a peak current of 9 MA must be transferred from six balanced homopolar generator (HPG) busbar circuits to the liquid nitrogen (LN/sub 2/) cooled magnet. To date the system has delivered pulses of up to 8.14 MA to the magnet, producing an on-axis field of 18.1 T. In order to properly synchronize current transfer, an explosive closing switch is used for each of the six independent HPG/busbar circuits. The switches operate by explosively driving a scalloped copper ring into a tapered annular gap made up of two copper alloy rings. With a jitter time of 10 mu s, parallel circuit synchronization is better than 0.03% relative to the current rise time. The excellent performance of the switches during discharges of up to 8.14 MA is attributed to several design features which assure proper current distribution. Busbar design considerations are discussed, and the performance of the switches and busbars is described.<>
{"title":"High current transmission and switching system for prototype 20 Tesla toroidal magnet","authors":"R. Sledge, K. Hsieh, W. Weldon, M. Werst","doi":"10.1109/FUSION.1991.218873","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218873","url":null,"abstract":"The Ignition Technology Demonstration (ITD) is a 0.06 scale prototype toroidal field magnet of the proposed full-scale IGNITEX (Ignition Experiment) tokamak. The goal of ITD is to achieve an on-axis magnetic confinement field of 20 T while demonstrating the magnet's ability to withstand high magnetic and thermal stresses. To accomplish this task, a peak current of 9 MA must be transferred from six balanced homopolar generator (HPG) busbar circuits to the liquid nitrogen (LN/sub 2/) cooled magnet. To date the system has delivered pulses of up to 8.14 MA to the magnet, producing an on-axis field of 18.1 T. In order to properly synchronize current transfer, an explosive closing switch is used for each of the six independent HPG/busbar circuits. The switches operate by explosively driving a scalloped copper ring into a tapered annular gap made up of two copper alloy rings. With a jitter time of 10 mu s, parallel circuit synchronization is better than 0.03% relative to the current rise time. The excellent performance of the switches during discharges of up to 8.14 MA is attributed to several design features which assure proper current distribution. Busbar design considerations are discussed, and the performance of the switches and busbars is described.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"22 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":"116486773","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.1991.218679
T. Haines
Three types of folding structures are described, together with concepts for use in fusion machines. The Storable Tubular Extendible Member (STEM) was conceived by the National Research Council of Canada and developed by Spar Aerospace Limited. The Astromast is a folding truss developed by Astro Aerospace Corporation. The X-Beam is an ultrastiff folding truss. X-Beam trusses can extend to considerable distances both horizontally and vertically and can support heavy loads. Two such trusses can be mounted on a remotely controlled vehicle to provide a mobile folding crane able to enter the many confined locations in a fusion machine. It can be used in these locations to eliminate the need for fixed cranes. Because it can be removed before reactor operation and is not exposed to the extreme environment that a fixed crane must withstand, it will have better reliability and maintainability. A brief history of the development of folding structure technology is included.<>
{"title":"The use of folding structures in fusion reactors","authors":"T. Haines","doi":"10.1109/FUSION.1991.218679","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218679","url":null,"abstract":"Three types of folding structures are described, together with concepts for use in fusion machines. The Storable Tubular Extendible Member (STEM) was conceived by the National Research Council of Canada and developed by Spar Aerospace Limited. The Astromast is a folding truss developed by Astro Aerospace Corporation. The X-Beam is an ultrastiff folding truss. X-Beam trusses can extend to considerable distances both horizontally and vertically and can support heavy loads. Two such trusses can be mounted on a remotely controlled vehicle to provide a mobile folding crane able to enter the many confined locations in a fusion machine. It can be used in these locations to eliminate the need for fixed cranes. Because it can be removed before reactor operation and is not exposed to the extreme environment that a fixed crane must withstand, it will have better reliability and maintainability. A brief history of the development of folding structure technology is included.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"138 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":"125750111","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.1991.218664
L. Stewart, M. Monsler, S. Humphries
The size and cost of an induction linac heavy ion beam driver for inertial confinement fusion will depend primarily on ion kinetic energy, current, mass, and charge. Energy, current, and mass are constrained by implosion physics, and all four quantities are constrained by beam propagation physics. The authors contrast two alternatives to beam propagation modeling: arbitrary specification of partial beam space charge cancellation and calculation of the beam envelope in the presence of an autoneutralizing cloud of hot electrons. They explore model-dependent sensitivities to beam kinetic energy, pulse energy, charge state, ion species, final focus geometry, and stripping in the reactor chamber. Preferred parameter regimes for a 3000-MW/sub t/ fusion power plant are indicated.<>
{"title":"Parametric studies of an induction linac heavy ion beam driver for ICF","authors":"L. Stewart, M. Monsler, S. Humphries","doi":"10.1109/FUSION.1991.218664","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218664","url":null,"abstract":"The size and cost of an induction linac heavy ion beam driver for inertial confinement fusion will depend primarily on ion kinetic energy, current, mass, and charge. Energy, current, and mass are constrained by implosion physics, and all four quantities are constrained by beam propagation physics. The authors contrast two alternatives to beam propagation modeling: arbitrary specification of partial beam space charge cancellation and calculation of the beam envelope in the presence of an autoneutralizing cloud of hot electrons. They explore model-dependent sensitivities to beam kinetic energy, pulse energy, charge state, ion species, final focus geometry, and stripping in the reactor chamber. Preferred parameter regimes for a 3000-MW/sub t/ fusion power plant are indicated.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"8 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":"126194823","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.1991.218807
M. Z. Hasan, F. Najmabadi, G. Orient, E. Reis, S. Sharafat, D. Sze, M. Valenti, C. Wong
The thermal-hydraulic and structural design of the ARIES-III divertor plate is presented. The divertor plate is made of small-diameter W-3Re tubes laid along the radial direction and coated with 4 mm of plasma-sprayed tungsten on the plasma-facing side to withstand one hard disruption. The plate is contoured to have the constant heat flux of 5.44 MW/m/sup 2/ on the entire surface. The total divertor thermal power of 629 MW is removed by the organic coolant HB-40 with the same inlet/exit temperatures (340 degrees C/425 degrees C) as in the first-wall/shield coolant circuit. The principal mode of heat transfer is by subcooled flow boiling. The inlet pressure is 5.34 MPa and the exit pressure is 4.3 MPa, which, by passing through an orifice, is reduced to 1 MPa, equal to the first-wall/shield exit pressure. The total coolant flow rate is 3.35 m/sup 3//s and the circulation power is 18 MWe. The maximum plate temperature is 821 degrees C. The safety factor with respect to the critical heat flux is >or=2, and it is approximately 3 with respect to the maximum allowable plate temperature. Maximum equivalent total, thermal, and pressure stresses are also given.<>
{"title":"Thermo-structural design of the ARIES-III divertor with organic coolant in subcooled flow boiling","authors":"M. Z. Hasan, F. Najmabadi, G. Orient, E. Reis, S. Sharafat, D. Sze, M. Valenti, C. Wong","doi":"10.1109/FUSION.1991.218807","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218807","url":null,"abstract":"The thermal-hydraulic and structural design of the ARIES-III divertor plate is presented. The divertor plate is made of small-diameter W-3Re tubes laid along the radial direction and coated with 4 mm of plasma-sprayed tungsten on the plasma-facing side to withstand one hard disruption. The plate is contoured to have the constant heat flux of 5.44 MW/m/sup 2/ on the entire surface. The total divertor thermal power of 629 MW is removed by the organic coolant HB-40 with the same inlet/exit temperatures (340 degrees C/425 degrees C) as in the first-wall/shield coolant circuit. The principal mode of heat transfer is by subcooled flow boiling. The inlet pressure is 5.34 MPa and the exit pressure is 4.3 MPa, which, by passing through an orifice, is reduced to 1 MPa, equal to the first-wall/shield exit pressure. The total coolant flow rate is 3.35 m/sup 3//s and the circulation power is 18 MWe. The maximum plate temperature is 821 degrees C. The safety factor with respect to the critical heat flux is >or=2, and it is approximately 3 with respect to the maximum allowable plate temperature. Maximum equivalent total, thermal, and pressure stresses are also given.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"132 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":"128438595","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.1991.218904
A. Brooks
The toroidal field (TF) coils for BPX are Bitter plate coils designed to be rapidly pulsed at high power levels, and then recooled by liquid nitrogen within an hour. The analysis of this cooldown process in support of the conceptual design of the BPX TF coils is discussed. There are two major aspects of the cooldown analysis: (1) the thermal conduction within the Bitter plate; and (2) two-phased nitrogen flow and heat transfer. Analysis results are presented.<>
{"title":"Cooldown analysis of the toroidal field coils for the Burning Plasma Experiment (BPX)","authors":"A. Brooks","doi":"10.1109/FUSION.1991.218904","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218904","url":null,"abstract":"The toroidal field (TF) coils for BPX are Bitter plate coils designed to be rapidly pulsed at high power levels, and then recooled by liquid nitrogen within an hour. The analysis of this cooldown process in support of the conceptual design of the BPX TF coils is discussed. There are two major aspects of the cooldown analysis: (1) the thermal conduction within the Bitter plate; and (2) two-phased nitrogen flow and heat transfer. Analysis results are presented.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"30 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":"129943474","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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