Pub Date : 1991-09-30DOI: 10.1109/FUSION.1991.218888
J. Minervini, M. Steeves, J. Schultz, D. Montgomery, M. Takayasu, T. Painter
A preliminary design of a central solenoid model coil for ITER (International Thermonuclear Experimental Reactor) has been made. The preliminary conductor and coil design described should achieve the model coil performance requirements with minimum cost and maximum reliability. The design is based on a Nb/sub 3/Sn cable-in-conduit conductor incorporating a conduit made from the new superalloy Incoloy 908. This alloy has excellent mechanical properties and a COE (cost of electricity) well matched to the COE of the superconducting wire. Major advantages in coil performance result, due to reduction of the thermal strain-induced degradation of the critical properties of the Nb/sub 3/Sn superconductor. The principal performance advantages include higher critical current density and higher critical temperature, particularly at high fields, which results in increased operating margins of critical current and stability. A prediction of enhanced operating performance is made relative to conductor designs by other ITER parties which rely on conduits made from austenitic stainless steels with relatively high coefficients of thermal expansion.<>
{"title":"Preliminary design of a US ITER model poloidal coil","authors":"J. Minervini, M. Steeves, J. Schultz, D. Montgomery, M. Takayasu, T. Painter","doi":"10.1109/FUSION.1991.218888","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218888","url":null,"abstract":"A preliminary design of a central solenoid model coil for ITER (International Thermonuclear Experimental Reactor) has been made. The preliminary conductor and coil design described should achieve the model coil performance requirements with minimum cost and maximum reliability. The design is based on a Nb/sub 3/Sn cable-in-conduit conductor incorporating a conduit made from the new superalloy Incoloy 908. This alloy has excellent mechanical properties and a COE (cost of electricity) well matched to the COE of the superconducting wire. Major advantages in coil performance result, due to reduction of the thermal strain-induced degradation of the critical properties of the Nb/sub 3/Sn superconductor. The principal performance advantages include higher critical current density and higher critical temperature, particularly at high fields, which results in increased operating margins of critical current and stability. A prediction of enhanced operating performance is made relative to conductor designs by other ITER parties which rely on conduits made from austenitic stainless steels with relatively high coefficients of thermal expansion.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"5 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":"128723800","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.1991.218711
T. Kimura
The initial experimental results in the JT-60 (JAERI Tokamak-60) upgrade are reported along with an overview of reconstruction of JT-60 and future plans for the upgrade. Reconstruction of the JT-60 machine was successfully completed at the end of March 1991. 4-MA stable NB (neutral beam)-heated discharges with a discharge duration of 15 s and a flat-top duration of approximately 1 s have been achieved in a short operational period of 4 months. The stored energy has reached 5.1 MJ with I/sub P/=4 MA, B/sub t/=4 T, and NBI (neutral-beam injection) power of 20 MW. The global energy confinement time of 0.9 s has been achieved in deuterium OH discharges with I/sub P/=3 MA, B/sub t/=4 T, and n/sub e/=2*10/sup 19/ m/sup -3/. The energy confinement time of 0.25 s has been achieved in 15-20-MW NB-heated deuterium plasmas with I/sub p/=4 MA. The confinement time is in good agreement with that in the L-mode scaling of the ITER89 power law. The discharges of high ion temperature mode with T/sub i/(0) over 20 keV have been obtained by injecting deuterium NB with low target density of approximately 0.5*10/sup 19/ m/sup -3/, where the maximum neutron yield is 1.3*10/sup 16/ n/s and the energy confinement time is about 2.5 times as large as that of the International Thermonuclear Experimental Reactor power law scaling.<>
{"title":"Initial results from the JT-60 upgrade","authors":"T. Kimura","doi":"10.1109/FUSION.1991.218711","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218711","url":null,"abstract":"The initial experimental results in the JT-60 (JAERI Tokamak-60) upgrade are reported along with an overview of reconstruction of JT-60 and future plans for the upgrade. Reconstruction of the JT-60 machine was successfully completed at the end of March 1991. 4-MA stable NB (neutral beam)-heated discharges with a discharge duration of 15 s and a flat-top duration of approximately 1 s have been achieved in a short operational period of 4 months. The stored energy has reached 5.1 MJ with I/sub P/=4 MA, B/sub t/=4 T, and NBI (neutral-beam injection) power of 20 MW. The global energy confinement time of 0.9 s has been achieved in deuterium OH discharges with I/sub P/=3 MA, B/sub t/=4 T, and n/sub e/=2*10/sup 19/ m/sup -3/. The energy confinement time of 0.25 s has been achieved in 15-20-MW NB-heated deuterium plasmas with I/sub p/=4 MA. The confinement time is in good agreement with that in the L-mode scaling of the ITER89 power law. The discharges of high ion temperature mode with T/sub i/(0) over 20 keV have been obtained by injecting deuterium NB with low target density of approximately 0.5*10/sup 19/ m/sup -3/, where the maximum neutron yield is 1.3*10/sup 16/ n/s and the energy confinement time is about 2.5 times as large as that of the International Thermonuclear Experimental Reactor power law scaling.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"90 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":"128265437","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.1991.218760
A. Navarro, L. Almoguera, J. Alonso Gozalo, J. Alonso Candenas, M. Blaumoser, J. Botija, A. García, M. Liniers, A. Martínez, M. Medrano
TJ-II is a medium-size stellarator, (R/sub 0/ = 1.5 m; =0.2 m; B/sub t/=1 T) under construction at CIEMAT. It will allow the exploration of helical magnetic axis plasmas in a wide range af' configurations (rotational transform ranges from 0.8 to 2.5, magnetic well up to 6%, and shear up to 10%) in discharges of 0.2 to 1 s, generated and heated by ECH (electron cyclotron heating) (400 kW at 53.2 GHz). < beta > limits of the configuration, theoretically predicted to be as high as 6%, will be explored by the addition of 4 MW of NBI. The design criteria for this device and the solutions adopted for the different components are presented. Modular structure for the vacuum vessel, splittable toroidal field coils, and an independent support structure for each element, together with a detailed assembly procedure, allow independent and parallel construction of the different components. Strict requirements on tolerances, shapes, and current densities have been carefully addressed and solved in the design of the two most complicated components, the vacuum vessel and the central coils.<>
{"title":"Engineering aspects and present status of the Spanish stellarator TJ-II","authors":"A. Navarro, L. Almoguera, J. Alonso Gozalo, J. Alonso Candenas, M. Blaumoser, J. Botija, A. García, M. Liniers, A. Martínez, M. Medrano","doi":"10.1109/FUSION.1991.218760","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218760","url":null,"abstract":"TJ-II is a medium-size stellarator, (R/sub 0/ = 1.5 m; =0.2 m; B/sub t/=1 T) under construction at CIEMAT. It will allow the exploration of helical magnetic axis plasmas in a wide range af' configurations (rotational transform ranges from 0.8 to 2.5, magnetic well up to 6%, and shear up to 10%) in discharges of 0.2 to 1 s, generated and heated by ECH (electron cyclotron heating) (400 kW at 53.2 GHz). < beta > limits of the configuration, theoretically predicted to be as high as 6%, will be explored by the addition of 4 MW of NBI. The design criteria for this device and the solutions adopted for the different components are presented. Modular structure for the vacuum vessel, splittable toroidal field coils, and an independent support structure for each element, together with a detailed assembly procedure, allow independent and parallel construction of the different components. Strict requirements on tolerances, shapes, and current densities have been carefully addressed and solved in the design of the two most complicated components, the vacuum vessel and the central coils.<<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":"129087458","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.1991.218764
L. Qiu, G. S. Luan, Q. Xu, Z. Guo, X. Gao, P. Duan, X.Q. Zhang, Y.Y. Li, W. Xu, Y. Wu, W.Y. Wu, C. Xia, Q. Huang, Y. Chen, N. Xiang, W.G. Zhang Yina, Y. Wu, Y. Chen, C. Qian
The authors present a new concept of a hybrid reactor: using a Joint-European-Torus-like device which works in the sub-breakeven condition as a source of high-energy neutrons makes it possible to induce a blanket fission of depleted-uranium, solid breeding material, and helium cooling so that one can produce 100 kg of nuclear fuel (/sup 239/Pu) per year. High temperature will be maintained by external ICRF and ECRF heating. A steady-state plasma current will be driven by LHCD. The plasma density will be maintained by pellet injection ICRF can produce a high-energy tail in the ion distribution function and lead to significant enhancement of D-T reaction rate by 2-5 times so that one can obtain a neutron source strength of 1*10/sup 19/ n/s, enough for the requirement of the hybrid reactor.<>
{"title":"Hefei tokamak Experimental Hybrid Reactor conceptual design","authors":"L. Qiu, G. S. Luan, Q. Xu, Z. Guo, X. Gao, P. Duan, X.Q. Zhang, Y.Y. Li, W. Xu, Y. Wu, W.Y. Wu, C. Xia, Q. Huang, Y. Chen, N. Xiang, W.G. Zhang Yina, Y. Wu, Y. Chen, C. Qian","doi":"10.1109/FUSION.1991.218764","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218764","url":null,"abstract":"The authors present a new concept of a hybrid reactor: using a Joint-European-Torus-like device which works in the sub-breakeven condition as a source of high-energy neutrons makes it possible to induce a blanket fission of depleted-uranium, solid breeding material, and helium cooling so that one can produce 100 kg of nuclear fuel (/sup 239/Pu) per year. High temperature will be maintained by external ICRF and ECRF heating. A steady-state plasma current will be driven by LHCD. The plasma density will be maintained by pellet injection ICRF can produce a high-energy tail in the ion distribution function and lead to significant enhancement of D-T reaction rate by 2-5 times so that one can obtain a neutron source strength of 1*10/sup 19/ n/s, enough for the requirement of the hybrid reactor.<<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":"130611008","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.1991.218673
L. El-Guebaly
During the conceptual design phase of the International Thermonuclear Experimental Reactor (ITER), several tritium breeding blanket designs were proposed by the four international parties, the US, the USSR, Japan (J), and the European Community (EC). The US, J, and EC designs utilize solid breeders with beryllium multipliers, while the USSR proposes an LiPb blanket design. All blankets are water cooled and use 316 SS as structural material. The shielding effectiveness of the different blanket designs was compared on the same basis and the radiation level at the superconducting toroidal-field magnets was assessed. The analysis shows that the inboard (i/b) blanket, in particular, has significant impact on magnet damage and total heating. For the US design, all magnet radiation limits are met with the current 84-cm-thick i/b region. For the other designs, the i/b blanket design should be modified to satisfy the magnet radiation limits, or 3-7 cm additional i/b shielding should be provided depending on the blanket type. The impact of the latest divertor design on the magnet damage was also analyzed, and the combined effect of both blanket and divertor designs on the overall magnet heat load was assessed.<>
{"title":"Magnet shielding effectiveness of the proposed blankets for ITER","authors":"L. El-Guebaly","doi":"10.1109/FUSION.1991.218673","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218673","url":null,"abstract":"During the conceptual design phase of the International Thermonuclear Experimental Reactor (ITER), several tritium breeding blanket designs were proposed by the four international parties, the US, the USSR, Japan (J), and the European Community (EC). The US, J, and EC designs utilize solid breeders with beryllium multipliers, while the USSR proposes an LiPb blanket design. All blankets are water cooled and use 316 SS as structural material. The shielding effectiveness of the different blanket designs was compared on the same basis and the radiation level at the superconducting toroidal-field magnets was assessed. The analysis shows that the inboard (i/b) blanket, in particular, has significant impact on magnet damage and total heating. For the US design, all magnet radiation limits are met with the current 84-cm-thick i/b region. For the other designs, the i/b blanket design should be modified to satisfy the magnet radiation limits, or 3-7 cm additional i/b shielding should be provided depending on the blanket type. The impact of the latest divertor design on the magnet damage was also analyzed, and the combined effect of both blanket and divertor designs on the overall magnet heat load was assessed.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"54 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":"123918562","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.1991.218800
P. Avanzini, M. Brossa, U. Guerreschi, G. Persano Adorno, N. Pietranera, F. Rosatelli, G. Simbolotti, V. Zampaglione
The proposed first wall (FW) for the NET/ITER (Next European Torus/International Thermonuclear Experimental Reactor) consists of a box structure, made of stainless steel, with an active cooling of the side facing the plasma by a flow of water through the tubes inserted into it. An innovative brazing process to improve the quality of the brazed joints between cooling tubes and plates has been developed and optimized within the framework of NET technological task PDT 1.2. The brazing technique is based on creep expansion of the tubes during the high-temperature holding time of the brazing treatment, which minimizes the gap that was initially present around the tubes. A flat, mockup FW panel with plate dimensions 622*216*33 mm and with eight cooling tubes was built using this technique. Thermal fatigue tests are scheduled to be carried out in the near future on this mockup. An automatic, computerized procedure for the nondestructive examination, of the brazed joints using ultrasonic probes has also been developed.<>
{"title":"Progress in manufacturing and testing of a first wall mockup for NET/ITER","authors":"P. Avanzini, M. Brossa, U. Guerreschi, G. Persano Adorno, N. Pietranera, F. Rosatelli, G. Simbolotti, V. Zampaglione","doi":"10.1109/FUSION.1991.218800","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218800","url":null,"abstract":"The proposed first wall (FW) for the NET/ITER (Next European Torus/International Thermonuclear Experimental Reactor) consists of a box structure, made of stainless steel, with an active cooling of the side facing the plasma by a flow of water through the tubes inserted into it. An innovative brazing process to improve the quality of the brazed joints between cooling tubes and plates has been developed and optimized within the framework of NET technological task PDT 1.2. The brazing technique is based on creep expansion of the tubes during the high-temperature holding time of the brazing treatment, which minimizes the gap that was initially present around the tubes. A flat, mockup FW panel with plate dimensions 622*216*33 mm and with eight cooling tubes was built using this technique. Thermal fatigue tests are scheduled to be carried out in the near future on this mockup. An automatic, computerized procedure for the nondestructive examination, of the brazed joints using ultrasonic probes has also been developed.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"20 31","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1991-09-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"120844635","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.1991.218784
P. Hsueh, M.Z. Khan, J. Swanson, T. Feng, S. Dinkevich, J. Warren
As presently designed, the Burning Plasma Experiment vacuum vessel will be segmentally fabricated and assembled by bolted joints in the field. Due to geometry constraints, most of the bolted joints have significant eccentricity, which causes the joint behavior to be sensitive to joint clamping forces. Experience indicates that, as a result of this eccentricity, the joint will tend to open at the side closest to the applied load, with the extent of the opening depending on the initial preload. Analytical models and a testing program were developed to investigate and predict the nonlinear behavior of the vacuum vessel bolted joint. The test results are comparable with the analytical solutions in general, showing about 15% less load capacity than the finite element analysis predicted because the test specimens contained certain manufacturing and fabrication tolerances. The bolted joint capacity can be predicted by the finite element analysis, provided that an appropriate factor of safety be applied to cover these tolerances. The imperfections of the flanges and spacer surfaces are sensitive to the bolted joint characteristics. The design tolerances for the surfaces should be carefully specified.<>
{"title":"Testing program for Burning Plasma Experiment vacuum vessel bolted joint","authors":"P. Hsueh, M.Z. Khan, J. Swanson, T. Feng, S. Dinkevich, J. Warren","doi":"10.1109/FUSION.1991.218784","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218784","url":null,"abstract":"As presently designed, the Burning Plasma Experiment vacuum vessel will be segmentally fabricated and assembled by bolted joints in the field. Due to geometry constraints, most of the bolted joints have significant eccentricity, which causes the joint behavior to be sensitive to joint clamping forces. Experience indicates that, as a result of this eccentricity, the joint will tend to open at the side closest to the applied load, with the extent of the opening depending on the initial preload. Analytical models and a testing program were developed to investigate and predict the nonlinear behavior of the vacuum vessel bolted joint. The test results are comparable with the analytical solutions in general, showing about 15% less load capacity than the finite element analysis predicted because the test specimens contained certain manufacturing and fabrication tolerances. The bolted joint capacity can be predicted by the finite element analysis, provided that an appropriate factor of safety be applied to cover these tolerances. The imperfections of the flanges and spacer surfaces are sensitive to the bolted joint characteristics. The design tolerances for the surfaces should be carefully specified.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"39 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":"121369738","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.1991.218833
C. Outten, J. C. Barbour, B. Doyle, D. Walsh
An ECR (electron cyclotron resonance) plasma and Colutron ion gun were used to study He self-pumping by several possible pump-limiter materials: Ni, V, Al, and Ni/Al multilayers. Ni and V exhibited similar pumping capacities (6*10/sup 15/ He/cm/sup 2/, 200 eV), whereas Al showed a reduced capacity (6*10/sup 14/ He/cm/sup 2/, 200 eV) due to increased sputtering. An He retention model based on ion implantation ranges and sputtering rates agreed with the experimental data. The pumping efficiency increased significantly with ion energy. A novel multilayer/bilayer pumping concept showed improved pumping above that for single-element films. D/He trapping site competition is more important in V than in Ni. However, D/He site competition in V was shown to be less important above 400 degrees C where hydride decomposition is enhanced.<>
{"title":"He self-pumping by tokamak pump limiter materials: Al, V, Ni, and Ni/Al alloys","authors":"C. Outten, J. C. Barbour, B. Doyle, D. Walsh","doi":"10.1109/FUSION.1991.218833","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218833","url":null,"abstract":"An ECR (electron cyclotron resonance) plasma and Colutron ion gun were used to study He self-pumping by several possible pump-limiter materials: Ni, V, Al, and Ni/Al multilayers. Ni and V exhibited similar pumping capacities (6*10/sup 15/ He/cm/sup 2/, 200 eV), whereas Al showed a reduced capacity (6*10/sup 14/ He/cm/sup 2/, 200 eV) due to increased sputtering. An He retention model based on ion implantation ranges and sputtering rates agreed with the experimental data. The pumping efficiency increased significantly with ion energy. A novel multilayer/bilayer pumping concept showed improved pumping above that for single-element films. D/He trapping site competition is more important in V than in Ni. However, D/He site competition in V was shown to be less important above 400 degrees C where hydride decomposition is enhanced.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"129 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":"114370429","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.1991.218730
W. Burke
The construction of C-MOD is described briefly with an emphasis on thermal considerations. The temperature limits for the various components and the overall performance requirements for the temperature control system are discussed. The design of the temperature control system hardware and software is presented in some detail. Instrumentation for the poloidal and toroidal field coils is described.<>
{"title":"Temperature control and magnet instrumentation in Alcator C-MOD","authors":"W. Burke","doi":"10.1109/FUSION.1991.218730","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218730","url":null,"abstract":"The construction of C-MOD is described briefly with an emphasis on thermal considerations. The temperature limits for the various components and the overall performance requirements for the temperature control system are discussed. The design of the temperature control system hardware and software is presented in some detail. Instrumentation for the poloidal and toroidal field coils is described.<<ETX>>","PeriodicalId":318951,"journal":{"name":"[Proceedings] The 14th IEEE/NPSS Symposium Fusion Engineering","volume":"196 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":"116499404","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.1991.218695
J. Wesley
Thermal effects of magnetically sweeping the separatrix strike point on the outer divertor target of the ITER are calculated. For the 0.2 Hz*+or-12 cm sweep scenario proposed for ITER operations, the thermal capability of a generic target design is found to be slightly inadequate (by approximately 5%) to accommodate the full degree of plasma scrape-off peaking postulated as a design basis. The principal problem identified is that the 5-s sweep period is long relative to the 1.4-s thermal time constant of the divertor target. An increase of the sweep frequency to approximately 1 Hz is suggested; this increase would provide a power handling margin of approximately 25% relative to present operational criteria.<>
{"title":"Thermal effects of divertor sweeping in ITER","authors":"J. Wesley","doi":"10.1109/FUSION.1991.218695","DOIUrl":"https://doi.org/10.1109/FUSION.1991.218695","url":null,"abstract":"Thermal effects of magnetically sweeping the separatrix strike point on the outer divertor target of the ITER are calculated. For the 0.2 Hz*+or-12 cm sweep scenario proposed for ITER operations, the thermal capability of a generic target design is found to be slightly inadequate (by approximately 5%) to accommodate the full degree of plasma scrape-off peaking postulated as a design basis. The principal problem identified is that the 5-s sweep period is long relative to the 1.4-s thermal time constant of the divertor target. An increase of the sweep frequency to approximately 1 Hz is suggested; this increase would provide a power handling margin of approximately 25% relative to present operational criteria.<<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":"126557243","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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