Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518528
P. Zaccaria
The mechanical behaviour of the RFX machine is monitored during the pulses and the baking operations, both to verify the agreement between the actual and the design values of strains and displacements and to assess the effects of possible faults or anomalous operations of the machine. A specific measurement system was designed to provide quasistatic and dynamic measurements of strains and displacements. The paper describes the criteria which guided the design of the system, pointing out the specific requirements which had to be met to obtain a complete and correct mechanical information of the machine, as confirmed by the first results obtained during the RFX operations.
{"title":"The mechanical measurement system of RFX","authors":"P. Zaccaria","doi":"10.1109/FUSION.1993.518528","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518528","url":null,"abstract":"The mechanical behaviour of the RFX machine is monitored during the pulses and the baking operations, both to verify the agreement between the actual and the design values of strains and displacements and to assess the effects of possible faults or anomalous operations of the machine. A specific measurement system was designed to provide quasistatic and dynamic measurements of strains and displacements. The paper describes the criteria which guided the design of the system, pointing out the specific requirements which had to be met to obtain a complete and correct mechanical information of the machine, as confirmed by the first results obtained during the RFX operations.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"15 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"130987507","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.1993.518449
S. Fairfax
Alcator C-MOD is a compact, high performance tokamak designed to address reactor-relevant issues including diverter operation, confinement, and auxiliary heating. It incorporates flexible shaping of non-circular, diverted plasmas, strong ICRF heating, and many innovative engineering features to achieve high performance in a compact device with modest cost. Like its predecessors, Alcator A and Alcator C, Alcator C-MOD uses cryogenically cooled copper magnets to produce high toroidal fields (9 Tesla at 0.67 m) and strong ohmic heating (up to 3 MA.) The thick wall of the vacuum vessel, while complicating the normal problems of plasma initiation and control, are relevant to virtually all next-generation designs. The facility operated briefly in late 1991 and again in early 1992, when a terminal failed on one of the PF magnets. The experiment resumed operations in May 1993. This report describes the start-up and early operational experience, comparing with both design scenarios and previous experience on Alcator A and C. Results from operations during the summer of 1993 are presented.
{"title":"Start-up and early results from Alcator C-MOD","authors":"S. Fairfax","doi":"10.1109/FUSION.1993.518449","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518449","url":null,"abstract":"Alcator C-MOD is a compact, high performance tokamak designed to address reactor-relevant issues including diverter operation, confinement, and auxiliary heating. It incorporates flexible shaping of non-circular, diverted plasmas, strong ICRF heating, and many innovative engineering features to achieve high performance in a compact device with modest cost. Like its predecessors, Alcator A and Alcator C, Alcator C-MOD uses cryogenically cooled copper magnets to produce high toroidal fields (9 Tesla at 0.67 m) and strong ohmic heating (up to 3 MA.) The thick wall of the vacuum vessel, while complicating the normal problems of plasma initiation and control, are relevant to virtually all next-generation designs. The facility operated briefly in late 1991 and again in early 1992, when a terminal failed on one of the PF magnets. The experiment resumed operations in May 1993. This report describes the start-up and early operational experience, comparing with both design scenarios and previous experience on Alcator A and C. Results from operations during the summer of 1993 are presented.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"19 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"132397780","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.1993.518515
J. Doane, C. Moeller
We describe a compact beam steering design with tall rectangular waveguide components having a variable distance between the side walls. The movable elements are located conveniently far from the vacuum vessel wall. Some design dimensions for application to ECH on ITER are shown, using the requirements specified in the ITER conceptual design.
{"title":"Compact ITER-relevant ECH beam steering antenna design","authors":"J. Doane, C. Moeller","doi":"10.1109/FUSION.1993.518515","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518515","url":null,"abstract":"We describe a compact beam steering design with tall rectangular waveguide components having a variable distance between the side walls. The movable elements are located conveniently far from the vacuum vessel wall. Some design dimensions for application to ECH on ITER are shown, using the requirements specified in the ITER conceptual design.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"30 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127915115","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.1993.518459
A. Pizzuto, C. Sangiovanni
The shielding of the heat flux coming from plasma is one of the most limiting factors in plasma-facing component (PFC) design. In fact, the performance of a cooled divertor plate system is mainly limited by the heat transfer capability (maximum value of the critical heat flux CHF) and by the capability to sustain thermal stresses, even if the maximum allowable heat flux is determined by the thermal conductivity of the protective material (maximum temperature value on the plasma facing surface). A new concept for cooled thermal shield design was devised and tested. Analyses and tests demonstrate that the new concept introduces very high improvement in PFC design, in terms of both heat removal capability (very high CHF) and related stress performance. Up to 80 MW/m/sup 2/ under steady state were successfully applied.
{"title":"A high performance water-cooled thermal shield device","authors":"A. Pizzuto, C. Sangiovanni","doi":"10.1109/FUSION.1993.518459","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518459","url":null,"abstract":"The shielding of the heat flux coming from plasma is one of the most limiting factors in plasma-facing component (PFC) design. In fact, the performance of a cooled divertor plate system is mainly limited by the heat transfer capability (maximum value of the critical heat flux CHF) and by the capability to sustain thermal stresses, even if the maximum allowable heat flux is determined by the thermal conductivity of the protective material (maximum temperature value on the plasma facing surface). A new concept for cooled thermal shield design was devised and tested. Analyses and tests demonstrate that the new concept introduces very high improvement in PFC design, in terms of both heat removal capability (very high CHF) and related stress performance. Up to 80 MW/m/sup 2/ under steady state were successfully applied.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"37 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129031645","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.1993.518326
C. Wong, K. Redler, E. Reis, R. Will, E. Cheng, C.M. Hasan, S. Sharafat
The ARIES-IV Nested Shell Blanket (NSB) design is an alternate blanket concept of the ARIES-IV low activation helium-cooled reactor design. The reference design has the coolant routed in the poloidal direction and the inlet and outlet plena are located at the top and bottom of the torus. The NSB design has the high velocity coolant routed in the toroidal direction and the plena are located behind the blanket. This is of significance since the selected structural material is SiC-composite. The NSB is designed to have key high performance components with characteristic dimensions of no larger than 2 m. These components can be brazed to form the blanket module. For the divertor design, we eliminated the use of W as the divertor coating material by relying on the successful development of the gaseous divertor concept. The neutronics and thermal-hydraulic performance of both blanket concepts are similar. The selected blanket and divertor configurations can also meet all the projected structural, neutronics and thermal-hydraulics design limits and requirements. With the selected blanket and divertor materials, the design has a level of safety assurance rate of 1 (LSA-1), which indicates an inherently safe design.
{"title":"ARIES-IV Nested Shell Blanket design","authors":"C. Wong, K. Redler, E. Reis, R. Will, E. Cheng, C.M. Hasan, S. Sharafat","doi":"10.1109/FUSION.1993.518326","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518326","url":null,"abstract":"The ARIES-IV Nested Shell Blanket (NSB) design is an alternate blanket concept of the ARIES-IV low activation helium-cooled reactor design. The reference design has the coolant routed in the poloidal direction and the inlet and outlet plena are located at the top and bottom of the torus. The NSB design has the high velocity coolant routed in the toroidal direction and the plena are located behind the blanket. This is of significance since the selected structural material is SiC-composite. The NSB is designed to have key high performance components with characteristic dimensions of no larger than 2 m. These components can be brazed to form the blanket module. For the divertor design, we eliminated the use of W as the divertor coating material by relying on the successful development of the gaseous divertor concept. The neutronics and thermal-hydraulic performance of both blanket concepts are similar. The selected blanket and divertor configurations can also meet all the projected structural, neutronics and thermal-hydraulics design limits and requirements. With the selected blanket and divertor materials, the design has a level of safety assurance rate of 1 (LSA-1), which indicates an inherently safe design.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"31 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"126738249","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.1993.518514
T. E. Harris, W. Cary
One of the primary components of the DIII-D radiofrequency (RF) program over the past seven years has been the 60 GHz electron cyclotron resonant heating (ECRH) system. The system now consists of eight units capable of operating and controlling eight Varian VGE-8006 60 GHz, 200 kW gyrotrons along with their associated waveguide components. This paper will discuss the operational upgrades and the overall system performance. Many modifications were instituted to enhance the system operation and performance. Modifications discussed in this paper include an improved gyrotron tube-fault response network, a computer controlled pulse-timing and sequencing system, and an improved high-voltage power supply control interface. The discussion on overall system performance will include operating techniques used to improve system operations and reliability. The techniques discussed apply to system start-up procedures, operating the system in a conditioning mode, and operating the system during DIII-D plasma operations.
{"title":"Operational upgrades to the DIII-D 60 GHz electron cyclotron resonant heating system","authors":"T. E. Harris, W. Cary","doi":"10.1109/FUSION.1993.518514","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518514","url":null,"abstract":"One of the primary components of the DIII-D radiofrequency (RF) program over the past seven years has been the 60 GHz electron cyclotron resonant heating (ECRH) system. The system now consists of eight units capable of operating and controlling eight Varian VGE-8006 60 GHz, 200 kW gyrotrons along with their associated waveguide components. This paper will discuss the operational upgrades and the overall system performance. Many modifications were instituted to enhance the system operation and performance. Modifications discussed in this paper include an improved gyrotron tube-fault response network, a computer controlled pulse-timing and sequencing system, and an improved high-voltage power supply control interface. The discussion on overall system performance will include operating techniques used to improve system operations and reliability. The techniques discussed apply to system start-up procedures, operating the system in a conditioning mode, and operating the system during DIII-D plasma operations.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"36 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"116565204","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.1993.518505
T. Bigelow, C.F. Fogelman, F. Baity, M. Carter, D. Hoffman, P. Ryan, J.J. Yugo, S.N. Golovato, P. Bonoli
The folded waveguide (FWG) launcher is being investigated as an improved antenna configuration for plasma heating in the ion cyclotron range of frequencies (ICRF). A development FWG launcher was successfully tested at Oak Ridge National Laboratory (ORNL) with a low-density plasma load and found to have significantly greater power density capability than current strap-type antennas operating in similar plasmas. To further test the concept on a high density tokamak plasma, a collaboration has been set up between ORNL and Massachusetts Institute of Technology (MIT) to develop and test an 80-MHz, 2-MW FWG on the Alcator C-Mod tokamak at MIT. The radiofrequency (RF) electromagnetic modeling techniques and laboratory measurements used in the design of this antenna are described in this paper, A companion paper describes the mechanical design of the FWG.
{"title":"RF modeling and design of a folded waveguide launcher for the Alcator C-Mod tokamak","authors":"T. Bigelow, C.F. Fogelman, F. Baity, M. Carter, D. Hoffman, P. Ryan, J.J. Yugo, S.N. Golovato, P. Bonoli","doi":"10.1109/FUSION.1993.518505","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518505","url":null,"abstract":"The folded waveguide (FWG) launcher is being investigated as an improved antenna configuration for plasma heating in the ion cyclotron range of frequencies (ICRF). A development FWG launcher was successfully tested at Oak Ridge National Laboratory (ORNL) with a low-density plasma load and found to have significantly greater power density capability than current strap-type antennas operating in similar plasmas. To further test the concept on a high density tokamak plasma, a collaboration has been set up between ORNL and Massachusetts Institute of Technology (MIT) to develop and test an 80-MHz, 2-MW FWG on the Alcator C-Mod tokamak at MIT. The radiofrequency (RF) electromagnetic modeling techniques and laboratory measurements used in the design of this antenna are described in this paper, A companion paper describes the mechanical design of the FWG.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"10 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"125162219","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.1993.518353
H. Ninomiya, Y. Kamada, N. Miya, S. Nakajima, S. Oguri, A. Oikawa, A. Sakasai, Y. Takahashi, T. Takizuka, N. Toyoshima, S. Nakagawa, K. Nakashima, M. Otsuka
A conceptual design study of a steady-state tokamak, JT-60 Super Upgrade, is being carried out. The capability of the present JT-60 facility will be fully utilized for this upgrade. The mission of JT-60SU is to establish integrated basis of physics and technology for steady-state tokamak reactors. In JT-60SU, steady-state physics will be evaluated in the intermediate parameter region between the present tokamaks and steady-state tokamak reactors. Technology development for long pulse operation and research for engineering safety will also be pursued.
{"title":"Concept of JT-60 Super Upgrade","authors":"H. Ninomiya, Y. Kamada, N. Miya, S. Nakajima, S. Oguri, A. Oikawa, A. Sakasai, Y. Takahashi, T. Takizuka, N. Toyoshima, S. Nakagawa, K. Nakashima, M. Otsuka","doi":"10.1109/FUSION.1993.518353","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518353","url":null,"abstract":"A conceptual design study of a steady-state tokamak, JT-60 Super Upgrade, is being carried out. The capability of the present JT-60 facility will be fully utilized for this upgrade. The mission of JT-60SU is to establish integrated basis of physics and technology for steady-state tokamak reactors. In JT-60SU, steady-state physics will be evaluated in the intermediate parameter region between the present tokamaks and steady-state tokamak reactors. Technology development for long pulse operation and research for engineering safety will also be pursued.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"49 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"123859180","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.1993.518328
B. Claude, T. Hua, David B. Black, I. Kirillov, S. I. Sidorenkov, Anatoly M. Shapiro, I. A. Evtushenko
A liquid metal MHD (magnetohydrodynamic)/heat transfer test was conducted at the ALEX (Argonne Liquid Metal Experiment) facility of ANL (Argonne National Laboratory), jointly between ANL and NIIEFA (Efremov Institute). The test section was a rectangular slotted channel geometry (meaning the channel has a high aspect ratio, in this case 10:1, and the long side is parallel to the applied magnetic field). Isothermal and heat transfer data were collected. A heat flux of /spl sim/9 W/cm/sup 2/ was applied to the top horizontal surface (the long side) of the test section. Hartmann numbers to 1050 (2 Tesla), interaction parameters to 9/spl times/10/sup 3/, Peclet numbers of 10-200, based on the half-width of the small dimension (7 mm), and velocities of 1-75 cm/sec. were achieved. The working fluid was NaK (sodium potassium eutectic). All four interior walls were bare, 300-series stainless steel, conducting walls.
{"title":"Liquid metal MHD and heat transfer in a tokamak blanket slotted coolant channel","authors":"B. Claude, T. Hua, David B. Black, I. Kirillov, S. I. Sidorenkov, Anatoly M. Shapiro, I. A. Evtushenko","doi":"10.1109/FUSION.1993.518328","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518328","url":null,"abstract":"A liquid metal MHD (magnetohydrodynamic)/heat transfer test was conducted at the ALEX (Argonne Liquid Metal Experiment) facility of ANL (Argonne National Laboratory), jointly between ANL and NIIEFA (Efremov Institute). The test section was a rectangular slotted channel geometry (meaning the channel has a high aspect ratio, in this case 10:1, and the long side is parallel to the applied magnetic field). Isothermal and heat transfer data were collected. A heat flux of /spl sim/9 W/cm/sup 2/ was applied to the top horizontal surface (the long side) of the test section. Hartmann numbers to 1050 (2 Tesla), interaction parameters to 9/spl times/10/sup 3/, Peclet numbers of 10-200, based on the half-width of the small dimension (7 mm), and velocities of 1-75 cm/sec. were achieved. The working fluid was NaK (sodium potassium eutectic). All four interior walls were bare, 300-series stainless steel, conducting walls.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"8 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127769127","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.1993.518330
E. Cheng, J. Garner, M. Simnad, J. Talbot
A preliminary design investigation was conducted for a low-activation D-T fusion power reactor blanket employing SiC as the structure and liquid lead-lithium alloy as the breeder/coolant. Developmental issues such as SiC and Pb-Li compatibility and bismuth removal from Pb-Li are outlined and discussed.
{"title":"A low-activation fusion blanket with SiC structure and Pb-Li breeder","authors":"E. Cheng, J. Garner, M. Simnad, J. Talbot","doi":"10.1109/FUSION.1993.518330","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518330","url":null,"abstract":"A preliminary design investigation was conducted for a low-activation D-T fusion power reactor blanket employing SiC as the structure and liquid lead-lithium alloy as the breeder/coolant. Developmental issues such as SiC and Pb-Li compatibility and bismuth removal from Pb-Li are outlined and discussed.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"20 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"1993-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"128049638","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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