Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518334
N. Sasaki, H. Ogawa, M. Obama, T. Honda, K. Shibanuma, S. Kakudate, M. Kondoh
Positioning accuracy is essential in remote maintenance of in-vessel components. To develop an accurate control system, mechanical analysis with the finite element method and dynamic simulation have been carried out. Furthermore, a 1/5 scale prototype model of the boom system has been constructed to develop its control method and to verify the appropriate analysis models. The deflection and the vibration of the boom were measured actually in experiments and were compared with the results of FEM analysis. When the stiffness of joints was defined as a spring element and bolted parts of the boom were simulated in the FEM models, the results of the analysis were quite similar to that of experiments. This indicates the FEM model is appropriate. We have also performed dynamic simulations taking into account the servo control of the boom. Through these comparison, it is proved that this kind of analysis models can be applied to the full scale articulated boom system.
{"title":"Development of the 1/5 scale model of articulated boom system for in-vessel remote maintenance of Fusion Experimental Reactor","authors":"N. Sasaki, H. Ogawa, M. Obama, T. Honda, K. Shibanuma, S. Kakudate, M. Kondoh","doi":"10.1109/FUSION.1993.518334","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518334","url":null,"abstract":"Positioning accuracy is essential in remote maintenance of in-vessel components. To develop an accurate control system, mechanical analysis with the finite element method and dynamic simulation have been carried out. Furthermore, a 1/5 scale prototype model of the boom system has been constructed to develop its control method and to verify the appropriate analysis models. The deflection and the vibration of the boom were measured actually in experiments and were compared with the results of FEM analysis. When the stiffness of joints was defined as a spring element and bolted parts of the boom were simulated in the FEM models, the results of the analysis were quite similar to that of experiments. This indicates the FEM model is appropriate. We have also performed dynamic simulations taking into account the servo control of the boom. Through these comparison, it is proved that this kind of analysis models can be applied to the full scale articulated boom system.","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":"127637745","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.1993.518393
W. Cary, J. A. Allen, R. Pinsker, C. Petty
On the basis of the rapidly increasing power, speed, and decreasing cost of the personal computer (microcomputer) it was felt that a real time data acquisition and control system could be configured quickly and very cost effectively. It was further felt that by using a high level or object-oriented programming language that considerable time and expense could be saved and at the same time increase system flexibility. This paper will attempt to address the desired system requirements and performance for both the control of the high power transmitters and for the data acquisition and presentation of the information.
{"title":"ICH RF system data acquisition and real time control using a microcomputer system","authors":"W. Cary, J. A. Allen, R. Pinsker, C. Petty","doi":"10.1109/FUSION.1993.518393","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518393","url":null,"abstract":"On the basis of the rapidly increasing power, speed, and decreasing cost of the personal computer (microcomputer) it was felt that a real time data acquisition and control system could be configured quickly and very cost effectively. It was further felt that by using a high level or object-oriented programming language that considerable time and expense could be saved and at the same time increase system flexibility. This paper will attempt to address the desired system requirements and performance for both the control of the high power transmitters and for the data acquisition and presentation of the information.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"21 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":"127702869","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.1993.518537
E. Reis, E. Chin, K. Redler, F. Williams
Structural design calculations and finite element analyses of the conceptual designs of the Tokamak Physics Experiment (TPX) divertor and inboard limiter were performed for thermal and electromagnetic induced mechanical loads. Finite element analyses of the divertor and inboard limiter support structures were performed for loads due to halo currents and eddy currents due to a plasma disruption. The results show the conceptual designs satisfy primary stress allowables. A number of scoping studies were performed to evaluate the thermal and structural response of various tile materials and designs for the TPX divertor. The purpose of these studies was to investigate what possible gains would occur if the present design requirements for the heat flux surfaces were eased. The studies were performed for beryllium and various carbon-carbon materials brazed to a dispersion strengthened copper tube. The studies included the effects of a soft copper compliant layer of varying thicknesses interfacing the copper tube and the tile. Elastic-plastic thermal stress analyses were performed of 1D, 2D, and 4D carbon-carbon and beryllium monoblock designs and for a saddleblock design with 1D carbon-carbon. The residual stresses and amount of plastic straining in the copper tube during the braze cycle are accounted for in computing the stress state after the brazing process and during steady state operating conditions.
{"title":"Structural analysis of the TPX plasma facing components","authors":"E. Reis, E. Chin, K. Redler, F. Williams","doi":"10.1109/FUSION.1993.518537","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518537","url":null,"abstract":"Structural design calculations and finite element analyses of the conceptual designs of the Tokamak Physics Experiment (TPX) divertor and inboard limiter were performed for thermal and electromagnetic induced mechanical loads. Finite element analyses of the divertor and inboard limiter support structures were performed for loads due to halo currents and eddy currents due to a plasma disruption. The results show the conceptual designs satisfy primary stress allowables. A number of scoping studies were performed to evaluate the thermal and structural response of various tile materials and designs for the TPX divertor. The purpose of these studies was to investigate what possible gains would occur if the present design requirements for the heat flux surfaces were eased. The studies were performed for beryllium and various carbon-carbon materials brazed to a dispersion strengthened copper tube. The studies included the effects of a soft copper compliant layer of varying thicknesses interfacing the copper tube and the tile. Elastic-plastic thermal stress analyses were performed of 1D, 2D, and 4D carbon-carbon and beryllium monoblock designs and for a saddleblock design with 1D carbon-carbon. The residual stresses and amount of plastic straining in the copper tube during the braze cycle are accounted for in computing the stress state after the brazing process and during steady state operating conditions.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"3 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":"126401456","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.1993.518452
I. Kondo
The performance of JT-60U has progressed with technical advances such as in wall conditionings, control abilities and heating performances as well as experimental optimization skills, while the issues coming from machine structure such as magnetic ripple, error field are of concern.
{"title":"Status of machine and operation of JT-60","authors":"I. Kondo","doi":"10.1109/FUSION.1993.518452","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518452","url":null,"abstract":"The performance of JT-60U has progressed with technical advances such as in wall conditionings, control abilities and heating performances as well as experimental optimization skills, while the issues coming from machine structure such as magnetic ripple, error field are of concern.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"47 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":"115819920","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.1993.518331
M. Enoeda, S. Satoh, T. Kurasawa, H. Takatsu
Effective thermal conductivity of Li/sub 2/O and Be sphere packed bed was measured for the design of pebble layered type breeding blanket of ITER (CDA). Measurement runs were performed with 1 mm/spl phi/ Be sphere in the temperature range of 150 C to 650 C for Li/sub 2/O and 130 C to 320 C for Be. As the result, the effective thermal conductivity of 1 mm/spl phi/ Li/sub 2/O was influenced by contact area between spheres and packing fraction due to its reactivity. The measured value was about 0.9 W/mK with the packing fraction of 48%. Effective thermal conductivity of 1,2 and 3 mm/spl phi/ Be sphere packed beds were observed as k=1.56+3.25/spl times/10/sup -2/ T, k=2.43+2.15/spl times/10/sup -3/ T, k=2.51+1.45/spl times/10/sup -3/ T, respectively, where k is effective thermal conductivity [W/mK] and T is temperature [C]. Measured packing fraction was 56, 61 and 64% for 1, 2 and 3 mm/spl phi/ Be.
{"title":"Measurement of effective thermal conductivity of lithium oxide and beryllium sphere packed bed","authors":"M. Enoeda, S. Satoh, T. Kurasawa, H. Takatsu","doi":"10.1109/FUSION.1993.518331","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518331","url":null,"abstract":"Effective thermal conductivity of Li/sub 2/O and Be sphere packed bed was measured for the design of pebble layered type breeding blanket of ITER (CDA). Measurement runs were performed with 1 mm/spl phi/ Be sphere in the temperature range of 150 C to 650 C for Li/sub 2/O and 130 C to 320 C for Be. As the result, the effective thermal conductivity of 1 mm/spl phi/ Li/sub 2/O was influenced by contact area between spheres and packing fraction due to its reactivity. The measured value was about 0.9 W/mK with the packing fraction of 48%. Effective thermal conductivity of 1,2 and 3 mm/spl phi/ Be sphere packed beds were observed as k=1.56+3.25/spl times/10/sup -2/ T, k=2.43+2.15/spl times/10/sup -3/ T, k=2.51+1.45/spl times/10/sup -3/ T, respectively, where k is effective thermal conductivity [W/mK] and T is temperature [C]. Measured packing fraction was 56, 61 and 64% for 1, 2 and 3 mm/spl phi/ Be.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"28 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":"130084408","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.1993.518295
A. Roquemore, D. Jassby, L. C. Johnson, J. Strachan, C. Barnes
TFTR will soon enter its D-T phase with the introduction of tritium. This will result in the production of neutrons having 14 MeV energy which is significantly greater than the 2.5-MeV neutrons encountered during D-D operation. In preparation for the D-T phase, a calibration of the four neutron detection systems was performed using a 14-MeV neutron generator producing 10/sup 8/ n/sec. To account for the spatial extent of the toroidally shaped plasma and for neutrons scattered from surrounding structures, detector responses were determined with the source positioned at many locations inside the vacuum vessel. Before the generator could be used as a calibration source, a characterization of its total yield and angular emission properties was obtained. The total yield was determined by aluminum activation methods to within /spl plusmn/6%, while the angular emission was found to be anisotropic in the forward and reverse cones along the generator axis. After the characterization was performed, the generator was mounted on a moveable track inside the vacuum vessel, where it could be remotely moved across the view of each detector. This paper presents details of the methods and results of the source characterization, together with initial results of the in-vessel D-T neutron calibration.
{"title":"Performance of a 14-MeV neutron generator as an in situ calibration source for TFTR","authors":"A. Roquemore, D. Jassby, L. C. Johnson, J. Strachan, C. Barnes","doi":"10.1109/FUSION.1993.518295","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518295","url":null,"abstract":"TFTR will soon enter its D-T phase with the introduction of tritium. This will result in the production of neutrons having 14 MeV energy which is significantly greater than the 2.5-MeV neutrons encountered during D-D operation. In preparation for the D-T phase, a calibration of the four neutron detection systems was performed using a 14-MeV neutron generator producing 10/sup 8/ n/sec. To account for the spatial extent of the toroidally shaped plasma and for neutrons scattered from surrounding structures, detector responses were determined with the source positioned at many locations inside the vacuum vessel. Before the generator could be used as a calibration source, a characterization of its total yield and angular emission properties was obtained. The total yield was determined by aluminum activation methods to within /spl plusmn/6%, while the angular emission was found to be anisotropic in the forward and reverse cones along the generator axis. After the characterization was performed, the generator was mounted on a moveable track inside the vacuum vessel, where it could be remotely moved across the view of each detector. This paper presents details of the methods and results of the source characterization, together with initial results of the in-vessel D-T neutron calibration.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"11 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":"130199404","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.1993.518534
M. Hollerbach, R. Lee, J.P. Smith, P. Taylor
An upgrade of the General Atomics DIII-D tokamak armor protection system has been completed. The upgrade consisted of armoring the outer wall and the divertor gas baffle with monolithic graphite tiles and cleaning the existing floor, ceiling, and inner wall tiles to remove any deposited impurity layer from the tile surfaces. The new tiles replace the graphite tiles used as local armor for neutral beam shine through, three graphite poloidal back-up limiter bands, and miscellaneous Inconel protection tiles. The total number of tiles increased from 1636 to 3200 and corresponding vessel coverage from 40% to 90%. A new, graphite armored, toroidally continuous, gas baffle between the outer wall and the biased divertor ring was installed in order to accommodate the cryocondensation pump that was installed in parallel with the outer wall tiles. To eliminate a source of copper in the plasma, GRAFOIL gaskets replaced the copper felt metal gaskets previously used as a compliant heat transfer interface between the inertially cooled tiles and the vessel wall. GRAFOIL, an exfoliated, flexible graphite material from Union Carbide, Inc., was used between each tile and the vessel wall and also between each tile and its hold-down hardware. Testing was performed to determine the mechanical compliance, thermal conductance, and vacuum characteristics of the GRAFOIL material. To further decrease the quantity of high Z-materials exposed to the plasma, the 1636 existing graphite tiles were identified, removed, and grit blasted to eliminate a thin layer of deposited metals which included nickel, chromium, and molybdenum. Prior to any processing, a selected set of tiles was tested for radioactivity, including tritium contamination. The tiles were grit blasted in a negative-pressure blasting cabinet using 37 /spl mu/m boron carbide powder as the blast media and dry nitrogen as the propellant. A 30 to 50 /spl mu/m-thick layer was removed from the surfaces of each tile. The tiles were cleaned in an ultrasonic alcohol bath and vacuum baked to 1000/spl deg/C prior to reinstallation. The installation of the graphite armor and gas baffle and the grit blasting of the tiles was performed in August 1992 through January 1993. Initial operational experience with the new armor system is summarized.
{"title":"Upgrade of the DIII-D vacuum vessel protection system","authors":"M. Hollerbach, R. Lee, J.P. Smith, P. Taylor","doi":"10.1109/FUSION.1993.518534","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518534","url":null,"abstract":"An upgrade of the General Atomics DIII-D tokamak armor protection system has been completed. The upgrade consisted of armoring the outer wall and the divertor gas baffle with monolithic graphite tiles and cleaning the existing floor, ceiling, and inner wall tiles to remove any deposited impurity layer from the tile surfaces. The new tiles replace the graphite tiles used as local armor for neutral beam shine through, three graphite poloidal back-up limiter bands, and miscellaneous Inconel protection tiles. The total number of tiles increased from 1636 to 3200 and corresponding vessel coverage from 40% to 90%. A new, graphite armored, toroidally continuous, gas baffle between the outer wall and the biased divertor ring was installed in order to accommodate the cryocondensation pump that was installed in parallel with the outer wall tiles. To eliminate a source of copper in the plasma, GRAFOIL gaskets replaced the copper felt metal gaskets previously used as a compliant heat transfer interface between the inertially cooled tiles and the vessel wall. GRAFOIL, an exfoliated, flexible graphite material from Union Carbide, Inc., was used between each tile and the vessel wall and also between each tile and its hold-down hardware. Testing was performed to determine the mechanical compliance, thermal conductance, and vacuum characteristics of the GRAFOIL material. To further decrease the quantity of high Z-materials exposed to the plasma, the 1636 existing graphite tiles were identified, removed, and grit blasted to eliminate a thin layer of deposited metals which included nickel, chromium, and molybdenum. Prior to any processing, a selected set of tiles was tested for radioactivity, including tritium contamination. The tiles were grit blasted in a negative-pressure blasting cabinet using 37 /spl mu/m boron carbide powder as the blast media and dry nitrogen as the propellant. A 30 to 50 /spl mu/m-thick layer was removed from the surfaces of each tile. The tiles were cleaned in an ultrasonic alcohol bath and vacuum baked to 1000/spl deg/C prior to reinstallation. The installation of the graphite armor and gas baffle and the grit blasting of the tiles was performed in August 1992 through January 1993. Initial operational experience with the new armor system is summarized.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"46 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":"134580243","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.1993.518477
H. Khater
A detailed activation analysis has been performed for the tokamak fusion power reactor ARIES-II. The reactor uses vanadium alloy as a structural material and liquid lithium as a coolant and tritium breeder. The total activity produced in the reactor at shutdown is 3848 MCi and drops to only 59 MCi during the first year following shutdown. Calculations of the decay heat showed that it is 53 MW at shutdown and it takes a relatively short time (<1 day) to decay by about a factor of 10. One week after shutdown, the values of the integrated decay heat generated in the structure are 548 and 1298 GJ for the reactor inboard and outboard regions, respectively. This heat represents less than 2% of the reactor thermal power and hence does not present a safety hazard. The biological hazard potential was calculated according to the NRC regulations specified in 10CFR20. The total BHP at shutdown is 388 x 10/sup 6/m/sup 3/ air. The radwaste classification of the ARIES-II structure has been evaluated according to both the NRC 10CFR61 and Fetter waste disposal concentration limits. Except for the reactor outboard blanket which would qualify as Class A low level waste, the rest of the reactor structure would only qualify for Class C rating. The outboard blanket has a Class A rating value of 0.95 which is based on allowing it to cool down for about 10 years following the end of the reactor lifetime.
{"title":"Activation and waste disposal of the ARIES-II tokamak fusion power reactor","authors":"H. Khater","doi":"10.1109/FUSION.1993.518477","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518477","url":null,"abstract":"A detailed activation analysis has been performed for the tokamak fusion power reactor ARIES-II. The reactor uses vanadium alloy as a structural material and liquid lithium as a coolant and tritium breeder. The total activity produced in the reactor at shutdown is 3848 MCi and drops to only 59 MCi during the first year following shutdown. Calculations of the decay heat showed that it is 53 MW at shutdown and it takes a relatively short time (<1 day) to decay by about a factor of 10. One week after shutdown, the values of the integrated decay heat generated in the structure are 548 and 1298 GJ for the reactor inboard and outboard regions, respectively. This heat represents less than 2% of the reactor thermal power and hence does not present a safety hazard. The biological hazard potential was calculated according to the NRC regulations specified in 10CFR20. The total BHP at shutdown is 388 x 10/sup 6/m/sup 3/ air. The radwaste classification of the ARIES-II structure has been evaluated according to both the NRC 10CFR61 and Fetter waste disposal concentration limits. Except for the reactor outboard blanket which would qualify as Class A low level waste, the rest of the reactor structure would only qualify for Class C rating. The outboard blanket has a Class A rating value of 0.95 which is based on allowing it to cool down for about 10 years following the end of the reactor lifetime.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"62 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":"126586486","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.1993.518423
C. Gung, M. Takayasu, J. Minervini
This paper demonstrates a practical process for evaluating AC losses of a full-size chrome-plated Nb/sub 3/Sn cable with key parameters determined from a laboratory-scale single-wire experiment. By using wire specifications given in the engineering design phase as experimental information, this process was applied to estimate AC losses in the inner-most layer of the ITER central solenoid.
{"title":"Experimental estimation of energy dissipation in ITER central solenoid superconductor","authors":"C. Gung, M. Takayasu, J. Minervini","doi":"10.1109/FUSION.1993.518423","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518423","url":null,"abstract":"This paper demonstrates a practical process for evaluating AC losses of a full-size chrome-plated Nb/sub 3/Sn cable with key parameters determined from a laboratory-scale single-wire experiment. By using wire specifications given in the engineering design phase as experimental information, this process was applied to estimate AC losses in the inner-most layer of the ITER central solenoid.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"9 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":"130692329","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.1993.518277
K.N. Sato, M. Onozuka, Y. Oda, H. Sakakita, R. Liang, S. Sudo, H. Kaneko, M. Sakamoto, S. Gotô
Prototype continuous pellet injector has been designed and is under construction for the Large Helical Device at National Institute for Fusion Science in Japan. This injector is a centrifuge type, aiming at long-pulse pellet injection. A centrifuge pellet injector is a mechanical device that utilizes centrifugal force to accelerate solid hydrogen-isotope pellets using a high-speed rotating rotor. The injector consists of a hydrogen-isotope filament extruder, a pellet forming device, and a pellet accelerator. It is expected that pellets of 1 mm and 1.5 mm in diameter and length will be accelerated up to 800 m/sec in this system.
{"title":"Development of continuous pellet injector for Large Helical Device","authors":"K.N. Sato, M. Onozuka, Y. Oda, H. Sakakita, R. Liang, S. Sudo, H. Kaneko, M. Sakamoto, S. Gotô","doi":"10.1109/FUSION.1993.518277","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518277","url":null,"abstract":"Prototype continuous pellet injector has been designed and is under construction for the Large Helical Device at National Institute for Fusion Science in Japan. This injector is a centrifuge type, aiming at long-pulse pellet injection. A centrifuge pellet injector is a mechanical device that utilizes centrifugal force to accelerate solid hydrogen-isotope pellets using a high-speed rotating rotor. The injector consists of a hydrogen-isotope filament extruder, a pellet forming device, and a pellet accelerator. It is expected that pellets of 1 mm and 1.5 mm in diameter and length will be accelerated up to 800 m/sec in this system.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"17 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":"132934260","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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