Pub Date : 1993-10-11DOI: 10.1109/FUSION.1993.518348
Y. Miura, R. Shimada
The study of the fusion reactor safety using the real reactor relevant plasma to examine the safety is essential for meeting new energy having the potential of public acceptance. If we have a high magnetic field technology that is three times higher than the present day tokamak, it is easy predict using the resent experimental result that the very small tokamak can reach the reactor grade plasma. These supporting experimental machines are important for the fusion reactor development to add the review method for safety and environmental view points. To obtain ultra-high field, the force free coil concept is developed. This force balanced coil (FBC) idea has a possibility to get higher field by canceling magnetic forces themselves. The magnetic configuration near plasma is tokamak-like one by using the multi-helical windings. The difference from ordinary tokamak is that the coils generate the toroidal and poloidal field all together and they ramp up with plasma current. Thus, this device needs the special operation to break down and ramp up phase of plasma. In this paper the high field ramp-up tokamak for fusion reactor safety research is proposed, and the feasibility applying the single layered multi-helical force balanced coil to tokamak is investigated with consideration to plasma confinement. The scenario to the nuclear burning phase in a pulse operation is simulated including the effect of the increase of the toroidal field with plasma current.
{"title":"Development of high field pulse tokamak using force free coil concept for fusion reactor safety research","authors":"Y. Miura, R. Shimada","doi":"10.1109/FUSION.1993.518348","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518348","url":null,"abstract":"The study of the fusion reactor safety using the real reactor relevant plasma to examine the safety is essential for meeting new energy having the potential of public acceptance. If we have a high magnetic field technology that is three times higher than the present day tokamak, it is easy predict using the resent experimental result that the very small tokamak can reach the reactor grade plasma. These supporting experimental machines are important for the fusion reactor development to add the review method for safety and environmental view points. To obtain ultra-high field, the force free coil concept is developed. This force balanced coil (FBC) idea has a possibility to get higher field by canceling magnetic forces themselves. The magnetic configuration near plasma is tokamak-like one by using the multi-helical windings. The difference from ordinary tokamak is that the coils generate the toroidal and poloidal field all together and they ramp up with plasma current. Thus, this device needs the special operation to break down and ramp up phase of plasma. In this paper the high field ramp-up tokamak for fusion reactor safety research is proposed, and the feasibility applying the single layered multi-helical force balanced coil to tokamak is investigated with consideration to plasma confinement. The scenario to the nuclear burning phase in a pulse operation is simulated including the effect of the increase of the toroidal field with plasma current.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"33 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":"132569267","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.1993.518479
S. K. Ho, M. D. Lowenthal
A computational module compatible with systems-level reactor design studies is being developed for incorporating the environmental and safety characteristics into design optimization. Preliminary and illustrative results for the analysis of HT-9 ferritic steel first wall structure are presented.
{"title":"A systems module for environmental and safety assessment in fusion reactor studies-illustrative results for HT-9 structures","authors":"S. K. Ho, M. D. Lowenthal","doi":"10.1109/FUSION.1993.518479","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518479","url":null,"abstract":"A computational module compatible with systems-level reactor design studies is being developed for incorporating the environmental and safety characteristics into design optimization. Preliminary and illustrative results for the analysis of HT-9 ferritic steel first wall structure are presented.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"59 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":"132583353","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.1993.518506
F. Carpignano, B. Coppi, M. Nassi
The ICRH system adopted for the Ignitor machine is characterized by a maximum power delivered to the plasma of 18 MW and a wide range of operating frequencies (100
{"title":"ICRH system for the Ignitor machine","authors":"F. Carpignano, B. Coppi, M. Nassi","doi":"10.1109/FUSION.1993.518506","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518506","url":null,"abstract":"The ICRH system adopted for the Ignitor machine is characterized by a maximum power delivered to the plasma of 18 MW and a wide range of operating frequencies (100</spl nu/<210 MHz). The main functions of the system are to control the time evolution of the plasma temperature and the toroidal current density profiles, to keep the region where g<1 small and to suppress the possible onset of sawtooth oscillations by fast particle stabilization in relatively low plasma density discharges. The ICRH system can also be used to accelerate the attainment of ignition, to extend the conditions under which ignition is possible, to explore the conditions under which the second stability region of finite /spl beta/ plasmas can be achieved, to produce significant levels of power from D-/sup 3/He fusion reactions, to reduce the Volt-sec requirement and to perform current drive experiments in low density discharges (n/sub e/<2/spl times/10/sup 20/ m/sup -3/). The wide range of frequencies has been adopted in order to operate in different regimes (/spl omega/=/spl omega//sub CD/, /spl omega/(C/sup 3/)/sub He/, 2/spl omega//sub CT/, /spl omega//sub CH/ at maximum toroidal magnetic field). This additional source of heating allows Ignitor to produce a significant level of /spl alpha/ power in low plasma current and low toroidal magnetic field discharges, while operating at lower mechanical and thermal stresses and sustaining the discharges for a longer period of time. The ICRH system is composed of 6 antennae, completely inserted in first wall recesses, that are driven by amplifiers through the equatorial ports of the machine. Each antenna module composed of straps grouped in poloidal pairs, is able to couple up to 3 MW of heating power to the plasma.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"208 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":"133700386","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.1993.518442
E. Mogahed, I. Sviatoslavsky
The SIRIUS-P conceptual design study is of a 1.0 GWe laser driven inertial confinement fusion power reactor utilizing near symmetric illumination of direct drive targets. Sixty laser beams providing a total of 3.4 MJ of energy are used at a repetition rate of 6.7 Hz with a nominal target gain of 118. The spherical chamber has an internal radius of 6.5 m and consists of a first wall assembly made from carbon-carbon composite material, and a blanket assembly made of SiC composite material. The chamber is cooled by a flowing granular bed of solid ceramic materials, non-breeding TiO/sub 2/ for the first wall assembly and breeding Li/sub 2/O for the blanket assembly. Helium gas (P=0.15 MPa) is used in a fluidized bed outside the reactor to return the particles to the top of the reactor. A moving bed is chosen over a fluidized bed because of its superior heat transfer capability. The heat transfer in a moving bed depends on the level of agitation and on the effective thermal conductivity of the solid material and the interstitial gas, whereas in a fluidized bed, it is entirely dominated by the thermal conductivity of the carrier gas. This paper describes the two-dimensional thermo-structural steady state analysis of the first wall elements at several critical locations utilizing the finite element analysis code, ANSYS, with r-/spl theta/ modeling. The stresses are dominated by bending due to the internal pressure of the He gas; modifying the shape of the tube from purely elliptical, while keeping the area constant reduces the stresses.
{"title":"Thermal and structural analysis of the first wall in the SIRIUS-P reactor","authors":"E. Mogahed, I. Sviatoslavsky","doi":"10.1109/FUSION.1993.518442","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518442","url":null,"abstract":"The SIRIUS-P conceptual design study is of a 1.0 GWe laser driven inertial confinement fusion power reactor utilizing near symmetric illumination of direct drive targets. Sixty laser beams providing a total of 3.4 MJ of energy are used at a repetition rate of 6.7 Hz with a nominal target gain of 118. The spherical chamber has an internal radius of 6.5 m and consists of a first wall assembly made from carbon-carbon composite material, and a blanket assembly made of SiC composite material. The chamber is cooled by a flowing granular bed of solid ceramic materials, non-breeding TiO/sub 2/ for the first wall assembly and breeding Li/sub 2/O for the blanket assembly. Helium gas (P=0.15 MPa) is used in a fluidized bed outside the reactor to return the particles to the top of the reactor. A moving bed is chosen over a fluidized bed because of its superior heat transfer capability. The heat transfer in a moving bed depends on the level of agitation and on the effective thermal conductivity of the solid material and the interstitial gas, whereas in a fluidized bed, it is entirely dominated by the thermal conductivity of the carrier gas. This paper describes the two-dimensional thermo-structural steady state analysis of the first wall elements at several critical locations utilizing the finite element analysis code, ANSYS, with r-/spl theta/ modeling. The stresses are dominated by bending due to the internal pressure of the He gas; modifying the shape of the tube from purely elliptical, while keeping the area constant reduces the stresses.","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":"131883270","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.1993.518365
J. Phillips, C. Baxi, R. Hong
The current DIII-D neutral beam system is a nominal five second pulse length upgrade of a previous design rated for only 500 msec operation. While the ion sources are rated for 60 sec operation, in practice pulse lengths are limited both by the beamline internal components ability to handle the fraction of the power which is scraped off, and by the power supplies ability to provide pulse lengths of greater than 5 sec. This paper examines the capability of the existing DIII-D neutral beamline heat removing components both in terms of present and desired operating parameters. To date, at 2.5 MW per ion source, pulses are limited to 3.5 sec by beamline internal components, while at lower ratings of 2.0 MW per ion source, up to 5 sec pulses have been achieved, Recent advances and demonstration of the extraction of 3.5 MW per DIII-D ion source give an even wider window of operating conditions. A full series of beamline thermocouple data has been collected to determine the heat loading and implied surface temperatures for the various DIII-D neutral beamline internal components. These data will be presented along with an analysis of the needs for specific component upgrades, given a desire for 10 sec operation.
{"title":"Investigation of the heat handling capabilities of DIII-D neutral beamline internal components","authors":"J. Phillips, C. Baxi, R. Hong","doi":"10.1109/FUSION.1993.518365","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518365","url":null,"abstract":"The current DIII-D neutral beam system is a nominal five second pulse length upgrade of a previous design rated for only 500 msec operation. While the ion sources are rated for 60 sec operation, in practice pulse lengths are limited both by the beamline internal components ability to handle the fraction of the power which is scraped off, and by the power supplies ability to provide pulse lengths of greater than 5 sec. This paper examines the capability of the existing DIII-D neutral beamline heat removing components both in terms of present and desired operating parameters. To date, at 2.5 MW per ion source, pulses are limited to 3.5 sec by beamline internal components, while at lower ratings of 2.0 MW per ion source, up to 5 sec pulses have been achieved, Recent advances and demonstration of the extraction of 3.5 MW per DIII-D ion source give an even wider window of operating conditions. A full series of beamline thermocouple data has been collected to determine the heat loading and implied surface temperatures for the various DIII-D neutral beamline internal components. These data will be presented along with an analysis of the needs for specific component upgrades, given a desire for 10 sec operation.","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":"134479265","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.1993.518297
B. McHarg
The General Atomics DIII-D tokamak fusion experiment is now collecting over 80 MB of data per discharge once every 10 min, and that quantity is expected to double within the next year. The size of the data files, even in compressed format, is becoming increasingly difficult to handle. Data is also being acquired now on a variety of UNIX systems as well as MicroVAX and MODCOMP computer systems. The existing computers collect all the data into a single shot file, and this data collection is taking an ever increasing amount of time as the total quantity of data increases. Data is not available to experimenters until it has been collected into the shot file, which is in conflict with the substantial need for data examination on a timely basis between shots. The experimenters are also spread over many different types of computer systems (possibly located at other sites). To improve data, availability and handling, software has been developed to allow individual computer systems to create their own shot files locally. The data interface routine PTDATA that is used to access DIII-D data has been modified so that a user's code on any computer can access data from any computer where that data might be located. This data access is transparent to the user. Breaking up the shot file into separate files in multiple locations also impacts software used for data archiving, data management, and data restoration.
{"title":"Access to DIII-D data located in multiple files and multiple locations","authors":"B. McHarg","doi":"10.1109/FUSION.1993.518297","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518297","url":null,"abstract":"The General Atomics DIII-D tokamak fusion experiment is now collecting over 80 MB of data per discharge once every 10 min, and that quantity is expected to double within the next year. The size of the data files, even in compressed format, is becoming increasingly difficult to handle. Data is also being acquired now on a variety of UNIX systems as well as MicroVAX and MODCOMP computer systems. The existing computers collect all the data into a single shot file, and this data collection is taking an ever increasing amount of time as the total quantity of data increases. Data is not available to experimenters until it has been collected into the shot file, which is in conflict with the substantial need for data examination on a timely basis between shots. The experimenters are also spread over many different types of computer systems (possibly located at other sites). To improve data, availability and handling, software has been developed to allow individual computer systems to create their own shot files locally. The data interface routine PTDATA that is used to access DIII-D data has been modified so that a user's code on any computer can access data from any computer where that data might be located. This data access is transparent to the user. Breaking up the shot file into separate files in multiple locations also impacts software used for data archiving, data management, and data restoration.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"122 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":"134633432","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.1993.518284
A.G. Heics, W. Shmayda
A prototypical metal hydride based recirculating glovebox cleanup system was commissioned and tested with tritium. Getter material SAES St 198 was selected for its ability to effectively remove tritium and trace impurities from inert or nitrogen glovebox cover gases and its ease of tritium recovery by heating to elevated temperatures. The Secondary Enclosure Clean-up (SEC) system utilizes a programmable controller for process control and system isolation and alarm in the event of an abnormal condition. The system was used to detect glovebox air in leakage by tracking the moisture level within the glovebox when the bed is bypassed. An aliquot of 0.5 Ci of tritium, intentionally released into a glovebox to demonstrate the system performance, was effectively removed by the getter bed in about 20 minutes or about 7 system time constants.
一个基于金属氢化物的循环手套箱清理系统原型进行了调试，并对其进行了氚测试。吸气材料SAES St 198之所以被选中，是因为它能够有效地从惰性气体或氮气手套箱盖气体中去除氚和微量杂质，并且易于通过加热到高温回收氚。次级外壳清理(SEC)系统利用可编程控制器进行过程控制和系统隔离，并在异常情况下发出警报。当床层被绕过时，该系统通过跟踪手套箱内的湿度水平来检测手套箱中的泄漏空气。将0.5 Ci的氚故意释放到手套箱中以演示系统性能，在大约20分钟或大约7个系统时间常数内被吸气床有效地去除。
{"title":"Development of a secondary enclosure clean-up system for tritium systems","authors":"A.G. Heics, W. Shmayda","doi":"10.1109/FUSION.1993.518284","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518284","url":null,"abstract":"A prototypical metal hydride based recirculating glovebox cleanup system was commissioned and tested with tritium. Getter material SAES St 198 was selected for its ability to effectively remove tritium and trace impurities from inert or nitrogen glovebox cover gases and its ease of tritium recovery by heating to elevated temperatures. The Secondary Enclosure Clean-up (SEC) system utilizes a programmable controller for process control and system isolation and alarm in the event of an abnormal condition. The system was used to detect glovebox air in leakage by tracking the moisture level within the glovebox when the bed is bypassed. An aliquot of 0.5 Ci of tritium, intentionally released into a glovebox to demonstrate the system performance, was effectively removed by the getter bed in about 20 minutes or about 7 system time constants.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"7 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":"114241527","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.1993.518324
Y. Baoshan, Yan Kailin, Liu Shiming, Tan Manqio, Wang Shangbing, C. Xiaoguang, Jin Qinghua, Qi Wanli
The HL-1M tokamak is the modification device of HL-1. In this paper, the consideration for plasma equilibrium and the design of the coil, power supply and control circuits for the plasma position control system in HL-1M tokamak are presented, four-quadrant thyristor converter (FQTC) with the circulating current for the power supply of the fast vertical field (FF) coils and the control circuits are described, the commissioning results for the FF power supply with a dummy coil are given.
{"title":"Plasma position control system in HL-1M tokamak","authors":"Y. Baoshan, Yan Kailin, Liu Shiming, Tan Manqio, Wang Shangbing, C. Xiaoguang, Jin Qinghua, Qi Wanli","doi":"10.1109/FUSION.1993.518324","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518324","url":null,"abstract":"The HL-1M tokamak is the modification device of HL-1. In this paper, the consideration for plasma equilibrium and the design of the coil, power supply and control circuits for the plasma position control system in HL-1M tokamak are presented, four-quadrant thyristor converter (FQTC) with the circulating current for the power supply of the fast vertical field (FF) coils and the control circuits are described, the commissioning results for the FF power supply with a dummy coil are given.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"1 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":"115835300","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.1993.518290
A. Cosler, E. Kemmereit, H. Soltwisch
A combined HCN laser interferometer/polarimeter comprising nine vertical probing beams has been in routine use on TEXTOR for simultaneous measurements of the electron density and poloidal magnetic field distributions and for feedback control of the gas injection and horizontal plasma positioning systems. Motivated by a growing interest in rapid changes of the internal magnetic field structure as well as by the need for reliable control diagnostics of the plasma shape and vertical position under long pulse conditions, the instrument has been complemented recently with a horizontal channel. The upgrading has been achieved by minor modifications of the existing optical arrangement and mechanical structure without impairing the function of the original apparatus. In this contribution, the constraints and technical solutions for the alterations are outlined and the improved performance is illustrated by practical examples.
{"title":"Upgrade of the far-infrared interferometer and polarimeter on TEXTOR","authors":"A. Cosler, E. Kemmereit, H. Soltwisch","doi":"10.1109/FUSION.1993.518290","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518290","url":null,"abstract":"A combined HCN laser interferometer/polarimeter comprising nine vertical probing beams has been in routine use on TEXTOR for simultaneous measurements of the electron density and poloidal magnetic field distributions and for feedback control of the gas injection and horizontal plasma positioning systems. Motivated by a growing interest in rapid changes of the internal magnetic field structure as well as by the need for reliable control diagnostics of the plasma shape and vertical position under long pulse conditions, the instrument has been complemented recently with a horizontal channel. The upgrading has been achieved by minor modifications of the existing optical arrangement and mechanical structure without impairing the function of the original apparatus. In this contribution, the constraints and technical solutions for the alterations are outlined and the improved performance is illustrated by practical examples.","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":"121639417","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.1993.518359
M. Nishi, H. Nakajima, T. Ando, T. Hiyama, Y. Takahashi, K. Okuno, K. Yoshida, T. Kato, K. Kawano, T. Isono, M. Sugimoto, N. Koizumi, K. Hamada, T. Sasaki, K. Imahashi, S. Iwamoto, H. Ebisu, T. Ouchi, K. Ohtsu, J. Okayama, M. Oshikiri, T. Kawasaki, S. Seki, S. Sekiguchi, T. Takahashi, Y. Takaya, F. Tajiri, H. Tsukamoto, H. Hanawa, F. Hosono, A. Miyake, Y. Yasukawa, K. Yamamoto, H. Wakabayashi, Y. Wadayama, H. Tsuji
As a high-field pulse-poloidal test coil in the development program of superconducting coils for fusion machines, a 1 m-inner diameter superconducting coil named DPC-EX wound with 10 kA Nb/sub 3/Sn cable-in-conduit forced-cooled conductor was fabricated and its various performances were investigated. In its second test, the ramp-rate limitation phenomenon, i.e. the dependence of the coil current at quench on the current ramp speed, was studied in detail and a lot of important data for the design of large pulse coils were achieved.
作为核聚变用超导线圈研制计划中的高场脉冲极向试验线圈，采用10 kA Nb/sub - 3/Sn管内强制冷却导线绕制了内径为1 m的超导线圈DPC-EX，并对其各项性能进行了研究。在第二次试验中，详细研究了缓变速率限制现象，即线圈在猝灭时电流对电流缓变速率的依赖关系，获得了许多大脉冲线圈设计的重要数据。
{"title":"Ramp-rate limitation test results of the Nb/sub 3/Sn Demo Poloidal Coil (DPC-EX)","authors":"M. Nishi, H. Nakajima, T. Ando, T. Hiyama, Y. Takahashi, K. Okuno, K. Yoshida, T. Kato, K. Kawano, T. Isono, M. Sugimoto, N. Koizumi, K. Hamada, T. Sasaki, K. Imahashi, S. Iwamoto, H. Ebisu, T. Ouchi, K. Ohtsu, J. Okayama, M. Oshikiri, T. Kawasaki, S. Seki, S. Sekiguchi, T. Takahashi, Y. Takaya, F. Tajiri, H. Tsukamoto, H. Hanawa, F. Hosono, A. Miyake, Y. Yasukawa, K. Yamamoto, H. Wakabayashi, Y. Wadayama, H. Tsuji","doi":"10.1109/FUSION.1993.518359","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518359","url":null,"abstract":"As a high-field pulse-poloidal test coil in the development program of superconducting coils for fusion machines, a 1 m-inner diameter superconducting coil named DPC-EX wound with 10 kA Nb/sub 3/Sn cable-in-conduit forced-cooled conductor was fabricated and its various performances were investigated. In its second test, the ramp-rate limitation phenomenon, i.e. the dependence of the coil current at quench on the current ramp speed, was studied in detail and a lot of important data for the design of large pulse coils were achieved.","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":"125387909","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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