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.1993.518324
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.1993.518347
S. K. Ingram
In the existing budgetary environment, low-cost, high-payoff fusion experiments are increasingly attractive. IGNITEX, a homopolar generator powered single-turn tokamak, has the potential to achieve ignition at relatively low cost, A 60 MJ homopolar generator facility (with an easy upgrade path to 90 MJ) is in place at CEM-UT. Using this facility, CEM-UT researchers have produced a 20 T on-axis magnetic field in a 1/16 scale IGNITEX prototype. This paper presents the results of optimization studies conducted to determine the lowest-cost homopolar generator facility which can power an IGNITEX prototype of any chosen scale operating at 20 T on-axis.
{"title":"Optimization studies in IGNITEX","authors":"S. K. Ingram","doi":"10.1109/FUSION.1993.518347","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518347","url":null,"abstract":"In the existing budgetary environment, low-cost, high-payoff fusion experiments are increasingly attractive. IGNITEX, a homopolar generator powered single-turn tokamak, has the potential to achieve ignition at relatively low cost, A 60 MJ homopolar generator facility (with an easy upgrade path to 90 MJ) is in place at CEM-UT. Using this facility, CEM-UT researchers have produced a 20 T on-axis magnetic field in a 1/16 scale IGNITEX prototype. This paper presents the results of optimization studies conducted to determine the lowest-cost homopolar generator facility which can power an IGNITEX prototype of any chosen scale operating at 20 T on-axis.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"41 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":"132093398","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.1993.518416
R. Bulmer
The Tokamak Physics Experiment (TPX) will have a poloidal field system capable of full inductive operation for approximately a 20-s flattop and, with superconducting toroidal and poloidal field coils and non-inductive current drive, it will be capable of true steady-state operation. The poloidal field design is based on the ideal MHD equilibrium model as implemented in the TEQ code developed at LLNL. The PF coils are arranged in an up-down symmetric configuration, external to the TF coils. The TPX diverted plasma will have an aspect ratio of 4.5 and is highly shaped with a nominal elongation of 2 and triangularity of approximately 0.8 as measured at the separatrix. The tokamak design is based on a high-current (q/sub /spl Psi//=3) plasma scenario and a low current scenario. Each scenario has an operational flexibility requirement which is defined as a region of plasma pressure and inductivity (/spl beta//sub N/-l/sub i/) space, where the plasma shape is constrained to keep the divertor configuration operational. Single-null plasma configurations are feasible, even with the same divertor hardware, by operating the PF coils asymmetrically. Recently applied optimization techniques have improved the capability of the PF system without additional cost.
{"title":"Tokamak Physics Experiment poloidal field design","authors":"R. Bulmer","doi":"10.1109/FUSION.1993.518416","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518416","url":null,"abstract":"The Tokamak Physics Experiment (TPX) will have a poloidal field system capable of full inductive operation for approximately a 20-s flattop and, with superconducting toroidal and poloidal field coils and non-inductive current drive, it will be capable of true steady-state operation. The poloidal field design is based on the ideal MHD equilibrium model as implemented in the TEQ code developed at LLNL. The PF coils are arranged in an up-down symmetric configuration, external to the TF coils. The TPX diverted plasma will have an aspect ratio of 4.5 and is highly shaped with a nominal elongation of 2 and triangularity of approximately 0.8 as measured at the separatrix. The tokamak design is based on a high-current (q/sub /spl Psi//=3) plasma scenario and a low current scenario. Each scenario has an operational flexibility requirement which is defined as a region of plasma pressure and inductivity (/spl beta//sub N/-l/sub i/) space, where the plasma shape is constrained to keep the divertor configuration operational. Single-null plasma configurations are feasible, even with the same divertor hardware, by operating the PF coils asymmetrically. Recently applied optimization techniques have improved the capability of the PF system without additional cost.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"123 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":"134265715","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.1993.518414
H. Miura, S. Nishio, T. Suzuki
One of the most critical issues for fusion in-vessel components is to reduce the transient electromagnetic force due to plasma disruption and to establish suitable support structures. It is indicated that the component deflection is damped by the secondary induced eddy current to reduce the primary electromagnetic force (coupling effects of electromagnetic induction and mechanical deflection). In order to perform the structural design of components, we must fully understand the transient electromagnetic phenomena. For the better fusion reactor design, we've developed a computer code for 3-D thin shell structure with non-ferrous and elastic conductors. Here, we employed a finite element numerical model for mechanical deformation and a wire-grid numerical model for eddy currents. In order to verify the computer code and to understand the influence of support condition in consideration of above coupling effects, we performed some experimental and numerical studies. The experimental results agree well with the numerical results, and we could grasp the influence of support condition for the vibration phenomena in the magnetic field.
{"title":"Experimental and analytical study of the electromagnetomechanics in fusion reactors","authors":"H. Miura, S. Nishio, T. Suzuki","doi":"10.1109/FUSION.1993.518414","DOIUrl":"https://doi.org/10.1109/FUSION.1993.518414","url":null,"abstract":"One of the most critical issues for fusion in-vessel components is to reduce the transient electromagnetic force due to plasma disruption and to establish suitable support structures. It is indicated that the component deflection is damped by the secondary induced eddy current to reduce the primary electromagnetic force (coupling effects of electromagnetic induction and mechanical deflection). In order to perform the structural design of components, we must fully understand the transient electromagnetic phenomena. For the better fusion reactor design, we've developed a computer code for 3-D thin shell structure with non-ferrous and elastic conductors. Here, we employed a finite element numerical model for mechanical deformation and a wire-grid numerical model for eddy currents. In order to verify the computer code and to understand the influence of support condition in consideration of above coupling effects, we performed some experimental and numerical studies. The experimental results agree well with the numerical results, and we could grasp the influence of support condition for the vibration phenomena in the magnetic field.","PeriodicalId":365814,"journal":{"name":"15th IEEE/NPSS Symposium. Fusion Engineering","volume":"2 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":"134273848","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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