Pub Date : 2002-11-07DOI: 10.1109/FUSION.2002.1027704
M. Lennholm, G. Agarici, A. Barbuti, G. Berger-by, F. Bouquey, J. Clary, C. Darbos, R. Dumont, G. Giruzzi, M. Jung, R. Magne, D. Roux, J. Ségui, X. Zou
The ECRH experiment on TORE SUPRA is designed to inject up to 3 MW of power at 118 GHz, using an antenna consisting of six fixed spherical mirrors and three mobile steering minors. The position of the mobile mirrors can be varied in real time using two stepper motors for each mobile mirror. In addition to controlling the injection angle, the position of the mobile mirrors also affects the polarisation of the injected wave. Accurate formulae to computer in real time, the stepper motor positions required to obtain the desired beam injection angles have been derived. Formulae to determine the effect on the wave polarisation, of the actual mobile mirror positions have also been determined. These formulae have been verified by precise laser measurements and by comparison of power deposition calculations and experimental results. Experiments with different values of polarisation and injection angles have been carried out in the 2001 campaign.
{"title":"ECRH power injection in TORE SUPRA","authors":"M. Lennholm, G. Agarici, A. Barbuti, G. Berger-by, F. Bouquey, J. Clary, C. Darbos, R. Dumont, G. Giruzzi, M. Jung, R. Magne, D. Roux, J. Ségui, X. Zou","doi":"10.1109/FUSION.2002.1027704","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027704","url":null,"abstract":"The ECRH experiment on TORE SUPRA is designed to inject up to 3 MW of power at 118 GHz, using an antenna consisting of six fixed spherical mirrors and three mobile steering minors. The position of the mobile mirrors can be varied in real time using two stepper motors for each mobile mirror. In addition to controlling the injection angle, the position of the mobile mirrors also affects the polarisation of the injected wave. Accurate formulae to computer in real time, the stepper motor positions required to obtain the desired beam injection angles have been derived. Formulae to determine the effect on the wave polarisation, of the actual mobile mirror positions have also been determined. These formulae have been verified by precise laser measurements and by comparison of power deposition calculations and experimental results. Experiments with different values of polarisation and injection angles have been carried out in the 2001 campaign.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":"12 1","pages":"329-332"},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78725407","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"艺术学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2002-11-07DOI: 10.1109/FUSION.2002.1027722
Y. Oh, C.H. Choi, J. Sa, H. Ahn, K. J. Cho, Y.M. Park, Y.S. Kim, K. Kim, D.K. Lee, S. Cho, N. Her, M. Kwon, J. Bak, G.S. Lee
The magnet structure system of the Korea Superconducting Tokamak Advanced Research (KSTAR) device consists of 16-segmented toroidal field (TF) coil structures encasing each D-shaped TF coil, a central solenoid (CS) structure surrounding 4 pairs of CS coils, modular poloidal field (PF) coil structures supporting each PF coil in 8 or 16 places, and a gravity support. The engineering design of the magnet structures has been conducted with related electromagnetic load calculations and structural analyses for various operation scenarios. A prototype TF coil structure will be fabricated to check the manufacturing feasibility. A prototype magnet supporting post has been fabricated and tested at 80 K up to 15,000 cycles of vertical load under 80 tons. In addition to the magnet structure development, winding and heat treatment of a real-sized prototype TF coil have been finished without any defect such as SAGBO. The fabrication of the coil will be completed by the middle of 2002. As an interface of the magnet system, a cryogenic facility and a current feeder system have been designed.
{"title":"Design overview of the KSTAR magnet structures","authors":"Y. Oh, C.H. Choi, J. Sa, H. Ahn, K. J. Cho, Y.M. Park, Y.S. Kim, K. Kim, D.K. Lee, S. Cho, N. Her, M. Kwon, J. Bak, G.S. Lee","doi":"10.1109/FUSION.2002.1027722","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027722","url":null,"abstract":"The magnet structure system of the Korea Superconducting Tokamak Advanced Research (KSTAR) device consists of 16-segmented toroidal field (TF) coil structures encasing each D-shaped TF coil, a central solenoid (CS) structure surrounding 4 pairs of CS coils, modular poloidal field (PF) coil structures supporting each PF coil in 8 or 16 places, and a gravity support. The engineering design of the magnet structures has been conducted with related electromagnetic load calculations and structural analyses for various operation scenarios. A prototype TF coil structure will be fabricated to check the manufacturing feasibility. A prototype magnet supporting post has been fabricated and tested at 80 K up to 15,000 cycles of vertical load under 80 tons. In addition to the magnet structure development, winding and heat treatment of a real-sized prototype TF coil have been finished without any defect such as SAGBO. The fabrication of the coil will be completed by the middle of 2002. As an interface of the magnet system, a cryogenic facility and a current feeder system have been designed.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":"209 1","pages":"400-403"},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"76070815","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"艺术学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2002-11-07DOI: 10.1109/FUSION.2002.1027723
L. Bromberg, T. Brown, F. Dahlgren, P. Heitzenroeder
This paper presents a conceptual design of the magnet systems for an advanced tokamak fusion reactor (ARIES-AT). The main emphasis of the paper is the extrapolation of the current state-of-the-art in high temperature superconductors and coil design, and their implementation in an advanced commercial fusion reactor concept. A conceptual design of both the poloidal field coils and toroidal field coils is presented. The current design point is described and supported with a preliminary structural analysis and a discussion of the merits, performance, and economics of high temperature vs. low temperature superconductors in an advanced fusion reactor design.
{"title":"ARIES-AT magnet systems","authors":"L. Bromberg, T. Brown, F. Dahlgren, P. Heitzenroeder","doi":"10.1109/FUSION.2002.1027723","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027723","url":null,"abstract":"This paper presents a conceptual design of the magnet systems for an advanced tokamak fusion reactor (ARIES-AT). The main emphasis of the paper is the extrapolation of the current state-of-the-art in high temperature superconductors and coil design, and their implementation in an advanced commercial fusion reactor concept. A conceptual design of both the poloidal field coils and toroidal field coils is presented. The current design point is described and supported with a preliminary structural analysis and a discussion of the merits, performance, and economics of high temperature vs. low temperature superconductors in an advanced fusion reactor design.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":"41 1","pages":"404-408"},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83658151","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"艺术学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2002-11-07DOI: 10.1109/FUSION.2002.1027736
S. Raftopoulos, G. Barnes, J. Chrzanowski, J. Dimattia, R. Fedder, D. Gallagher, E. Gilsenan, G. Gettelfinger, R. Jakober, J. Laneski, P. Lamarche, R. Lindenberg, R. Parsells, E. Perry, K. Rule, J. Semler, J. Stacy, M. Viola, I. Zatz, W. Zimmer
The Tokamak Fusion Test Reactor (TFTR) operated for 15 years, from 1982 to 1997. From November 1993 to April 1997 a mixture of deuterium-tritium (D-T) was used to fuel experiments, leaving the reactor system in an activated and contaminated state. In this condition, the removal of the TFTR and its subsystems is greatly complicated. In 1999, PPPL commenced the TFTR Decontamination and Decommissioning (D&D) Project. The objectives of the D&D projects are to completely remove the TFTR in a safe, efficient and cost effective manner. To date all diagnostics, structural components and ancillary systems have been removed, and the tokamak is in the process of being sectioned into smaller pieces. This paper will describe the criteria, the methods employed and the successes of the D&D Project.
{"title":"Overview of the TFTR D&D program","authors":"S. Raftopoulos, G. Barnes, J. Chrzanowski, J. Dimattia, R. Fedder, D. Gallagher, E. Gilsenan, G. Gettelfinger, R. Jakober, J. Laneski, P. Lamarche, R. Lindenberg, R. Parsells, E. Perry, K. Rule, J. Semler, J. Stacy, M. Viola, I. Zatz, W. Zimmer","doi":"10.1109/FUSION.2002.1027736","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027736","url":null,"abstract":"The Tokamak Fusion Test Reactor (TFTR) operated for 15 years, from 1982 to 1997. From November 1993 to April 1997 a mixture of deuterium-tritium (D-T) was used to fuel experiments, leaving the reactor system in an activated and contaminated state. In this condition, the removal of the TFTR and its subsystems is greatly complicated. In 1999, PPPL commenced the TFTR Decontamination and Decommissioning (D&D) Project. The objectives of the D&D projects are to completely remove the TFTR in a safe, efficient and cost effective manner. To date all diagnostics, structural components and ancillary systems have been removed, and the tokamak is in the process of being sectioned into smaller pieces. This paper will describe the criteria, the methods employed and the successes of the D&D Project.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":"8 1","pages":"465-468"},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89650821","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"艺术学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2002-11-07DOI: 10.1109/FUSION.2002.1027671
C. Makariou, B. Bray, C. Hsieh
The Thomson scattering system measures electron density and temperature with the aid of eight pulsed Nd:YAG lasers on the DIII-D tokamak. DIII-D is the United States National Fusion Facility for magnetic confinement fusion experiments. The Thomson system probes the tokamak in three separate regions. The measurements from the different regions produce density and temperature profiles during plasma discharges. The laser light must be aimed through the DIII-D vessel with 1 mm of spatial accuracy to produce correct density profiles. Recent upgrades to the alignment system reduce the effort required to perform the initial alignment before plasma operations, and improve the monitoring and control of the alignment. The upgraded hardware allows for monitoring and control of the alignment from the Thomson control room during operations and from multiple key locations along the 35 m long optical path. In future upgrades the adjustment of the mirrors will be automated by utilizing feedback from a computerized beam analysis system that relays beam position information to the control computer. The newly established methods for setting up the YAG beams using the CCD cameras instead of burn paper offer a faster and more reliable method to prepare the system for plasma-operations. The YAG lasers are aimed with repeatable accuracy and any drifts can be detected and corrected during the small warm up period before each shot.
{"title":"Upgraded alignment control for the DIII-D Thomson scattering laser system","authors":"C. Makariou, B. Bray, C. Hsieh","doi":"10.1109/FUSION.2002.1027671","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027671","url":null,"abstract":"The Thomson scattering system measures electron density and temperature with the aid of eight pulsed Nd:YAG lasers on the DIII-D tokamak. DIII-D is the United States National Fusion Facility for magnetic confinement fusion experiments. The Thomson system probes the tokamak in three separate regions. The measurements from the different regions produce density and temperature profiles during plasma discharges. The laser light must be aimed through the DIII-D vessel with 1 mm of spatial accuracy to produce correct density profiles. Recent upgrades to the alignment system reduce the effort required to perform the initial alignment before plasma operations, and improve the monitoring and control of the alignment. The upgraded hardware allows for monitoring and control of the alignment from the Thomson control room during operations and from multiple key locations along the 35 m long optical path. In future upgrades the adjustment of the mirrors will be automated by utilizing feedback from a computerized beam analysis system that relays beam position information to the control computer. The newly established methods for setting up the YAG beams using the CCD cameras instead of burn paper offer a faster and more reliable method to prepare the system for plasma-operations. The YAG lasers are aimed with repeatable accuracy and any drifts can be detected and corrected during the small warm up period before each shot.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":"120 1","pages":"180-183"},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88988578","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"艺术学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2002-11-07DOI: 10.1109/FUSION.2002.1027711
I. Ricapito, G. Cambi, A. Aiello, G. Benamati, O. Kveton, M. Futterer
A summary is here presented concerning the first results in evaluating the safety features of the European water-cooled lead lithium blanket in which tritium permeation barriers on the external surface of the breeder cooling pipes are not used or, if used, a lower efficiency is required for them. As a consequence, in this design concept, a significant amount of tritium must be recovered through the water detritiation system. The technical and economical feasibility of such a tritium management strategy was demonstrated up to a tritium permeation rate of 10+15 g/day, accepting, at the same time, a higher tritium activity in the primary coolant in order not to overload the water detritiation system. A safety analysis has then been performed to evaluate the impact of a higher tritium specific activity in the primary coolant, both in normal operation and in accidental condition. Tritium release to the environment in case of a LOCA was found to satisfy the limits recommended by the ITER Project Design Guidelines for accidental tritium release, while the respect of the limits for chronic release strongly depends on the water leak rate from the primary to the secondary circuit.
{"title":"Impact of low PRF tritium permeation barriers-on the water cooled lead lithium blanket: A safety analysis","authors":"I. Ricapito, G. Cambi, A. Aiello, G. Benamati, O. Kveton, M. Futterer","doi":"10.1109/FUSION.2002.1027711","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027711","url":null,"abstract":"A summary is here presented concerning the first results in evaluating the safety features of the European water-cooled lead lithium blanket in which tritium permeation barriers on the external surface of the breeder cooling pipes are not used or, if used, a lower efficiency is required for them. As a consequence, in this design concept, a significant amount of tritium must be recovered through the water detritiation system. The technical and economical feasibility of such a tritium management strategy was demonstrated up to a tritium permeation rate of 10+15 g/day, accepting, at the same time, a higher tritium activity in the primary coolant in order not to overload the water detritiation system. A safety analysis has then been performed to evaluate the impact of a higher tritium specific activity in the primary coolant, both in normal operation and in accidental condition. Tritium release to the environment in case of a LOCA was found to satisfy the limits recommended by the ITER Project Design Guidelines for accidental tritium release, while the respect of the limits for chronic release strongly depends on the water leak rate from the primary to the secondary circuit.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":"162 1","pages":"356-359"},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"76936353","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"艺术学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2002-11-07DOI: 10.1109/FUSION.2002.1027677
T. Sakai, K. Noborio, Y. Yamamoto
In this paper, we study the mechanism of a self-maintaining discharge in the Inertial Electrostatic Confinement Fusion (IECF) with D/sub 2/ gas. We developed a 1-D particle code with Monte Carlo collision scheme including atomic and molecular processes of ion, energetic neutral, and electron impact based on the PDS-1 code. Also we developed the energy dependent transparency model of the cathode, and applied constant discharge current control to simulate a steady discharge. As electrons are accelerated quickly just outside of the cathode, electron impact ionization hardly occurred outside the cathode. Therefore, the quantity of D/sub 2//sup +/ supplied by electron impact ionization is not enough to maintain a discharge. From the simulation, we found that D/sub 2//sup +/ impact charge-exchange and D/sub 2//sup 0/ impact re-ionization contribute greatly to maintain a discharge.
{"title":"Analysis of discharge characteristics of the inertial electrostatic confinement fusion using a particle code with Monte Carlo collision scheme","authors":"T. Sakai, K. Noborio, Y. Yamamoto","doi":"10.1109/FUSION.2002.1027677","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027677","url":null,"abstract":"In this paper, we study the mechanism of a self-maintaining discharge in the Inertial Electrostatic Confinement Fusion (IECF) with D/sub 2/ gas. We developed a 1-D particle code with Monte Carlo collision scheme including atomic and molecular processes of ion, energetic neutral, and electron impact based on the PDS-1 code. Also we developed the energy dependent transparency model of the cathode, and applied constant discharge current control to simulate a steady discharge. As electrons are accelerated quickly just outside of the cathode, electron impact ionization hardly occurred outside the cathode. Therefore, the quantity of D/sub 2//sup +/ supplied by electron impact ionization is not enough to maintain a discharge. From the simulation, we found that D/sub 2//sup +/ impact charge-exchange and D/sub 2//sup 0/ impact re-ionization contribute greatly to maintain a discharge.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":"6 1","pages":"209-212"},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81155224","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"艺术学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2002-11-07DOI: 10.1109/FUSION.2002.1027629
R. Aymar
Following the end of the Engineering Design Activities (EDA) in July 2001, all the essential elements are now available to make a decision to construct ITER. The design has been elaborated to the extent necessary to allow a realistic assessment of its feasibility, performance and cost at a generic site, and R&D has been carried out to underpin the design choices. The project is now entering a phase of preparation for construction which addresses the issues arising from the creation of an ITER Legal Entity (ILE) at a particular site, and prepares for its establishment and formal acceptance of all commitments by host and participating countries during construction, exploitation and decommissioning. In parallel to the development of legal and contractual agreements, this phase will move the project technically from the drawing boards of the EDA Joint Central Team and ITER Parties' Home Teams to a state of readiness for the procurement by industry of the longest lead items, and for formal application for a construction license with the host country.
{"title":"Towards ITER construction","authors":"R. Aymar","doi":"10.1109/FUSION.2002.1027629","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027629","url":null,"abstract":"Following the end of the Engineering Design Activities (EDA) in July 2001, all the essential elements are now available to make a decision to construct ITER. The design has been elaborated to the extent necessary to allow a realistic assessment of its feasibility, performance and cost at a generic site, and R&D has been carried out to underpin the design choices. The project is now entering a phase of preparation for construction which addresses the issues arising from the creation of an ITER Legal Entity (ILE) at a particular site, and prepares for its establishment and formal acceptance of all commitments by host and participating countries during construction, exploitation and decommissioning. In parallel to the development of legal and contractual agreements, this phase will move the project technically from the drawing boards of the EDA Joint Central Team and ITER Parties' Home Teams to a state of readiness for the procurement by industry of the longest lead items, and for formal application for a construction license with the host country.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":"67 1","pages":"1-9"},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74954504","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"艺术学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2002-11-07DOI: 10.1109/FUSION.2002.1027715
Junhui Yu, S. Wu, Y. Song, P. Weng
The cryostat of HT-7U Tokamak is a large vacuum vessel surrounding the entire Basic Machine with cylindrical shell, dished top and flat bottom. The main function of HT-7U cryostat provides the thermal barrier between the ambient temperature testing hall and the liquid helium cooled superconducting magnet. The loads applied to the cryostat are vacuum pressure, dead weight, seismic events and electromagnetic forces originated by eddy currents. It also provides feed through penetrations for all the connecting elements inside and outside the cryostat. The main material selected for the cryostat is stainless steel 304L. The structural analyses including buckling for the cryostat vessel under the plasma operation condition have been carried out using a finite element code. Stress analysis results show that the maximum stress intensity was below the allowable value. In this paper, we emphasize on the structural analyses of HT-7U cryostat. The preliminary design of the cryostat is also described.
{"title":"Structural analyses and preliminary design of HT-7U cryostat","authors":"Junhui Yu, S. Wu, Y. Song, P. Weng","doi":"10.1109/FUSION.2002.1027715","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027715","url":null,"abstract":"The cryostat of HT-7U Tokamak is a large vacuum vessel surrounding the entire Basic Machine with cylindrical shell, dished top and flat bottom. The main function of HT-7U cryostat provides the thermal barrier between the ambient temperature testing hall and the liquid helium cooled superconducting magnet. The loads applied to the cryostat are vacuum pressure, dead weight, seismic events and electromagnetic forces originated by eddy currents. It also provides feed through penetrations for all the connecting elements inside and outside the cryostat. The main material selected for the cryostat is stainless steel 304L. The structural analyses including buckling for the cryostat vessel under the plasma operation condition have been carried out using a finite element code. Stress analysis results show that the maximum stress intensity was below the allowable value. In this paper, we emphasize on the structural analyses of HT-7U cryostat. The preliminary design of the cryostat is also described.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":"11 1","pages":"372-375"},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80741801","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"艺术学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2002-11-07DOI: 10.1109/FUSION.2002.1027714
R. Kaita, S. Jardin, B. Jones, C. Kessel, R. Majeski, J. Spaleta, R. Woolley, L. Zakharov, B. Nelson, M. Ulrickson
Liquid metal walls have the potential solve to first-wall problems for fusion reactors, such as heat load and erosion of dry walls, neutron damage and activation, and tritium inventory and breeding. In the near term, such walls can serve as the basis for schemes to stabilize magnetohydrodynamic (MHD) modes. Furthermore, the low recycling characteristics of lithium walls can be used for particle control. Liquid lithium experiments have already begun in the Current Drive eXperiment-Upgrade (CDX-U). Plasmas limited with a toroidally localized limiter have been investigated, and experiments with a fully toroidal lithium limiter are in progress. A liquid surface module (LSM) has been proposed for the National Spherical Torus Experiment (NSTX). In this larger ST, plasma currents are in excess of 1 MA and a typical discharge radius is about 68 cm. The primary motivation for the LSM is particle control, and options for mounting it on the horizontal midplane or in the divertor region are under consideration. A key consideration is the magnitude of the eddy currents at the location of a liquid lithium surface. During plasma start up and disruptions, the force due to such currents and the magnetic field can force a conducting liquid off of the surface behind it. The Tokamak Simulation Code (TSC) has been used to estimate the magnitude of this effect. This program is a two dimensional, time dependent, free boundary simulation code that solves the MHD equations for an axisymmetric toroidal plasma. From calculations that match actual ST equilibria, the eddy current densities can be determined at the locations of the liquid lithium. Initial results have shown that the effects could be significant, and ways of explicitly treating toroidally local structures are under investigation.
{"title":"Modeling of spherical torus plasmas for liquid lithium wall experiments","authors":"R. Kaita, S. Jardin, B. Jones, C. Kessel, R. Majeski, J. Spaleta, R. Woolley, L. Zakharov, B. Nelson, M. Ulrickson","doi":"10.1109/FUSION.2002.1027714","DOIUrl":"https://doi.org/10.1109/FUSION.2002.1027714","url":null,"abstract":"Liquid metal walls have the potential solve to first-wall problems for fusion reactors, such as heat load and erosion of dry walls, neutron damage and activation, and tritium inventory and breeding. In the near term, such walls can serve as the basis for schemes to stabilize magnetohydrodynamic (MHD) modes. Furthermore, the low recycling characteristics of lithium walls can be used for particle control. Liquid lithium experiments have already begun in the Current Drive eXperiment-Upgrade (CDX-U). Plasmas limited with a toroidally localized limiter have been investigated, and experiments with a fully toroidal lithium limiter are in progress. A liquid surface module (LSM) has been proposed for the National Spherical Torus Experiment (NSTX). In this larger ST, plasma currents are in excess of 1 MA and a typical discharge radius is about 68 cm. The primary motivation for the LSM is particle control, and options for mounting it on the horizontal midplane or in the divertor region are under consideration. A key consideration is the magnitude of the eddy currents at the location of a liquid lithium surface. During plasma start up and disruptions, the force due to such currents and the magnetic field can force a conducting liquid off of the surface behind it. The Tokamak Simulation Code (TSC) has been used to estimate the magnitude of this effect. This program is a two dimensional, time dependent, free boundary simulation code that solves the MHD equations for an axisymmetric toroidal plasma. From calculations that match actual ST equilibria, the eddy current densities can be determined at the locations of the liquid lithium. Initial results have shown that the effects could be significant, and ways of explicitly treating toroidally local structures are under investigation.","PeriodicalId":44192,"journal":{"name":"NINETEENTH CENTURY MUSIC","volume":"161 1","pages":"368-371"},"PeriodicalIF":0.4,"publicationDate":"2002-11-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86395765","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"艺术学","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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