A Lower Rigid Support Structure for the HT—7U Vacuum Vessel

Vacuum vessel of the HT-7U is a fully welded toroidal structure with a noncircularcross-section nested in the bore of the TF coils. According to the requirement of the physicsdesign, sixteen horizontal ports on outboard mid-plane and thirty-two vertical ports on the topand bottom are designed for diagnostics, plasma heating, current driving, vacuum pumping andgas puffing. Bellows on these port necks are used for flexible components to absorb the relativedisplacement in radial and vertical directions due to external load, thermal expansion or contrac-tion and assembly tolerance, and also used for isolation of mechanical vibration. For the supportsystem of vacuum vessel it should be not only strong enough to withstand forces acting on thevessel interior components and the vessel itself due to the dead weight and electromagnetic inter-actions during plasma disruption, but also sufficiently flexible to be suited to thermal expansionduring baking. In order to solve this contradiction a new kind of low rigid s     

Linear Buckling Analysis of the HT-7U Vacuum Vessel

1. IntroductionHT-7U superconducting Tokamak, as a nationalproject, is an advanced steady-state plasma exper-imental device to be built in 2Oo3. Now all kindsof engineering design have been begun. The vac-uum vessel is one of the key parts for HT-7U de- vice, which can provide not only a clean and ultra-high vacuum env1ronment in the operatlon of plasmafor the production quilibrium and heating of thelasma, but also reliable supporting structure andvarious channels for a series of equipme…     

Electrical Parameters of the Vacuum Vessel in HT-7U Tokamak

1. IntroductionThe time-t'arying currents in plasma and poloidalfield coils induce the eddy current on the vacuumvessel of tokamak- The magnetic field produced bythe eddy currents has an effect on the penetrationof poloidal field, the position and the shape of theplasma: t'.hich is related to the structure of the yao-uum vessel and the distribution of the eddy currentsThe level of the effect can be described by the electrical parameters of the vessel, being of very importance in the study Of t…     

Temperature Field and Thermal Stress Analysis of the HT-7U Vacuum Vessel

1.IntroductionHT-7UsuperconductingTokamakisanadvancedsteady-stateplasmaphysicsexperimentaldevice.ThedevicewillbebuiltattheinstituteofPlasmaPhysics,theChineseAcademyofSciencesintheyear2003.Itscessionistodevelopthescientific.basisandtechnologicbajsisforthefutureTokamakfusionreactorsandtostudyphysicalissuesonthesustenanceOfanon-burningplasmascenarioforthesteady-stateoperation.TheHT--7Uhasacapabilityforlongpulse(60-1000s)operation.Theprojecthasbeenapprovedandfundedasanationalmegscience-engin…     

Superconducting Busline Design for HT—7U Tokamak

Superconducting (SC) tokamak HT-7U has seven pairs of buslines connecting toroidal/Poloidal coils and the current leads,These SC buslines(SCBLs) share a common cryostat and are made of the calbe in conduit conductors(CICCs) arranged as a decoupling configuration.In order to reduce the heat loads conducted from the seven current leads with a capacity of 15kA during the magnets cooldown.the buslines with a much lower thermal conduction were employed in comparison with the current leads,and a special cooling loop was designed.     

Design of the Magnetic Field for HT-7U Tokamak

1. Main parameters ofHT-7U deviceThe scientific cession of the HT－7U(HefeiTOkamak-7 Upgrade ), being a next generation advanced tokamak, is to study physical issues on thesteady-state operation of a non-burning plasma. The'. -.engineering mission of the HT-7U tokamak is to establish the technology basis of full superconductingtokamaks for the future reactor. The HT-7U toka-mak will have a long pulse (60 ~ 1000 s ) capability,a flexible PF system, an auxiliary heating and current drive …     

Design of the Cryostat for HT—7U superconducting Tokamak

The cryostat of HT-7U tokamak is a large vacuum vessel surrounding the entire basic machine with a cylindrical shell,a dished top and a flat bottom.The main function of HT-7U cryostat is to provide a thermal barrier between an ambient temperature test hall and a liquid helium-cooled superconducting magnet.The loads applied to the cryostat are from sources of vacuum pressure,dead weight,seismic events and electromagnetic forces originated by eddy currents.It also provides feed-through penetrations for all the conecting 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 by using a finite element code.Stress analysis results show that the maximum stress intensity was below the allowable value.In this paper,the structural analyses and design of HT-7U cryostat are emphasized.     

核聚变实验装置HT-7U一维及二维辐射防护设计研究

主要介绍一维、二维中子输运程序ANISN,DOT3.5在核聚变实验装置HT-7U辐射屏蔽物理设计中的应用。计算和分析了该装置实验大厅内外中子注量/能谱、γ注量/能谱、中子剂量率、γ剂量率的空间分布,对屏蔽材料的选取及屏蔽层厚度进行了优化设计,为HT-7U装置的辐射屏蔽物理设计提供了建议性意见及理论依据。     

Feasibility Study on Welding Structure of the HT-7U Toroidal Field Coil Case

1. IntroductionThe TF magnet system of HT-7U consists of 16coils arrayed toroidally and spaced 22.5" apart oneanother. The designed TF coils provide a magneticfield, which is necessary to maintain plasma in atokamak configuration with a current up to I MA.The TF coils have a D--shaped geometry of 3.5 meters high and 2.5 meters wide, producing a magneticfield of 3.5 T on 1.7 m of main radius. The TF coilcases, enclosing the TF winding packs, are the mainstructural components of the magn…     

Design and Structural Analysis of Support Structure for ITER Vacuum Vessel

The International Thermonuclear Experimental Reactor (ITER) vacuum vessel (VV) is a safety component confining radioactive materials such as tritium and activated dust. An independent VV support structure with multiple flexible plates located at the bottom of VV lower port is proposed as a new concept, which is deferent from the current design, i.e., the VV support is directly connected to the toroidal coils (TF coils). This independent concept has two advantages comparing to the current one: (1) thermal load due to the temperature deference between VV and TF coils becomes lower and (2) the TF coils are categorized as non-safety components because of its independence from VV. Stress Analyses have been performed to assess the integrity of the VV support structure using a precisely modeled VV structure. As a result, (1) the maximum displacement of the VV corresponding to the relative displacement between VV and TF coils is found to be 15 mm, much less than the current design clearance of 100 mm, and (2) the stresses of the whole VV system including VV support are estimated to be less than the allowable ones defined by ASME Section III Subsection NF, respectively. Based on these assessments, the feasibility of the proposed independent VV support has been verified as a VV support.     

Preliminary Design of Control Network for HT-7U Tokamak Cryogenic System

In the course of the cryoplant modernization,a control network will be set up inorder to facilitate the control,the supervision,the centralized data acquisition and the alarmhandling of the cryogenic system for HT-7U tokamak.The paper introduces the preliminarydesign of control network based on the Controller Link Network for HT-7U tokamak cryogenic     

Recent Progress of HT-7U Superconducting Tokamak

HT-7U is a superconducting tokamak. which is being constructed in Institute of Plasma Physics, Chinese Academy of Sciences. The mission of the HT-7U project is to develop a scientific and engineering basis of the steady state operation of advanced tokamak.The engineering design of the device has been optimized. The R&D program is going on. Short samples of the conductor and a CS model coil were tested. All the TF and PF coils will be manufactured and tested in Institute of Plasma Physics. Therefore, a 600-meter long jacketing line for cable-in-conduit conductors along with two winding machines, a set of VPI equipment and a test facility for the TF and PF coils are ready in ASIPP now. In this paper, the recent progress of the HT-7U is described.     

Preliminary Design of Poloildal Field Power Supply for HT-7U Super-conducting Tokamak

1. IntroductionBeing distinct from other power supplies, poloidalfield power supply (PFPS) is the major part in powersupply system of HT--7U superconducting tokamak,with characteristics of hi$h current derivative, highpeak power, high power variation, and sophisticatedcontrol. The main function and task of PFPS are torealize the energy transfer and power conversion ofplasma initiation, maintenance, heating: and control(shape, position and disruption control) in experiment.Based on the main…     

Fast Thermography for HT-7U Superconducting Tokamak

1. IntroductionThe tokamak configuration is the most favorable approach for the confinement of nuclear fusionplasma with magnetic field. But the tokamak suffers from three main limitations at present f disruption, pulsed operation and low fi values. HT--7U isa full superconducting tokamak device designed toaddress these three problems, especially for an advanced steady--state operation. FOr disruption study,It is very important to know the temperature profilesacross the first wall during dis…     

New CICC Design Versions with High Copper Fraction for HT-7U Tokamak

1. IntroductionHT-7U TOkamak is a fusion experiment devicewith a major radius of 1.7 m and a minor radiusof 0.4 x 0.72 m (D--shaped cross--section of plasma).The magnets of the tokamak consist of 16 D--shapedcoils for the toroidal field (TF), 6 coils for the central solenoid, and another 6 coils for divertor, shaping and equilibrium fields. Fig. 1 shows the TF andpololdal field (PF) coil arrangement. As CICC hasmany advantages of excellent winding insulation andintegrity, lower ac losse…     

Structural Analysis of Central Solenoid for HT—7U

The central solenoid(CS) composed of cable-in-conduit(CIC) type conductor is an important part of the HT-7U device,The CS coils can be considered from the viewpoint of micromechanics as a compostie material.And then,the residual stiffness is computed according to the micro-damage modeling of continuum damage mechanics,These material properties have been used as input data for finite element method(FEM) analysis.In this paper the computational analysis of the stress and the displacement on the central solenoid are made by the finite element analysis system COSMOS/M2.0 under operating temperature.According to the analytical results.the CS coils are all satisfied with designed safety criteria.     

The Design of Circuit for Checking Short in HT-7U Superconducting Tokamak Device

1.IntroductionHT-7Usuperconductingtokamakisanationalsci-entificresearchprojectofChina,inwhichthislargefusionexperimentdevicewillbebuiltattheInstituteofPlasmaPhysicsofChineseAcademyofSciences.ThemissionofHT-7Uistodevelopascientificbasisforacontinuouslyoperatingtokamakfusionre-actor,withasteady-stateandextremelylong-pulseoperation.AsHT-7Uisafullsuperconductingtokamakdevice,itisimportanttokeeptheHT-7Ude-viceasafeoperation.BeforetheoperationofHT-7Udevice,theoperatorhastogettheoperationalp…     

Preliminary Engineering Design of Toroidal Field Magnet System for Superconducting Tokamak HT-7U

1.IntroductionHT-7Usuperconductingtokamakisanationalsci-entificresearchprojectofChina.ThislargefusionexperimentaldevicewillbebuiltinthelnstituteofPlasmaPhyslcsofChineseAcademyofSciences.ThemissionofHT-7Uistodevelopthescientificba-sisforacontinuouslyoperatingtokamakfusionreac-tor,achievingasteadystateandanextremelylongpuIseoperatlonwlthhigh-betadouble-null,aswellassingle-nullplasmasatfllllcllrrent.HT-7Uisafullsuperconductingtokamakdevlce.ltstoroidalfieldsystemisoneoftheimportantpartsofthe…     

Design of the Pumping Stations and the Gas Puffing Systems for the Superconducting Tokamak HT-7U

1. IntroductionHT-7U is an upgrade device for the superconducting' tokamak HT--7 in Hefei, C,hina." It will be t.helargest TOkamak in China, with a major radius of l.7m and a minor cross section of 0.6 m x 0.4 m. Thetypical plasma pulse length will be 1000 s, and thetoroidal superconducting magnetic field on plasmaaxis will be 3.5 T [1].The parameters of the vacuum system of TokamakTore Supra in France and Tokamak JET in the U.Kare taken as our design references. The volume ofJET tor…