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Joint TEXT (J-TEXT) is an iron-core tokamak. The iron core can reduce the requirement of heating power supplies, and it also has a significant effect on the poloidal magnetic field during J-TEXT discharge. However, it is difficult to simulate an iron-core tokamak discharge process for many tokamak simulation codes (TSCs), such as TSC and DINA. Thus, the influence of the iron core on the J-TEXT magnetic field is analyzed by the theoretical model and the ANSYS model and compared with experimental results. ANSYS simulation shows that the influence of the iron core on the toroidal magnetic field can be ignored, but it has a significant impact on the poloidal magnetic field. By the comparison of ANSYS simulation and experimental results, the correction coefficient of the poloidal field with and without iron core is obtained, which can be used in some non iron-core TSC.

The plasma confinement of the International Tokamak Experimental Rector (ITER) is provided by the magnetic field generated by 18 toroidal field (TF) coils while 6 poloidal field (PF) and 6 central solenoid coils have the function to drive, shape and pre-heat the plasma. Fusion for Energy (F4E), the European Domestic Agency for ITER, is responsible for the supply of 10 TF coils and 5 PF coils to the ITER project. The ITER Organization (IO) team is instead responsible for the design of such coils as well for the coordination of the activities of the different Domestic Agencies (DAs) producing the different components, and their assembly into the Tokamak. The PF coils utilize NbTi Cable-in-Conduit-Conductor and have different diameters between 8 and 24 meters and weights of up to 400 tons. Regarding the PF coils produced by F4E, so far one has been completed by the Institute of Plasma Physics Chinese Academy of Sciences (ASIPP) under a collaboration agreement with F4E. The other 4 PF coils are being produced at the ITER site in Saint Paul lez Durance, France, under F4E supervision. The first of these (PF5) will be completed by July 2020 while the last coil (PF3) will be ready be the end of 2023. The TF coils utilize Nb3Sn conductor and are manufactured with the “Wind, React & Transfer” method. The first TF coil is close to completion and will be delivered to the ITER site in early 2020. Subsequent TF coils will follow at a rate of about one every 3-4 months. In this article we will report on the production status of both PF and TF coils and, in particular, the different manufacturing strategies employed. The main challenges faced so far and the results obtained are also described.

Disruption simulations with DINA code are performed for JT-60SA design. The simulation results have been applied for the design of many components, not only for the vacuum vessel and in-vessel components, but also for peripheral components. For instance, for the design of in-vessel coils, stabilizing plate and magnetic sensors, EM force induced by halo current and eddy current at disruption were calculated. For design of poloidal field (PF) coils, the power supply of PF coils and refrigerator system for super conducting coils, eddy current of PF coils and AC loss of superconducting coils were evaluated.

In order to simulate a full tokamak discharge, a fast integrated tokamak modeling tool for scenario design, METIS, is coupled with a quasi-static free-boundary magnetic equilibrium code, FEEQS. The first code is able to compute at each time the plasma equilibrium, the current density and plasma pressure profiles. The generated outputs are iteratively used by the second code that computes the poloidal field coils currents needed in order to obtain a given plasma shape. This is used to assess the feasibility of scenarios designed by integrated modeling simulations. This approach is much faster with respect to full simulators, equipped with specific feedback controllers, and can be regarded as complementary or preparatory to much more time-consuming scenario control optimisation. The results obtained with the equilibrium code are benchmarked with data already available and computed using the TOSCA code. To illustrate how the coupled codes work, the optimization of one of the reference scenarios of the JT-60SA tokamak is carried out. The use of this new simulation tool for tokamak scenario design is discussed.

Tokamaks are fusion machines that are under development to produce baseload power. Baseload power is power that is produced 24/7 and provides the base for powering the electric grid. The International Tokamak Experimental Reactor (ITER) is an international project that will produce net power from a Tokamak. Net power means the Tokamak produces more energy than it consumes. Consumption includes heating the plasma, controlling it, and powering all the auxiliary systems needed to maintain the plasma. It will allow researchers to study the physics of the Tokamak which will hopefully lead the way toward operational machines. A Tokamak is shown in Figure 6.1. The inner poloidal field coils act like a transformer to initiate a plasma current. The outer poloidal and toroidal coils maintain the plasma. The plasma current itself produces its own magnetic field and induces currents in the other coils.

Merging compression is an efficient and robust method for plasma start-up which can achieve both high plasma current and high temperatures at fusion relevant densities. It involves the formation of plasma around two in-vessel poloidal field coils followed by a magnetic reconnection event after which the plasma can be compressed. In this paper we extrapolate from START/MAST experimental data to predict the performance of ST40 (currently under construction). In our analysis we found that induced eddy currents within passive elements greatly affected the scenario.

An upgrade to the lower divertor is currently being planned for EAST superconducting tokamak, aiming at reaching over 400 s long-pulse H-mode operations with a full metal wall and a divertor heat load of ˜10 MW/m2. A new divertor concept for EAST, “Tightly Baffled Divertor”, suited to water- cooled W/Cu plasma face components (PFC) with minimized divertor volume, has been proposed to achieve Te,target

One of the most crucial issues in the design of the ITER machine is the electromagnetic (EM) loads associated with eddy currents induced in the conducting structures during plasma disruptions. The ITER database contains tens of calculation scenarios for possible plasma disruption events. The duration of these scenarios is usually limited by the end of the plasma current quench when the toroidal plasma current decreases to zero. For further EM analysis, the currents in the poloidal field coils are conventionally assumed constant after the end of the current quench. Such approach is not good for estimation of electromagnetic loads acting on the tokamak components located outside vacuum vessel. A possible way to solve this problem is to employ pre-determined coil current variations over the entire period of observation including time interval after the current quench. This paper describes an approach to such electromagnetic calculations and comparative results for selected cases with/out an extended current-time specification.

The ITER poloidal field (PF) magnet system is a set of six circular coils attached to the periphery of the toroidal field coil (TFC) structure. While the manufacturing of the PF coils is being launched in three different countries, the ITER Organization (IO) is looking ahead at assembly of PF coils in the tokamak. The PF coils can be grouped in three pairs with respect to their similarities in design and attachment to the TFC, which leads to preparation of different strategies and tooling for the transportation and assembly. The procedures shall also consider the integration, assembly sequence, and environment of the whole tokamak. For example, the two lower coils PF6 and PF5 must be brought in storage position in the tokamak pit and stay there until all the TFCs are assembled, before being attached to them. This paper presents the assembly scenario of the PF coils, including logistics, analysis of assembly tolerances, aligning strategy, and detailing the particularities of each PF coil, from the moment they are delivered to IO until they are connected to the feeder, waiting for the commissioning of the machine.

The ITER magnet system is made up of four main subsystems: the 18 toroidal field (TF) coils, the central solenoid, the six poloidal field (PF) coils, and the correction coils (CCs). The feeder system, with its main busbar (MB) and CC busbar (CB), represents one of the main magnet components as well. All coils and busbars with different dimensions used cable-in-conduit conductors. China has signed three conductor packages, which are the so-called procurement arrangements, between ITER and the Chinese Domestic Agency (CN DA): a TF conductor package, a PF conductor package, and a CC and feeder conductor package. They include 7.5% of the TF conductors (11); all the PF2 (12), PF3 (16), PF4 (16), and PF5 (16) conductors; all the CC (18) conductors; and the MB (3) and CB (2) conductors for the feeders. Complex technologies have been developed by ASIPP for the serial production of all ITER conductors, in terms of cabling parameter design, welding, and elaboration of cable insertion, compaction, and winding processes. China has finished all qualification phases and is well into the main series production. All conductor samples required for quality control have successfully passed the SULTAN tests with good performance. The status of the production of ITER conductors in China is described in this paper.

This chapter deals with the design concepts of power supply systems (PSSs) for energising the magnet systems of tokamaks, as well as with the main problems arising in the course of their development and the approaches to solutions of these problems. The peculiarities of the requirements imposed on the PSSs of two main groups of the magnet system’s coils, which have essentially different functions, namely, the toroidal field and poloidal field coils, are considered. Particular attention is given to the description of the means for the protection of the superconducting coils. The last section describes high-current switching devices used for initiation of plasma discharge and for emergency extraction of energy from magnet systems in case of coil superconductor-to-normal transition. The focus is on the switches, superior to commercial products in the key parameters, developed at the Efremov Institute in the framework of the ITER project.

For the advanced tokamak, the particle deposition and thermal load on the divertor is a big challenge. By moving the strike points on divertor target plates, the position of particle deposition and thermal load can be shifted. We could adjust the Poloidal Field (PF) coil current to achieve the strike point position feedback control. Using isoflux control method, the strike point position can be controlled by controlling the X point position. On the basis of experimental data, we establish relational expressions between X point position and strike point position. Benchmark experiments are carried out to validate the correctness and robustness of the control methods. The strike point position is successfully controlled following our command in the EAST operation.

The Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) is experienced in the design and fabrication of high-temperature superconductor (HTS) current leads such as those successfully used in the Experimental Advanced Superconducting (EAST) Tokamak. The aim of the 52-kA trial lead was to develop a prototype for the leads supplying the poloidal field and central solenoid magnets of ITER. This lead differs from former HTS current leads and from another trial (68 kA) lead developed in parallel for the toroidal field magnet, in that it features a foil-type heat exchanger, a helium-cooled transition section, and a binary shunt made of stainless steel and phosphor bronze. The 52-kA trial lead was fabricated and tested in the EAST Tokamak facility. Results of the test are presented. An analysis of these results indicates that the design concepts introduced are validated. The test also served to validate the HTS stacks that equipped the lead. These were supplied by InnoST.

HL-2M (Li, 2013 [1] ) is a tokamak device that is under construction. Based on the magnetic coils design of HL-2M, four kinds of divertor configurations are calculated by CORSICA code (Pearlstein et al., 2001 [2] ) with the same main plasma parameters, which are standard divertor, exact snowflake divertor, snowflake-plus divertor and snowflake-minus divertor configurations. The potential properties of these divertors are analyzed and presented in this paper: low poloidal field area around X-point, connection length from outside mid-plane to the primary X-point, target plate design and magnetic field shear. The results show that the snowflake configurations not only can reduce the heat load at divertor target plates, but also may improve the magneto-hydrodynamic stability by stronger magnetic shear at the edge. A new divertor configuration, named “tripod divertor”, is designed by adjusting the positions of the two X-points according to plasma parameters and magnetic coils current of HL-2M.

The remote maintenance schemes in a DEMO reactor are categorized by insertion direction, blanket segmentation, and divertor maintenance scheme, and are quantitatively evaluated by analysing the plasma equilibrium. The positions of the poloidal field (PF) coil are limited by the size of the toroidal field (TF) coil and the maintenance port layout of each remote maintenance scheme. Because the PF coils are located near the larger TF coil and far from the plasma surface, the horizontal sector transport maintenance scheme requires the largest part of total PF coil current, 25% larger than that required for separated sector transport using vertical maintenance ports with segmented divertor maintenance (SDM). In the unsegmented divertor maintenance (UDM) scheme, the total magnetic stored energy in the PF coils at plasma equilibrium is about 30% larger than that stored in the SDM scheme, but the time required for removal and installation of all the divertor cassettes in the UDM scheme is roughly a third of that required in the SDM scheme because the number of divertor cassettes in the UDM scheme is a third of that in the SDM scheme. From the viewpoint of simple maintenance operations, the merit of the UDM scheme has more merit than the SDM scheme.

This paper presents the eddy current heating effect on the surrounding components of high-current dc reactor, which is applied to ITER poloidal field converter. According to the normal operation analysis of the PF converter system, the ripple current can be up to 3 kA with the frequency of 300 Hz. Based on the finite element method, the eddy current heating effect on the surrounding magnetic components with high permeability, focusing on the floor rebar, steel column and roof beam, are analyzed in this paper, which provides an essential reference for the installation design of the stacked dc reactors.

International Thermonuclear Experimental Reactor (ITER) Poloidal Field (PF) converter fault will lead to the major accident of power supply system or even damage it. In this paper, detailed analysis on typical fault types of ITER PF AC/DC converter system under different operation mode is presented. Three typical fault were listed and analyzed, such as three-phase short circuit in second side of converter transformer, short circuit upstream DC reactor and thyristor failure. Theoretical analysis and simulation for each fault are carried respectively.

The ITER correction coils (CCs) are designed to be arranged toroidally around the machine inside the poloidal field coils and mounted on the toroidal field coils in order to compensate field errors arising from misalignment of the coils and winding deviations from the nominal shape as a result of fabrication tolerances, joints, leads, busbars CCs, and assembly tolerances. China is responsible for manufacturing all the CCs and their conductor. The paper presents the qualification progress and results of ITER CCs and CC conductor, including the test results of CC SULTAN samples, double pancakes winding experiments, CC case welding techniques, and manufacturing experiment of CC helium inlet/outlets.

The design of poloidal field (PF) coils and correction coils (CCs) for the International Thermonuclear Experimental Reactor (ITER) relies on the use of 45-kA NbTi cable-in-conduit conductors. All PF5 and CC conductors are produced in China. Research and development programs are needed to acquire knowledge on the behavior of such conductors. Since the conductors are new, full-size copper dummy conductors have been produced in advance for testing the cabling parameters, the definition of the automatic tungsten-inert-gas welding of a seamless jacket section, the elaboration of cable insertion, compaction, etc. Then, two short qualification conductor samples (the PF5 and the CC) are manufactured at the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP), Hefei, China, with the NbTi advanced strands produced by the Western Superconducting Technologies Company, Ltd. In this paper, the manufacturing procedures for the two PF5 and CC conductor samples are described in detail.

In order to obtain enhanced plasma parameters a complete new tokamak HL-2M is now under construction in Southwestern Institute of Physics. To assure the structural safety of the device for the entire operation cycle, one of the most important issues is the lifetime-limiting effects due to the pulsed operation mode. Fatigue is one of the major failure modes to be considered in mechanical design, and pulsed operation imposes stress with significant alternating components on the support structure (SS). Therefore, the reliability of the whole device is strongly affected by the stress and fatigue characteristic of the SS as the interface structure. This article introduces the SS design and details the fatigue life calculation methods based on the different characteristics of the sub-structures. The fatigue life in hazardous areas of the toroidal field coils anti-torque structure (TFCs-ATs) has been determined by non-linear analysis results. And with the stress-time history data of the vacuum vessel & poloidal field coils support structure (VV&PFCs SS), the fatigue analysis of the hot spots has been completed based on rain-flow counting method and linear cumulative damage method. The calculated minimum fatigue life on TFCs-ATs and VV&PFCs SS is 4.743E+05 and 1.805E+06 cycles, respectively. And the calculated fatigue life on sub-structures can meet the required life for HL-2M tokamak: 1.0E+05 cycles.

A snowflake divertor magnetic configuration (Ryutov in Phys Plasmas 14(6):064502, 2007) with the second-order poloidal field null offers a number of possible advantages for tokamak plasma heat and particle exhaust in comparison with the standard poloidal divertor with the first-order null. Results from snowflake divertor experiments are briefly reviewed and future directions for research in this area are outlined.

This paper presents evaluation of applicability of 2D iron core model for highly non-axisymmetric two limb configuration of GOLEM tokamak (former CASTOR). Presented results explain the long-term discrepancy between measured magnitudes of external poloidal field and those calculated by air-core approach on this tokamak. The model has been applied to two poloidal planes at different toroidal angles in the vacuum vessel region and has shown that close to central column of the transformer, it is possible to correct for 3D effects by variation of chosen dimensions of axisymmetric iron core model. Satisfactory agreement of the 2D model results with the measured distribution of BR field component was achieved.

Large magnetic field generated by reactor applied for the switch network unit of Experimental Advanced Superconducting Tokamak (EAST) poloidal field can seriously disrupt the proper operation of discharge circuit. This paper presents a detailed analysis of the reactor burnout accident during its operation, and to solve the problem, three different types of shielding structure are proposed and shielding performances are compared. And in order to obtain the best shielding performance, an optimal structure is investigated by finite element method (FEM) taking into account eddy current.

The China Fusion Engineering Test Reactor (CFETR) is a new tokamak device, whose magnet system includes the toroidal field, central solenoid (CS), and poloidal field coils. To develop the manufacturing technique for the full-size CS coil, the Central Solenoid Model Coil (CSMC) project was launched first. The cable-in-conduit conductor used for CFETR CSMC refers to the International Thermonuclear Experimental Reactor CS conductor with the same short twist pitch cable pattern. As the short twist pitch and low void fraction, a higher compaction ratio during cabling is required, and the possibility of damages occurring on strands is increased. After numerical analysis and experimental trials on cabling technology, one short cable sample was manufactured, which shows good results compared with former experimental cables.

The ITER Conductor Procurement Arrangement requires that all the conductors have to be qualified before the production phase. The objects of the work here presented are the Poloidal Field Coils conductors, specifically the PF1/6 and the PF2 conductors. The realization of the samples comprised the jacketing, the compaction and the straightening of the CICC, together with the design and manufacturing of the bottom hairpin box and the upper termination. The instrumentation and the related jacket machining completed the preparation of the samples. The samples have been assembled according to the specification defined by ITER and have been shipped to SULTAN facility. This paper describes all the activities performed during the preparation, the features of the samples and all the related issues.

The EAST is an Experimental Advanced Superconducting Tokamak. The mission of the EAST Project is to bring out scientific issues on the continuous nonburning plasma scenario of steady-state operation and engineering issues on establishing the basis of technology for superconducting tokamak. Superconducting magnets were chosen for all poloidal field (PF) and toroidal field (TF) systems since the engineering mission is to establish the technology basis of full superconducting Tokamak for future fusion reactors. The superconducting magnets of EAST consist of sixteen TF coils and fourteen PF coils (seven coil-pairs). To obtain the good performance of the superconducting magnets, all TF magnets and most PF magnets have been tested before assembly. The assembly of the main device was completed in the end of 2005 and at the beginning of 2006, we made successfully the first engineering commissioning of the EAST system. Up to now the EAST device has been used in 4 operation campaigns and has achieved good experimental results.

The EAST plasma control system (PCS) is in continuous development to satisfy the EAST experimental requirements. In order to realize low latency and distortion-free signal transmission between PCS and servo systems such as the poloidal field power supply, in-vessel coil power supply and real-time scope, reflective memory boards (RFM) were applied. The new hardware layout and enhanced performance are reported. Newly implemented PCS control algorithms for gas control and real-time data display are also presented.

The joint TEXT (J-TEXT) tokamak, formerly TEXT/TEXT-U tokamak, operated by the University of Texas at Austin in USA, has been re-built up in Huazhong University of Science and Technology in China. The machine has been run for two experimental campaigns since 2007. Discharges with plasma current beyond 200 kA and duration of 300 ms have been obtained. At present all sub-systems, such as poloidal field (PF) and toroidal field (TF) power supplies, vacuum system, diagnostics systems etc, are successfully integrated into the routine operation. The plasma position can be elaborately controlled within two centimeters. The operational scenario of J-TEXT are typical of the machine and easily understandable compared to the former TEXT/TEXT-U results.

Europe is involved in the procurement of most of the high-technology items for the ITER device (e.g. parts of the superconducting Toroidal (TF) and Poloidal Field (PF) coils, the vacuum vessel (VV), the in-vessel components, the remote handling, the additional heating systems, the tritium plant and cryoplant and finally parts of the diagnostics). In many cases the technologies required to manufacture these components are well established, in others there is still ongoing design and R&D work to select and optimise the final design solutions and to consolidate the underlying technologies as, for example, in the areas of heating and current drive, plasma diagnostics, shield blanket and first wall, remote handling, etc. A design review has recently been conducted by the ITER Organisation, with the support of the Domestic Agencies (DAs) established by the countries participating to ITER, to address the remaining outstanding technical issues and understand the associated implications for design, machine performance, schedule and cost. This paper provides an update of the design and technical status of EU contributions to ITER.

Joint-TEXT (J-TEXT) is a medium-sized tokamak constructed at Huazhong University of Science and Technology. At present, the poloidal field power supply consists of three subsystem sets, namely Ohmic heating field power supply, vertical field power supply, and horizontal field power supply. The structure and function of the poloidal field (PF) power supply as well as the updated control system of the power supply is described. The new scheme uses the realtime feedback control method for the poloidal field power supply. The experimental results show that the poloidal field energized by poloidal field power supply meets the J-TEXT operational requirement perfectly.

Industrial studies have been carried out in Europe to prepare the manufacture of the five poloidal field (PF) coils, which will be manufactured on the ITER site. A first study, carried out by Ansaldo Superconduttori, addressed the manufacturing sequence, assuming the manufacture of the PF coils inside the two buildings which will further host the cryogenic system. A second study, carried out by Alstom investigated how to achieve the manufacture of some crucial points. A new layout of the manufacturing line was proposed, aiming at manufacture of the PF2–6 coils within 36 months. A recent study performed by Alstom, assumes the manufacture of the PF coils in a single dedicated building, releasing so the constraint of meeting the deadline fixed in ITER reference scheme by the starting point of installation of the cryogenic components.

AC operation with multiple full cycles has been successfully performed on the superconducting tokamak HT-7. In the experiment, it was discovered that the saturation of the transformer magnetic flux with DEF, a signal name, was one of key aspects that affected the AC operation. The conditions of DEF were examined through the DEF feedback control system. By controlling the working patterns of the subsystems, namely the poloidal field control system and density control system, it was guaranteed that DEF would remain in the non-saturated status.

The National Compact Stellarator Experiment (NCSX) will study the physics of low-aspect ratio, high-β, quasi-axisymmetric stellarators. To achieve the scientific goals of the NCSX mission, the devi...

This study demonstrates a remarkable flexibility of advanced divertor configurations created with the remote poloidal field coils. The emphasis here is on the configurations with three poloidal field nulls in the divertor area. We are seeking the structures where all three nulls lie on the same separatrix, thereby creating two zones of a very strong flux expansion, as envisaged in the concept of Takase’s cusp divertor. It turns out that the set of remote coils can indeed produce a cusp divertor, with additional advantages of: (i) a large stand-off distance between the divertor and the coils and (ii) a thorough control that these coils exert over the fine features of the configuration. In reference to these additional favourable properties acquired by the cusp divertor, the resulting configuration could be called ‘a super-cusp’. General geometrical features of the three-null configurations produced by remote coils are described. Issues on the way to practical applications include the need for a more sophisticated control system and possible constraints related to excessively high currents in the divertor coils.

This paper describes the latest design of poloidal field coil configuration in the national centralized tokamak (NCT), that is a new design based on the former superconducting coil tokamak JT-60SC. The most notable design change was made for the outer equilibrium field (EF) coils to increase a plasma shaping parameter S(=q95*Ip /a*Bt) and to make off-axis heating using N-NBI possible. As a result, the maximum plasma shaping parameter of ~ 7 was obtained by double null divertor configuration with a lower aspect ratio of A ~ 2.6. It was confirmed that two additional EF coils are useful for the plasma squareness control, but only one is finally adopted for ITER plasma simulation and to keep compatibility of off-axis N-NBI heating

This paper gives a description of the poloidal field coil quench protection (QP) and control of the EAST poloidal field power supplies. A configuration of the QP hardware and an operating scenario are presented; the QP design principle in the EAST tokamak is also introduced.

Two short lengths of the full-size NbTi cable-in-conduit (CIC) conductor used to fabricate the ITER poloidal field coil insert were tested in the SULTAN facility. The two investigated conductors, otherwise identical, are distinguished only by the presence and absence of the last stage subcable wraps. The voltage-current (V-I) characteristics and the quench behaviour of the conductors with and without subcable wraps are compared. The dc performance of the conductors is limited by the occurrence of sudden take-offs above approximate to 35 kA in the leg with wraps and above approximate to 45 kA in the leg without wraps. Hall probes used to study the current distribution indicate that the better performance of the conductor without subcable wraps is closely related to the lower transverse resistance providing a more efficient current redistribution. (c) 2005 Elsevier B.V. All rights reserved.

The force flow cooled superconducting cable-in-conduit conductor (CICC) is used in both of EAST toroidal field (TF) and poloidal field (PF) coils. The conductor consists of multi-stage NbTi superconducting cable and 1.5 mm thick square jacket. The cable is pulled through in a thin wall circular jacket and then compacted to square cross-section conductor. The jacket material is SUS316LN austenitic stainless steel seamless tubes (about 10 m each), which is assembled by butt-welding up to 600 m. The results of the welding procedure investigation and quality assurance procedures carrying out are described in this paper.

In the paper the consequences of an unmitigated quench in the equatorial poloidal field coil (PF3) of ITER are investigated. This coil is crucial because its thermal failure would pose a hazard to all equatorial ports of ITER. The analysis shows that about 0.08 m 3 of conductor copper will get molten temporarily but the steel jacket will contain the melt.

The design of the poloidal field (PF) coils of the International Thermonuclear Experimental Reactor (ITER) relies on the use of 45 kA NbTi cable-in-conduit conductors. An R&D programme is carried out jointly between CEA and ENEA to acquire knowledge on the behaviour of such conductors. In addition to the manufacture and tests of two subsize joint samples at CEA Cadarache, the design of a poloidal field full-size-joint sample (PF-FSJS) has been carried out. This sample is composed of two conductor legs made of stainless steel-jacketed NbTi full size cables, of a lower joint built according to the “twin box” concept developed at CEA, and of two upper terminals to be connected to the facility current leads. A description of the sample design, the techniques used for the different manufacturing steps of the conductor legs, the terminations and the assembly is related. This prototype sample will be tested in the Sultan test facility at Villigen (Switzerland) and will be a model for the manufacture of joints for the PF coils.

The design of the JT-60U tokamak, the configuration of the coil power supplies, and the operational experiences gained to date are reviewed. JT-60U is a large tokamak upgraded from the original JT-...

The problem of better choice of the input signal for the resistive wall mode (RWM) feedback control is analysed. Three cases are compared: with radial field sensors, poloidal field sensors inside the vessel, and poloidal sensors outside. It is shown that, among the signals thus measured, the largest is the internal poloidal one. This makes the system with internal poloidal field sensors more attractive than the other two. The combination of poloidal and radial sensors is proposed as an even better choice. Several tests are proposed for experimental verification in DIII-D.

This paper deals with the preliminary design of power supply system of HT-7U super-conductive Tokamak. With the emphasis on poloidal field power supplies, quench protection and electromagnetic compatibility, the design principle and features are introduced, the design scheme and R&D progress are described, the simulation studies and laboratory test are presented too.

A new concept of radial passive and active press, to replace the previous axial design is presented in this paper. This solution reduces the stresses in the components concerned and at the same time it improves the assembly, operation and maintenance of the machine.

Through theoretical analysis of thyristor switch, criterion of turn-off was derived for the design of thyristor switch. The expression of parameter design and its math model during the turn-off were deduced. The simulation and experiment have been accomplished to validate the analysis.

The linear behavior of resistive ballooning modes in the edge and scrape-off layer of diverted tokamaks is explored in the context of a collisional fluid model. It is shown that the large magnetic shear and small poloidal field in the X-point region act to increase the wave number, and hence the importance of resistivity, near the X point. The resulting “disconnection” of the eigenmodes across the X point profoundly influences the unstable spectrum. A new class of modes called resistive X-point (RX) modes exploits this synergism between resistivity and the X-point geometry, giving rise to robust growth rates at moderate-to-low mode numbers. Relative to an equivalent limited plasma, the diverted plasma is shown to be more unstable in the edge (inside the separatrix), and more stable in the scrape-off layer.

This paper reviews how we have designed the upgraded poloidal field control system on HT-7 superconducting Tokmak by applying advanced computer technologies and Multivariable decoupling control theory. This paper emphasizes on designing the method of the control system, including system components and their functions.

The TF model coil (TFMC) is one of the seven large RD the tests of the W 7-X stellarator prototype coil in an arrangement similar to the ITER TF model coil testing.