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HL-2M is a middle-sized copper conductor tokamak, which aims to operate close-burning and high-performance plasma for experimental magnetic fusion researches. The magnetic field coil system is an important part of HL-2M machine, consisting of a toroidal field (TF) coil system and a poloidal field (PF) coil system. In the operation of HL-2M, the TF coils operate in the environment of high current, strong magnetic field, and high stress. The structure, electromagnetic, electrical and thermal issues of TF coils were solved in design and construction phase. In this paper, the structure design principle of the demountable TF coils including some key components, was introduced, and then the construction of the TF coils, especially the industry processing, was established. [ABSTRACT FROM AUTHOR]

The construction of a magnet system for the tokamak device of JT-60 super-advanced (JT-60SA) was completed in March 2020. The manufacturing of central solenoid (CS) had been finished in March 2019. The circularity of 4.0 mm is a requirement for CS manufacturing from the point of view of plasma control. The high accurate manufacturing method and jigs had been developed, and the circularity of CS achieves 1.44 mm, which meets the requirement of 4.0 mm. The clearance between the CS and the TF coils is very small, only 14 mm in design. In case the CS touches the TF coils and is subjected to load during operation, the CS can be damaged. Thus the surface dimensions of the CS and the TF coils have been measured before the installation of CS to confirm if the clearance is sufficiently large to avoid the CS crashing into the TF coils. The CS was successfully installed to the tokamak center without any damages in December 2019 as the final step of the magnet assembly of JT-60SA. In this paper, the manufacturing results and the installation of the CS are described. [ABSTRACT FROM AUTHOR]

Test blanket modules (TBMs) are key components of the ITER for the demonstration of tritium breeding. During plasma disruption, the TBMs and other in-vessel components suffer large electromagnetic forces and vibrations accompanied by a significant magnetostructural coupling effect. Aiming to evaluate the coupled vibration of a test module in J-TEXT Tokamak, a simplified analytical model was developed in this paper. The model has one degree of torsional freedom and was established by the idealization of the geometry and the confining magnetic field. The analytical solution of the model indicates that the magnetic stiffness effect is not significant for this problem, and the model gives a conservation estimation of the magnetic damping effect. To demonstrate the feasibility of the analytical model, the dynamic response of the test module in J-TEXT during a typical plasma disruption event was simulated by using an FEM numerical code of the Lagrangian sequential coupling method and was found to be in good agreement with the analytical solution. To prove the validity of the numerical model and code, the coupled vibration of a cubic block in a large constant magnetic field and a pulsed magnetic field of an electromagnet was measured by fabricating an experimental setup in a large magnet and was also calculated by using the FEM code under the experimental conditions. The measured results agree well with the numerical ones, which proved the validity of the analytical model indirectly. Due to the explicit form of the coupling term, the analytical solution gives a good reference when considering the magnetostructural coupling effect of a typical in-vessel structure in the tokamak device. [ABSTRACT FROM AUTHOR]

The effective control of magnetohydrodynamic instabilities in fusion experiments by means of actively controlled coils calls for challenging dynamic requirements to the relevant power supply (PS) systems. The first part of this paper gives an overview on typical requirements and technological solutions adopted in some present fusion experiments. Then, the advantages of SiC devices with respect to the Si ones are described in this paper, including lower switching losses, higher switching speed, higher operating junction temperature, and higher voltage capability. As an application example, this paper reports on the case of the PS system for resistive wall mode control in the JT-60SA satellite tokamak, where analyses showed the difficulty of satisfying the demanding requirements in terms of high current bandwidth (3 kHz) and short latency ( <50~\mu \texts ) with a simple H-bridge topology adopting standard IGBT. On the contrary, the use of Si–SiC IGBT with the same topology can allow meeting the specifications, as demonstrated by the development and test of such a power amplifier, rated for $300~A_{{\mathrm{pk}}}$ and 240 V. [ABSTRACT FROM AUTHOR]

The plasma configuration in tokamaks is created and controlled by magnetic fields generated by the external coils. Similarly, plasma macroscopic parameters are determined using sensors measuring magnetic field or magnetic flux. Some algorithms require separation of the plasma contribution from the contribution of poloidal field coils and currents induced in passive structures. In this paper, an efficient, real-time applicable method is introduced. The method is based on the identification of a state-space model in vacuum shots using both experimental and finite-element model data. The tests were performed for ISTTOK and RFX-mod. In , a simple method for plasma position determination for ISTTOK is proposed. We show that the centroid position measurement by magnetic diagnostics is feasible using our algorithm for separation of the plasma signal. The correctness of the resulting method is proven by comparison with the results of the heavy-ion beam. [ABSTRACT FROM AUTHOR]

This paper presents the study and analysis of the error field of Joint TEXT tokamak. Error fields are first calculated through both the finite-element method and the theoretical method. Then, a Fourier decomposition of the error field is performed on the plasma surface to evaluate the lower mode ( $m$ , $n$ ) = (1, 1), (2, 1), and (3, 1) error fields. The results of the two methods are notably consistent. Superposition and linearity are also verified for small misalignments. The major sources of error fields are discussed by comparing the error fields caused by different misalignments of various coils. A Monte Carlo simulation, which combines the contribution of all the coils, is performed to obtain the total error fields. A comparison between the calculated and measured error fields is also conducted. The results indicate that the analysis method presented in this paper is very effective in predicting the error field. [ABSTRACT FROM PUBLISHER]

Nowadays, divertors are used in the main tokamaks to control the magnetic field and to improve the plasma confinement. In this paper, we present analytical symplectic maps describing Poincaré maps of the magnetic field lines in confined plasmas with a single-null poloidal divertor. Initially, we present a divertor map and the tokamap for a diverted configuration. We also introduce the Ullmann map for a diverted plasma, whose control parameters are determined from tokamak experiments. Finally, an explicit, area-preserving, and integrable magnetic field line map for a single-null divertor tokamak is obtained using a trajectory integration method to represent toroidal equilibrium magnetic surfaces. In this method, we also give examples of onset of chaotic field lines at the plasma edge due to resonant perturbations. [ABSTRACT FROM AUTHOR]

Control of the plasma current and the position of its centroid and its shape is vital for obtaining advanced performance regimes in tokamak operation. For that aim, the use of poloidal magnetic coils for feedback control is essential. In this paper, a multiple-input multiple-output (MIMO) optimal control algorithm for the vertical and horizontal control position of the plasma current centroid at IST tokamak (ISTTOK) is presented. Using the ISTTOK magnetic system, the real-time control is made possible by the execution of generic application modules (GAMs), developed to be executed on the Multithreaded Application Real-Time executor ISTTOK controller. The design of the proposed MIMO controller is based on a state-space model of the ISTTOK magnetic response that has been identified from experimental data. Both the control design and the identification procedure rely on the reconstruction of the plasma centroid position computed by the magnetic probes (Mirnov coils) installed at ISTTOK. The centroid position is reconstructed in the Position GAM by numerical integration of the magnetic probes signals in the recently upgraded hardware based on the Advanced Telecommunications Computing Architecture. The effectiveness of the proposed control approach is presented comparing simulation results with experimental ones obtained using single-input single-output PID controllers. Moreover, due to the modularity of the ISTTOK control system, the proposed real-time reconstruction algorithm has been easily deployed on the plant, and the same approach will be adopted to test the MIMO controller in the future. [ABSTRACT FROM AUTHOR]

Tokamak devices such as ITER, EAST, and HL-2M have been designed and constructed (or under construction) to challenge the magnetic-confined nuclear fusion technology for the sustainable energy problem of human beings. The magnet system of the tokamak devices is mainly composed with poloidal field (PF) coils, toroidal field (TF) coils, and central solenoid coils. Among them, the PF coils play a key role in adjusting the confining magnetic field for the control of the fusion plasma. In order to avoid the plasma disruption, it is necessary to know the correlation of the magnetic field in a vacuum vessel (VV) with the current in PF coils and to predict the current distribution of PF coils which can generate a required configuration of confining magnetic field. In this paper, taking the HL-2M tokamak device as an example, an inverse analysis scheme to predict and identify the perturbation of current in PF coils from the target magnetic field signals in the VV was proposed. First, based on the Biot–Savart law, a forward code to calculate the magnetic field due to PF coil currents and TF coil currents was developed. Second, an inversion code to predict and identify current distribution in PF coils from the target magnetic field was developed based on the conjugate gradient optimization method and the singular value decomposition method. The validity of the proposed inversion scheme and the corresponding numerical codes was investigated by reconstructing PF coil current distributions from several groups of perturbed magnetic fields. In addition, the effect of the measurement points and the measurement noises on the reconstruction results was also discussed. Finally, a simplified experiment on a coil system was conducted to verify the validity and efficiency of the proposed forward and inversion analysis code. [ABSTRACT FROM AUTHOR]

Magnetic compression (MC) technology was recommended for tokamak to study compressed plasma and a power module have been designed here for developing its server power supplier. The minor-radius compression is one of the most effective methods to improve the performance parameters of existing tokamaks, enabling the fusion plasmas operated at high density, high temperature, and high beta. In this paper, a high-frequency and high-power supply system is proposed, used for powering coils of MC within the vacuum chamber of the experimental advanced superconducting tokamak. The basic of the power module is an ac/pulse converter, implemented by a series-resonant full-bridge circuit and controlled by pulsewidth modulated. Also, control method adopts current closed-loop proportional–integral control, which has less than 3-ms current response time in real time. The converter is analyzed and the design procedure is discussed. Experimental results obtained from a 3-kW converter prototype are presented to validate the converter’s performance with the redesigned control board. [ABSTRACT FROM AUTHOR]

The poloidal field (PF) coils are one main subsystem of the ITER magnets. The PF6 coil is being manufactured by the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) as per the PF coils cooperation agreement between ASIPP and Fusion for Energy. The ITER PF6 winding pack is composed by stacking nine double pancakes. Dummy double pancake fabrication aims to fully qualify the critical fabrication processes, before series production. This paper describes ITER PF6 dummy double pancake winding process and results, which was performed with a “two-in hand” configuration. Conductors were wound sequentially in the following steps: despooling, straightening, ultrasonic cleaning, sandblasting, bending, and turn insulation wrapping until deposition on the winding table. Joggle bending, helium inlet manufacturing, and tail assembly were also carried out during winding. The dummy double pancake winding was finally completed with the qualification requirements well met, which demonstrated good synchronization of the control system, fine conductor positioning accuracy, tight radial buildup, and so on. [ABSTRACT FROM AUTHOR]

The test about simple central difference average (CDA) + MIK ( $\text{M}_{ik}$ mutual inductance between ith and kth component) quench detection method implemented a total of about 100 charging runs on EAST in the past two years, aiming at verifying this new method whether to apply it to ITER CS coils or not. This project is supported by the ITER Organization. To obtain this target, the whole instrumentation hardware have been developed and installed on EAST and used to generate experimental data. Considering the whole new CS coils connection configuration and the possible risk for EAST device, the typical achievement is to accomplish successfully plasma scenarios (the longest shot of plasma current around 250 kA and pulse 3 s) by using a simple CDA + MIK quench detection method on EAST for the first time. Moreover, the numerical quench detection model has been designed, and the numerical signals are in good agreement with experimental data. Recent experimental results showed that CDA + MIK technology is applicable to the ITER CS coils as a backup method but the sensitivity of this system should be improved greatly in order to meet experiment requirements for ITER operation in the future. This paper introduces the test program, typical achieved operation, and the results of preliminary analysis. [ABSTRACT FROM AUTHOR]

In JT-60SA, the control of resistive-wall-mode (RWM) instabilities will also be realized with a dedicated active control system based on 18 in-vessel sector coils. Each coil will be fed independently by a dedicated fast inverter. Due to the outstanding dynamic performance required, new insulated-gate bipolar transistor modules based on silicon carbide (SiC) have been adopted. Being capable to switch at 30 kHz, these components, together with a very fast control system, allow reaching the required high dynamic performance with the simple and compact H-bridge topology. The RWM-PS will be the first power supply system for fast control of plasma instabilities in fusion experiments adopting SiC semiconductors. This paper gives an overview of the final design of the RWM-PS, with particular emphasis on its special features and the solutions implemented to satisfy the critical requirements. The issues related with the high switching frequency and the peculiar nature of the load will be treated in detail. [ABSTRACT FROM AUTHOR]

In this paper, the influence of stray magnetic fields (SMFs) in the ITER electron cyclotron resonance heating and current drive (ECRH-CD) building on the operation of the European (EU) 170 GHz, 1 MW, continuous wave gyrotrons is investigated. There are mainly two sources of SMF in the ITER ECRH-CD building: 1) the tokamak and 2) the neighboring gyrotrons which are placed in close proximity. The possible electron beam degradation caused by the influence of the SMF is numerically investigated. Furthermore, the collector operation under the influence of an SMF is also studied. A proposal to compensate the influence of the SMF is discussed. [ABSTRACT FROM PUBLISHER]

China Fusion Engineering Test Reactor (CFETR) is a tokamak reactor; one design option under the consideration of the China National Integration Design Group employs superconducting magnets. The fusion power is at the range of 50–200 MW and the duty cycle (or burning time) was envisioned as 30%–50%. The plasma current will be 10 MA and the major and minor radii are 5.7 and 1.6 m, respectively. The concept engineering design including the magnet system, vacuum vessel (VV) system, and maintenance method has been carried out in the past years. The toroidal magnetic field strength at R0 is 5 T and the maximum flux swing provided by central solenoid winding will be 160 VS. The main design work, including the electromagnetic analysis of the magnet system, has already been carried out. This paper mainly probes into the VV design and optimization based on three types of maintenance ports. Furthermore, the maintenance method, counted as one of the most important design studies, is presented specifically in this paper. It includes the design of the maintenance ports and the remote handling system design, and so on. The next design stage will mainly include mechanical design, conductor stability, different types of divertor system design, and so on. [ABSTRACT FROM PUBLISHER]

As Tokamak toroidal field (TF) coils operate at a high magnetic field and experience a variety of transient heating loads, the applicability of no-insulation (NI) techniques to wind the TF coil is taken into consideration here in an effort to secure high current density and high thermal stability. This paper presents the construction and test results of a NI high temperature superconducting (HTS) toroidal magnet. We fabricated the magnet using BSCCO wire manufactured by Sumitomo. The magnet consists of ten racetrack double pancake (DP) coils. Each DP coil was tested separately in liquid nitrogen at 77 K to obtain its critical current and characteristic resistance, then, after assembly and soldering, the magnet was tested at 77 K to measure the variations in voltage and magnetic field during charging, discharging, and current disturbancing. Finally, we discuss the thermal runaway behavior of the toroidal magnet during overcurrent operation as well as the influence of the quenched coil on the other coils in the system. [ABSTRACT FROM PUBLISHER]

The subassembly procedure and relevant tools for the complete 40^\circ sector assembly as a foundation of the ITER machine have been developed to meet technical and functional requirements for years. Based on the concept of subassembly procedure and tools provided by ITER Organization, detailed procedure and relevant purpose-built assembly tools are developed by Korea Domestic Agency. The tools described in this paper contain an upending tool, a lifting tool, a sector subassembly tool, and a radial beam and vacuum vessel supports with alignment units. To verify the 6-degree-of-freedom system for fine alignment of components in the sector subassembly tool, scale-down mock-up has been manufactured and tested. This paper presents the overall 40^\circ sector subassembly procedure, the fundamental specifications of the relevant assembly tools, and the current status of the mock-up developments. [ABSTRACT FROM PUBLISHER]

A toroidal field coil is composed of several individual toroidal coils (ITCs) which are connected in a series and distributed in a toroidal and symmetrical form. Cross section of ITCs is rectangular or negligible. This paper presents analytical equations of mutual inductance of two ITCs applicable to tokamak reactors using the filament method. These equations are based on those formulated by Neumann. The numerical analysis of the integrations resulting from these equations is solved using the extended three-point Gaussian algorithm. The finite element method (FEM) is employed to verify the mutual inductance equations of ITCs. The results obtained using FEM, when dimensional parameters of ITCs are changed, confirm the analytical and empirical results showing an error of less than 0.2043% in the worst case. This indicates the reliability of the presented equations. This paper also employs FEM to obtain depth of penetration of electromagnetic waves in high-conductivity conductors (HCCs). The results show that the ratio of real depth of penetration, obtained by FEM, to ideal depth of penetration in HCCs is independent of frequency. This ratio shows that the magnitude of traveling plane waves in HCC decreases with e^-1.7916 rather than e^-1. [ABSTRACT FROM AUTHOR]

An increase in the magnetic strength of the vertical field will, in general, compress and heat plasmas, which thus allows high temperature, high density, and high beta to be reached. In this paper, according to the scaling laws for the magnetic compression, the scaling range of both the major-radius compression and the minor-radius compression is figured out. Then, the comparison between the major-radius compression and the minor-radius compression is carried out. It demonstrates that the minor-radius compression can obtain higher \beta p of plasma. Consequently, we try to significantly enhance discharge parameters by the minor-radius compression by inserting the sets of the vertical-field coils in the experimental advanced superconducting tokamak (EAST) vacuum chamber. Afterward, the minor-radius compression based on EAST shot #34128 is deduced with Ca=1.5 and CR=0.925 , which increases \beta p by 76.19%. It notes that the minor-radius compression is a promising way to effectively improve \beta p in a tokamak. Finally, the preliminary design of the minor-radius compression, which includes the control system, the in-vessel vertical-field coils, and its fast-control power supply, is described. [ABSTRACT FROM PUBLISHER]

Rotating magnetic field (RMF), which interacts with plasma through electromagnetic force, is increasingly used for plasma instability control in magnetic confinement fusion experiment devices. In general, energizing magnetic field coils to obtain the desired RMF requires multiphase currents with an amplitude of up to several kiloamperes at a frequency ranging from a hundred hertz to several kilohertz. In this paper, we propose a two-phase ac power supply (ACPS) system based on full-bridge resonant inverters as an extension of the application of multiphase resonant converters. A novel phase control scheme for the two-phase resonant inverter system using a digital phase-locked loop is also proposed to synthesize RMF with enough spectral purity. The completed power system consists of two units. Each unit can supply a load with a rated output power of 1200 kW and a sinusoidal current in the frequency range of 1–5 kHz. The proposed ACPS can effectively cope with the physical requirements. [ABSTRACT FROM PUBLISHER]

Plasma energy, poloidal beta, plasma pressure, plasma temperature, plasma resistance, and plasma effective atomic number are important parameters in tokamak plasma experiments. In this paper, we present a theoretical and experimental approach especially based on diamagnetic loop for measurement of these parameters in the presence of resonant helical field (RHF) and biased electrode in IR-T1 tokamak. For this purpose, an electrode biasing system and a diamagnetic loop with its compensation coil designed, constructed, and installed on the IR-T1. In addition, we measured the plasma current, plasma voltage, and the plasma density from the Rogowski coil, the poloidal flux loop, and the Langmuir probe measurements, respectively. Experimental results show that the addition of a relatively small amount of RHF with \(L=2\) and \(L=3\) modes could be effective for improving the quality of tokamak plasma discharge by creating the steady-state plasma. [ABSTRACT FROM PUBLISHER]

The Procurement of Feeder system in Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) for The International Thermonuclear Experimental Reactor (ITER) Tokamak device has been carrying out after the signature of Procurement Arrangement in October 2011. The relevant main tasks in 2012 and 2013 are the qualification of some key subcomponents, including high temperature superconducting current leads (HTS-CL), superconducting joints, 80-K thermal shield (TS), vacuum barrier, in-cryostat feeder, cold mass support, etc. The production of manufacturing drawings, preparation of manufacturing plan, manufacture and qualification of tooling, and qualification of subcomponents are all involved in this phase. This paper describes the progress made by ASIPP and some sub-contractors for the feeder's qualification work. [ABSTRACT FROM PUBLISHER]

The Chinese Fusion Engineering Test Reactor (CFETR) is under conceptual design, which will complement ITER. One of the critical issues for the long pulse or steady-state operation in CFETR is how to reduce heat loads below a limit on plasma facing components, especially on divertor plates. This paper presents the selected optimized results of the snowflake diverted equilibria in CFETR. [ABSTRACT FROM PUBLISHER]

The main purpose of this paper is the determination of plasma equilibrium parameters, as well as current density and safety factor profiles, in IR-T1 tokamak. For this purpose, a diamagnetic loop with its compensation coil and an array of magnetic probes were designed, constructed, and installed on the outer surface of the IR-T1 chamber, and then, the poloidal beta and poloidal and radial magnetic fields were measured. Moreover, a few approximate values of the internal inductance for different possible profiles of the plasma current density were also calculated. Then, from these two techniques results, the plasma horizontal displacement was determined. In addition, from the results, the current density and q-profiles were obtained. [ABSTRACT FROM PUBLISHER]

The feeder design has been improved by the feeder teams at the ITER Organization (IO) and the Institute of Plasma Physics, Chinese Academy of Science (ASIPP) by incorporating the results of mechanical and thermal analyses as well as the system integration and assembly tolerances in the present CAD model. The feeder design is being finalized progressively, and will be delivered to the Chinese Domestic Agent (CNDA) for further procurement arrangement (PA) activities. Pre-PA manufacturing studies and tests performed at ASIPP have been effective in clarifying feeder design feasibility and component manufacturability. This paper reports the recent advancements on feeder design, analysis and manufacturing studies. [ABSTRACT FROM PUBLISHER]

The JT-60SA experiment will be the world's largest superconducting tokamak when it is assembled in Naka, Japan (R=3\ m, a=1.2\ m). This paper describes the approach taken to define appropriate manufacturing tolerances and metrology points for each toroidal field (TF) coil and in particular the proposed procedure for the final assembly of the TF magnet system. [ABSTRACT FROM PUBLISHER]

The Procurement of the feeder system for The International Thermonuclear Experimental Reactor (ITER) tokamak device started in ASIPP (China) in 2006. The relevant tasks of structural design and analysis have now come to an end, but the R&D activities started in 2009 for critical feeder components and key technology are still ongoing. Following the signature of the Procurement Arrangement between China and the ITER Organization (IO) in 2010, a great number of R&D tasks have been launched with the purpose of testing and verifying the original design, leading to design changes and optimizations. This paper mainly discusses the progress of the feeder design and R&D activities; it also presents some test results. [ABSTRACT FROM PUBLISHER]

To investigate the interactions between external resonant magnetic perturbations (RMPs) and a tokamak plasma, a set of saddle coils are designed for the J-TEXT. A method for designing RMP field structure based on Fourier transform is used in this paper. Since the coils will be mounted inside the vacuum vessel where the pressure is about 10^-7\ Pa, an airproof multi-layer structure is adopted to avoid contaminating the vacuum environment. Besides, stress and thermal analysis and the AC response of magnetic field are also given. [ABSTRACT FROM PUBLISHER]

The Korea Superconducting Tokamak Advanced Research (KSTAR) Central Solenoid (CS) magnet system coils are made of Cable-In-Conduit Conductors (CICC) that contains 240 Nb3Sn strands and 120 copper strands inside an Incoloy 908 jacket. It consists of 4 symmetric pairs of coils to the equatorial plane. Each coil is wound with no internal joints by the continuous pancake winding method. It operates with pulsing current, which naturally induces inductive voltages across coils. Minimizing the inductive voltages is critical for the quench detection. To suppress the inductive voltages, each coil voltage was measured by using conductive tapes that are wound at the outer surface of the jacket with the same pitch length of the final sub-cable. During the KSTAR campaigns, the voltages were collected and analyzed. In addition, more rejection schemes were applied to enhance the stability and reliability of the quench detection. The paper deals with the up-to-date quench detection method of the KSTAR CS magnet system and its experimental results. [ABSTRACT FROM PUBLISHER]

The International Thermonuclear Experimental Reactor is an international scientific project with the aim of building a tokamak fusion reactor capable of producing at least 10 times more energy than that spent to sustain the reaction. In a tokamak the fusion reaction is magnetically confined and the toroidal field coil system plays a primary role in this confinement. Fusion for Energy, the European Domestic Agency for ITER, is responsible for the supply of 10 out the 19 toroidal field coils. Their procurement has been subdivided in three main work packages: the production of 70 radial plates (the structural components which will house the conductors), the manufacture of 10 winding packs (the core of the magnet) and cold test and insertion into the coil cases of 10 winding packs. The cold test/insertion work package presents significant technological challenges. These include the cold test of the winding packs 14 m high, 9 m wide and weighing 110 t, the welding and inspection of the 316 LN stainless steel coil case, with welded thicknesses of up to 144 mm accessible only from one side combined with the need to minimize the deformation during the welding process (more than 70 m of weld per coil and up to 90 passes to fill the chamfer) and the resin filling of the coil case after insertion of the winding pack (the total volume to be filled up is about one cubic meter per coil). From 2009 up to mid 2011, F4E has carried out an R&D program in order to investigate the most challenging steps of the manufacturing processes associated to this work package, both to meet the demands of the ITER schedule and to minimize technological risks; in this paper an overview of the results obtained is presented. [ABSTRACT FROM PUBLISHER]

JAEA has conducted qualification and rationalization activities in Japan since March 2009 in order to rationalize manufacturing procedure of ITER toroidal field coil (TFC) structures. These following activities has been performed: qualification of structural materials and qualification and development of welding procedure according to the JSME code constituted for fusion devices, demonstration of the manufacturing method and procedures through manufacture of 1 m mock-ups and full-scale segments of TFC structure. From results of these activities, JAEA confirmed quality of materials and weld joints of gas tungsten arc welding and gas metal arc welding. In addition, JAEA obtained knowledge of welding deformation of actual TFC structures. This paper describes results of these qualification and development activities for ITER TFC structure. [ABSTRACT FROM PUBLISHER]

The ITER Magnet System contains 18 Toroidal Field Coils (TFC). These are large D-shaped coils of about 300 t, 17.5-m height and 9-m width. They consist of a Winding Pack (WP) enclosed in a rigid structural steel case, the Toroidal Field Coil Case (TFCC). The WP is a bonded structure of 7 Double Pancakes (DP), each made up of a radial plate (RP) housing the reacted Nb3Sn cable-in-conduit superconductor (CICC), which operate at 4.5 K in supercritical helium. The conductor carries a current of 68 kA in operation to produce a nominal peak field of 11.8 T. The total stored magnetic energy in the 18 TFCs is 41 GJ. While the Japanese and European Domestic Agencies that are in charge of the procurement of the TFCs are progressing with the manufacturing design and the fabrication trials prior to launch the production of the real coils, the ITER Organization (IO) is completing the development and qualification of the most critical items, e.g. cyanate ester and resin blends for the conductor and WP insulation system, the terminal region, the helium inlet, a charged resin system for the filling of the gap between the WP and the TFCC and the general tolerancing especially at the interfaces between the neighboring systems. This paper presents the final design of the TFCs and the results of the developments carried out in the aforementioned areas in the last 2 years. [ABSTRACT FROM PUBLISHER]

Testing each of the SST-1 Toroidal Field (TF) Magnets under representative conditions in cold with nominal currents was one of the primary and mandatory requisite in the refurbishment of the Steady-State Superconducting Tokamak (SST-1). Under this mandate prior to the SST-1 machine shell re-assembly, each of the 16 SST-1 TF magnets have been tested fully and successfully in a dedicated test stand. The campaign began on June 10, 2010 and was concluded on Jan 24, 2011 spanning over seven and half months without any interruption in-between. These campaigns ensured that all the sixteen TF magnets could be charged to their nominal currents of 10000 A in either two-phase or supercritical cooling conditions with leak-tight inter-double pancake resistances being in the range of 150 pico-ohms to 1200 pico-ohms. Sensors and diagnostics signal conditioning and data acquisitions, magnet quench detection and magnet protection aspects in these campaigns have been precise, extensive and redundant wherever feasible. Test team had many failures as well as several learning in the course of these challenging tasks. Magnet preparations for the tests, the test designs and their executions, some exciting and confidence building experimental results as well as some failure experiences have been presented in this paper. [ABSTRACT FROM PUBLISHER]

JT-60SA is the satellite tokamak for ITER in the Broader Approach agreement. The JT-60SA uses 18 toroidal field coils, a central solenoid with 4 modules, and 6 equilibrium field coils, they are all superconducting coils with forced flow cooled conductors. All detailed designs of these superconducting coils have been completed. The manufacturing of conductors and coils are progressing in Japan and EU. This paper shows the latest manufacturing activities and final design adjusting of its magnet system and their utilities. [ABSTRACT FROM PUBLISHER]

As a part of refurbishment of Steady-state Superconducting Tokamak-1 (SST-1) at the Institute for Plasma Research, India, all 16 toroidal field (TF) magnets of SST-1 have been individually tested in cold with nominal current in a dedicated experimental cryostat. The magnets were cooled down to 4.5 K using either supercritical or two-phase helium, after which they were charged up to 10 kA of transport current. Precise cryogenic temperature and magnetic field measurements in the experimental configuration were mandatory. Temperatures were required to be accurately measured at several locations in the magnet and hydraulic circuits, whereas the field measurements were carried out at few predefined locations on the magnet case. Highly accurate in-house modular signal conditioning electronics had been developed for low temperature and high magnetic field measurements for such large magnets. This paper describes the scheme of the measurement, precautions taken for enhanced reliability, precautions taken for long-term offset stability in cryogenic environment, and the results of SST-1 TF magnet tests. [ABSTRACT FROM PUBLISHER]

This paper describes the detailed design of the quench protection circuits (QPCs) for the superconducting toroidal field (TF) and poloidal field (PF) magnets of the satellite tokamak JT-60SA, which will be installed in Naka, Japan. The nominal currents to be interrupted and the maximum reapplied voltages are 25.7 kA and 2.8 kV for the TF QPCs and 20 kA and 5 kV for the PF QPCs. The innovative solution proposed in the QPC design is based on a hybrid circuit breaker (CB) composed of a mechanical bypass switch for conducting the continuous current, in parallel to a static CB for current interruption. The main choices of the final design are presented and discussed, either to confirm or to update and complete the study performed at the conceptual design level. [ABSTRACT FROM PUBLISHER]

In the framework of the JT-60SA project, aiming at upgrading the present JT-60U tokamak toward a fully superconducting configuration, the detailed design phase led to adopt for the three main magnet systems a brand new design. Europe (EU) is expected to provide to Japan (JA) the totality of the toroidal field (TF) magnet system, while JA will provide both Equilibrium field (EF) and Central Solenoid (CS) systems. All magnet designs were optimized trough the past years and entered in parallel into extensive experimentally-based phases of concept validation, which came to maturation in the years 2009 and 2010. For this, all magnet systems were investigated by mean of dedicated samples, e.g. conductor and joint samples designed, manufactured and tested at full scale in ad hoc facilities either in EU or in JA. The present paper, after an overall description of magnet systems layouts, presents in a general approach the different experimental campaigns dedicated to qualification design and manufacture processes of either coils, conductors and electrical joints. The main results with the associated analyses are shown and the main conclusions presented, especially regarding their contribution to consolidate the triggering of magnet mass production. The status of respective manufacturing stages in EU and in JA are also evoked. [ABSTRACT FROM AUTHOR]