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The construction of magnet system for the tokamak device of JT-60 super advanced (JT-60SA) was completed in March 2020. The manufacturing of CS had been finished in March 2019. The circularity of 4.0 mm is 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. The CS touches the TF coils and is subjected to load, the CS can be broken. Thus the surface dimension of the CS and the TF coils was measured before the installation of CS to confirm the clearance sufficient large to avoid the CS crashing in-to the TF coils. The CS was successfully installed to the tokamak center without any damages in December 2019 as the final step of magnet assembly of JT-60SA. In this paper, the manufacturing results and the installation of the CS are described

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.

JT-60SA装置の真空排気設備の運転を開始した。目標の到達真空度は、真空容器で10-5Pa、超伝導コイル冷却開始前のクライオスタットで10-3Paである。真空容器では排気開始後、ヘリウムリーク試験、200℃ベーキング運転を行い、1.7×10-5 Paの真空度を達成している。クライオスタットでは、排気開始後、ヘリウムリーク試験を行い、超伝導コイル冷却前に3.2×10-3Paに到達した。本報告では、真空排気設備の運転実績について報告する。, プラズマ核融合学会第38回年会

In fusion reactor, large transient heat load attributed to the edge localized mode (ELM) could damage the diverter plate. Therefore, it is needed to mitigate or supress the ELM amplitude and an application of the resonant magnetic perturbation (RMP) field is thought to be one of effective solutions. Because the RMP field has three-dimensional (3D) geometry, 3D plasma fluid transport and kinetic neutral transport code, ''EMC3-EIRENE'' is suitable to model the RMP effect on steady-state parameters. In previous researches, EMC3-EIRENE simulations of RMP effects are conducted in several devices such as NSTX, AUG, and ITER. The construction of JT-60SA was completed in 2020, and various experiments will be conducted with RMP. Recently, although an EMC3-EIRENE modeling was firstly performed in JT-60SA, the applied RMP condition was only one case. Therefore, this study perform numerical simulations with several RMP strengthes in JT-60SA. The divertor heat flux profiles with RMP coil currents of Ic = 0 (without RMP), 10, 20, and 30 kA are simulated. Here, plasma responce is not considered. Except for Ic = 0 case, the strike-point splitting patterns are found with the toroidal mode number of n = 3, which is the same of that of the RMP. Furthermore, by increasing the RMP coil current, enhancement of the splitting is found. We will present detailed relationships between the RMP strength and edge and divertor plasma parameters., The 30th International Toki Conference on Plasma and Fusion Research

The thermal shield (TS) of JT-60SA which is a superconducting tokamak, is installed to reduce radiation heat load from a vacuum vessel (VV), a cryostat vessel and ports at ambient temperature to superconducting coils at 4 K. The TS is double walled structure with He cooling pipe at 80 K, and is divided electrically in toroidal direction and poloidal direction to suppress eddy currents flowing through the TS during disruption. The TS is required to be installed in a narrow space which is between the VV, the ports, and the superconducting coils. Manufacturing and assembly accuracy of the TS are required to ensure the sufficient space for the relative displacement caused by thermal displacement and seismic load. Customization of mechanical joints in the divided section of the TS is effective process for keeping the required accuracy. Assembly of 340 ° sector of the vacuum vessel thermal shield has been completed.

Two types of alignment tools has been developed for hydraulic connection between cooling pipes in divertor cassette and vacuum vessel in JT-60SA. These tools were designed to be used by a remote handling system that are planned for the maintenance and repair of the divertor cassette in the latter research phase of JT-60SA. Plasma facing components on the divertor cassette will be actively cooled because they are exposed to high heat flux. Before connecting two cooling pipes by a laser welding tool, an alignment of the groove is required. Here, the outer diameter, thickness and material of the cooling pipe are 59.8 mm, 2.8 mm and SUS316 L, respectively. The axial misalignment at the groove position is predicted to be 3 mm as the maximum owing to the installation accuracy of the pipes in the vacuum vessel. The axial misalignments were successfully aligned less than the welding range of

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.

The thermal shield (TS) of JT-60SA which is a superconducting tokamak, is installed to reduce radiation heat load from a vacuum vessel (VV), a cryostat vessel and ports at ambient temperature to superconducting coils at 4 K. The TS is double walled structure with He cooling pipe at 80 K, and is divided electrically in toroidal direction and poloidal direction to suppress eddy currents flowing through the TS during disruption. The TS is required to be installed in a narrow space which is between the VV, the ports, and the superconducting coils. Manufacturing and assembly accuracy of the TS are required to ensure the sufficient space for the relative displacement caused by thermal displacement and seismic load. Customization of mechanical joints in the divided section of the TS is effective process for keeping the required accuracy. Assembly of 340 ° sector of the vacuum vessel thermal shield has been completed.

Disruption is a critical phenomenon in a tokamak reactor. Although disruption causes serious damage to the reactor, its physical mechanism remains unclear. To realize a tokamak reactor, it is necessary to understand and control the disruption phenomenon. The present research constructs a disruption predictor using experimental high-beta plasma data in the JT-60U tokamak. The predictor was constructed using a support vector machine as a linear discriminant, and we focus on a variable selection problem for the binary classification by sparse modeling, specifically, exhaustively searching the best combinations of variables which maximize the predictor performance. By the sparse modeling, we found that the six input parameters as the best combinations. The selected parameters were the n = 1 mode amplitude | B r n = 1 | and its time derivative d | B r n = 1 | / d t , the plasma density (relative to the Greenwald density limit) and its time derivative, and the time derivatives of the plasma internal inductance and plasma elongation. In particular, it was identified that the parameter d | B r n = 1 | / d t , plays a key role on plasma disruption. We should notice that the combination with other plasma parameters is indispensable and remarkably make it possible to improve the performance of disruption prediction.

The inboard first wall for initial operation of JT-60SA is installed to protect vacuum vessel and magnetic sensors from plasma, under no-cooling-water condition for in-vessel components. The first wall is designed to withstand forces such as magnetic forces due to eddy and halo currents, acceleration due to plasma disruption and earth quake, and heat loads such as radiation from plasma, limiter configuration and ECRH direct irradiation. The inboard first wall consists of graphite tiles, stainless pedestals and base rails. The base rails are welded on the surface of the inboard vacuum vessel. The stainless pedestals are bolted on the base rails. Two graphite tiles are bolted on each support base. Shed roof shape of graphite tiles is employed to avoid heat concentration at the edge of the tiles and to mitigate heat load during limiter configuration. Thermo-mechanical numerical analyses were carried out, and it was confirmed that the stress was below the design stress against the predicted heat loads and the predicted forces.

The JT-60 Super Advanced (JT-60SA) tokamak construction has been achieved respecting the requirements of very tight tolerance for the assembly and by handling very heavy components in a very close space environment. The construction of this large superconducting tokamak represents a big step forward in the world nuclear fusion history, opening the road for ITER and DEMO. Precise assembly is required, not only to avoid mechanical interference, but also to obtain good plasma performance by less magnetic error field. To complete this work, unique and well-considered procedures were introduced. In this paper, the developed technologies and their results are reported, focusing on the assembly of the final sector of vacuum vessel, central solenoid and in-vessel components., 28th IAEAFusion Energy Conference

This report focuses on the development of the thermal insulation devices including thermal shield (TS) and cryostat for the superconducting tokamak JT-60SA. 1. Design, manufacturing and acceptance test were successfully completed by 2019 and installation will be done by March 2020. 2. The technique and knowledge to realize high accuracy manufacturing and short time installation of these devices will contribute to the ITER construction and DEMO design., 28th IAEAFusion Energy Conference

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 beenfinished 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 sufficient 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., Applied Superconductivity Conference 2020 (ASC2020)

The stabilizing plate of JT-60SA which is the largest superconducting tokamak, has been designed based on an electromagnetic and structural analysis. The stabilizing plate plays a role of both a passive stabilizer of magnetohydrodynamics (MHD) instabilities such as vertical displacement event (VDE) and resistive wall mode, and the first wall at low field side in combination with carbon tiles. The stabilizing plate is made of SS316L with low cobalt content and has double skin structure with 10mm thickness each in order to have simultaneously high resistivity in the toroidal direction and high strength against plasma disruption as well as seismic events. A finite element method for the calculation of the electromagnetic force induced by disruption and structural analysis has been applied. The most serious event which is fast (~4 ms) major disruption, is considered. The eddy current reaches up to 100 MA/m2, which induces electromagnetic force < 120 MN/m3. The distribution of eddy current in the stabilizing plate is determined by openings (ports) for diagnostics and heating which cause separation and combination of the eddy current. The stabilizing plate has been modified in order to satisfy allowable membrane, bend and peak stress of SS316L. Trial manufacture of a part of the stabilizing plate has been done to investigate the effect of weld on the deformation of the stabilizing plate resulting from the contraction of the weld metal. We have chosen how to weld, e.g. amount of welding, groove shape of each metallic plate consisting of the stabilizing plate to minimize the deformation and maximize the allowable stress. The arrangement of heat sinks and coolant pipes, and carbon tiles has also been done, taking into account the long pulse operation of JT-60SA., 31st Symposium of Fusion Technology (SOFT2020)

JT-60SAのトロイダル磁場コイルと中心ソレノイドが干渉しないよう、トロイダル磁場コイルの内径と、中心ソレノイドの外径を測定し、運転中でも4mm以上のクリアランスが確保できるように各コイルの最終加工を行った。中心ソレノイドのトカマク中心への挿入時は、レーザートラッカで位置を測定し、トロイダル磁場コイルとの隙間を管理しながら実施した。最終的に、水平方向0.6mm、高さ1.2mm、垂直度1.6mmの精度（公差はそれぞれ2.0mm）で据え付けることに成功した。, 第99回低温工学・超電導学会

In JT-60U tokamak, low frequency modes (LFMs) have been reported recently when the m/n = 2/1 neoclassical tearing mode grows sufficiently. LFMs show the low mode frequencies (< 20 Hz) with the rotation direction to the counter-current direction (ctr-direction) toroidally even if the strong injection power to the co-current direction (co-direction) is applied by tangential neutral beam injections. As a candidate to drive the rotation to the ctr-direction, the neoclassical toroidal viscosity (NTV) torque is investigated in this study. Indeed, when a LFM is observed, the estimated "offset velocity" of the NTV torque is larger than the observed toroidal velocity around the q = 2 surface and results in the torque to the ctr-direction (ctr-torque). The torque balance between the NTV torque and the torque induced by the eddy current on the resistive wall is investigated. It is found that the rotation to the ctr-direction having the low mode frequency is obtained as the result of the ctr-torque to the offset velocity. Our investigation suggests the requirement to include the NTV torque in the modelling on the mode locking of neoclassical tearing modes in tokamaks.

Eighteen error field correction coils (EFCCs) consisting of copper conductor and epoxy resin are installed in the vacuum vessel of JT-60SA in order to correct magnetic fields caused by manufacturing and installation errors of superconducting coils. The EFCCs are supported by pedestals made of stainless steel 316 L, and the pedestal has spring bars made of NCF625. The pedestals need to withstand electromagnetic force induced by EFCC current and magnetic field during plasma operations. They also need to allow relative displacement between the EFCC and the vacuum vessel during baking operations caused by the temperature difference between the vacuum vessel of 200 °C and EFCC of 40 °C. Diameter of the spring bars and connection structure for fixing the spring bars were designed through FEM analysis.

Eighteen error field correction coils (EFCCs) consisting of copper conductor and epoxy resin are installed in the vacuum vessel of JT-60SA in order to correct magnetic fields caused by manufacturing and installation errors of superconducting coils. The EFCCs are supported by pedestals made of stainless steel 316 L, and the pedestal has spring bars made of NCF625. The pedestals need to withstand electromagnetic force induced by EFCC current and magnetic field during plasma operations. They also need to allow relative displacement between the EFCC and the vacuum vessel during baking operations caused by the temperature difference between the vacuum vessel of 200 ◦C and EFCC of 40 ◦C. Diameter of the spring bars and connection structure for fixing the spring bars were designed through FEM analysis.

The stabilizing plate (SP) of JT-60SA has been designed based on an electromagnetic and a structural analysis. The SP plays a role of both a passive stabilizer of magnetohydrodynamics (MHD) instability and a first wall at low field side in combination with a graphite tile. The SP has a double skin structure with 10 mm thickness each in order to have simultaneously high resistivity in the toroidal direction and high strength against plasma disruption as well as a seismic event. A finite element method for the calculation of the electromagnetic force induced by disruption and the structural analysis has been applied. The most serious event which is fast major disruption, is mainly considered. The eddy current reaches up to 100 MA/m2, which induces electromagnetic force

Minimum sets of in-vessel components were installed for initial operation of JT-60SA in order to product and control the plasma and to know the plasma basic information, including inboard first wall, upper divertor, protection limiter, glow electrodes and magnetic sensors. All of magnetic sensors were installed within the required accuracy with in-situ measurement with laser tracker. The inboard first wall and the upper divertor were installed for the limiter configuration at the plasma initiation and the divertor configuration, respectively. They have carbon tiles as the first wall, which must be installed with accuracy of ± 1 mm in order to avoid the heat concentration. However, the vacuum-vessel wall had some deformation. Therefore, we measured in advance with laser tracker the position of the vacuum vessel where the bases of the inboard first wall and upper divertor would be welded. By using these position data and 3D CAD, we customized all bases. Owing to these procedures, we could install the first wall with required accuracy. Also, all other components were installed with required accuracy.

JT-60SAのファーストプラズマに向けて真空容器内機器の製作と設置を行った。真空容器内機器は、プラズマ対向壁、保護リミタ、グロー電極、および磁気センサーで構成される。これらは、主に有限要素法を用いた電磁力及び熱解析などに基づき設計されており、試作や製造までも含めその詳細について記述する。

Active control and stabilization of locked modes (LM) via rotating external magnetic perturbations are numerically investigated under a realistic low resistivity condition. To explore plasma responses to rotating and/or static external magnetic perturbations, we have developed a resistive magnetohydrodynamic code ‘AEOLUS-IT’. By using AEOLUS-IT, dependencies of mode behavior on frequency and polarity of the rotating magnetic perturbation are successfully clarified. Here, the rotational direction of the rotating magnetic perturbation to the equilibrium plasma rotation in the laboratory frame is referred to as ‘polarity’. The rotating magnetic perturbation acts on the background rotating plasma in the presence of a static field. Under such circumstances, there exist bifurcated states of the background rotating plasma, which should be taken into account when studying the dependence of the mode behavior on the rotating magnetic perturbation. It is found that there exist an optimum frequency and polarity of the rotating magnetic perturbation to control the LM, and that the LM is effectively stabilized by a co- polarity magnetic perturbation in comparison with a counter-polarity one.

In order to obtain information for control of JT-60SA plasma and physics research on the plasma, many types of magnetic sensors are developed. We have designed a new architecture of magnetic sensors and connection methods for JT-60SA taking installation and maintenance into account. The former include a newly designed Mirnov coil, a Rogowski coil and a diamagnetic loop. They enable us to make simple and robust sensors. The latter include a newly developed connector between a sensor and a mineral insulation cable, and a connection box. An unsealed connection box enables easy installation in a vacuum vessel; however, it should have a withstand voltage greater than 1 kV at intermediate gas pressures. We successfully achieved a high withstand voltage for the unsealed connection box in the gas pressure region around the Paschen minimum voltage. After R&D of JT-60SA magnetic sensors, almost all manufacturing steps are complete. This paper reports the manufacturing and tests of the magnetic sensors for JT-60SA.

Helical cores (HCs) in plasmas are one of the important phenomena because the helical structure affects impurity transport and induces loss of confined energetic particles or loss of toroidal momentum. Substantial studies on experimental observation of HCs were reported, because HCs would be excited in hybrid scenarios in ITER or DEMO. In addition to the effect on transport by HCs, recent studies suggested that MagnetoHydroDynamics (MHD) dynamo accompanied with HCs redistributes the current profile in the core and realizes the experimentally observed sawtooth-free plasmas of hybrid scenarios in DIII-D, in which the minimum value of the safety factor profile, qmin, is kept slightly above unity by "flux pumping". Several theoretical models to explain the excitation mechanism of HCs have been proposed, such as the generation of the magnetic island by the excitation of tearing modes (TMs) or HC equilibria induced by internal kink modes. Though substantial experimental observations were reported in previous studies, the experimental observations focusing on the detailed relationship between the mode structures of the HC and other MHD modes have not been reported well. In this paper, we report experimental finding of n = 1 HCs accompanied with saturated m/n = 2/1 TMs having low rotation frequencies in JT-60U with various diagnostics. Here, m is the poloidal mode number and n is the toroidal mode number. The TMs with HCs are observed after an increase of the mode amplitude and a decrease of the mode frequency of m/n = 2/1 precursors having the tearing parity. The decreased mode frequency is lower than 20 Hz typically. With various diagnostics, the coupling of n = 1 helical cores and m/n = 2/1 TMs have been clearly observed. Because the coherent oscillations in the ion temperature are observed in the core region and in the edge region, the flux surfaces including the m/n = 2/1 magnetic island appear to have m = 1 helical deformation. It has also been suggested that the m/n = 2/1 TM and the HC rotate in the electron diamagnetic direction keeping fm/n=1/1(HC) = 2fm/n=2/1(TM) in several plasmas even with different bulk rotation speed. Here, the fm/n=1/1(HC) is the rotation frequencies of HCs and fm/n=2/1(TM) is the rotation frequency of TMs. In addition, the core seems to be shifted to the high field side when the “O” points of the m/n = 2/1 magnetic island line up in the midplane, which is confirmed by the reconstruction of MHD equilibria with Motional Stark Effect measurement and MEUDAS code. Our observations on m/n = 2/1 TMs having HCs would contribute to the understanding of the excitation mechanism of HCs in tokamak plasmas., The 28th International Toki Conference on Plasma and Fusion Research

Prediction and likelihood identification of high-beta disruption in JT-60U has been discussed by means of feature extraction based on sparse modeling. In disruption prediction studies using machine learning, the selection of input parameters is an essential issue. A disruption predictor has been developed by using a linear support vector machine with input parameters selected through an exhaustive search, which is one idea of sparse modeling. The investigated dataset includes not only global plasma parameters but also local parameters such as ion temperature and plasma rotation. As a result of the exhaustive search, five physical parameters, i.e., normalized beta βN, plasma elongation κ, ion temperature Ti and magnetic shear s at the q = 2 rational surface, have been extracted as key parameters of high-beta disruption. The boundary between the disruptive and the non-disruptive zones in multidimensional space has been defined as the power law expression with these key parameters. Consequently, the disruption likelihood has been quantified in terms of probability based on this boundary expression. Careful deliberation of the expression of the disruption likelihood, which is derived with machine learning, could lead to the elucidation of the underlying physics behind disruptions.

First divertor cassette covering a 10° sector in the toroidal direction has been manufactured for JT-60SA. Modularized plasma facing components such as inner and outer targets, baffles, and domes are assembled into the divertor cassette. All plasma facing components are actively cooled. Carbon armor tiles bolted to water-cooled copper alloy heat sinks are used to remove heat loads of 0.3–1MW/m2 for 100 s and 10 MW/m2 for less than 5 s. The divertor cassette has two pipe joints between it and the inlet and outlet pipes on the vacuum vessel side. After removing the divertor cassette, the inner target can be replaced by cutting and welding the other two pipe joints with the divertor cassette. A remote pipe welding tool has been developed both to install the divertor cassette into the vacuum vessel and to connect the inner target with the divertor cassette. The space around the cooling pipes is so limited that they are cut and welded remotely from the inside for replacement. The laser welding method was used, and the focusing mirror inside the pipe was rotated to perform circumferential welding. As a countermeasure for welding groove misalignment, the aiming accuracy of the laser was improved by simultaneously controlling the rotation and lifting of the tool. Thereby, reliable welding is achieved even with large gaps of up to 0.5 mm and grooves with a maximum angular deviation of 0.5 ° between connecting pipes. The outer diameter of the cooling pipe is 59.8 mm, the wall thickness is 2.8 mm, and the material is SUS316L. After connecting the inner target, the other plasma facing components have been assembled into the first lower divertor cassette. When installing the divertor cassette towards the vacuum vessel, the laser welding tool described above is used to connect the cooling pipes between divertor cassette and vacuum vessel in JT-60SA, because the outer diameter and wall thickness of the cooling pipe are the same as the inner target.

In JT-60U, simultaneous excitation of n = 1 helical cores (HCs) and m/n = 2/1 Tearing Modes (TMs) was observed [T. Bando et al., Plasma Phys. Control. Fusion 61 115014 (2019)]. In this paper, we have investigated the excitation mechanism of n = 1 HCs with m/n = 2/1 TMs based on the experimental observations and a simple quasi-linear MHD model. In the previous study, it was reported that a "coupling" on the phase of the MHD mode is observed between n = 1 HCs and m/n = 2/1 TMs. In this study, it is found that the coupling is observed with the mode frequency from several Hz to 6 kHz. This indicates that the resistive wall and the plasma control system do not induce the coupling because the both time scales are different from the mode frequency. In addition, n = 1 HCs appear to be the non-resonant mode from the two observations: n = 1 HCs do not rotate with the plasma around the q = 1 surface in the core and the coupling is also observed even when qmin > 1. It is also observed that the electron fluctuation due to an n = 1 HC in the core region disappears with the stabilization of an m/n = 2/1 neoclassical tearing mode by electron cyclotron current drive, implying that n = 1 HCs are driven by m/n = 2/1 TMs. This perspective, n = 1 HCs are driven by m/n = 2/1 TMs, is supported by the observation that the saturated amplitude of the m/n = 1/1 component of the radial displacement in the core is smaller than that of the m/n = 2/1 component. Finally, we revisit a quasi-linear MHD model where the m/n = 1/1 HC is induced directly by the sideband of the current for the m/n = 2/1 TM, which allows to excite the non-resonant m/n = 1/1 mode. The model also describes the characteristic of the coupling, fm/n=1/1(HC) = 2fm/n=2/1(TM).

The inhomogeneous poloidal magnetic field of tokamak device, which is called error field, has to be reduced because the error field degrades the plasma performance. There are 18 sets of EFC coils installed inside the vacuum vessel for JT-60SA to compensate the error field. The conceptual design of EFC coils has been completed. The water-cooled hollow copper conductor was selected to reduce the conductor size since the available space for EFC coils is small. The outer size of the conductor and the diameter of the cooling channel were optimized in considering hydraulic and thermal characteristics. The design of the conductor was validated by the testing of a mock-up coil. The bar springs are used for the structure of EFC coils. The structural analysis was performed to optimize the parameters of bar springs. The results of structural analysis suggest that the structure of EFC coils can be used for both the conditions of plasma operation and baking operation with the use of Inconel625 for bar springs. In this paper, the specification of the JT-60SA EFC coils, the test results of mock-up coils, and the structural analysis results of the EFC coil structure are described. The manufacture of EFC coils has started based on these designs.

Generally, DEMO requires larger toroidal field (TF) coils than ITER, resulting in one of the major difficulties, the tolerance in TF coil fabrication. This paper presents the possible solutions based on the design study on Japan’s DEMO (JA DEMO). It was confirmed that, in the case of adopting a mitigated tolerance by a factor of 2.5–5 compared with that of ITER, the resulting error field of TF coils is correctable to an acceptable level in terms of locked mode avoidance. In addition, the design of the error field correction coil (EFCC) on JA DEMO was investigated.

Overview of plasma facing components (PFCs) in JT-60SA is introduced, including upper divertor, lower divertor, inboard first wall and outboard first wall. These PFCs are upgraded in 4 stages along plasma experiment of JT-60SA; In the Stage1, only partial inboard first wall and upper divertor are installed with graphite tiles and inertial cooling. In the Stage 2, remaining PFCs including outboard first wall and lower divertor are installed, and the inboard surface is fully covered. In the Stage 3, vertical targets of lower divertor are replaced by CFC monoblock targets with water cooling. In the Stage 4, whole carbon wall is replaced by tungsten wall, and remote handling for lower divertor is introduced. Three R&D on major issues concerning PFCs are presented. Mock-ups of CFC monoblock target were fabricated and it showed good heat removal of the expected maximum heat load of 15 MW/m2 without degradation of performance. Samples tungsten coating on CFC were fabricated and they showed no severe damage by surface temperature up to 2000 C. Laser welded samples showed good appearance and no defect was found by RT.

The operation scenarios for plasma breakdown and current ramp-up phases in JT-60SA tokamak have been developed and verified in simulation using the TOSCA code. The induced current in the conducting elements such as vacuum vessel and stabilizing plate increases to the comparable level of plasma current of ∼600 kA during the breakdown phase and thus enhances the strength of stray field. The optimized scenarios for half and full pre-magnetization cases satisfied the conditions required for the plasma initiation. At the initial plasma, the vertical magnetic field required to sustain the plasma position was controlled by the outer equilibrium field (EF) coil currents which compensate for a vertical field due to a large eddy current. The condition for the formation of divertor configurations given by the combination of the magnetic flux for plasma and the plasma current has been satisfied which enables us to develop the operational scenarios with a smooth transition from a limiter to a divertor configuration.

In the present study, the thermal endurance of epoxy-based, bismaleimides, and cyanate ester resins for the current design of the in-vessel coils was measured by performing acceleration tests to assess their insulation properties using the thermal endurance defined by the International Electrotechnical Commission (IEC-60216 Part1–Part 6) for a minimum of 5,000 h in the 180–240 °C temperature range. It was found that none of the resin insulators could tolerate the baking conditions of 40,000 h at ∼200 °C in the JT-60SA vacuum vessel. Therefore, the design of the in-vessel coils, including the error field correction coils (EFCC), was changed from the type without water cooling to with water cooling on JT-60SA.

This paper reports the results of detailed 3D modelling of the JT60-SA tokamak. Different computational tools have been used, ranging from a purely electromagnetic description to models including the plasma response. Detailed 3D finite elements meshes have been developed, including key conducting structures of JT60-SA. The positive comparison of results produced with different assumptions and independent codes increases confidence in results. Frequency-domain electromagnetic characterization of active coils has been achieved, as well as resistive wall modes growth rate computation.

We have designed in-vessel coils for a correction of magnetic error fields in JT-60SA. In order to design the in-vessel coils, namely, error field correction coils (EFCCs), error fields from several sources such as manufacturing and assembly errors of superconducting coils are calculated by Monte-Carlo approach. Required EFCC currents to correct error fields are evaluated by a least square method. Additionally, by the field line tracing, it is found that resonant magnetic perturbations (RMPs) by the EFCC enable to produce stochastic magnetic structures at the edge region, that is applicable to an ELM control.