Superconducting (SC) tokamak JT-60SA plays an essential role in fusion research and development by supporting and complementing the ITER project, providing directions to the DEMO design activity and fostering next generation scientists and engineers. Since the short circuit incident at the terminal joints of equilibrium field coil #1 during the integrated commissioning (IC) in March 2021, both EU and JA implementing agencies (IAs) have examined how to ensure safe operation of JT-60SA by mitigating the risk of possible discharge occurrence inside the cryostat. Based on the experience of the global Paschen tests, the IAs have established a strategy of risk mitigation measures, which is a combination of (i) reinforcement of insulation, (ii) avoiding unnecessary voltage application to the coil systems and (iii) immediate de-energization of the coils when deteriorated vacuum conditions are detected. Thanks to the considerable efforts of the Integrated Project Team members, the IC restarted in May 2023. After confirmation of the SC state of the coil systems (TF, EF and CS), the coil energization test and the plasma operation phase 1 (OP-1) started. The first plasma was successfully achieved on 23 October 2023 with a limited value of voltage and current applied to the coils. The plasma configuration control was also confirmed with low plasma current and low auxiliary heating power conditions. Based on the IO-F4E-QST collaboration, activities of JT-60SA have been shared with the IO and provided an important lesson for ITER assembly and commissioning, and will provide an outstanding contribution to fusion research at large. After OP-1, maintenance & enhancement phase 1 (M/E-1) starts from January 2024, in which in-vessel components are installed, and heating and diagnostic systems are extensively upgraded to allow a high power heating experiment planned in OP-2. In order to make the best use of JT-60SA, a newly organized JT-60SA experiment team will refine the research plan for the future high heating power operation phase.

WEST is an MA class superconducting, actively cooled, full tungsten (W) tokamak, designed to operate in long pulses up to 1000 s. In support of ITER operation and DEMO conceptual activities, key missions of WEST are: (i) qualification of high heat flux plasma-facing components in integrating both technological and physics aspects in relevant heat and particle exhaust conditions, particularly for the tungsten monoblocks foreseen in ITER divertor; (ii) integrated steady-state operation at high confinement, with a focus on power exhaust issues. During the phase 1 of operation (2017-2020), a set of actively cooled ITER-grade plasma facing unit prototypes was integrated into the inertially cooled W coated startup lower divertor. Up to 8.8 MW of RF power has been coupled to the plasma and divertor heat flux of up to 6 MW m(-2) were reached. Long pulse operation was started, using the upper actively cooled divertor, with a discharge of about 1 min achieved. This paper gives an overview of the results achieved in phase 1. Perspectives for phase 2, operating with the full capability of the device with the complete ITER-grade actively cooled lower divertor, are also described.

The general and detailed design of the JT-60SA toroidal field (TF) coil system was done by all the Voluntary Contributors in the project, CEA, ENEA, and F4E. The French part including the supply of 9 + 1 spare of the 18 TF winding packs and their integration in casings was entrusted mid-2011 to Alstom (Belfort, France), now General Electric. Manufacturing flow definition, drawings, and quality assurance documentation were achieved prior qualification of the critical processes. The procurement and commissioning of the needed tooling was led in parallel and the production was started in January 2014. The first winding pack was ready in the end of 2014. The integration inside the casing started in March 2015 for coil completion end 2015. Then, the coil was delivered to the coil test facility for testing at nominal operation condition (T = 4.5 K, I = 25, 7 kA) and for qualifying the coil resilience to quench events. In parallel, the successive coils were engaged in the manufacture process. The paper reports the status of the manufacture and gives a feedback on the main issues and on the technical solutions implemented. A focus is made on the mastering of the manufacturing processes, which is highlighted by the learning curve of the TF coils manufacturing.

The JT-60SA experimental device will be the world's largest superconducting tokamak when it is assembled in 2019 in Naka, Japan (R = 3 m, a = 1.2 m). It is being constructed jointly by institutions in the EU and Japan under the Broader Approach agreement. Manufacturing of the six NbTi equilibrium field (EF) coils, which have a diameter of up to 12 m, has been completed. So far, 13 of the 18 NbTi toroidal field (TF) coils, each 7-m high and 4.5-m wide, have also been manufactured and tested at 4 K in a dedicated test facility in France. The first three of four Nb3Sn central solenoid (CS) modules have been completed, as have all of the copper in-vessel error field correction coils. Installation of the TF magnet, around the previously welded 340° tokamak vacuum vessel and its thermal shield, started at the end of 2016 and is currently underway. The TF magnet will in turn support the EF and CS coils.

The 18 D-shaped Nb-Ti toroidal field (TF) coils for the JT-60SA tokamak will each be 7 m high and 4.5 m wide. Together they will generate an on-axis field of 2.25 T. All the main contracts for their manufacture are now in place, with manufacturing split primarily between sites in Japan (superconducting strand), Italy (conductor cabling and jacketing, casings fabrication and coil winding, and integration), and France (support structures, coil winding and integration, and final coil cold testing). This paper will summarize the key aspects of the design of the coils and the current status of manufacture on each area of the manufacture of the TF coils. A simple overview of the overall schedule for their completion is included.

•We describe the status of the JT-60SA TF coils manufacture at General Electric.•The completion of the manufacturing is described.•The main issues during manufacture are highlighted.•The cold tests main results and status are described.•The pre-assembly operations and status of the TF coils sending to Japan is reported.

The general design of the JT-60SA TF system was defined by all the participants in the project (CEA, ENEA, F4E); the detailed design was issued by the voluntary contributors. For the French part, including the procurement of 9 of the 18 TF winding packs and their integration in the casings, an industrial contract was signed with Alstom (France). After the first phases, includingmanufacturing flow definition, manufacturing drawings and QA documentation, critical processes qualification, as well as procurement and commissioning of the tooling, the production-was started in January 2014. After solving the setting issues linked to industrialization of complex tooling such as the winding machine or impregnation mold, the winding operations, as well as the joint area manufacture and the impregnations, were led. The first winding pack insertion process, as well as casing welding using an industrial robot, was performed during the year 2015 with the target to deliver the first coil to the cold test facility before the end of the year. Of course, in parallel, the successive coils manufacturing was engaged. This paper presents the status of the production with a focus on the issues encountered on the first coil and the production objectives.

The most distinctive feature of the superconducting magnet system for JT-60SA is the optimized coil structure in terms of the space utilization as well as the highly accurate coil manufacturing, thus meeting the requirements for the steady-state tokamak research: a conceptually new outer inter-coil structure separated from the casing is introduced to the toroidal field coils to realize their slender shape, allowing large-bore diagnostic ports for detailed plasma measurements. A method to minimize the manufacturing error of the equilibrium-field coils has been established, aiming at the precise plasma shape/position control. A compact butt-joint has been successfully developed for the Central Solenoid, which allows an optimized utilization of the limited space for the Central Solenoid to extend the duration of the plasma pulse.

Within the framework of the Broader Approach, the French voluntary contributor represented by CEA awarded a contract for the production of 9 toroidal field coils for the JT-60SA project to Alstom, Belfort, France in mid-2011. A first preparatory phase was led to establish the team, produce the manufacture drawings, define the manufacturing process, procure the required tooling and prepare the quality documentation. In parallel, a qualification phase on the critical major processes has proved Alstom's ability to master the processes and reach the requirements. After reviewing of the qualification results and modification of the processes and tooling to overcome the encountered difficulties, a Production Readiness Review has authorized Alstom to start the production winding. A prototype double pancake was wound as the first of series. In addition to complying with the pancake width all around the D shape, the straightness of the centreline in the critical straight leg part was correct. The production of the successive double pancakes to constitute the first winding pack was then completed and the joints and terminals were manufactured. The paper describes the completion of the last qualifications and the status of the winding production. (C) 2015 Elsevier B.V. All rights reserved.

The JT-60SA experiment will be the world's largest superconducting tokamak when it is assembled in 2019 in Naka, Japan (R=3m, a=1.2m). The superconducting magnet system includes 18 D-shaped toroidal field coils, each 7m high and 4.5m wide, 6 pulsed equilibrium field coils up to 12m in diameter and 4 central solenoid modules. Manufacturing of the superconducting magnets for JT-60SA is well established in Japan and in Europe. Conductor manufacturing is almost complete, half of the superconducting coils have been wound and the first cold test results for production coils will be available later in 2015. Challenges remain to integrate the coils with their mechanical structures and to assembly them into the tokamak.

The general design of the JT-60SA toroidal field system was defined in agreement with all the participants in the project (CEA, ENEA, F4E), the detailed design was issued by the Voluntary Contributors. For the French part including the procurement of 9 of the 18 TF winding packs and their integration in the casings, an industrial contract was signed mid-2011 with Alstom (France). After agreement on manufacturing drawings and QA documentation, the manufacturing process was defined giving the guidelines for the workshop organization and the definition of the required tooling. The critical manufacturing points were identified in the process and, regarding technical requirements, have led to the definition of a set of qualification mockups. They are related to helium inlets, conductor winding and insulation, local conductor bending, electrical joint and terminal areas for the winding pack (WP), as well as winding embedding, case welding, and impregnations for WP integration in the casing. The fabrication processes have been improved and shall be qualified thanks to the manufacture and testing of 12 corresponding mockups. The successful achievement of several key mock-ups gives confidence in the feasibility of the manufacture, and their completion will give the green light to the industrial coils manufacture.

In parallel to the direct contribution to the procurement phase of ITER and Broader Approach, CEA has initiated research & development programmes, accompanied by experiments together with a significant modelling effort, aimed at ensuring robust operation, plasma performance, as well as mitigating the risks of the procurement phase. This overview reports the latest progress in both fusion science and technology including many areas, namely the mitigation of superconducting magnet quenches, disruption-generated runaway electrons, edge-localized modes (ELMs), the development of imaging surveillance, and heating and current drive systems for steady-state operation. The WEST (W Environment for Steady-state Tokamaks) project, turning Tore Supra into an actively cooled W-divertor platform open to the ITER partners and industries, is presented.

Following the completion in May 2011 of the shutdown for the installation of the beryllium wall and the tungsten divertor, the first set of JET campaigns have addressed the investigation of the retention properties and the development of operational scenarios with the new plasma-facing materials. The large reduction in the carbon content (more than a factor ten) led to a much lower Zeff (1.2–1.4) during L- and H-mode plasmas, and radiation during the burn-through phase of the plasma initiation with the consequence that breakdown failures are almost absent. Gas balance experiments have shown that the fuel retention rate with the new wall is substantially reduced with respect to the C wall. The re-establishment of the baseline H-mode and hybrid scenarios compatible with the new wall has required an optimization of the control of metallic impurity sources and heat loads. Stable type-I ELMy H-mode regimes with H98,y2 close to 1 and βN ∼ 1.6 have been achieved using gas injection. ELM frequency is a key factor for the control of the metallic impurity accumulation. Pedestal temperatures tend to be lower with the new wall, leading to reduced confinement, but nitrogen seeding restores high pedestal temperatures and confinement. Compared with the carbon wall, major disruptions with the new wall show a lower radiated power and a slower current quench. The higher heat loads on Be wall plasma-facing components due to lower radiation made the routine use of massive gas injection for disruption mitigation essential.

Following the first conceptual design activity in which the general design of the JT-60SA TF system was defined in agreement with all the participants in the project (CEA, ENEA, F4E), a second phase dealing with the detailed design was engaged by each of the voluntary contributors. For the French part which includes the procurement of 9 of the TF winding packs and their integration in the casing, an industrial contract was signed mid 2011 with Alstom (France). Several actions have been carried out to prepare the manufacturing phase.To precisely define one of the main interfaces which is the temporary electrical connection of the coils to the current leads during cold test in the CEA facility as well as their final connection to the feeders at the Naka site, a design compatible with both requirements was developed by CEA, supported by the previous developments led on the joints and assembly techniques.In addition to prepare the coils manufacture, hydraulic qualification was led on the first conductor qualification length to set the parameters which will be used by the coils manufacturer for conductor acceptance.At last, mechanical characterizations of both the conductor and of the empty compacted jacket were performed in order to define as precisely as possible the elastic and plastic properties of these components. These are crucial properties used during the bending process which is one of the most important operations during the winding pack manufacture. These data will be very helpful for the winding machine parameters settings as well as for designing the local bending tooling needed to shape the conductors extremities at the connection area and at the double pancakes joggles. (C) 2013 Elsevier B.V. All rights reserved.

Tore Supra routinely addresses the physics and technology of very long-duration plasma discharges, thus bringing precious information on critical issues of long pulse operation of ITER. A new ITER relevant lower hybrid current drive (LHCD) launcher has allowed coupling to the plasma a power level of 2.7 MW for 78 s, corresponding to a power density close to the design value foreseen for an ITER LHCD system. In accordance with the expectations, long distance (10 cm) power coupling has been obtained. Successive stationary states of the plasma current pro le have been controlled in real-time featuring (i) control of sawteeth with varying plasma parameters, (ii) obtaining and sustaining a 'hot core' plasma regime, (iii) recovery from a voluntarily triggered deleterious magnetohydrodynamic regime. The scrape-off layer (SOL) parameters and power deposition have been documented during L-mode ramp-up phase, a crucial point for ITER before the X-point formation. Disruption mitigation studies have been conducted with massive gas injection, evidencing the difference between He and Ar and the possible role of the q = 2 surface in limiting the gas penetration. ICRF assisted wall conditioning in the presence of magnetic eld has been investigated, culminating in the demonstration that this conditioning scheme allows one to recover normal operation after disruptions. The effect of the magnetic eld ripple on the intrinsic plasma rotation has been studied, showing the competition between turbulent transport processes and ripple toroidal friction. During dedicated dimensionless experiments, the effect of varying the collisionality on turbulence wavenumber spectra has been documented, giving new insight into the turbulence mechanism. Turbulence measurements have also allowed quantitatively comparing experimental results with predictions by 5D gyrokinetic codes: numerical results simultaneously match the magnitude of effective heat diffusivity, rms values of density uctuations and wavenumber spectra. A clear correlation between electron temperature gradient and impurity transport in the very core of the plasma has been observed, strongly suggesting the existence of a threshold above which transport is dominated by turbulent electron modes. Dynamics of edge turbulent uctuations has been studied by correlating data from fast imaging cameras and Langmuir probes, yielding a coherent picture of transport processes involved in the SOL.

Following a first conceptual design activity in which the general design of the JT-60SA TF system was defined and frozen in agreement with all the participants in the project (CEA, ENEA, F4E), a second phase had to be launched to deal with the detailed design. In this paper, we present the work performed at CEA on the TF coil design during this second phase. Part of this work, concerns the determination of conductor hydraulic performances during operation as well as in factory. The thermohydraulic of the conductor was also assessed to confirm the need of helium inlets and a specific design was developed and qualified to be compatible with the available hydraulic performance of the cryoplant. The mechanical behavior is still to be assessed and qualified. Last but not least, the inner electrical joints of the coil have been modified with respect to the original twin-box design developed by CEA for the ITER coils in order to simplify the fabrication process. A dedicated qualification program for their manufacture is ongoing.

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.

For the superconducting tokamak JT-60SA, one of the major projects within the Broader Approach Agreement between Europe and Japan, CEA is in charge of the cryoplant system. Preliminary studies were performed in order to prepare the specifications. The presentation focuses on the modelling of two operating modes, the baking and the cool down, and on a study of the economics of the electrical power distribution of the warm compression station.Baking mode consists in heating the vacuum vessel to 473 K during several days to "clean" surfaces. Because of the increased radiation heat loads from the vacuum vessel on the thermal shields, the helium mass flow at 80K has to be increased by 50%. To avoid designing the cryogenic plant for this occasional overload, different scenarios were studied.During cool down mode, the mass flows have been calculated to ensure the required cool down speed. The limiting factors are the pressure drop in the coil windings and the magnet cool down speed.The warm compressor station represents a large part of the investment cost and is the main consumer of electrical power of the cryogenic system. Different compressors and motors have been analysed to optimise power consumption vs. investment costs.