JT-60SA is the world's largest superconducting tokamak in operation jointly built and exploited by Europe and Japan in the framework of the Broader Approach. JT-60SA aims at addressing some of the technological and physics challenges, such as the long pulse steady-state plasma operation at high beta. The start-up of JT-60SA, which culminated in the first JT-60SA plasma achieved on 23 October 2023 and Operation-1 (OP-1) until the end of 2023, including the achievement of >1 MA diverted plasmas, paves the way for a new generation of large superconducting tokamaks, such as ITER. Several key scientific topics were investigated during this initial phase. Similarly to ITER, the available parallel electric field (E-||) is low and yet plasma initiation was quickly obtained by means of the trapped particle configuration (TPC) with the assistance of similar to 1.5 MW of electron cyclotron resonance heating (ECRH). A first analysis and classification of the causes for disruptions have been done after the results of OP-1. Vertical displacement events (VDEs) were responsible for the vast majority of disruptions in increasing elongated plasmas, as the stabilization plate was not yet installed in this phase. Therefore, VDE predictors and control algorithms were developed using machine learning techniques with magnetics probe data, showing that these novel techniques are also suitable for the start-up tokamak phases characterized by scarce input data. JT-60SA will restart operation in 2026 following a series of upgrades. The experimental programme for future operations is guided by significant modelling 'predict first' activity, which shows that access to and development of H-mode in conditions of future burning plasmas will be possible with high negative neutral beam injection (N-NBI) and ECRH input power. The integration of such elements into a steady-state long pulse operation will be done with the installation of W plasma facing components (PFC) after the initial campaigns.

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.

The JT-60SA superconducting tokamak is being constructed under dual projects, the EU-JA international collaboration framework and Japanese national programme. To suppress magnetic field error to less than 0.01% with respect to the toroidal magnetic field for good confinement of plasma, QST has developed a precise onsite assembly technology of tokamak components such as vacuum vessel (VV) and toroidal magnetic field coils (TFC) with size of over 10-m scale. Prior to an onsite assembly, the assembly process is carefully simulated in a three dimension CAD model, through which the location and number of the reference points for the assembly are evaluated to determine each component position. In the onsite assembly, the tokamak components are positioned by measuring with the laser tracker of a 0.5 mm spatial resolution and assembled by adjusting with shims and splice plates. The onsite assembly of the VV 340 degrees sector has been succeeded in an allowable accuracy of +8/-4 mm with the 10 m diameter by welding sectors, each of which was circumferentially segmented and manufactured by taking into account the deformation due to the welding. The remaining 20 degrees sector will be installed after the TF coils installation. The metrology developed for the onsite assembly of the JT-60SA is expected to be applicable to ITER whose size is the double of JT-60SA.

The disassembly of JT-60 tokamak device and its peripheral equipments, where the total weight was about 5400 tons, started in 2009 and accomplished in October 2012. This disassembly was required process for JT-60SA project, which is the Satellite Tokamak project under Japan-EU international corroboration to modify the JT-60 to the superconducting tokamak. This work was the first experience of disassembling a large radioactive fusion device based on Radiation Hazard Prevention Act in Japan. The cutting was one of the main problems in this disassembly, such as to cut the welded parts together with toroidal field coils, and to cut the vacuum vessel into two. After solving these problems, the disassembly completed without disaster and accident. This report presents the outline of the JT-60 disassembly, especially tokamak device and ancillary facilities for tokamak device. (author)

Manufacturing of JT-60SA Vacuum Vessel (VV) is under the onsite sector assembly, where ten VV sectors are to be welded into a torus. The vessel is described as 10 m diameter, 6.6 m height and a double wall structure with type 316L stainless steel of 18 mm thickness and locally 30 mm, and total dimensional tolerance is designed to avoid any interference due to operational displacements. This report introduces two issues such as sector manufacturing and onsite assembly. The sector manufacturing is accomplished in high precision, and the assembly concept is obtained with combination of the direct butt joint and splice plate joint. Moreover, status of the onsite sector assembly are presented.

Disassembly of the JT-60U torus was started in 2009 after 18 years of D-2 operations and was completed in October 2012 for assembling the JT-60SA torus at the same position. The JT-60U torus was featured by the complicated and welded structure against the strong electromagnetic force, and by the radioactivation due to deuterium-deuterium (D-D) reactions. Since this work is the first experience of disassembling a large radioactivated fusion device in Japan, careful preparations of disassembly activities, including treatment of the radioactivated materials and safety work, have been made. During the disassembly period over 3 years, careful measures against exposure were taken and stringent control of exposure dose were implemented, and as a result, accumulated collective effective dose of similar to 41,000 person-day to workers was only similar to 22 mSv in total and no internal exposure was observed. About 13,000 components cut into pieces with measuring the contact dose were removed from the torus hall and stored safely in storage facilities. The total weight of the disassembly components reached up to similar to 5400 tonnes. Most of the disassembly components will be treated as non-radioactive ones after the clearance level inspection under the Japanese regulations in the future. The assembly of JT-60SA has started in January 2013 after this disassembly of JT-60U torus. (C) 2014 Elsevier B.V. All rights reserved.

Fuel retention rates and carbon re-deposition rates in the plasma shadowed areas, or tile gaps and remote areas, in JT-60U were measured. The total fuel retention rate in the plasma shadowed areas was more than two times higher than that in the carbon re-deposited layers on the plasma facing surfaces, or the divertor tiles. This is because of lower temperature in the plasma shadowed areas than in the plasma facing surfaces, which leads to high hydrogen saturation concentration, although the amount of the carbon re-deposited on the plasma shadowed areas was only 60% of that on the plasma facing surfaces. The total fuel retention rate in JT-60U, including previously determined for all the plasma facing areas, was evaluated to be 1.3 × 1020 H + D s−1, and this retention rate was lower than that in the other devices, due probably to high baking temperature operation in JT-60U. Distributions of the fuel retention and the carbon re-deposition in the whole in-vessel of a large tokamak were determined for the first time in the world.

Retention and depth profile of all hydrogen isotopes (hydrogen, deuterium and tritium) in the carbon tiles of outer divertor, outboard baffle plate and outboard first wall in JT-60U, which were mostly eroded, were measured by TDS, SIMS and TIPT, and discussed from view points of tile temperature, incident energy and flux of impinging H, D and T. The distribution of T in the analyzed tiles was compared with the impinging flux of tritium produced by DD reactions, which was calculated by the OFMC code. In this paper, H, D and T retention mechanisms and isotopic replacement effects are summarized.

T retention and its depth profile in the graphite tiles used for first wall of JT-60U have been measured by a tritium imaging plate technique and a full combustion method. T was found only limited depth beneath the plasma facing surface and little in both the surface region shallow than 1 mu m and in bulk more than 1 mm in depth. Although most of T produced by DD reactions are thermalized and neutralized in plasma and impinge on the plasma facing surface and penetrate into the inner surface, they are isotopically replaced by subsequently incoming D. Only some of high energy T escaping from plasma are directly implanted beneath the surface and retained escaping from the isotopic replacement until attainment of a saturation concentration.

Supersonic molecular beam injection (SMBI) was successfully operated in JT-60U after improving the performance of the vacuum seal used inside the injector head. Frequent density jumps were clearly observed in the main plasma against the SMBI pulses with filling gas pressures (P FG) of 0.2–0.6 MPa. The fuelling efficiency exhibited a weak dependence on P FG and the injection direction (high- and low-field-side injections). The amount of fuelling necessary for achieving the same density level is much smaller for SMBI than for gas puffing. It is comparable for SMBI and pellet injection even with shallower penetration of SMBI as discussed. The SMBI ionization area was estimated based on emission measured using a fast TV camera with a time resolution of 0.167 ms. The estimations indicated a similar penetration position for P FG = 0.6 and 0.2 MPa, although the ionization area was larger for 0.6 MPa. This result supports the weak P FG dependence of the fuelling efficiency. The front of the ionization area moved between the first and second frames of the fast TV camera and it reached just inside the separatrix in the second frame. The ionization area was significantly expanded even in the first frame from the expected SMB size and the expansion was enhanced in the second frame. These relatively slow changes between the two frames suggest that the interaction between the SMB and the plasma significantly influences the fuelling characteristics.

The characteristics of the internal transport barrier (ITB) have been investigated under reactor relevant conditions with edge fuelling and electron heating in JT-60U weak shear plasmas. In order to investigate the effects of edge fuelling and electron heating separately, two independent classes of experiments were performed, i.e. one with edge fuelling and ion dominant heating and the other with central beam fuelling and additional electron heating. High confinement was sustained at high density with edge fuelling by shallow pellet injection or supersonic molecular beam injection. The ion temperature (T i) in the central region inside the ITB decreased due to cold pulse propagation even with edge fuelling. By optimizing the injection frequency and the penetration depth, the decreased central T i recovered and a good ITB was sustained with enhanced pedestal pressure. The T i-ITB also degraded significantly with electron cyclotron heating (ECH), when the stiffness feature was strong in the electron temperature (T e) profile. The ion thermal diffusivity in the ITB region increased with the electron thermal diffusivity, indicating the existence of a clear relation between ion and electron thermal transport. On the other hand, the T i-ITB remained unchanged or even grew, when the stiffness feature was weak in the T e profile. The density fluctuation level at the ITB seemed unchanged during ECH; however, the correlation length became longer in the T i-ITB degradation case and shorter in the T i-ITB unchanging case.

The tokamak machine monitoring and control system has been developed to avoid machine damages in JT-60 safety operation. It monitors the JT-60 mechanical structures and other sub-components to maintain their conditions. The system adopts the CAMAC digitizers, the Signal Transfer Unit (STU), and the hard-wired interlocks. Over 20 year, it has never been affected by any major troubles except a water leakage in toroidal field coil (TFC). The TFC monitoring system has been developed for the TFC safety with the water leakage from a cooling water channel. It has improved the reliability and efficiency of the TFC operation.

Characteristics of internal transport barrier (ITB) have been investigated under reactor relevant condition with edge fuelling and electron heating in JT-60U weak shear plasmas. High confinement was sustained at high density with edge fuelling by shallow pellet injection or supersonic molecular beam injection (SMBI). The ion temperature (Ti) in the central region decreased even with edge fuelling. The Ti decrease with edge fuelling was larger inside the ITB than that outside the ITB, which can be described by cold pulse propagation using the ion thermal diffusivity (χi) estimated from power balance analysis in the SMBI case. By optimizing the injection frequency and the penetration depth, the decreased Ti was recovered and good ITB was sustained with enhanced pedestal pressure. The Ti-ITB also degraded significantly when stiffness feature was strong in the electron temperature (Te) profile against electron cyclotron heating (ECH). The value of χi in the ITB region increased with the electron thermal diffusivity (χe), indicating existence of clear relation between ion and electron thermal transport. On the other hand, Ti-ITB unchanged or even grew, when stiffness feature was weak in the Te profile. Density fluctuation level seemed to be unchanged during ECH, however, correlation length became longer in the Ti-ITB degradation case and shorter in the Ti-ITB unchanging case.

Hydrogen isotope distributions of the outer divertor target CFC tile, in the W-shaped divertor of JT-60U where erosion dominated, were analyzed by secondary ion mass spectroscopy (SIMS) to study the correlation between the concentrations of hydrogen isotopes and erosion/deposition rate. The chemical states of the CFC tile surfaces were analyzed by X-ray photoelectron spectroscopy (XPS). The existence of re-deposition layers was observed by scanning electron microscope (SEM). The erosion/deposition depth was also evaluated by a surface profilometer, namely dial gauge. The hydrogen and deuterium retention profiles were compared with the tritium pro. le obtained by the tritium imaging plate technique (TIPT). It was found that almost all the deuterium once retained in the near surface region of the erosion dominated area was replaced by hydrogen during the final HH discharge and/or exposure to the atmosphere. The hydrogen retention pro. le was controlled by surface temperature and plasma flux, and the total hydrogen retention was much less than that in the re-deposition layers observed at the plasma shadowed area in this particular tile.

Low-Z impurity (7Be) on the JT-60U divertor tiles was analyzed to study the impurity behavior in the divertor region. The amount of the 7Be increased approximately one hundred times after B4C-tile installation in the outer divertor. The 7Be was probably produced by 10B(p,α)7Be nuclear reaction on the divertor tiles in the hydrogen experiment with ion cyclotron range of frequency heating. The 7Be was distributed asymmetrically in the poloidal and the toroidal direction. The highest 7Be concentration was found at the inner divertor whose boron content (B/(B+C)∼20%) was lower than the B4C tiles (B/(B+C)∼80%) of the outer divertor. This result may imply impurity transport from the outer divertor to the inner divertor.

Modification of JT-60 as a full superconducting tokamak (JT-60SC) is planned. The objectives of the JT-60SC programme are to establish scientific and technological bases for steady-state operation of high performance plasmas and utilization of reduced-activation materials in an economically and environmentally attractive DEMO reactor. Advanced fusion technologies relevant to the DEMO reactor have been developed for the superconducting magnet technology and plasma facing components of the JT-60SC design. To achieve a high current density in a superconducting strand, Nb3Al strands with a high copper ratio of 4 have been newly developed for the toroidal field coils (TFCs) of JT-60SC. The R&D to demonstrate the applicability of the Nb3Al conductor to TFCs by a react-and-wind technique has been carried out using a full-size Nb3Al conductor. A full-size NbTi conductor with low ac loss using Ni-coated strands has been successfully developed. A forced cooling divertor component with high heat transfer using screw tubes has been developed for the first time. The heat removal performance of the carbon fibre composite target was successfully demonstrated on an electron beam irradiation stand.

Recent JT-60U results leading to high integrated performance are reported with emphasis on the projection to the reactor-relevant regime. Negative-ion-based neutral beam (NB) and electron cyclotron (EC power increased up to 6.2 MW and 3 MW, respectively. A high beta(p) H-mode plasma with full non-inductive current drive has been obtained at 1.8 MA, and the fusion triple product reached 3.1 x 10(20) m(-3) keV s. A high beta with beta(N) = 2.7 was maintained for 7.4 s. Neoclassical tearing mode suppression with EC was accomplished using a real-time feedback control system, and improvement in beta(N) was obtained. The stable existence of a current hole was observed. A high DT-equivalent fusion gain of 0.8 was maintained for 0.55 s in a plasma with a current hole. Current profile control in high bootstrap current reversed shear plasmas was demonstrated using N-NB and LH. A new operation scenario has been established in which a plasma with a high bootstrap current fraction and internal transport barriers (ITBs) is produced without the use of an OH coil. An ECCD study was undertaken in a reactor-relevant high T-e regime. A new type of Alfven eigenmode mode has been proposed and has been found to explain the observed frequency chirp quite well. High confinement reversed shear plasmas with T-e > T-i were obtained. Ar exhaust with EC heating was obtained in a high beta(p) mode plasma. Impurity accumulation related to strong ITBs in a reversed shear plasma and degradation of ITB by ECH in a weak positive shear plasma have been found. Dedicated measurement of edge localized mode dynamics and scrape-off-layer plasma flow have advanced understanding of the physics. N-NB heating in an Ar-seed plasma extended the density region to 95% of the Greenwald density, with HHy2 = 0.9. Enhancement of pedestal pressure was obtained with an increase in beta(p) in a high triangularity configuration.

Detailed tritium profiles on the JAERI Tokamak-60U (JT-60U) W-shaped divertor and the first wall tiles were examined by a tritium imaging plate technique (TIPT) and full combustion method. The highest tritium level (60 kBq/cm2) was observed at the dome top tiles. The tritium level of the divertor target was lower (2 kBq/cm2). The result of the triton deposition simulation using orbit following Monte-Carlo code was consistent with the tritium distribution obtained by TIPT and full combustion method. These results indicate that the tritium distribution of the JT-60U W-shaped divertor reflects mainly the distribution of the energetic triton impinging on the wall. According to the simulation, the tritium atoms produced by D–D nuclear reaction in JT-60U are not loosing completely their initial energy of 1 MeV and around 1/3 of them are implanted into the wall.

Designs and operations of the gas system and pellet injection systems for JT-60 and JT-60U are described. A gas injection valve that is a key component of the gas injection system was developed using a multilayer piezo-electric element. The maximum flow rate of this system is 43.3 Pa (.) m(3)/s. The valve has mechanism for adjustment at atmospheric side meaning that a repair and an adjustment can be conducted without ventilation inside a vacuum vessel. It was confirmed that the effect of magnetic field and temperature change on the valves in the JT-60U environment was negligible. In JT-60U, two systems of pellet injector-a pneumatic drive and a centrifugal one-were developed. The pneumatic type attained a pellet velocity of 2.3 km/s, which was the world record at the time in 1988. On the other hand, the centrifugal one was developed in 1998. This injector can eject trains of up to 40 cubic (2.1 mm(3)) pellets at frequencies of 1 to 10 Hz and speed of 0.1 to 1.0 km/s. A guide tube for a magnetic high field side injection (HFS) (top) was also developed in 1999. The pellet injection experiment with the HFS system started in 2000. In addition, another guide tube for HFS(mid) injection was newly developed and installed in March 2001. These systems are working well.

The modification of JT-60 is planned as a fully superconducting tokamak (JT-60SC). The mission of JT-60SC program is to establish scientific and technological bases for an advanced operation in an economically and environmentally attractive DEMO reactor and ITER. The research objectives are to accomplish high performance steady state operation with high beta and non-inductive full current drive, with high bootstrap current fraction, and demonstrate the plasma applicability of reduced-activation material for a plasma of break-even class relevant to the reactor plasma. Basic design of JT-60SC has been completed and the detailed design is under way. The engineering design for main components of JT-60SC is described.