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Port structure in the JT-60SA is designed to access vacuum vessel inside for vacuum exhausts, plasma heating systems and diagnostics etc. from cryostat outside. The port consists of a rectangular pipe, bellows and additional materials such as electrical insulations and connection parts for onsite assembly. The vacuum space as the thermal insulation for superconducting coils is penetrated by the port as the vacuum interfaces between the vacuum vessel and the cryostat so that port structure is required to be structurally integrated against plasma and baking operations and accidental events. This structure must be installed in narrow space, and electrical insulation is functionally designed to avoid the unexpected large eddy current loop. This paper reports the design of port connection structure for the assembly. Port connection parts “Port Extension” and “adjustable plate” are introduced to absorb vertical of 10 mm, toroidal of 8 mm, and port axial of 10 mm error due to the manufacturing and assembly. The bellows are expected to absorb relative displacement of 39 mm only between the vacuum vessel and the cryostat during the operation. The structure is confirmed to withstand the load conditions because stress is estimated to be lower than allowable level by FEM analysis.

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° 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° 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.

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.

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 has been designed and is being constructed to demonstrate and develop steady-state high-beta operation. Resistive wall mode (RWM) control, error field correction, and edge-localized mode (ELM) control will be performed using in-vessel coils. For these controls, we have developed a biaxial magnetic sensor to determine 3D magnetic configuration of the plasma. Moreover, for obtaining basic information about JT-60SA plasma, magnetic sensors, in particular, one-turn loops, Rogowski coils, diamagnetic loops, and saddle coils have been developed. Because the length of the vacuum vessel in the poloidal direction of JT-60SA is 16 m and almost twice as long as that of JT-60U, the length of the Rogowski coil and the diamagnetic loop of JT-60SA are also twice as long as those on JT-60U. We have devised new types of sensors and a connector for the mineral-insulated cable because construction and installation of these sensors are much more difficult in JT-60SA. We will report the design and specification of the magnetic sensors for JT-60SA from the physics and engineering aspects.

Disassembly of the JT-60U torus was started in 2009 after 18 years of D2 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 ∼41,000 person-day to workers was only ∼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 ∼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.

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.

Low- Z impurity ( 7 Be) on the JT-60U divertor tiles was analyzed to study the impurity behavior in the divertor region. The amount of the 7 Be increased approximately one hundred times after B 4 C-tile installation in the outer divertor. The 7 Be was probably produced by 10 B(p,α) 7 Be nuclear reaction on the divertor tiles in the hydrogen experiment with ion cyclotron range of frequency heating. The 7 Be was distributed asymmetrically in the poloidal and the toroidal direction. The highest 7 Be concentration was found at the inner divertor whose boron content (B/(B+C) ∼ 20%) was lower than the B 4 C tiles (B/(B+C) ∼ 80%) of the outer divertor. This result may imply impurity transport from the outer divertor to the inner divertor.

A centrifugal pellet injector and a guide tube for magnetic high-field side (HFS) injection were developed in JT-60U. The injector can eject trains of up to 40 cubic (1.9 mm) 3 deuterium pellets at frequencies of 1–10 Hz and velocities of 0.2–1.0 km/s. Pellet injection experiments into ohmic heating (OH) plasma were performed. From plasma density measurement, central fueling and enhanced fueling rate compared to gas puffing were observed for magnetic low-field side (LFS) injections. The results of HFS injection through the guide tube connected to the upper port of the vessel showed that the pellet velocity exceeding 220 m/s caused pellet destruction inside the tube.

In JT-60U, the modification to a W-shaped pumped divertor will be completed in May 1997, aiming to realize sufficient reduction in heat flux to the targets and good H-mode confinement simultaneously. W-shaped geometry is optimized not only for forming radiative divertor plasmas and reducing the back flow of neutral particles but also for allowing various experimental configurations. Toroidally and poloidally segmented divertor plates, dome and baffles are arranged in a W-shaped poloidal configuration. The pumping speed can be changed during a shot by variable shutter valves in the three pumping ports under the outer baffle. The net throughput is enough for particle control in the steady radiative operations with high power NBI heating. Carbon fiber composite (CFC) tiles are used for the divertor targets and the divertor throat where large heat flux is expected. Gaps between two adjacent segments are carefully sealed to suppress the leak of neutral gas from the exhaust duct below the divertor and baffles. The strength of the whole structure is confirmed by an electromagnetic force analysis and structural analysis carried out for disruptions of 3 MA discharges with a halo current.

B 4 C conversion-coated carbon fiber composites (B 4 C-converted CFC) have been installed on the outboard strike point of the divertor plate in JT-60U to form a complete toroidal ring. The plasma-surface interactions of the B 4 C-converted CFC have been investigated in high power divertor operation with neutral beam (NB) heating power of - 30 MW for 2 s. It is found that the carbon impurity is reduced without a significant increase of the boron impurity using a 100 μm thick B 4 C layer with a B /C ratio of 2.7. The oxygen impurity is reduced significantly by the evaporation of a 300 μm thick B 4 C layer with a B /C ratio of 3.8. A reduction in the detrimental effects on plasma-surface interactions with carbon-based materials has been demonstrated for more than 1000 shots by the use of surface-boronized carbon tiles in JT-60U

The mechanical fracture of three carbon fiber composite (CFC) first wall tiles was observed. This damage was probably caused by the electromagnetic force due to halo current during disruption. The required current to break the CFC tile is estimated to be 25 kA. The broken tile was rotated poloidally around the plasma with a speed of about 10 m/s during the following discharge. A possible driving force of this rotation might be the electromagnetic force due to the scrape-off layer (SOL) current. The required current to rotate the piece of the broken tile is 1 kA. These results indicate that electromagnetic interaction between SOL plasma and the plasma facing component is important in the research on the plasma wall interactions in fusion devices.

Essential divertor functions — density and energy control — have been investigated in the JT-60 divertor with metal walls. Neutral pressure in the divertor chamber increases in proportion to n2e and high recycling state for particle exhaust is realized. At ne ≤ 6 × 1019m−3, wall pumping by the divertor plates is dominant in particle exhaust. However, at ne ≥ 6 × 1019m−3, particle exhaust by active divertor pumping systems becomes essential. Because of the effectiveness of the divertor, radiation loss in the main plasma is reduced to 5–10% of the absorbed power and impurity concentrations are significantly suppressed at very low level (Zeff = 1.5-2.0). In the experiments with graphite walls, short periods of H-mode phases were found in the outside X-point divertor discharges although improvement in energy confinement is limited to within 10%. In limiter discharges, combination of high vessel temperature (210–300 °C) and low current limiter discharges (1.5 MA) without gas puffing is effective as a wall conditioning in obtaining reproducible high density discharges. The maximum parameters, Ip = 3.2 MA, ne = 1.2 × 1020m−3, qeff = 2.2 at Bt = 4.8T, a Murakami parameter of 7.5, and a stored energy of 3.1 MJ have been attained. The maximum density is limited by the occurence of the MARFE. During high power neutral beam heating, enhanced carbon influx was observed possibly due to chemical sputtering. The wall pumping by the inner wall is still effective in decreasing the electron density from the high density region of ne ~ 1 × 1020m−3. An empirical energy confinement scaling of JT-60 has been drawn in terms of an offset linear function as τE = 0.19 I1.9PPabs + 0.062 a1.8p