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Remote pipe welding tool accessing from inside of the pipe has been newly developed for JT-60SA. Remote handling (RH) system is necessary for the maintenance and repair of the divertor cassette in JT-60SA. Because the space around the cooling pipe connected with the divertor cassette is very limited, the cooling pipe is to be remotely cut and welded from inside for the maintenance. A laser welding method was employed to perform circumferential welding by rotating the focusing mirror inside the pipe. However, the grooves of connection pipes are not precisely aligned for welding. The most critical issue is therefore to develop a reliable welding tool for pipe connection without a defect such as undercut weld due to a gap caused by angular and axial misalignments of grooves. In addition, an angular misalignment between two pipes due to inclination of pipe has to be taken into account for the positioning of the laser beam during welding. In this paper, the followings are proposed to solve the above issues: (1) Cooling pipe connected with the divertor is machined to have a jut on the edge so as to expand the acceptable welding gap up to 0.5 mm by filling the gap with welded jut. (2) Positioning accuracy of the laser beam for reliable welding is realized to be less than ±0.1 mm along the circumferential target for welding by a position control mechanism provided in the tool even in the case of up to angular misalignment of 0.5° between connection pipes. Based on the above proposals, we have achieved robust welding for a large gap up to 0.5 mm as well as the maximum angular misalignment of 0.5° between connection pipes by using this newly developed tool.

This paper focuses on one of the JT-60SA vacuum vessel manufacturing R&D, onsite welding technology of the port joint. The work space is limited inside the vessel, and manipulator application is examined through the simulation and the mock-up trial. As a result, the most difficult port of the upper vertical is successfully weld-jointed. The quality as the product joint and the controllability of the manipulator are assured and perspectives to the other ordinary port joints are discussed with issues gained from this R&D.

The assembly scenarios and assembly tools of the major tokamak components for JT-60SA are studied in the following. (1) The assembly frame (with a dedicated 30-tonne crane), which is located around the JT-60SA tokamak, is adopted for effective assembly works in the torus hall and the temporary support of the components during assembly. (2) Metrology for precise positioning of the components is also studied by defining the metrology points on the components. (3) The sector segmentation for weld joints and positioning of the vacuum vessel (VV), the assembly scenario and tools for VV thermal shield (TS), the connection of the outer intercoil structure (OIS) and the installation of the final toroidal field coil (TFC) are studied, as typical examples of the assembly scenarios and tools for JT-60SA.

The JT-60SA vacuum vessel (VV) has a D-shaped poloidal cross section and a toroidal configuration with 10° segmented facets. A double wall structure is adopted to ensure high rigidity at operational load and high toroidal one-turn resistance. The material is 316L stainless steel with low cobalt content ( The manufacturing of the VV started in November 2009 after a fundamental welding R&D and a trial manufacturing of 20° upper half mock-up. The manufacturing of the first VV 40° sector was completed in May 2011. A basic concept and required jigs of the VV assembly were studied. This paper describes the design and manufacturing of the vacuum vessel. A plan of VV assembly in torus hall is also presented.

JT-60 is planned to be upgraded to a JT-60SA superconducting tokamak machine. This project is the JA-EU satellite tokamak program under both broader approach and Japanese domestic programs. The JT-60SA tokamak is composed of the following main components: vacuum vessel (VV), thermal shield, superconducting coils (toroidal field coil, equilibrium field coil, and central solenoid), cryostat, and heating facilities. The VV has a D-shaped poloidal cross section and a double-wall structure to ensure high rigidity and toroidal one-turn resistance simultaneously. The material of the VV is 316L stainless steel with a low cobalt content of

Construction has just begun in the JT-60 Super Advanced (JT-60SA) tokamak project, which is part of the Japanese domestic program and the Japan – EU international program “ITER Broader Approach”. All plasma facing components (PFCs) will be actively cooled because they are exposed to high heat flux for long duration. A lower single-null closed divertor with a vertical target (VT) will be installed when experiments begin. A divertor cassette integrated with coolant pipe connections will be used for remote handling maintenance. Each cassette covers a 10° sector in the toroidal direction. Modularized PFCs such as VTs, baffles, and domes will be assembled on divertor cassettes. Static structural analysis shows that the displacement and stress in the divertor module are generally small for determining the dead weight, coolant pressure, and electromagnetic forces (EMFs). However, the support structure of the outer baffle requires improvement because eddy currents generate concentrated overturning EMFs.

The ITER blanket (BL) is composed of about 400 modules in the vacuum vessel (VV). The BL has to be maintained by remote handling means due to high radiation levels in the VV after D-T operation. The remote handling (RH) equipment for BL maintenance consists of articulated rail, supports, a rail-mounted vehicle, a telescopic arm, an end-effecter, tools and related systems such as transfer casks and umbilical system. Towards the construction, the BL RH equipment design has been improved and developed in more detail, based on the 2001 FDR design. The overview of design results is introduced in this paper. The design of rail deployment system of the BL RH has been updated to enable the rail connection in the transfer cask in order to minimize occupation space at storage area. For this purpose, design work has been performed for concept, sequence and typical simulation of BL replacement in the VV and rail deployment/storage of the RH equipment in the cask, including cask docking. In particular, the technical issues of the rail connection in the cask are (1) tight tolerance of a pin at a hinge, (2) limited space for the connection inside a cask and (3) tight positioning accuracy. This paper summarizes the idea to solve these issues and the results of the design work. The paper also introduces new cable handling equipment, rail support equipment and BL module transporter.

The maintenance operation of the ITER in-vessel component, such as a blanket and divertor, must be executed by the remote equipment because of the high gamma-ray environment. During the Engineering Design Activity (EDA), the Japan Atomic Energy Agency (then called as Japan Atomic Energy Research Institute) had been fabricated the prototype of the vehicle manipulator system for the blanket remote handling and confirmed feasibility of this system including automatic positioning of the blanket and rail deployment procedure of the articulated rail. The ITER agreement, which entered into force in the last year, formally decided that Japan will procure the blanket remote handling system and the JAEA, as the Japanese Domestic Agency, is continuing several R&Ds so that the system can be procured smoothly. The residual key issues after the EDA are rail connection and cable handling. The mock-ups of the rail connection mechanism and the cable handling system were fabricated from the last year and installed at the JAEA Naka Site in this March. The former was composed of the rail connecting mechanism, two rail segments and their handling systems. The latter one utilized a slip ring, which implemented 80 lines for power and 208 lines for signal, because there is an electrical contact between the rotating spool and the fixed base. The basic function of these systems was confirmed through the mock-up test. The rail connection mechanism, for example, could accept misalignment of 1.5–2 mm at least. The future test plan is also mentioned in the paper.

The JT-60 Super Advanced (JT-60SA) tokamak project starts under both the Japanese domestic program and the international program “Broader Approach”. The maximum heat flux to JT-60SA divertor is estimated to ∼15 MW/m 2 for 100 s. Japan Atomic Energy Agency (JAEA) has developed a divertor armor facing high heat flux in the engineering R&D for ITER, and it is concluded that monoblock-type CFC divertor armor is promising for JT-60SA. The JT-60SA armor consists of CFC monoblocks, a cooling CuCrZr screw-tube, and a thin oxygen-free high conductivity copper (OFHC-Cu) buffer layer between the CFC monoblock and the screw-tube. CFC/OFHC-Cu and OFHC-Cu/CuCrZr joints are essential for the armor, and these interfaces are brazed. Needed improvements from ITER engineering R&D are good CFC/OFHC-Cu and OFHC-Cu/CuCrZr interfaces and suppression of CFC cracking. For these purposes, metalization inside CFC monoblock is applied, and we confirmed again that the mock-up has heat removal capability in excess of ITER requirement. For optimization of the fabrication method and understanding of the production yield, the mock-ups corresponding to quantity produced in one furnace at the same time is also produced, and the half of the mock-ups could remove 15 MW/m 2 as required. This paper summarizes the recent progress of design and mock-up test results for JT-60SA divertor armor.

The authors developed a simulator for the remote maintenance system of the ITER blanket using a general 3D robotic simulation software, ENVISION. The simulator is connected to the control system of the manipulator, which was developed as part of the blanket maintenance system during the Engineering Design Activity (EDA), and can reconstruct the positions of the manipulator and blanket module using position data transmitted from motors through a LAN. In addition, it can provide virtual visual information (e.g., about the interface structures behind the blanket module) by making the module transparent on the screen. It can also be used for confirming a maintenance sequence before the actual operation. The simulator will be modified further, with addition of other necessary functions, and will finally serve as a prototype of the actual simulator for the blanket remote handling system, which will be procured as part of an in-kind contribution.

Demonstration tests for manufacturing and assembly of the International Thermonuclear Experimental Reactor (ITER) vacuum vessel have been conducted to confirm manufacturing and assembly process of the vacuum vessel (VV). The full-scale partial mock-up fabrication was planned and is in progress. The results will be available in the near future. Field-joint assembly procedure has been demonstrated using a test stand. Due to limited accessibility to the outer shell at the field joint, some operations, including alignment of the splice plates, field-joint welding, and examination, were found to be very difficult. In addition, a demonstration test on the selected back-seal structures was performed. It was found that the tested structures have insufficient sealing capabilities and need further improvement. The applicability of ultrasonic testing methods has been investigated. Although side drilled holes of 2.4 mm in diameter were detected, detection of the slit-type defects and defect characterization were found to be difficult. Feasibility test of liquid penetrant testing has revealed that the selected liquid penetrant testing (LPT) solutions have sufficient low outgas rates and are applicable to the VV.

International audience; The JT-60SA experiment is one of the three projects to be undertaken in Japan as part of the Broader Approach Agreement, conducted jointly by Europe and Japan, and complementing the construction of ITER in Europe. The superconducting magnet system for JT-60SA consists of 18 Toroidal Field (TF) coils, a Central Solenoid (CS) and six Equilibrium Field (EF) coils. The TF magnet generates the field to confine charged particles in the plasma, the CS provides the inductive flux to ramp up plasma current and contribute to plasma shaping and the EF coils provide the position equilibrium of plasma current and the plasma vertical stability. The six EF coils are attached to the TF coil cases through supports with flexible plates allowing radial displacements. The CS assembly is supported from the bottom of the TF coils through its pre-load structure. The design status of the JT-60SA superconducting magnetic system is reviewed.

The paper provides an overview of the ITER transfer cask system design. The system is required during the assembly and maintenance periods of ITER. The system is designed with safety as the primary goal, to ensure minimum exposure risk to personnel and to allow the reliable transfer of components between the machine and the hot cell.

In International Thermonuclear Experimental Reactor (ITER), blanket maintenance requires the 4-tonne module handling with high positioning accuracy of ±2 mm. In order to meet this requirement, it is essential to suppress the dynamic deflection and vibration of the remote handling equipment due to sudden transfer of the module weight from/to the back-plate supports to/from the equipment itself during installation and removal. A new control scheme was proposed and tested so as to suppress the dynamic behaviors. As a result, the dynamic deflection of the rail and the acceleration of the manipulator were sucesessfully decreased to nearly zero. Based on the test results, the proposed control scheme was concluded to be effective so as to suppress this kind of dynamic effect during heavy component handling.

A periscope-type viewing system has been chosen as a reference of the ITER in-vessel viewing system due to its wide viewing capability. According to the ITER research and development program, development of a radiation hard periscope has been conducted by the Japan Home Team in collaboration with the JCT. As an intermediate stage, a sub-scaled radiation hard periscope with a length of 6 m was fabricated and irradiated at a dose rate about 10 kGy h –1 . In this periscope, three types of radiation hard lenses made of alkaline barium glass, lead glass containing CeO 2 and OH doped synthetic quartz were adopted on the basis of the irradiation experiments on various glasses. The irradiation test results of the sub-scaled periscope show no degradation of the viewing performance up to the accumulated dose of 50 MGy, while a standard-type (non-radiation hardness) periscope becomes invisible after 2 h irradiation. Based on this, a full-scale radiation hard periscope with a length of 15 m was fabricated and tested. As a result, it has been verified that the developed 15-m-long periscope has sufficient viewing capability and the focal adjusting mechanisms are capable of accommodating thermal expansion due to high temperature operation up to 250°C.

In the International Thermonuclear Experimental Reactor (ITER), remote handling of the blanket is a key issue since scheduled maintenance is foreseen, including a complete replacement of the shielding blanket by a breeding blanket. According to the ITER R&D program, a rail-mounted vehicle-type remote handling (RH) system has been developed, and its applicability to blanket maintenance has been demonstrated through fabrication and testing of prototypes. Based on this, a full-scale blanket RH test platform is being fabricated in order to demonstrate the remote replacement of a full-scale blanket module of about 4 tons dead weight. The test platform is basically composed of a vehicle manipulator operating on a toroidal rail, for handling a blanket module in a 180° in-vessel region, a receiver for transporting a module from the manipulator to a transfer cask for delivery to the hot cell, rail deployment and supports. These RH equipment / tools were due to be completed by the end of June 1997 and, thereafter, assembled into the blanket RH test platform. This paper describes the progress and outlines the main RH equipment / tools for the blanket RH test platform.

JT-60SA is a fully superconducting coil tokamak upgraded from the JT-60U, and this paper focuses on the vacuum vessel (VV, 150 tons) design concept with the related technology of welding and its manufacturing status. The design concept is developed from the ASME Boiler and Pressure Vessel Code Section VIII Division 2, and the damage tolerant concept is adopted into the welding part hardly inspected. These weld joints are treated as a partial penetration and residual strength is evaluated in the design process. Consequently, accuracy of VV weld structure is enhanced due to reduction of the welding heat input and also less process of weld work is achieved. The vacuum vessel has been manufactured since Nov. 2009. Welding technology related to the VV is surveyed and developed in three stages; element test, 1-m size trail and 20-degree upper half mock-up manufacture. The welding methodologies are selected from element tests, and then the typical vessel segments as a 1-m size are tested. The welding conditions to control its quality and the distortion are evaluated, and the lower heat input condition is additionally researched. On the basis of these results, the manufacturing sequences are established through the large size trial of the mock-up manufacturing.

The time has come for fusion experimental reactor programmes (ITER/FER) to enter the engineering design phase, where the large scale models which are capable of extrapolation to the construction of reactors will be developed as the core of the main activities. A brief review of the present status of the required R&D for experimental reactors, and the technological realization perspectives, are described.

Recent lower hybrid current drive (LHCD), and heating (LHH) experiments on JT-60 are reported. The current drive product of neRpIRF approximately 12.5*1019 m-2 MA was achieved at the LH power of approximately 4.5 MW, and the CD efficiency, the energy confinement, the global power balance and the heat load on divertor plates were investigated in high power LHCD plasmas. Nearly steady state H-mode discharges were found during LHCD with two different frequency injections. Sawtooth suppression in NB heated plasmas by LHCD have shown an improvement in confinement near the plasma center. Parametric instabilities in LH heating experiments were significantly reduced by increasing the plasma current, and the stored energy increased linearly with heating power of up to approximately 9 MW at ne approximately 7*1019 m-3 and Ip=2.75 MA. Parametric instabilities near the plasma edge in the ion heating regime were also reduced in peaked density plasmas produced by pellet injection and LH waves increased the central plasma pressure at ne(0) > 1.4*1020 m-3.

The Fusion Experimental Reactor (FER) is a D-T-burning tokamak machine currently being designed. It is expected to provide physical information and technical experiences that will be sufficient to proceed towards the DEMO Fusion Reactor which will demonstrate electric power generation by fusion energy. An efficient ash exhaust, a hybrid current drive operation, the use of a 3% ripple field, the technological achievements in R&D of the magnets, and the negative-ion beam system are expected to allow the FER to achieve its cost-effectiveness. >

This paper proposes a force control scheme for a remote maintenance robot for nuclear fusion reactors. In the International Thermonuclear Experimental Reactor (ITER), a remote maintenance system which can perform the assembly of heavy objects with high positional accuracy is required. For this kind of task, force control is much more preferable than position control. We applied force control to an experimental blanket-handling system and performed several experiments on the assembly of a dummy blanket. The experimental results illustrate the validity of the force control for the task.

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

Remote welding/cutting of cooling pipes for divertor plates, which is necessary for in-vessel replacement of the plates in the Fusion Experimental Reactor(FER), was discussed. A concept of internal-access pipe welding/cutting was proposed to save the access space around the pipes. Welding and cutting experiments using a 10kW class output carbon-dioxide laser beam were performed with 200A(sch.40, 8.2mm thick) stainless steel pipes. As the result, fully penetrated welding and completed cutting were demonstrated, so that the feasibility of internal-access pipe welding/cutting concept for divertor cooling pipes was confirmed.

Diamond-like carbon (DLC) coating was tested as a candidate solid lubricant for transmission gears of the maintenance equipment of the blanket of the ITER instead of an oil lubricant. The wear tests using the pin-on-disk method were performed on disks with SCM440 and SNCM420 as the base materials and coated with soft, layered, and hard DLCs. All cases satisfied the required allowable contact stress (2 GPa) and lifetime (104 cycles), and therefore the feasibility of the DLC coating was validated. Among the three types of DLCs, the soft DLC showed the best performance.

In the last year, ohmic heating experiments were performed in JT-60 for about three months; May–June 1985 and March 1986. The major experimental effort has been concentrated on the divertor configuration in anticipation of H-mode operation of JT-60 in high power supplementary heating experiments starting in August 1986. Stable discharges with a divertor have been obtained at Ip = 1.85 MA and ne = 5.7 × 1019m−3. Radiated power from the main plasma in divertor discharges is reduced to one third of that in material limiter discharges. The electron density has a broad radial profile and the effective safety reaches qeff = 2.5. The neutral gas pressure in the divertor chamber becomes significantly high and the control of particle recycling has been attempted.

Emphases in JT-60 experiments are placed on (1) lower-hybrid (LH) current drive characteristics with a multi-junction type launcher, and (2) the confinement study with combination of neutral beam injection LH current drive and pellet injection. The new multi-junction LH launcher provides a sharp N/sub /// spectrum with high directivity for N/sub ///=1-3.4. The current drive efficiency and the radial distribution of high energy electron production show clear correlation with injected N/sub ///: the current drive efficiency has the maximum at low N/sub ///( approximately 1.3) while flattening of plasma current is more effective in large N/sub ///. A broad radial distribution of high energy electron current and approximately 30% reduction in sawtooth inversion radius were obtained by high N/sub /// ( approximately 2.5) LH injection. To fully suppress the sawtooth activity, low N/sub /// ( approximately 1.3) injection was found to be more effective. Improved energy confinement has been obtained with hydrogen pellet injection. Energy confinement time was enhanced up to 40% relative to usual gas fuelled discharges. The discharge has a strongly peaked electron density profile with ne(0)/(ne) approximately 5 and ne(0) approximately 2.0*1020 m-3. The improved discharges are characterized by a strongly peaked pressure profile within the q=1 magnetic surface, and degrades when a large sawtooth recovers or the pressure gradient may reach a critical value. When large (3 mm, 4 mm) and fast (2.2 km/s) pellets were injected, 30% energy confinement improvement was obtained even during the NB heating of 14 MW. Further investigations of IDC characteristics have been made. The oxygen impurity lines from the main plasma and the main radiative loss drop first. Then the plasma stored energy starts to rise. The particle recycling is reduced around the main plasma, and is localized in the neighborhood of the X-point with a time lag of approximately 0.2 sec. Eventually the discharge shows a significant remote radiative cooling power at the divertor region.

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

From June to October 1987, JT-60 achieved fusion product (ne(0). tau E*.Ti(0)) of 6*1019 m-3.keV.s with hydrogen plasma at plasma current of 2.8 to 3.1 MA with neutral beam power of approximately 20 MW. The central electron density of 1.3*1020 m-3 was obtained at plasma current of 3 MA with 13 approximately 20 MW neutral beam power and the confinement time reached 0.14-0.18 s. An offset linear scaling law like the Shimomura-Odajima scaling on confinement time will be able to reproduce experimental data better than that of the Goldston type scaling. With low beam energy injection approximately 40 keV, confinement degradation was found. Many short periods (0.05 approximately 0.1 s) of H-mode phase were found in outside X-point divertor discharges with NB or NB+RF(LH or IC) heating power above 16 MW. However, improvement in energy confinement time was limited to 10 %. The ballooning/interchange stability analyses were also made for the outside X-point divertor equilibrium in connection with H0-phase capability. Heating powers of 9.5 MW and 1.9 MW were obtained by LHRF, ICRF injection, respectively. In combined LHRF and NB heating, the incremental energy confinement time of 0.064 s was obtained, which is the same level of that of NB heating only. In combined NB and on-axis ICRF heating of low ne discharge, an incremental energy confinement time of 0.21 s was obtained, which is three times as long as those of NB or ICRF heating only. High energy beam ions were accelerated by ICRF in the central region of the plasma.