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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 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.

A conceptual design for a pellet injection system will be worked out, capable to support key missions of the new tokamak device JT-60SA. For exploitations in view of ITER and to resolve key physics and engineering issues for DEMO, several tasks were assigned to this system. Physics investigations aim at operation at high density in ITER and DEMO relevant plasma regime above Greenwald density, power exhaust techniques with radiation layers, particle balance studies, and ELM control and mitigation. The postulated engineering requirement is to quantify pellet actuation on electron density for application within the advanced real-time control scheme by controlling density gradients. Our intended pellet system comprises three major components: pellet source, accelerator, and guiding system. The guiding system must be installed inside the torus vessel already under construction; hence still possible launch geometry was pursued first. Three different options have been identified: inboard, outboard, and top launch. The first one is most promising with respect to fuelling performance but will impose pellet speed restrictions to about 470 m/s for adequate pellet sizes. Both others offer headroom for significantly higher injection speed but under less favourable physics boundary conditions. In order to evaluate expected performances for all relevant plasma scenarios, detailed modelling efforts for every launch geometry option have been made. For a suitable pellet source covering all requirements, several options are at hand including commercial providers. For the accelerator, the high speed option up to about 4000 m/s could be covered by a two-stage gas gun. Single-stage gas guns and centrifuges can cover the speed range up to about 1000 m/s for the basic work load since both fulfil the requirements for pellet size and speed. Due to a higher speed precision resulting in less timing jitter, a centrifuge would be better suited for control requirements.

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.

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.

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.

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.

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.

In this work, the exhausted neutral gas flow is modeled for two cases of a simplified JT-60SA sub-divertor geometry and compared via three different approaches, namely (i) a collisionless approach based on the ProVac3D code, (ii) the DSMC approach based on the DIVGAS code that can be run with and without consideration of particle collisions, and (iii) the NEUT2D approach which has been extensively used in the past for the JT-60 design. In a first case study, the transmission probability was calculated by the 3 approaches and very good agreement is found between NEUT2D-ProVac3D whereas discrepancies between DIVGAS and NEUT2D are found and further analyzed. In the second case, the assessment of collisions is done by means of DIVGAS. Simulations showed that the flow is found in the transitional regime with Kn numbers between 0.1 and 0.4. The DIVGAS collisionless case yielded lower values of temperature than the collisional one by factors of 0.5–0.8 in regions near the inlets of the sub-divertor whereas in regions near the pump and the chevron, the temperature difference is marginal. Moreover, a relative percentage difference of 16–40% in pressure values was found between collisionless and collisional approaches.

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 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.

Vacuum vessel of JT-60 Super Advanced (JT-60SA) is composed of 18 port sections with 73 ports. The vacuum vessel has a D-shaped poloidal cross-section with a double wall structure, which is filled by shielding water for neutron. The vessel is made of AISI316L with low Cobalt content of

The modification of JT-60 to a fully superconducting coil tokamak, JT-60SA (JT-60 Super Advanced) device, has been programmed to contribute and supplement ITER toward to DEMO. Lower divertor design with the ITER-like lower single null divertor configuration is studied to obtain high heat and particle controllability using SOLDOR/NEUT2D code. With anticipated total power flux into SOL of 37 MW (90% of input power), the peak heat load on outer divertor target can be reduced to 5.8 MW/m 2 at the detached condition by gas puffing in the vertical divertor target with the “V-shaped corner”. It is ∼2/5 of the allowable level of 15 MW/m 2 . On the other hand, particle controllability such as control of detached to attached condition by divertor pumping is improved by increase the strike point distance from 20 to 120 mm with above divertor geometry, suggesting that recover from severe detachment at the small distance case can be achieving by elevation of the strike point locations. Optimization of upper divertor design is in progress for high β steady-state operation using upper single null divertor configuration.

This paper describes the conceptual design of cryostat for the JT-60SA, which is a research device for the commercial production of electricity from the controlled fusion reaction in the future. JT-60SA is designed to be a fully superconducting device and cryostat is one of the main components to allow the normal operation. Cryostat covers up the tokamak device, which is 15 m of total height and 7 m of radius, and supports the total weight of 25 MN. Cryostat components consist of vessel body, gravity support and auxiliary systems, such as 80 K thermal shield and vacuum exhaust. The functions required of cryostat are these three, thermal insulation for superconducting magnets, gravity support for the tokamak device, and bio-shielding. The design conditions for each cryostat component are outlined and the features of auxiliary systems such as capacity of vacuum exhaust related to 80 K thermal shield design are summarized.

The Japanese national project toward DEMO reactor and a satellite tokamak project for ITER in a Broader Approach with Japan and EU collaboration are combined as the JT-60SA (Super Advanced) project. Design of PFC for JT-60SA is widely modified from original national plan due to the increase of plasma heating power and annual neutron yield. Mono-block divertor target aiming at power handling of 15 MW/m 2 should be adopted to an outer divertor, because expected heat flux will exceed 10 MW/m 2 . Since high coolant flow velocity of 10–12 m/s is required to enhance heat transfer and critical heat flux for swirl or screw tube, primary coolant water system is increased from 1300 to 4600 m 3 /h. Remote handling system similar with ITER blanket manipulator is adopted to maintain plasma-facing components. Divertor cassette has 10° width in toroidal direction and weight of 300 kg itself due to the limitation of port width and manipulator.

The radiation shielding, nuclear heating, absorbed dose of the insulator and dose rate have been evaluated for the superconducting tokamak JT-60SA. The shielding structure consists of double wall using the 24 mm thickness of SS316L filled with the 140 mm thickness of borated water (40 °C, 95% 10B enriched). The maximum nuclear heating of the winding at the inboard side of the superconducting TF coil was suppressed to be lower than 0.3 mW/cc. The absorbed dose of the insulator on the in-vessel coil was estimated to be 8.5 MGy for 10 years. The insulators are not necessary to be exchanged with 10 years operation on JT-60SA. The dose rate of JT-60SA site boundary is estimated to be 144 μSv/3 months, and it satisfies the legal limit of 250 μSv/3 months without any additional shielding performance for the JT-60 building.

Three options of fusion DEMO plant are proposed characterized by functions of the center solenoid (CS). The prime option uses a downsized CS, which does not provide sufficient V-s for plasma current ramp-up but supplies enough coil current for plasma shaping. This option produces a fusion output of 3 GW with a major radius of 5.5 m, aspect ratio of 2.6, normalized beta of 4.3 and maximum field of 16.4 T. The estimated reactor weight is lighter than that of other conventional tokamak reactors, suggesting an economic advantage. The plant uses rather conservative technologies such as Nb3Al superconductor, water-cooled solid breeder blanket, low activation ferritic steel as the structural material and tungsten monoblock divertor plate. The design philosophy and key issues related to the constituent technologies of the plant are described in the present paper.