Design and manufacturing of JT-60SA vacuum vessel

JT-60 is planned to be upgraded to JT-60SA tokamak machine with fully superconducting coils, which is a project of the JA-EU satellite tokamak program under both Broader Approach program and Japanese domestic program. The JT-60SA vacuum vessel (VV) has a D-shape poloidal cross section and a toroidal configuration with 10° facet segmented in toroidal direction. The material of the VV is 316L stainless steel with low cobalt content of <0.05 wt%. A double wall structure is adopted for the VV to ensure high rigidity and high toroidal one-turn resistance simultaneously.Fundamental welding R&D and a trial manufacturing of the 20° upper half of the VV have been performed to study the manufacturing procedure. After the confirmation of the quality of the mock-up, manufacturing of the actual VV started in November 2009.     

JT-60SA vacuum vessel manufacturing and assembly

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 (<0.05%). The design temperatures of the VV at plasma operation and baking are 50 °C and 200 °C, respectively. In the double wall, boric-acid water is circulated at plasma operation to reduce the nuclear heating of the superconducting magnets. For baking, nitrogen gas is circulated in the double wall after draining of the boric-acid water.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.     

Full scale prototype of the JT-60SA Quench Protection Circuits

This paper deals with the development, manufacturing and testing of the full scale prototype of the Quench Protection Circuit (QPC) for the superconducting magnets of the JT-60SA Satellite Tokamak, which will operate in Naka, Japan.After the completion of the system detailed design in summer 2011, the manufacture of the poloidal and toroidal prototypes was launched and completed at the beginning of 2012. Several factory type tests on the main components have been done at the manufacturers’ premises and are described in this paper. Then, two main campaigns have been performed to test the operation of the overall poloidal and toroidal QPC prototypes; the main results are reported in the paper too.     

Assembly study for JT-60SA tokamak

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.     

Design and Related Experiments of JT-60 Divertor Coils

The JT-60 divertor coils produce a separatrix configuration in divertor operations of JT-60. A suitable separatrix configuration was obtained for a plasma current of 2.1 MA with coil ampere turns of ± 0.755 MAT. A high primary membrane stress of 52 MPa was permissible at the welded joints of the copper conductor made on the site. The mechanical strength of the joints welded in a factory was also improved by means of a press treatment. Electric insulation materials were selected considering a degradation of with stand voltage characteristics due to high cyclic mechanical strain. Vacuum-tight coil cases were composed of rigid rings and U-shaped bellows made of Inconel-625 alloy, and designed to withstand plasma disruption with a current decay time constant of 3 ms. The maximum temperature of the conductor in the periodic operation of divertor discharges was below 155°C which was the allowable temperature of the coil insulation. Molybdenum armor plates coated with titanium carbide and Inconel-625 bellows cover plates were attached against high heat flux from plasma. Thermal and mechanical load tests were carried out using component models to evaluate their performance in advance of the final fabrication of the actual coils. The satisfactory performance of the divertor coils were demonstrated in the pre-operational power test.     

Overcurrent analyses in JT-60SA poloidal circuits due to plasma disruption and quench protection intervention

The identification of the maximum amplitude of the currents circulating in the circuits is a useful indication for the design both of magnet and power supply components in fusion experiments. This paper evaluates the maximum level of coil overcurrents in the poloidal superconducting magnets of JT-60SA, the satellite tokamak that will be built in Naka, Japan, in the framework of EU-JA “Broader Approach” Agreement and that is expected to perform first plasma on 2016.To derive these information, a complete model capable to take into account all the mutually coupled elements was worked out, including the poloidal superconducting coils, the plasma position control in-vessel coils, the vacuum vessel, the stabilizing plates and the plasma.The model was utilized to analyze plasma disruption and quench protection circuit intervention in a large variety of different conditions to identify the possible overcurrent levels. The paper describes the model and the analyses performed, and presents and discusses the results.     

Design study of a wide-angle infrared thermography and visible observation diagnostic on JT-60SA

Design study of a wide-angle infrared (IR) thermography (surface temperature measurement) and visible observation diagnostics for JT-60SA are reported. The new design offers an optical solution without a “blind spot” which is one of the advantages. In order to image a large section inside the vacuum vessel (both in poloidal and toroidal directions), the optical system of endoscope is to provide a wide-angle view in the IR and visible wavelength ranges. The estimated IR optical spatial resolution is approximately 2 cm at a distance of 7.6 m from the front optics with a pupil diameter of 4 mm. For a surface temperature measurement it would be larger (∼4 cm for a surface temperature error less than 5%). The optics of this system can be divided into three parts: (1) a mirror based optical head (two set of spherical mirrors plus two flat mirrors) that produces an intermediate image, (2) a Cassegrain telescope system, and (3) a relay group of lenses, being adapted to the two kinds of detectors for IR and visible observations.     

Overview of engineering design,manufacturing and assembly of JT-60SA machine

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 JT-60SA device is a fully superconducting tokamak capable of confining break-even equivalent deuterium plasmas with equilibria covering high plasma shaping with a low aspect ratio at a maximum plasma current of I_p = 5.5 MA. This makes JT-60SA capable to support and complement ITER in all the major areas of fusion plasma development necessary to decide DEMO reactor construction. After a complex start-up phase due to the necessity to carry out a re-baselining effort with the purpose to fit in the original budget while aiming to retain the machine mission, performance, and experimental flexibility, in 2009 detailed design could start. With the majority of time-critical industrial contracts in place, in 2012, it was possible to establish a credible time plan, and now, the project is progressing on schedule towards the first plasma in March 2019. After careful and focused R&D and qualification tests, the procurement of the major components and plant is now well advanced in manufacturing design and/or fabrication. In the meantime the disassembly of the JT-60U machine has been completed and the engineering of the JT-60SA assembly process has been developed. The actual assembly of JT-60SA started in January 2013 with the installation of the cryostat base. The paper gives an overview of the present status of the engineering design, manufacturing and assembly of the JT-60SA machine.     

Present Status of JT-60SA Project and Development of Heating Systems for JT-60SA

Present status of the JT-60SA (JT-60 Super Advanced) project, implemented jointly by Europe and Japan since 2007, is described. The design of the main tokamak components was completed in late 2008, and all the scientific missions are preserved to contribute to ITER and DEMO reactors. The construction of the JT-60SA has begun with procurement activities for the superconducting magnet systems, vacuum vessel, in-vessel components and other components under the relevant procurement arrangements between the implementing agencies of JAEA (Japan Atomic Energy Agency) in Japan and Fusion for Energy in Europe. Designs and developments of the auxiliary heating systems for JT-60SA have been progressing at JAEA so as to provide the total injection power of 41 MW for 100 s.     

On a scenario assessment tool for the operation of the JT-60SA superconducting tokamak

JT-60SA is a superconducting tokamak to be assembled and operated at the JAEA laboratories in Naka (Japan). The tokamak is designed, manufactured and operated under the funding of the Broader Approach Agreement (between the government of Japan and the European Commission) and of the Japan Fusion National Programme; JT-60SA aims to prepare, support and complement the ITER experimental programme. The European contribution to the JT-60SA is, for a large fraction, procured by France, Germany, Italy, Spain and Belgium.This paper summarizes the activities carried out at F4E to develop a user-friendly software tool able to assess in real-time if an operational scenario could be structurally withstood by the magnet system of JT-60SA. Such tool is based on a theoretical formulation which is supported by a series of dedicated finite element method (FEM) calculations, and is able to provide a comparative assessment of any candidate scenario with respect to the baseline scenarios, and a quantitative assessment of all electro magnetic (EM) forces acting on the magnet system at any time during the candidate scenario. The tool as it is presented is specifically designed to be used for the JT-60SA tokamak, though it is designed so to that its usage could be extended easily to any other tokamak.     

Safe disassembly and storage of radioactive components of JT-60U torus

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.     

Application of Galerkin Method to Hydraulic Analysis for Turbulent Fluid Flow around LMFBR Fuel Rod

Wide range of parameter surveys are made on the DT fusion tokamak experimental reactor next to JT-60. Various physics and engineering requirements are taken into account, e.g. self-ignition, available maximum toroidal β value, α-particle confinement, total fusion power, neutron wall loading, heat flux to divertor plate, structural restriction on major radius, device size, maximum toroidal magnetic field, poloidal field power supply and so on. Theoretical scaling law for the available maximum toroidal β value determined by ballooning mode instability is used. The toroidal magnetic field on plasma axis can be expressed by the aspect ratio A for a given maximum field at the toroidal field coil conductor. Empirical scaling law for the electron energy confinement and neoclassical heat conductivity for the ion are employed. These confinement times can be expressed by the plasma minor radius a and A through the maximum available β value and the toroidal field on axis. In the similar way, most of the physics and engineering requirements can be mapped on the a-A diagram. This diagram enables us to make systematic and wide range of parameter surveys of the device. In particular, this offers a clear perspective on the device parameters, which can mitigate the engineering difficulties and can also realize the required plasma performances.     

Interaction of Ruthenium Tetroxide with Stainless Steel

Research and development (R&;D) on the selection of molybdenum first wall during FY1975–1976 are described. The JT-60 machine parameters are plasma current of 2.7 MA, toroidal magnetic field of 4.5 T, duration time of 5 to 10 s and additional heating power of 20 to 30 MW. From the viewpoint of first wall design, these parameters are more stringent in JT-60 than in medium size tokamaks. Therefore, R&;D on selection of material and structure of the JT-60 first wall was carried out. Initially, comparison between candidate materials were made regarding material, thermal, mechanical and vacuum properties. Molybdenum, pyrolytic graphite (PyG) and CVD-Sic coated graphite (SiC/C) were primary candidate materials. Of these three materials, full-sized trial productions of the first wall were made. High heat load tests with electron beam were carried out to compare thermal shock and thermal cycle properties. Test conditions were heat fluxes of 350 to 1,000 W/cm², duration of 10 s and cycle numbers from 10 to 320. From the test results, many cracks and “crater-like” damage were observed on the surfaces of PyG and SiC/C, but no damage was observed on the Mo surface. Following evaluation of all properties including these results, Mo was selected as primary first wall material for JT-60. Moreover, a trial production of Mo honeycomb structure was done. However, the honeycomb structure was not applied because of the expensive fabrication cost. After the operation of JT-60, the first wall materials (limiter, armor plates and magnetic limiter plate) were changed to graphite in FY1987 in order to reduce severe plasma contamination.     

Characterization of industrial NbTi strands at variable field for JT-60SA toroidal field coils

The mission of the JT-60SA Tokamak, to be built in Japan, is to contribute to the early realization of fusion energy by its exploitation in support of the ITER program. JT-60SA project is an important part of the “broader approach” activity as a satellite program for ITER. The toroidal field (TF) coils are a European “in kind” contribution and they will partly be built by France. JT-60SA TF coil uses the Cable In Conduit Conductor (CICC) with NbTi superconductor strands. TF conductors will have to operate at 5.7 T, 5 K and at current density of 450 A/mm² with sufficient margins. In the framework of JT-60SA TF coil manufacture, the variable temperature characterization is an important step to select NbTi strand. At an early stage of design, we had to choose the strand with acceptable performances. During the design qualification and validation stage, it is important to qualify strands in conditions close to the operation conditions. The industry has proposed various strands manufactured with different processes. This work and publication examines a strand with an internal CuNi barrier, which is expected to lead to better current distribution between strands, by more precise calibration and control of the inter-strand resistance. The strands were tested at the Grenoble High Magnetic Field Laboratory facility. The domain (B, T, J) explored was in the range of 4.5–11 T for the magnetic field intensity, 4.2–6.5 K for the temperature and between 40 A/mm² and 1200 A/mm² for the current density. For each strand, “critical current density” and “current sharing temperature” measurements have been carried out, with a temperature precision of few tens of mK. Once the measurements performed, the fitting parameters (of type J_C = f(B, T)) of each strand have been found, by performing regression analysis. This work will lead to select the strand with the best characteristics. In this paper, we present the results of this measurement task, the data and regression analysis (fits, T_cs, etc.) and the conclusion about the strand choice.     

Progress in development and design of the neutral beam injector for JT-60SA

Neutral beam (NB) injectors for JT-60 Super Advanced (JT-60SA) have been designed and developed. Twelve positive-ion-based and one negative-ion-based NB injectors are allocated to inject 30 MW D⁰ beams in total for 100 s. Each of the positive-ion-based NB injector is designed to inject 1.7 MW for 100 s at 85 keV. A part of the power supplies and magnetic shield utilized on JT-60U are upgraded and reused on JT-60SA. To realize the negative-ion-based NB injector for JT-60SA where the injection of 500 keV, 10 MW D⁰ beams for 100 s is required, R&Ds of the negative ion source have been carried out. High-energy negative ion beams of 490–500 keV have been successfully produced at a beam current of 1–2.8 A through 20% of the total ion extraction area, by improving voltage holding capability of the ion source. This is the first demonstration of a high-current negative ion acceleration of >1 A to 500 keV. The design of the power supplies and the beamline is also in progress. The procurement of the acceleration power supply starts in 2010.     

Design Study of Top Lid with Clamp Structure in JT-60SA Cryostat

JT-60SA is a fully superconducting coil tokamak upgraded from the JT-60U. This paper focused on the integrity of the top lid of cryostat in JT-60SA. The design requirement for the cryostat in normal operations is to achieve vacuum insulation of 10 3 Pa, and the top flange of the top lid is lightly welded onto its body flange. The weld is tensile-loaded by bending deformation of the top lid due to vacuum pressure of external 0.1 MPa. This weld integrity is evaluated with tensile-load reduction, which results in clamp reinforcement. The structural integrity of the top lid is validated.     

Electromagnetic analysis, structural integrity and progress on mechanical design of the ITER ferromagnetic insert

Ferromagnetic material is used to reduce the toroidal field ripple in JFT-2M [H. Kawashima, et al., Demonstration of ripple reduction by ferritic steel board insertion in JFT-2M, Nucl. Fusion, 41 (2001) 257-263] and JT-60U [H. Takenaga, the JT-60 Team, Overview of JT-60U results for development of steady-state advanced Tokamak scenario, Proceedings of the 21st IAEA Fusion Energy Conference, Chengdu, China, 2006]. In ITER, since the ferromagnetic material is inserted in the space between the double walls of ITER Vacuum Vessel (VV), it is called “ferromagnetic inserts”. Suitable material is selected to satisfy the design requirements of ITER. The proper location and amount of the ferromagnetic inserts are optimized with the goal of reduction of the toroidal field ripple. The ferromagnetic inserts are designed to minimize electromagnetic forces acting on them. The electromagnetic forces have been calculated with the latest disruption scenarios. Magnetization forces due to magnetic fields have also been calculated. Structural integrity has been validated by a structural analysis.     

Plan of ITER remote experimentation center

Plan of ITER remote experimentation center (REC) based on the broader approach (BA) activity of the joint program of Japan and Europe (EU) is described. Objectives of REC activity are (1) to identify the functions and solve the technical issues for the construction of the REC for ITER at Rokkasho, (2) to develop the remote experiment system and verify the functions required for the remote experiment by using the Satellite Tokamak (JT-60SA) facilities in order to make the future experiments of ITER and JT-60SA effectively and efficiently implemented, and (3) to test the functions of REC and demonstrate the total system by using JT-60SA and existing other facilities in EU. Preliminary identified items to be developed are (1) Functions of the remote experiment system, such as setting of experiment parameters, shot scheduling, real time data streaming, communication by video-conference between the remote-site and on-site, (2) Effective data transfer system that is capable of fast transfer of the huge amount of data between on-site and off-site and the network connecting the REC system, (3) Storage system that can store/access the huge amount of data, including database management, (4) Data analysis software for the data viewing of the diagnostic data on the storage system, (5) Numerical simulation for preparation and estimation of the shot performance and the analysis of the plasma shot. Detailed specifications of the above items will be discussed and the system will be made in these four years in collaboration with tokamak facilities of JT-60SA and EU tokamak, experts of informatics, activities of plasma simulation and ITER. Finally, the function of REC will be tested and the total system will be demonstrated by the middle of 2017.     

Tokamak edge plasma rotation in the presence of the biased electrode

Electrode biasing system was designed, constructed, and installed on the IR-T1 tokamak, and then biasing experiments were carried out. Also, using a Mach probes the effects of radial electric field (produced by biased electrode) on the poloidal and toroidal components of the edge plasma velocity were investigated. The results showed an increase in both toroidal and poloidal components of the edge plasma velocity during biasing regime. Results compared and discussed. During positive biasing, increased E_r tends to slow the poloidal rotation in the electron diamagnetic drift direction, i.e., to speed up rotation in the ion diamagnetic drift direction. An increased toroidal rotation velocity has the opposite effect on the poloidal rotation.