Progress in JT-60 Experiments

The Japan Atomic Energy Research Institute completed in April 1985 the construction of the large tokamak JT-60, which constitutes a focus of the Second Phase of the Japanese fusion development program under the Atomic Energy Commission started in 1975. Initial experiments of JT-60 were carried out in hydrogen plasma after the completion of the device. Full installation and testing of the heating devices was completed in July 1986 and subsequently the heating experiment was initiated. The target parameters were achieved in deuterium-equivalent values in September 1987 by high power heating of high density hydrogen plasmas at high plasma current. JT-60 has entered the phase of advanced experiments in 1988.     

Development of Negative-Ion Based NBI System for JT-60

The negative-ion based neutral beam injector (N-NBI) for JT-60 has been developed for plasma core heating and neutral beam current drive in higher density plasmas. Construction of the N-NBI system was completed in 1996, and just after this completion, the efforts to increase beam power and beam energy started. The N-NBI system has already operated with negative ion beams with 14.3 A at 380 keV of deuterium and with 18.5A at 360 keV of hydrogen. During N-NBI experiments on JT-60, a deuterium neutral beam power of 5.2MW at 350keV has been injected for 0.7s stably, and the response of the JT-60 plasma to high energy beam injection with the N-NBI has been confirmed to be in agreement with a theoretical prediction.     

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.     

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.     

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.     

Progress in ECRF antenna development for JT-60SA

Structural, mechanical and optical design work on antennas/launchers for the electron cyclotron range of frequency heating and current drive system in JT-60 Super Advanced (JT-60SA) have been advanced based on a linear motion antenna concept. A CAD model of the launcher was built with realistic component sizes. A mock-up of the steering structure consisting of two different bellows sections for poloidal and toroidal beam scanning was fabricated to test movement of the bellows. The poloidal (?40° to +20°) and toroidal (?15° to +15°) injection angle ranges required in JT-60SA were shown to be realized by this steering structure and mirrors.     

Studies on the requirements of the power supply system for the resistive-wall-mode control in JT-60SA

The JT-60SA satellite tokamak will be built in Naka, Japan. One of the main aims of this machine is to achieve steady-state high-beta plasmas. To reach this result, passive stabilizing plate (SP) and resistive wall modes (RWM) active control system based on 18 in-vessel coils will be installed. In the present design, these coils are placed on the plasma side of the SP, behind the first wall. This solution maximizes the efficiency in producing fast magnetic fields into the plasma by minimizing the shielding effect of the passive structures. Then, if the power supply (PS) and the control system have sufficient dynamic performance, it is possible to control the RWM with very low magnetic fields. This allows minimizing the Ampere-turns and the power requested to control the RWM. Conversely, the very fast dynamics required represents one of the main issues for the design of the RWM control system. This paper, after having recalled the main specification data for the RWM control system deriving from the physics studies, describes the analyses performed to complete the set of requirements necessary for the PS design. The characterization of coils and feeders is shown and the voltage necessary to produce the required current and bandwidth is quantified. Possible connections among PS and coils are analyzed in order to achieve the highest possible flexibility in controlling the RWM with a reduced set of independent PS. Finally, considerations on reasonable voltage margins to cope with load uncertainties are given.     

Steady state ECRH for next step superconducting fusion devices

Controlled plasma start-up, heating towards thermonuclear fusion temperatures and steady state discharge control in advanced configurations is a major challenge for both, stellarators and tokamaks. Electron Cyclotron Resonance Heating (ECRH) and current drive (ECCD) plays a key role in the steady state operation scenarios of the ITER and JT-60 SA tokamaks as well as for the W7-X stellarator. The physics demands as well as the key technology of the different ECRH-systems, which are similar in frequency and have continuous wave (cw) capability, are presented. Advanced solutions for future ECRH-systems are discussed.     

Plasma Centrifuge Heat Engine – a Route to Non-thermal p- 11 B Fusion

An invention [US Patent and Trademark Office App. Nos. 60/596567 (2005) and 60/766791 (2006)] combines centrifugal and dipole confinement, with recent oscillating plasma theory. The plasma undergoes compression/expansion (C/E), parallel to B by centrifugal force and perpendicular to B by B variation, providing a thermal cycle which recovers most (>95%) of heating as mechanical energy. This gives a “Q-amplifier” for beam-target systems. Centrifugally confined Boron plasma undergoes C/E by slow, cross-B interchange activity. Parallel and perpendicular C/E are matched by the rotation profile which arises naturally. Hot plasma is heated and cold plasma is cooled. Beam-target fusion reactions occur in the hot plasma region and expansion returns most of the heat energy as rotation energy. Rotation energy, in turn, produces waves which drive protons to an energy near the fusion peak cross section. A possible machine, including the arrangement of magnets and HV, is described.     

Minimization of Reactive Power Fluctuation in JT-60SA Magnet Power Supply

This paper describes an asymmetric control method for the firing angle and a start/stop timing shift control of four thyristor converters called "Booster PS" to minimize the reactive power fluctuation during plasma initiation in JT-60SA. From the simulation using the "PSCAD/EMTDC" code, it is found that these control methods can drastically reduce the reac- tive power induced by the four units of the "Booster PS". In addition, the voltage fluctuation of the motor-generator connected to the "Booster PS" is expected to be suppressed. This can also contribute to achieve stable control of the JT-60SA magnet power supplies.     

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.     

Quench detection of fast plasma events for the JT-60SA central solenoid

The JT-60 is planned to be modified to a full-superconducting tokamak referred to as the JT-60 Super Advanced (JT-60SA). The maximum temperature of the magnet during its quench might reach the temperature of higher than several hundreds Kelvin that will damage the superconducting magnet itself. The high precision quench detection system, therefore, is one of the key technologies in the superconducting magnet protection system.The pick-up coil method, which is using voltage taps to detect the normal voltage, is used for the quench detection of the JT-60SA superconducting magnet system. The disk-shaped pick-up coils are inserted in the central solenoid (CS) module to compensate the inductive voltage. In the previous study, the quench detection system requires a large number of pick-up coils. The reliability of quench detection system would be higher by simplifying the detection system such as reducing the number of pick-up coils. Simplifying the quench detection system is also important to reduce the total cost of the protection system. Hence the design method is improved by increasing optimizing parameters. The improved design method can reduce the number of pick-up coils without reducing the sensitivity of detection; consequently the protection system can be designed with higher reliability and lower cost. The applicability of the disk-shaped pick-up coil for quench detection system is evaluated by the two dimensional analysis. In the previous study, however, the analysis model only took into account the CS, EF (equilibrium field) coils and plasma. Therefore, applicability of the disk-shaped pick-up coil for the quench detection system remains open question because the fast plasma events, such as disruption, mini collapse and ELM (edge localized mode), directly influences on the voltage of pick-up coil making the quench signal undetectable. Consequently, a new analysis model proposed in the present paper was designed to avoid this difficulty by introducing the passive coil series such as vacuum vessel and stabilizer. The influence of fast plasma events is absorbed by passive coil series like real system, and the evaluation of applicability can be examined in detail. The analysis results show that the disk-shaped pick-up coil is applicable whenever the standard operation, disruption, mini collapse and ELM.     

High-Beta Steady-State FRC Plasma Sustained by Rotating Magnetic Field with Spatial High-Harmonic Components

Field-reversed configurations (FRCs) driven by rotating magnetic fields (RMFs) with spatial high-harmonic components have been studied in the metal flux conserver of the FRC injection experiment (FIX). The high-harmonic RMF method has some unique features; (1) field lines of the RMF do not penetrate or cross the vessel wall, (2) selective penetration/exclusion of the fundamental/high-harmonic RMF component will result in a generation of effective magnetic pressure near the separatrix, which helps to keep the separatrix away from the vessel wall, (3) strong azimuthal non-uniformity of the RMF will cause the n = 4 deformation of the core FRC plasma, which will eliminate the destructive modes caused by the rotation of the plasma column. The RMF method with high harmonics will provide quasi-steady current drive of high-beta FRC plasmas without destructive n = 2 rotational mode and will be helpful in reducing the particle loss and thermal load when applied to the fusion core plasma.     

Advanced Control Scenario of High-Performance Steady-State Operation for JT-60 Superconducting Tokamak

Plasma control on high-βN steady-state operation for JT-60 superconducting modification is discussed. Accessibility to high-βN exceeding the free-boundary limit is investigated with the stabilising wall of reduced-activated ferritic steel and the active feedback control of the in-vessel non-axisymmetric field coils. Taking the merit of superconducting magnet, advanced plasma control for steady-state high performance operation could be expected.     

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.     

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.     

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.     

Progress in Physics and Technology Developments for the Modification of JT-60

Recent progress in the physics and engineering design study for the modification programme of JT-60 is presented. In order to achieve a steady state high-β plasma operation, which is the dominant issue of this programme, physics design for the MHD control and the stability analysis is investigated. Engineering design and the R & D for the superconducting coils, irradiation shield are performed well towards the mission of programme.     

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.     

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.