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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.

In the present RF-driven (ECCD) steady-state plasma on QUEST (Bt = 0.25 T, R = 0.68 m, a = 0.40 m), plasma current seems to flow in the open magnetic surface outside of the closed magnetic surface in the low-field region according to plasma current fitting (PCF) method. We consider that the current in the open magnetic surface is due to orbit-driven current by high-energy particles in RF-driven plasma. So based on the analysis of current density profile based on the orbit-driven current, plasma equilibrium is to be calculated. We calculated high energy particles guiding center orbits as a contour plot of conserved variable in Hamiltonian formulation and considered particles initial position with different levels of energy and pitch angles that satisfy resonance condition. Then the profile of orbit-driven current is estimated by multiplying the particle density on the resonance surface and the velocity on the orbits. This analysis shows negative current near the magnetic axis and hollow current profile is expected even if pressure driven current is considered. Considering the hollow current profile shifted toward the low-field region, the equilibrium is fitted by J-EFIT coded by MATLAB.

In the present RF-driven (ECCD) steady-state plasma on QUEST, the plasma current seems to flow in the open magnetic surface outside of the closed magnetic surface in the low-field region according to the plasma current fitting (PCF) method. The current in the open magnetic surface is due to the orbit-driven current by high-energy particles in the RF-driven plasma. High-energy particles guiding center orbits are calculated as contour plots of conserved variables in the Hamiltonian formulation considering particles in the initial position with different energies and pitch angles satisfying the resonance condition. A negative current appears near the magnetic axis, and a hollow current profile is expected even if the pressure driven current is considered. The equilibrium is fitted within nested magnetic surfaces by J-EFIT coded by MATLAB using the hollow current profile shift toward the low-field region. Although the plasma boundary shape reflects the plasma current density profile, the equilibrium shape fitted by J-EFIT does not coincide with the orbit-driven current profile. However, introducing an extension of the current profile without assuming nested contours into the J-EFIT code appropriately fits the plasma shape with the hollow current profile to the measured magnetic data.

JT-60SA is the satellite tokamak under construction in Naka, Japan, in the framework of the EU-JA -Broader Approach? Agreement. In JT-60SA, to attain steady-state high-beta plasmas, suppression of Resistive Wall Modes (RWM) is necessary. At this purpose, a passive stabilizing plate (SP) and an active control system based on 18 in-vessel sector coils (SC) are foreseen. In the past years, physics studies have been carried out to produce the inputs for the design of the RWM control system and the SC design evolved accordingly. In the present arrangement, the coils are placed on the plasma side of SP, to minimize the magnetic shielding effects of passive structures. Electromagnetic analyses on SC have been carried out in support to the evaluation of the different coil design solutions and to derive the requirements for the related power supply system (RWM-PS). To achieve the highest flexibility in RWM control, it has been devised to feed each SC independently, with currents following 18 real-time references. In 2012, the SC design has been further improved by increasing the number of turns from 2 to 8. This allowed reducing the rated current, thus the voltage drops on the feeders and the power required from RWM-PS. To keep the RWM under control with relatively low Ampere-turns (2.2 kAt), the RWM-PS has to guarantee very fast dynamics (current bandwidth: 3 kHz; maximum latency: 50 us), largely beyond that of standard industrial applications. This paper firstly reports the main requirements for the RWM control system. Then, the reference design of the RWM-PS is described. It includes an ac/dc conversion system, dc-link capacitor banks and a set of 18 fast inverters. The advantages of the proposed scheme are discussed and the main electrical parameters are shown in detail. The main requirements of the control section are given, with details on possible implementation and interfaces with JT-60SA central 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.

A new database system for the JT-60SA tokamak has been developed. This system meets the following requirements: to hold all the JT-60SA data and prevent loss of data, to allow researchers to use data without knowing internal database configuration and structure, to manage data revision so that stored data can be traced back to the older revision anytime. Depending on the timing of data transfer for storage or reference, the experiment data are categorized into three types: (i) discharge data, (ii) plant monitoring data, and (iii) unprocessed data. In the first step in constructing a new database system, we have designed the basic structure of the JT-60SA database system to manage all three data types. The dedicated tools for data transfer or retrieval prevent unregistered computers from accessing the experiment data. The RAID6 configuration of a hard disk drive (HDD) and the magnetic tapes improve data reliability. This paper reports details about the functions of the new JT-60SA database system. Furthermore, the elapsed time of data transfer and storage is measured to evaluate the performance of the JT-60SA database system. It takes 20 s to transfer and store 6 GB of data, which is enough to meet the required performance.

CCS (Cauchy Condition Surface) method is a numerical approach to reproduce plasma shape, which has good precision in conventional tokamaks. In order to apply it in plasma shape reproduction of ST (Spherical Tokamak), the calculation precision of the CCS method in a spherical tokamak CPD (Compact PWI experimental Device) (Bt = 0.25 T, R = 0.3 m, a = 0.2 m) has been analyzed. The precision was confirmed also in ST and decided to be applied to a spherical tokamak QUEST (Bt = 0.25 T, R = 0.68 m, a = 0.40 m).. In present stage from the magnetic measurement, it is known that the eddy current effect is large in QUEST experiment, and there are no special magnetic measurements for eddy current now, so some proper model should be selected to evaluate the eddy current effect. The eddy current density by not only CS (Center Solenoid) coil but also plasma current is calculated using EDDYCAL (JAEA). The eddy current magnitudes are taken as unknown variables and solved together with plasma shape reconstruction in ohmic discharge and ECCD (Electron Cyclotron Current Drive) discharge.

The JT-60 operations were stopped in 2008 in order to be upgraded to a superconducting tokamak device, JT-60 Super Advanced (JT-60SA). The design activity of JT-60SA is going on under the ITER-BA p...

A fast charge exchange recombination spectroscopy (CXRS) system has been developed for the real-time measurement and feedback control of ion temperature ( T i ) profile and toroidal rotation velocity ( V t ) in JT-60U. In order to control T i and V t in real-time, the charge exchange recombination spectroscopy with high time resolution, the real-time processor system, and the real-time control system have been developed. Utilizing this system, real-time control of the T i gradient between r / a ∼ 0.25–0.5 has been demonstrated with neutral beams at high beta plasmas (normalized beta β N ∼ 1.6–2.8). The strength of the internal transport barrier is controlled. Moreover, the real-time control of V t has been demonstrated from counter (anti-parallel to the plasma current, I p ) to co (parallel to the I p ) direction. Then the behavior of edge localized mode (ELM) is changed by controlling the V t .

The design activity of JT-60SA (JT-60 Super Advanced) addresses the remodeling of JT-60 to a superconducting tokamak device under the JA-EU collaborative ITER–BA project. For the JT-60SA control system, the existing system should be reused as much as possible from the viewpoint of cost-effectiveness. Initial considerations of the configuration of the advanced supervisory control system (SVCS) including the following systems are: (1) an ultimate flexibility real-time control system, (2) a precision timing system able to diagnose a plasma event having a time range of ∼1 μs, and (3) a safety shutdown control system. In this report, we present the design study of the JT-60SA SVCS with the focus on the above systems.

For safety and efficient operation of a fusion device, appropriate information should be provided to the personnel participating in the experiment. In a continuing effort to meet this requirement, we have recently developed highlight displays showing the plasma movie, the status of the discharge sequence, the discharge schedule, the status of present troubles, etc. These highlight displays are presented on the large (65 in.) TV monitor in front of the central consoles. These displays are automatically changed according to the timing signals of the JT-60 discharge sequence and/or at the JT-60 session leader's request. We call this system “the supervisory discharge operation monitoring system.” This report describes the technical details of both this system and the discharge information management system (DIMS).

The real-time plasma movie with the computer graphics (CG) of plasma shape is one of the most effective methods to know what discharge have been made in the experiment. For an easy use of the movie in the data analysis, we have developed the plasma movie database system (PMDS), which automatically records plasma movie according to the JT-60 discharge sequence, and transfers the movie files on request from the web site. The file is compressed to about 8 MB/shot small enough to be transferred within a few seconds through local area network (LAN). In this report, we describe the developed system from the technical point of view, and discuss a future plan on the basis of advancing video technology.

Burn control simulation experiments were conducted in non-burning DD plasmas by using a burning plasma simulation scheme in JT-60U. In the burning plasma simulation scheme, two neutral beam (NB) groups were used, where one simulates α-particle heating and the other simulates external heating. The stored energy and the neutron yield rate were controlled at nearly constant values in both ELMy H-mode plasma and reversed shear (RS) plasma with an Internal Transport Barrier. In these plasmas, the NB power for the external heating simulation increased/decreased when the NB power for the α-particle heating simulation decreased/increased. Variation of the NB power for the external heating simulation was larger in the RS plasma than that in the ELMy H-mode plasma, indicating that larger control margin is necessary for the RS plasmas. In order to understand the detailed physical mechanism for the difference between ELMy H-mode and RS plasmas, we conducted numerical analysis using a 1.5-dimensional transport code. This calculation indicates the larger variation experimentally observed in the RS plasma could not be explained by the difference of the thermal diffusivity profiles or its temperature dependence.

The large tokamak fusion device JT-60 is expected to explore more advanced tokamak discharge scenarios toward the ITER and a future power reactor. Since various experimental issues are to be adequately discussed, and possibly to be solved in JT-60, the plasma real-time control system has been drastically improved with remodelling in hardware as well as in software. To satisfy the requirements, a “multiple networks” structure is employed as a basic principle of the remodelling. Distributed processors for diagnostics, actuators, and supervisory controllers are linked through a reflective memory (RM) network for fast, real-time communication. Similarly, advanced and complex calculations to reproduce plasma shape and profiles are performed by several processors connected to the same RM network. In this report, we discuss the developments to improve the JT-60 plasma control and data processing systems. In addition, a future plasma control system leading to a standard design for a power reactor is outlined, which is based on the 20-year plasma operation experience.

The JT-60U control system, composed of a UNIX-based workstation and a VME-bus system, has been successfully performing advanced plasma control for the JT-60U plasma discharges. Various hardware modifications and advanced plasma control algorithms have been continuously integrated into the JT-60U control system since its upgrade in 1991. The ZENKEI, the supervisory control system, has been entirely converted from its original 16-bit minicomputers and CAMAC system to the combination of a workstation and VME-bus system linked through an Ethernet network. The plasma equilibrium control system utilizes a full VME-bus system composed of three DEC alpha CPU modules that are currently running at control cycle times of 250 μs. The parameters for plasma equilibrium control are calculated by a newly developed real-time plasma shape calculation and reconstruction system. These parameters are fed to the plasma equilibrium control system synchronized by 250-μs clock pulses. To progress toward the advanced plasma control concept, which includes adoption of a global plasma parameter control algorithm and suppression of MHD instabilities, improvements were extensively consolidated into the plasma control system (PCS). Currently, additional modifications of the JT-60U control system have extended the pulse duration of the discharge without a major change in JT-60 equipment.

The present CAMAC based timing system has been used for synchronizing sequential events of the discharge and the data collection of the interesting JT-60U experiment plasma phenomena. However, a more flexible and sophisticated state-of-the-art timing system now is required to realize advanced plasma control with minimal maintenance costs. In this context, the versa module Europe (VME-bus) system with a high-speed data communication network using reflective memory (RM) modules and user-friendly application software based on MATLAB™ tools has been selected to develop the new prototype timing system. In the ZENKEI, the supervisory control system of the JT-60, the supervisory timing system provides the 50-μs master clock pulses, the various timing signal preparation logic, which is built into the digital signal processing (DSP) module in conjunction with the discharge sequence event signals, and the 6.2 MB/s high-speed communication data link provided by the RM module. Except the clock pulse generator (CPG) module, no other special timing module is necessary for this new timing system. The timing signal is prepared by software logic in conjunction with sequential events and the preset timer, is transferred to the subsystems through the RM module, where it is synchronized to the 50-μs clock pulses. The timing system of the subsystems also consists of hardware similar in structure to the ZENKEI timing system. The fundamental timing system configuration, the necessary functions, and the preliminary test results of the prototype system are reported in this presentation.

The real-time feedback control of the plasma stored energy has been developed for control of the plasma MHD stability in the JT-60U tokamak. The plasma stored energy can be detected with high accuracy in real-time by a function parameterization method for various plasmas available in JT-60U, such as Ohmic plasmas, the L-mode, the H-mode, the high poloidal beta mode and the reversed shear mode over a wide range of the plasma parameters. By manipulating the neutral beam injection power, the plasma stored energy has been successfully controlled along the preprogrammed reference waveform. Especially in the reversed shear mode, this feedback control scheme has improved the reproducibility of the formation of the internal transport barrier, and MHD activities could be suppressed keeping the normalized beta in a stable region. A DT equivalent fusion amplification gain of 0.5 was sustained for 0.8 s in a reversed shear plasma by employing this feedback control scheme.

The significance of real-time feedback control is emphasized in this paper as an indispensable method to improve and sustain the improved plasma characteristics in JT-60U high fusion performance discharges as well as to operate the fusion reactor under the optimal divertor conditions with respect to the heat load and exhaust pumping. In accordance, substantial improvement in the equivalent fusion amplification gain of over unity has been reproducibly achieved at the JT-60U tokamak in the reversed shear mode of operation with the robust feedback controls, where the value of target density was deliberately optimized for the reliable internal transport barrier formation, and the magneto-hydrodynamic stability control was performed with the stored energy feedback. The feedback control techniques also demonstrated the effectiveness to produce quasi-steady-state high-performance plasmas. In addition, three major parameters associated with the fusion reactor instrumentations, namely the neutron production rate, operating density, and divertor radiation power, were simultaneously feedback controlled in the ELMy H-mode plasmas. Here, the matrix response function was evaluated to identify the limitations involved with the linear combination of independent controls. Other advanced feedback schemes, such as the feedback suppression of the neoclassical tearing mode required to sustain high plasma pressure in a steady-state, are alsomore » described. Finally, the controversial issues for the future intelligent plasma control necessary for the advanced steady-stated tokamak reactor are addressed.« less

The sustainment of radiative divertor plasma in JT-60U is performed by the feedback control of the gas puffing rate into the divertor region with an actuator of the divertor radiation power. The feedback control system switches the piezoelectric gas valve every 10 ms, so that the proper amount of gas flux is fed for divertor radiation power to follow the reference waveform. The divertor radiation of 6 MW, which is about 40% of the heating power, is sustained for about 3 s by the hydrogen gas puffing with feedback control. The present logic of the feedback system, however, is not sufficient to fully control the divertor radiation which is modified by the change in heatings and plasma parameters. In order to prevent the loss of control, multi actuators for the feedback control, for example, the involvement of the heating power, and the selection of observed bolometer channels may be required.

This paper describes the central control system of the JT-60 tokamak. This control system not only has to organize plasma operation, but also has to utilize devices under isolated operation. Therefore the control system has been designed in a special way.

Burning plasma simulation experiments were performed for bum control study on ELMy H-/L-mode plasmas and reversed shear (RS) plasmas with an internal transport barrier in JT-60U. In a burning plasma simulation scheme, two neutral beam (NB) groups were used: one that simulates alpha-particle heating and another that simulates external heating. For the alpha-particle heating simulation, the heating power proportional to the deuterium-deuterium (D-D) neutron yield rate was injected. The behavior of the part of the NB heating simulating alpha-particle heating was varied by increasing the proportional gain relating the applied power to the measured neutron yield rate in both ELMy H-mode and RS plasmas, while the part of the NB power in the role of external heating was held constant i.e., no-bum-control case. Above a certain value of the proportional gain, a runaway effect was triggered where excursive increases in the neutron yield rate and stored energy were observed. With burn control, where the stored energy was controlled at a constant value by a feedback control system using the external heating, the runaway was not triggered, and the neutron yield rate was kept at a constant value in the L-mode plasmas. Zero-dimensional calculation indicated that the runaway triggered by increasing the proportional gain well simulates the runaway triggered by improved confinement. The limitations came from differences between deuterium-tritium and D-D plasmas, such as the dominant reaction for the neutron yield and the temperature dependence of the fusion reaction rate, which were discussed together with improvement on the burning plasma simulation scheme.

A real-time control function of neutron emission rate has been developed as a first step towards fusion power control. By controlling the number of neutral beam injection units, the neutron emission rate was successfully kept constant for ∼ 5 s until the feedback control was turned off. The plasma stored energy was also kept constant during the same period owing to the correlation between the stored energy and the neutron emission rate. This function is expected to be effective in extending the steady-state performance and in controlling the reactivity of a fusion reactor at Q ≤ 1.