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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.

The first half of this paper briefly describes the major achievements of the existing JT-60U machine, i.e., high-performance plasma with full noninductive current drive, long time sustainment of high-beta plasma, and real-time neoclassical tearing modes (NTM) suppression experiments. In addition, the ongoing minor modification of the pulse length prolongation for further enhancement of numerous plasma properties is mentioned. The second half describes the major modification program for the next generation full superconducting coil tokamak called JT-60SC. The object of JT-60SC is to help in the design of DEMO reactor which will improve the attractiveness of fusion power plants from the economical and environmental viewpoints. The present status of the machine design and major technological developments are also presented.

Modification of JT-60 as a full superconducting tokamak (JT-60SC) is planned. The objectives of the JT-60SC programme are to establish scientific and technological bases for steady-state operation of high performance plasmas and utilization of reduced-activation materials in an economically and environmentally attractive DEMO reactor. Advanced fusion technologies relevant to the DEMO reactor have been developed for the superconducting magnet technology and plasma facing components of the JT-60SC design. To achieve a high current density in a superconducting strand, Nb3Al strands with a high copper ratio of 4 have been newly developed for the toroidal field coils (TFCs) of JT-60SC. The R&D to demonstrate the applicability of the Nb3Al conductor to TFCs by a react-and-wind technique has been carried out using a full-size Nb3Al conductor. A full-size NbTi conductor with low ac loss using Ni-coated strands has been successfully developed. A forced cooling divertor component with high heat transfer using screw tubes has been developed for the first time. The heat removal performance of the carbon fibre composite target was successfully demonstrated on an electron beam irradiation stand.

Hydrogen retention in carbon dust produced by hydrogen discharges in JT-60 was investigated to estimate the tritium inventory for the next generation of fusion reactors such as ITER. Several types of dust-like pieces were collected from the vacuum vessel, flake-like carbon dust, graphite, and metal pieces. The amount of retained hydrogen in the flake-like carbon dust was two orders of magnitude larger than that of the other samples. The hydrogen concentration of the flake-like carbon dust was less than 0.04 in the atomic ratio of H/C. This value is one order of magnitude smaller than that of carbon dust produced in the large tokamak, JET. This difference can be explained by the wall temperature in JT-60 being higher than in JET. The surface morphology and atomic compositions were also examined for characterization of the collected samples.

A fully superconducting tokamak named JT-60SC is designed for the modification programme of JT-60 to enhance economical and environmental attractiveness in tokamak fusion reactors. JT-60SC aims at realizing high-β steady-state operation in the use of low radio-activation ferritic steel in a low ν* and ρ* regime relevant to the reactor plasmas. Objectives, research issues, plasma control schemes and a conceptual design for JT-60SC are presented.

The modification of JT-60 is programmed to be a superconducting tokamak (JT-60SC) that has superconducting toroidal field (TF) and poloidal field coils. The TF coil system, which consists of 18 D-shaped coils, makes the field of 3.8 T at plasma center (R = 2.8 m). The stored magnetic energy in total is 1.7 GJ. A squared Nb/sub 3/Al cable-in-conduit (CIC) conductor using SS316LN jacket allows us to fabricate the TFC by a react-and-wind (R&W) technique because of low strain sensitivity of Nb/sub 3/Al J/sub c/ characteristic. In order to support TF coils against the centering force, the inboard side of the TF coil case is wedged. Shear panels and keys between the TF coils are placed against the overturning force induced by poloidal fields. Stress analysis of the TF coil support structure was carried out. In the case of the maximum displacement caused by overturning force, the maximum stress was lower than the allowable stress of SS316LN at 4 K. The stress analysis of insulator and conduit at the inboard side of the TF coil with zooming model was also carried out. It is clear from zooming analysis that the stresses of conduit and insulator are within the allowable stress. Regarding a progress of R&D, demonstration of R&W technique using a D-shaped double layer coil with the full-scale CIC conductor is planned in this year in order to confirm its designed performance.

The capability of detachment control in JT-60SC is demonstrated with a divertor code (SOLDOR/NEUT2D). Under the standard operation of the core edge density at 95% of the minor radius, n edge = 3.2 x 10 19 m 3 , the power flow from the core plasma Q c = Q i = 6 MW, and the conductance of inner and outer cryopanels C in = C out = 50 m 3 /s, the inner divertor plasma is partially detached and the outer divertor plasma attached. The inner divertor can be changed from partially detached to attached plasma by inner pump of 150 m 3 /s. The fully detached inner divertor is obtained in high density operation with n sep > 4 x 10 19 m 3 , where n sep is the density at the separatrix outer midplane, The elastic collision in a dense and cold divertor plasma (n c ≥ 5 x 10 20 m 3 . T c < 5 eV) play an important role on detachment.

This paper describes the modification program of JT-60 towards the tokamak fusion DEMO-reactor in support of ITER. The JT-60 is planned to be modified for the superconducting tokamak, JT-60SC, to establish steady-state operation at high beta and ensure plasma applicability of ferritic steel as reduced activation materials in a reactor-relevant regime with a break-even class plasma. Main parameters of the JT-60SC device are the plasma current of 4 MA, the toroidal field of 3.8 T, the major and minor radii of 2.8 and 0.85 m, respectively, the elongation and triangularity of 1.8 and 0.35, respectively, a current flattop duration of 100 s and the total heating power of 15 MW for 100 s. Nb 3 Al and Nb 3 Sn superconducting strands with a high Cu/non-Cu ratio and NbTi strands with ∼10 μm diameter of filaments have been developed for the toroidal and poloidal superconducting coils requiring compactness and low AC losses. Full-scale cable-in-conduit cables were manufactured by automatic welding of SS316 conduits and jacketing the cables. An irradiation test for a mockup of the cost-effectively designed CFC divertor target plate has demonstrated the required high heat handling capability of ∼10 MW/m 2 on the JAERI Electron Beam Irradiation Stand.

The radiation shielding, nuclear heating and dose rate have been evaluated for JT-60 superconducting modification. The double-walled structure of a vacuum vessel was adopted so that the maximum nuclear heating at the inboard side of the superconducting TF coil could be suppressed to be lower than 2.5 mW/cm3. The 316 stainless steel (SS316L) boards outside the vacuum vessel are installed for effective gamma-ray shielding. The dose rate with a reduced activation ferritic steel pedestal for the first wall in the vacuum vessel was estimated. This pedestal reduced the in-vessel dose rate by 40% compared to SS316L pedestal. The dose rate outside the boron carbide doped concrete cryostat was estimated at shutdown after 10 years operation.

A flat carbon fiber composite (CFC) tile mock-up with screw tubes, which have helical fins like a nut, was fabricated aiming at further improvement of the heat removal performance of the cost-effectively manufactured divert or target for JT-60SC (modified JT-60 as a superconducting coil tokamak). The heat removal performance of the mock-up was successfully demonstrated on the JAERI Electron Beam Irradiation Stand. The estimated heat transfer coefficient of the screw tube at the non-boiling region was roughly three times higher than that of the smooth tube. This corresponds to 1.5 times that of the swirl tube. A heat cycle test of 10 MW/m2 showed that the mock-up with the screw tubes could withstand for 1400-cycles. These results indicate that the divertor target plate with the flat CFC tile and the screw tube can be a promising candidate for the JT-60SC divertor target.

The positive-ion based neutral beam injection (NBI) system for JT-60, which consists of 14 beamline units and has a beam energy of 70 to 100 keV, started operation in 1986 with hydrogen beams and injected a neutral beam power of 27 MW at 75 keV into the JT-60 plasma. In 1991, the NBI system was modified to be able to handle deuterium beams as part of the JT-60 upgrade modification. After executing some research and developments, deuterium beams of 40 MW at 95 keV were injected in 1996. As a result, NBI has contributed to the achievement of the highest performance plasmas, a DT-equivalent fusion power gain of 1.25 and a fusion triple product of 1.55 x 10{sup 21} keVs/m{sup 3}, in the world on JT-60U.

Fusion plasma performance and confinement studies on JT-60 and JT-60U are reviewed. With the main aim of providing a physics basis for ITER and the steady-state tokamak reactors, JT-60/JT-60U has b...

JT-60 is planned to be modified to a fully superconducting coil tokamak (JT-60SC) to establish scientific and technological bases for an attractive DEMO reactor. A divertor geometry for JT-60SC is designed to achieve enough exhaust capability for high triangularity configuration, which is essential to obtain high plasma pressure and type II ELMs. For the improvement of detachment control capability, vertical targets are applied with baffle plates into which 3?cm SOL flux surface on the outer midplane can be led. Cryopanels are installed under both a private dome and an outer divertor to allow separated pressure control for inner and outer divertor legs, which is important for the detachment control. The preliminary results of the numerical simulation with SOLDOR/NEUT2D code show that the pumping speed of ?60?m3?s-1 at an outer pumping slot would be enough to control the detachment at the outer divertor without an X-point MARFE under the condition of the edge density of 3.2?1019?m-3 and the SOL power flow of qi = qe = 6?MW.

Effluent of tritium in vacuum exhaust in JT-60 through the stack to environment always remains a level below detectable level (

Two types of cable-in-conduit conductors (CICCs) for the JT-60 toroidal field coil (TFC) have been developed. The operational conditions of this conductor are 7.4 T in a maximum magnetic field and 19.4 kA in operational current. Each conductor consists of a stainless steel conduit (17 mm in inner diameter) containing 216 high performance superconducting strand (Nb/sub 3/Al or Nb/sub 3/Sn) and 108 copper wires. The diameter of strand is 0.74 mm including 2-/spl mu/m Cr plating. In order to attain a compact and stable coil, strands with high Cu/non-Cu ratio and high critical current density (Jc) were developed. The Cu/non-Cu ratio of Nb/sub 3/Al and Nb/sub 3/Sn strands are 4.1 and 3.6, respectively. The noncopper J/sub c/ of developed Nb/sub 3/Al and Nb/sub 3/Sn strands were 1914 A/mm/sup 2/ and 1843 A/mm/sup 2/ at 7.4 T, 4.2 K, respectively. Measurements of critical current (I/sub c/) of full size conductors showed that the I/sub c/ of Nb/sub 3/Al and Nb/sub 3/Sn CICCs were 98% and 79% of I/sub c/ estimated from I/sub c/ in strands. These values exceed the target values indicating that the developed conductors are quite appropriate for TFC of JT-60.

The full-size samples of 20-kA Nb/sub 3/Sn and NbTi cable-in-conduit conductors for the superconducting poloidal field coils of the fusion experimental tokamak JT-60 were manufactured. Since these coils operate in pulse mode, reduction of ac losses is one of the key issues. For the Nb/sub 3/Sn conductor, in order to separate the strands sintered with Cr coat after heat treatment, bending strains were applied on it. As a result, ac loss was effectively reduced, and it was found that 0.2% bending strain is enough to realize the design value level of 60 ms. Its critical current was also measured, and it was estimated that the compression strain due to the difference between thermal contractions of the stranded wire and stainless steel jacket was -0.6 %. For the NbTi conductor, comparison between the ac losses of the SnAg- and Cr-coated conductors was carried out. The measured coupling time constant of the Cr-coated conductor satisfied the design requirement of 50 ms, however the SnAg-coated conductor was more than five times that of the Cr-coated one.

At JAERI a plan of remodeling JT-60 tokamak fusion device by using superconducting coil system is being considered to investigate performance of break-even plasma in long pulse operation. The coil system are composed of Toroidal Field (TF), Central Solenoid (CS) and Equilibrium Field (EF) coils. The TF coil is designed to have a maximum field of 7.4 T at winding with a nominal current of 19.4 kA and a magnetic stored energy of 1.7 GJ which is larger than that of every superconducting coil constructed so far. The coil uses a cable-in-conduit conductor that a Nb/sub 3/Al (or Nb/sub 3/Sn) strand cable is inserted into a square stainless steel conduit with a circular hole and is wound with a react-and-wind technique. The design operating temperature is 4.6 K. The CS which consists of 4 modules is operated with a maximum field of 7.4 T and a maximum ramp up rate of 1.0 T/s. For the conductor, a square Nb/sub 3/Sn cable-in-conduit conductor with stainless steel conduit is used. The EF coil consists of 6 coils that the maximum size coil has an outer diameter of 11 in. The EF4 coil which has a maximum field of 7.4 T and a ramp up rate of 2.2 T/s uses the same conductor as CS and other field coils use a NbTi cable-in-conduit conductor.

The 500 keV negative-ion based neutral beam injector for JT-60U has been operated since 1996 and has injected so far a neutral beam power of 5.2 MW at 350 keV for 0.7 s. In order to further increase the beam power and extend the beam pulse duration time, a few issues concerning the negative ion source have to be solved. The most serious issue of them is non-uniformity of source plasma in the arc chamber. Two out of some countermeasures for improving the non-uniformity have been tried. The first is changing the spatial distribution of the arc discharge through regulating arc current resistors which connect in series to each of eight filament groups. The second is masking both ends of the plasma grid with thin molybdenum plates to block beam extraction at the edges. Owing to these countermeasures, the acceleration efficiency of the deuterium negative ion beams in the accelerator has improved by more than 30% from 0.55 to 0.74.

The JT-60 facility is now planning to replace the existing machine with a superconducting machine (JT-60SC) in order to proceed with advanced studies of steady state plasmas with high performance core conditions close to those of fusion reactor. A maximum plasma current of 4 MA and duration of 100 s are expected. Existing toroidal and poloidal field coils and vacuum vessel will be completely replaced. Since the present JT-60 system has been designed for a plasma current duration of 15 s, the coil and power supply systems need to be reconstructed to ones with a long pulse capability. This paper describes the new JT-60 machine and the design of the coil power supply system.

High current and high voltage DC circuit breakers have already been used in the field of fusion technology, in the Joint European Torus (JET) (P. Dokopulos and K. Kriechbaum, DC Circuit Breaker Arrangement 73 kA, 24 kV for the Joint European Torus, Elektrotechnische Zeitschrift A, 1976; P.L. Mondino, The JET ohmic heating circuit, 9th Symposium on Engineering Problems of Fusion Research, Chicago, US, Vol. 1, October 1981, pp. 386–389) and JT60 (R. Shimada et al., JT 60 power supplies, Fus. Eng. Des. 5 (1987) 47–68; R. Shimada et al., Development of high current DC circuit breaker for large Tokamak device, Fusion Technol. 1 (1978) 401–406) Tokamaks, with pulsed currents up to 90 kA or more, but were not able to simultaneously carry steady state currents of 60 kA and to interrupt current under high voltage. Therefore, RD M. Huguet, The Integrated Design of the ITER Magnets and Their Auxiliary Systems, 17th IAEA Fusion Energy Conference, Yokohama, Japan, October 1998) in case of a quench (Y. Shimomura et al., ITER Overview, 17th IAEA Fusion Energy Conference, Yokohama, Japan, October 1998). For this purpose, the coil magnetic energy (in the range of several GJ) is discharged in a large resistor connected in parallel with the DC circuit breaker. Most of the coil circuits operate at 60 kA maximum DC current. Nevertheless, the Central Solenoid (CS) coil (Y. Shimomura et al., ITER Overview, 17th IAEA Fusion Energy Conference, Yokohama, Japan, October 1998) of the ITER Tokamak was designed, till the middle of 1998, to be operated at 168 kA. The assessment of such a circuit breaker concept, using parallel operation of 60 kA rated mechanical switches, will be developed in this paper. This theoretical analysis assumes the use of pure mechanical switches connected in parallel without any use of semi-conductor devices.

Some preparations for pulse operation tests of the ITER CS model coil in the JT-60 power supply system have been performed. Since the ITER CS model coil takes a longer current ramp-up time, the pulse duration is extended up to 70 s from 15 s by decreasing the output voltage of the motor-generator. In the protection system a new grounding resistor with higher resistance has been installed and a new protection make-switch has also been developed. An analysis for the local resonances in the coil has been performed, and measurements of the coil impedance is planned to confirm the validity of the analysis results.

The JT-60 project concentrates research efforts to contribute to the ITER EDA and to establish the basis of a steady-state tokamak reactor. Negative shear, as well as high-βp-mode, are the major operation schemes in the recent JT-60 experiments. Equivalent-QDT=1.05 with a negative shear discharge and the fusion triple product nD·τE·Ti=1.5×1021 m−3·s·keV with a high-βp H-mode have been achieved. A high recycling and strong radiative boundary plasma has been obtained in negative shear high confinement discharges. Long sustainment of high performance discharges has satisfied the conditions required for a steady-state tokamak reactor concept such as SSTR. JT-60 started operation in 1985 and had a significant modification to JT-60U in 1989–1990. The most frequent machine troubles occurred due to control software bugs, air-leaks of the vacuum vessel, and electric insulation breaks in the tokamak machine. Machine behaviors beyond our expectations have been the strong membrane vibrations of the vacuum vessel in the original machine JT-60, the cooling water leaks from the TF coils, divertor tile breaks due to halo current, and the cooling pipe damages due to high energy electrons penetrating under the divertor tiles. These should be considered the important engineering issues even with regard to future fusion research.

As an R&D task of the ITER engineering design activities (EDA), we carried out the irradiation tests on the basic materials of transmission components such as ceramics, windows, fiber optics and mirrors, and in-vessel diagnostics sensors such as magnetic probes and bolometers. Ceramics will be used as general insulating materials. Here we investigated the radiation induced conductivity (RIC) of ceramics for 14 MeV neutrons. Optical elements are the most sensitive to radiation damages among the transmission components. Radiation induce transmission loss of window materials and fiber optics have been measured in the Japan material testing reactor (JMTR). Off line irradiation tests were carried out for molybdenum corner cube reflector (CCR) and gold coated mirrors in JMTR also. Fourteen MeV-neutron-induce luminescence of window material was evaluated in the fusion neutronics source (FNS). Bolometers should be installed inside the vacuum vessel to measure the radiation losses in the infra red to soft X-ray range. We demonstrated the performance of the JT 60 type bolometer under 60Co gamma-rays.

The JT-60 Super Upgrade (JT-60SU) is an upgraded tokamak device of JT-60U for developing the steady-state reactor and advanced tokamak operation in the International Thermonuclear Experimental Reactor. The device is planned to utilize the JT-60 facilities fully and to minimize the needed modification. The major radius is 4.8 m and the maximum plasma current is 10 MA. Neutral beam injection with 750 keV beam energy is the primary heating method. The machine is capable of steady-state operation with high density up to 8.8 × 10 19 m −3 at 5 MA plasma current. The high operating density, over the Greenwald et al. limit, is critically important in order to achieve high bootstrap current fraction. Ballooning mode and low n ideal magnetohydrodynamic (MHD) mode including the bootstrap current were analyzed for steady-state operation. The current profile must be optimized to obtain a normalized beta up to 3. The plasma configuration with high triangularity was adopted in order to get good MHD stability and high energy confinement. A compact divertor was designed in order to get the large plasma space.

日本原子力研究所の大型トカマク装置JT-60は, 1996年秋,大きな目標であった臨界プラズマ条件を達成した。これは,負磁気シアモードと呼ばれる定常核融合炉に適した運転方式で得たものである。この方式は,原研が提案した動力炉概念を契機に,その後も実験および理論の両面において,原研が世界的に研究をリードしてきた定常核融合炉構想から生まれた。本稿では,これらの炉心プラズマにかかわる実験および理論の最近の進展と核融合炉開発への意義,展望を概観する。

Ripple enhanced banana drift loss of fast ions is studied on the basis of heat load measurements at the first wall near the outer equatorial plane in JT-60 U. Hot spots in the ion cyclotron range of frequency (ICRF) and neutral beam injection (NBI) heating are observed at the outboard wall by using an infrared camera. In ICRF heating, the total heat load to the wall is around 5% of the injected ICRF power when the plasma-wall clearance gap is less than 0.1 m. The heat flux on the wall increases with the ICRF power and decreases strongly with the plasma density, which is consistent with Fokker-Planck calculations. In NBI heating, hot spots are observed at almost the same positions, and the heat flux and position agree with the banana drift loss as predicted by an orbit following Monte Carlo code

Damage to a TiC coated Mo divertor plate used in JT-60 for about two years was re-examined from a material viewpoint to obtain data concerning utilization of high- Z plasma facing components in future fusion reactors. Serious damage was observed on the edge of several divertor plates located next to a toroidal gap. Due to very high heat fluxes a apart of the edge was melted and a large crack of about 12 cm wide and 2.5 cm deep was caused near the melted edge along a lamination of the Mo plate. Recrystallization by the heat loading embrittled the subsurface region and resulted in cracking and melting at a lower heat loading. The present work suggests that the maximum temperature of the powder metallurgical Mo should be maintained below the recrystallization temperature to prevent material degradation and plasma radiation losses.

Various features of H-mode observed in the JT-60U tokamak have been intensively investigated with respect to the effective edge ion collisionality vi*eff, which takes the impurity contribution into consideration. Cumulated data exhibits vi*eff at the transition does vary, depending upon target density, magnetic field, equilibrium profile and the heating beam power. Scatters in the vi*eff have been categorized into three regions: low field, hot-ion mode and high- beta p mode regime. In all the regimes reduced turbulent density fluctuation was observed across the transition, which is a result of the decrease of coherence in the edge density fluctuation and increase in the radial shearing of the poloidal flow velocity. Reflectometric technique and fast-sample magnetic probes were used for the turbulence diagnostics in this work. It was also documented that turbulent density fluctuation inversely scales with shear in the poloidal flow velocity in the low-field H-mode. Particularly in high- beta p H-mode discharge, the poloidal rotation is in the ion diamagnetic direction, and enhanced magnetic turbulence was observed, which may invoke the theoretical considerations regarding the role of magnetic turbulence as a cause of the H-mode transition.

The recent experimental results on H-mode in JT-60U and JFT-2M are reported. The enhanced confinement mode with the edge transport barrier ( H-mode) in JT-60U is characterized by the high edge ion temperature. The confinement improvement is well correlated with the edge ion temperature. The H-mode at low magnetic field ( approximately 2 T) in JT-60U has similar features to the H-mode in JFT-2M: rapid reduction of Dalpha emission. However, in the H-mode at high magnetic field ( approximately 4 T) and at low density, no clear transition is observed in JT-60U, although the confinement is still improved with the edge transport barrier. Collisionality and the width of the transport barrier are also investigated in both tokamaks, and are compared with Shaing's theory (1989). Improvement of the high-pp mode discharge with the H-transition was demonstrated in JT-60U. Owing to this improvement, the record value of the fusion triple product (1.1*1021 m-3 s keV) has been achieved. Quasi-steady H-mode operation with H-factor of approximately 1.8 has been also demonstrated in JT-60U by controlling the repetition rate of edge localized modes (ELMs). In JFT-2M, ELM activity has been successfully induced by a new and active method using helical magnetic field perturbation, and as a result, a steady H-mode has been achieved.

A review of the results from JT-60 ICRF heating experiments is presented. The JT-60 ICRF experiments (120 and 131 MHz) are characterized by higher-harmonic heating with phase control of the compact 2*2 loop antenna array. Second-harmonic heating is most intensively investigated with antenna phase control for various cases where the resonant species (hydrogen) are either a majority component or a minority component, and with or without hydrogen HBI heating. It is found that antenna phase control plays a key role in optimizing antenna power injection characteristics as well as heating efficiency. The out-of-phase mode (( pi ,0) mode) can avoid RF sheath and parametric decay instabilities which cause harmful effects on antenna power injection capability with the in-phase mode ((0,0 ) mode) is some operating conditions. The incremental energy confinement time ( tau Einc) of the ( pi ,0) mode is about 50% better than that of the (0,0) mode. Minority hydrogen second-harmonic heating with ( pi ,0) phasing in helium discharges shows the best results over a wide range of electron densities (ne=2.5-6.5*1019 m-3) and plasma currents (Ip=1.5-2.4 MA (qa=4.2-2.6)). The best confinement enhancement factor ( tau E/ tau E(L-mode)) is 1.3, which is obtained with ohmic target plasmas. Higher-harmonic beam ion acceleration is observed up to fourth harmonics in a combined ICRF (( pi ,0) mode) and NBI heating scenario. Strong central-electron heating associated with beam ion acceleration is observed in combined third-harmonic ICRF and NBI heating.

In JT-60, hot ion modes (Ti(0) = 9-13 keV, H0 beam into H+ plasma) exhibiting peaked profiles of the ion temperature, Ti(r), and the toroidal rotation velocity, Vt(r) have been obtained with an effective safety factor qeff ~ 10. Experiments have been performed to study the characteristics of these peaked profiles in which qeff was scanned over the range 4.5 < qeff < 15. Peaking of the profiles of Ti(r) and Vt(r) occurs within a limited region. Moreover, the boundary of the region of improved performance is well characterized by the q = 1, 2 and 3 surfaces. The improvement in global confinement in JT-60 hot ion modes is attributed to the substantial increase in the volume of the enhanced confinement region within the q = 3 surface, in which region the peaked Ti profile is generated

Plasma coupling tests of the RF heating system in JT-60 were successfully performed at JAERI prior to the substantial injection for the plasma heating experiment. About 24 MW LHRF at 2 GHz band and 6 MW ICRF at 120 MHz band system were installed in JT-60. The total system is remotely controlled via a mini-computer, which is linked with the central control system named “ZENKEI” and “CAMAC” systems. The RF power and phase difference can be controlled according to a set preprogram and the plant data can be collected and recorded during RF injection. These are carried out at both LHRF and ICRF injections and simultaneous NBI injection. It is shown that this RF heating system can contribute to both LHRF and ICRF heating experiments for JT-60.

Requirements for collision data on helium, beryllium and boron are reviewed in the light of the directions of present and planned tokamak fusion experiments. The occurrence of the atoms and ions of these species and their roles in plasma behaviour and diagnostic measurements are described. Special emphasis is placed on alpha particle detection in reacting plasmas and on beryllium and boron in divertor configuration machines. The atomic reactions required to exploit the species in models and diagnostic analysis are gathered together and their relative importance indicated. The paper is an introduction to the detailed studies of collision cross-sections presented in the other contributions to this volume.

The analysis of particle excursion in the effect of both toroidal and poloidal magnetic field confinement and its energy loss has been performed. The other physical parameters such as, particle gyro radius, velocity, particle excursion, and the loss of energy rate in plasma are calculated numerically by using ITER simulation method. In order to verify the quality of another fusion reactor design, benchmarking to International Thermonuclear Experimental Reactor (ITER), the best design of fusion reactor on earth, has also been performed. This is based on parameter reactor research design that has been performed previously, which is built in U.K. and Japan. The gradient in magnetic confinement region follows the direction of magnetic field curvature. It means that particle will move by the same direction as the magnetic field, but in most cases there will always be particle excursion from the expected result. The magnitude of particle displacement in ITER design is around 1.85 cm in which assumes of the particle is marginally trapped in reactor chamber. In another design, Joint European Torus (JET), this value is around 3.95 cm, whereas the other design comes from Japan, it is named JT-60, in which particle excursion circa 1.7 cm. Almost all of the TOKAMAK design, there is a continuous loss of energy from the plasma which has to be replenished by plasma heating. Its value is depending inversely on the time confinement of plasma. All of the result just represents the simplified Large TOKAMAK model that is used in order to design the plasma fusion reactor by using ITER simulation method.

The direct interception of D− ion beam by the acceleration grid was reduced by modifying field shaping plates (FSPs) in JT‐60 negative ion sources to inject powerful neutral beams for long pulse duration of >10 s. The modified FSPs are designed by a 3D simulation code to properly steer the outermost beamlets that are deflected outward by space charge of the inward beamlets. The measured steering angles were −1.2 mrad and −6.4/−4.5 mrad in the vertical and horizontal direction, which were in agreement with the designed value calculated by the 3D simulation code. The proper steering of the outermost beamlets allowed a significant reduction of the power loading of the grounded grid. The power loading was successfully reduced to an allowable level of 5% with respect to the acceleration beam power.

This paper summarizes major plasma physics obtained during the period of JT‐60 operation for 23 years with special emphasis on research towards steady‐state operation of tokamak. Topics included are observation of large bootstrap current fraction, negative shear scenario, demonstration of efficient beam and EC current drive, discovery of current hole, stabilization of the RWM, discovery of ITB (internal transport barrier).

The stripped electron trajectories in a large negative ion accelerator with multi‐apertures and three acceleration stages, where non‐uniform stray magnetic field is horizontally created, are calculated in the JT‐60 negative ion source by a 3‐D numerical code. The horizontal non‐uniform stray field results in a significant power loss of the stripped electrons in the outmost acceleration channel on the grounded grid (GRG). The power loss in the outmost acceleration channel is more than twice higher than that in the central channel due to the weaker stray field although the total power loading on the GRG is by 25% larger than that by assuming a uniform stray field.

The JT-60 tokamak was modified wherein the original outer-single null vacuum vessel and poloidal field coils, were exchanged with those of large dee-shaped cross sections. The basic dimensions of the new vessel allows plasmas of 6 MA, 70 to 100 m 3 volume and vertical elongations of 1.4 to 1.7. The vessel was designed along a new concept of a continuous chamber in a double skin construction. Change of the gas to deuterium from hydrogen is being prepared and will enable neutral beams of 40 MW. The first wall was designed to utilize carbon fiber composite for the divertor and isotropic graphite for the other regions. The divertor tiles are cooled by water, while the other regions are cooled by conduction to the vessel wall.

The first experiment simulating helium ash exhaust in a DT fusion reactor has been performed in the JT-60 tokamak with {ital L}-mode divertor discharges. He was injected with a 30-keV neutral beam to realize core fueling. Divertor neutral pressures of hydrogen and He increased in proportion to the cube of the electron density of the main plasma. The He enrichment factor was 0.5 at {ital {bar n}}{sub {ital e}}=6{times}10{sup 19} m{sup {minus}3} and increased in proportion to {ital n}{sub {ital e}}. In an {ital L}-mode plasma of a fusion reactor, thermalized {alpha} particles (He ash) generated in the core plasma can be easily exhausted with a pumping speed of several tens m{sup 3}/s.

The transport of pellet fuelled plasmas in JT-60 has been investigated with a predictive tokamak transport code. Peaked density profiles have been observed in pellet fuelled plasmas during the sawtooth free phase. Simulation analysis shows that an inward pinch of ~0.2 m/s at the plasma half-radius and a reduced particle diffusion coefficient of 0.1 m2/s in the central region are necessary to explain the peaked density profile. These particle transport properties yield improved energy confinement in a plasma with a strong particle source in the central region under sawtooth free conditions, i.e. in a plasma fuelled by pellet injection. The dependence of the improved stored energy on the plasma current can be explained by reduced particle diffusion within the q = 1 surface.

The density limit for a series of gas fuelled and pellet fuelled limiter discharges in JT-60 has been studied. With pellet injection into high current/low q (qcyl = 2.1–2.4) discharges, the Murakami factor reaches (10–13) × 1019 m−2·T−1. The values are factors of 1.5–2.0 higher than those for gas fuelled discharges. For pellet fuelled discharges the central density is high, whereas in the outer region (a/2 < r) the electron density is limited to the same level as that for gas fuelled discharges. The density limit is confirmed to be an edge density limit; this can be explained by the power balance in the outer region of the plasma, for which the plasma purity (Zeff), the heating power (Pabs) and the electron temperature are the key parameters. The onset of a disruptive event can occur when Prad(total) ~ 12–80% of Pabs(total), and the disruptive limit of the density can be explained by Pabs and ne2 (r = 50 cm) Zeff (r = 50 cm). The thermal stability in the edge region is weaker for lower (

The JT-60 LHRF launcher was designed with broad frequency capability (1.74 to 2.23 GHz) to inject multi-megawatt, multi-second RF waves into plasmas. The launcher is characterized by 32 long vacuum waveguides of ∼ 8 m without front windows to avoid plasma bombardment of the ceramic windows in the large tokamak. The front of the launcher is of a detachable grill so it can be changed easily to optimize the wave spectrum. Three types of grills have been developed, and one of them employs a multijunction technique to drive the plasma current efficiently. Three launchers have been installed on JT-60, and the total injection power has reached up to ∼ 10 MW. A good coupling of reflection coefficients, ∼20%, has been obtained for yarious plasma configurations. The development and operation of the LHRF launcher on JT-60 are described.

The tokamak, JT-60, was upgraded to double the plasma current to 6 MA by replacing the vacuum vessel and poloidal field coils with those of large D-shaped cross-sections. The vessel is unique since it employs a three-dimensionally curved double-skin construction. The vessel, as well as the poloidal field coil, are nestled very closely to the original toroidal field coil, and minimization of the wali thickness to below 40 mm is inherent in order to ensure that the plasma current exceeds 6 MA. In order to design a thin but strong vessel wali, detailed structural analyses were conducted, including dynamie effects, and a multi-arc-shaped vessel of double-skin construction with a midplane support was found to be the best and was adopted. The poloidal field coils support structures were used for the vessel support. Three-dimensional shell analysis results, including effects from the support condition, are described. Local stresses near the base of ports and support rods were highest, and were confirmed...

The J1-60 neutral beam system has been successfully operated for 4 years under a wide range of operation conditions: beam energy of 30-75 keV, beam pulse up to 6 s, injection power up to 26 MW with hydrogen beams. The maximum injection power of 26 MW was obtained at 73 keV with a two-stage accelerator. In a lower energy beam injection with a single-stage accelerator, the beam power at 38 keV reached 18 MW. The system could routinely inject a nominal power of 20 MW with high reliability. The beam energy could be changed during a beam pulse, e.g. from 40 keV to 70 keV for 1.5 s. Helium beams were injected with one of the beamlines for a simulation experiment of helium ash, and the injected power was 0.4 MW at 31 keV. The helium beams could deposit in the vicinity of the plasma center column. Helium gas in the beamline was evacuated by SF6 gas condensed cryo-sorption pumps whose pumping speed was about 800 m3/s.

The JT-60 operation has run very smoothly for about 6 years since the first plasma was obtained in April 1985. The total number of shots amounts to about 9000, including both ohmic heating and additional heating shots. For all the equipment, over 900 troubles, including trivial ones, were reported in this period. Several effective countermeasures have been developed to reduce the number of troubles and to raise the operational efficiency. As a result, JT-60 regularly accomplished about 25 heating shots per day.

The electron gun of the 1 MW-klystron for JT-60 was redesigned to suppress the degradation of the withstand voltage due to deterioration with age. The key aims were to reduce the concentration of the electric field near the electrodes by improving the configuration of the electrodes, and to suppress the barium evaporation from the cathode by replacing the non-metal-coated dispenser cathode (S-type) with an iridium (Ir)-coated one allowing a iower operating temperature. The maximum beam operating voltage was inereased from 84 kV to 94 kV and an RF output power of 1.4 MW was generated for 10 sec at 2.17 GHz. Preliminary results of the klystron operation in JT-60 experiments showed that DC breakdown frequency of the improved klystron was reduced to less than 10% of the previous one.