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Insufficient helium exhaust can seriously degrade the performance of any fusion reactor and has recently been identified as a potentially significant issue for the International Thermonuclear Experimental Reactor (ITER). The effects of variations of the ratios DHe/χi, DHe/DD and vHe/DHe and of the sawtooth period on helium exhaust in ITER have been studied with BALDUR code transport simulations. Recent measurements of DHe/χi in TFTR, of DHe/DD in TEXTOR and of vHe/DHe in TFTR, TEXTOR and JT-60 are found to be compatible with the requirements of sustained ignition and helium exhaust for ITER. Ignition is found to be critically sensitive to the ratio vHe/DHe, particularly at the plasma edge, and to a reduced sawtooth period. These critical dependences are moderated for reduced helium recycling at the separatrix and for hollow particle diffusivity profiles.

A four-pellet pneumatic injection system has been developed for plasma fueling of the JT-60. The JT-60 pellet injector is capable of accelerating separately four cylindrical pellets 3.0 mm in diameter × 3.0 mm long for two pellets and 4.0 mm in diameter × 4.0 mm long for the remaining two. The JT-60 pellet injector was installed on the JT-60 tokamak machine at the end of 1988. Obtained pellet velocity was higher than 2300 m/s by propellant gases of up to 100 bar and the pellet fueling efficiency achieved was around 70% for both dimensions of pellets. This paper describes the design, injection operation and performance test results of the JT-60 pellet injector.

The two-electron transfer reaction was applied to the detection of helium ions in the thermal and epithermal energy range in JT-60. Using this technique, the helium ion density and the energy distribution were determined. The ion density and energy distribution obtained were consistent with the results from other diagnostics and from theory. The potential of this technique for measurements of the helium ash density and the energy distribution in fusion plasmas has been demonstrated.

This paper describes an overview of three major magnetic fusion programmes in Japan. The first is the JT-60 machine upgrade (JT-60U) programme at Japan Atomic Energy Research Institute (JAERI). The large tokamak JT-60 is being modified in order to improve the plasma performance. The vacuum vessel and the poloidal field coils are replaced for large D-shape plasma productions. The main machine parameters of JT-60 upgrade are plasma current of 6 MA, plasma volume of 100 m3. The second is the next step programme at JAERI. Design studies are being conducted of the Fusion Experimental Reactor (FER) as a domestic plan and of the International Thermonuclear Experimental Reactor (ITER) under the IAEA. The FER is a DT burning tokamak machine with a fusion power of approximately 400 MW and a 4.7-m major radius. This machine is to achieve adequate plasmas and reactor technologies prior to a Demo reactor. In parallel with the design studies, R&D works such as superconducting coils, negative ion beam and plasma facing component have been carried out. The third is the Large Helical Device (LHD) programme at National Institute for Fusion Science. The LHD programme is a major fusion programme of joint universities in Japan belonging to Ministry of Education, Science and Culture. This is the 7 years' programme, and the LHD is a Heliotron/Torsatron type helical device (R = 4 m, B = 4 T) which contains 2 GJ superconducting helical coils.

Operational experiences on the performance and material behavior of the JT-60 divertor plates are presented. The outside X-point operation using TiC-coated Mo (TiC/Mo) plates was performed with absorbed power up to 20 MW and duration of 1 s without degradation of the plasma performance by sweeping the separatrix to control the surface temperature below the melting point. The use of graphite divertor plates extended the absorbed power up to 25 MW with duration of 4 s in the lower X-point operation. However, the graphite was eroded and redeposited on the divertor plates and the carbon content in the plasma was increased compared to the outer X-point operation with carbon tiles. A few lower graphite tiles were seriously damaged and radioactivated. Improvements in thermal shock resistance and erosion characteristics of the graphite are necessary. High heat concentration at tile edges and disruption-induced heat loads should be avoided in the future devices.

A new Monte Carlo analytical technique is proposed to calculate the pressure distribution in an arbitrary three-dimensional structure. This technique has two features : First, an analytical model composed of complicated three-dimensional components can be built easily by the set theory and coordinate transformations, and second, pressures in the model can be calculated by counting the particles passing through the imaginary walls without interference with the particles. This technique was applied to the calculation of pressure distribution in the JT-60 neutral beam injector (NBI). The analytical results were in good agreement with the experimental ones, and hence the analytical technique was shown to be valid for the calculation of pressure distribution in an arbitrary three-dimensional structure composed of many complicated components.

Pumping characteristics of condensed SF6 layers for He gas have been measured in a small-size cryosorption pump. From the experimental results, He pressure dependence of the effective adsorption conefficient of He on condensed SF6 layers is estimated with the aid of numerical simulation.In the next stage of developing a large-scale cryosorption pump, which is a modification of the cryopump of the JT-60 neutral beam injector (NBI), the experimental results and numerical simulations are used to control the pumping performance of the SF6 layers on pumpmodules. The He pressures measured as a function of the He gas flow rate in the NBI chamber are in good agreement with the numerical calculations.

Results are presented of the second harmonic heating experiments in the ion cyclotron range of frequencies (ICRF) on JT-60. Heating and confinement characteristics are investigated by using the hydrogen second harmonic (H-majority 2ωcH scheme and the hydrogen minority second harmonic (H-minority 2ωcH scheme (nH/nHe ≈0.1) in two toroidal phasing modes of antenna currents, in phase ((0,0)) and out of phase ((π, 0)), at PIC ≤ 3 MW, ne =(1.3−6.6) × 1019 m−3 and Ip = 1–2.4 MA. Efficient plasma heating has been demonstrated in both heating schemes. Intense non-Maxwellian ion tails are observed at low density. Sawtooth oscillations with an enhanced period in the electron temperature are found over a wide range of electron densities. The incremental energy confinement time, τEinc, for (π, 0) phasing is larger than for (0,0) phasing, irrespective of the heating scheme. In particular, H-minority 2ωcH heating with (π, 0) phasing shows excellent plasma heating with τEinc ≈ 100–120 ms, which is larger than that of neutral beam injection (NBI) heating by a factor of about two. Different dependences of τEinc on Ip are obtained in the two heating schemes. The (0,0) phasing data of τEinc in H-majority 2ωcH heating increase with Ip up to 1.9 MA, while the (π, 0) phasing data in H-minority 2ωcH heating are almost constant with Ip up to 2.4 MA. The global energy confinement shows an L-mode behaviour. The energy confinement time for H-minority 2ωcH heating at high density agrees well with the Shimomura-Odajima scaling.

Temperature measurements and inverse thermal analyses of the heat flux on the divertor plates of JT-60 have been performed in both outside X-point and lower X-point operations. These investigations showed that heat load behavior during normal operation is quite different depending on parameters such as plasma density, the shape of the magnetic surface near the divertor plate, polarity of the plate (ion-side vs. electron-side) and the plate materials (TiC-coated Mo vs. graphite). Due to contracting magnetic fluxes near the divertor plate in the outside X-point operation, the peak heat flux on the outside divertor plate was higher than that on the lower divertor plate. Large heat concentration was produced at the edge of the plate due to a small gradient of the magnetic field line incident on the plate, which resulted in serious damage to the divertor plates. Heat deposition due to high energy runaway electrons produced during disruption was observed on the lower divertor plate. Based on these experimental results, heat load and design issues of the JT-60U divertor plate are discussed.

Improved energy confinement for pellet fuelled plasmas on JT-60 is mainly due to the peaked density and pressure profiles inside the q = 1 rational surface, where the confinement characteristics seem to be better than those in the outer (q > 1) region. In centrally fuelled pellet injection discharges, the sawtooth activity can be completely suppressed during 0.4−1 s, or the frequency of the sawteeth can be reduced by up to one order of magnitude during 0.5−1.5 s after pellet injection. For high current, low-q (Ip = 2.5−3.1 MA, q(a) < 3) discharges, reduction of the sawtooth frequency is strongly related with the enhanced confinement and the peakedness of the electron density profile. The contribution of the sawtooth activity to the global energy confinement increases systematically with decreasing safety factor at the plasma surface.

Beam acceleration by heating in the second-harmonic ion cyclotron range of frequency (ICRF) in combination with heating by hydrogen neutral beam injection (NBI) was investigated in the JT-60 tokamak. The energy spectra of the accelerated fast ions were measured by a charge exchange neutral energy analyser whose line of sight was intersected by specific beam lines in the plasma core in order to obtain the required information. The dependences of the tail ion temperature on various parameters (electron density, NBI power and the toroidal phasing of the antenna) were examined. Optimized conditions for beam acceleration were evaluated. For combined ICRF heating and NBI heating, an incremental energy confinement time of 210 ms was achieved, which was three times higher than that obtained with NBI heating alone or with ICRF heating alone. This improvement of the energy confinement during combined NBI and ICRF heating can be explained by a build-up of the fast ions accelerated by the ICRF wave. The scaling of the incremental energy confinement time during combined ICRF+NBI heating was obtained.

This paper reports that in the JT-60 neutral beam injectors (NBIs), an active method using a set of coils is adopted to shield unneutralized beams from stray magnetic fields, while the usual passive method using high-mu materials is used to shield the ion sources and neutralizers. This active shielding method is a unique solution for the unneutralized beams in the JT-60 NBI under the constraints of the limited space available and minimizing the error file induced by the shielding. A passive shielding method is permissible for the ion sources and the neutralizers because the space to be shielded is limited. The active shielding system is designed by making a one-fourth model of the magnetic system and calculating ion orbits using magnetic fields measured in the model. The shielding characteristics are checked by arrays of thermocouples buried in the beam dump where the unneutralized beams are thermalized. The thermocouple outputs are consistent with those predicted from the ion orbit calculations.

Helium ash exhaust characteristics have been studied in the JT-60 lower divertor with ohmic heating (OH) and neutral beam heating (NBH) of PNB ≦ 18 MW and Ip = 1 MA. Helium fueling was done by gas puff. The divertor neutral pressures of hydrogen (H2) and helium were measured by a residual gas mass analyzer. The helium enrichment in the divertor was also measured. In OH discharges with helium gas puff, no buildup of helium neutral pressure was observed and helium was depleted. On the other hand, in NBH discharges with helium gas puff, although enrichment of helium is not observed, the neutral pressures of H2 and helium increase in proportion to . The density dependence of the helium neutral pressures in NBH discharges is promising for efficient helium ash exhaust in fusion reactors.

The behaviour of the heat flux to the JT-60 divertor are investigated in the lower hybrid range of frequency (LHRF) heated and neutral beam heated discharges from the view point of the plasma wall interaction by using an IRTV camera. It is found that the power balance is determined by the distance between the launcher and the separatrix of the plasma with LHRF heating. As the LHRF power increases, the peakedness of the heat flux increases. The parameter dependence of the peaking ratio is investigated for neutral beam heated discharges. It is found that the peaking ratio is a function of the divertor geometry. This behaviour is explained by a simple model that takes into account the response of the divertor plasma to the transport of the heat flux and particle flux.

This paper describes the experimental investigation of the heat load on a lower hybrid (LH) launcher in divertor plasma. It was found that a good coupling of the reflection coefficient, ρ = 15–20%, was obtained by pushing the launcher head a few mm from the first wall even though the distance δ between the separatrix and the launcher head was more than 10 cm. In this condition, the heat load on the launcher head rapidly decreased at δ > 5 cm and was estimated to be about 0.4–0.6 kW/cm 2 at δ = 10 cm, n ¯ e 13 cm −3 during only LH injection. A higher heat load was observed on the ion side of the launcher guard limiter during combined NB + LH heating at n ¯ e > 3 × 10 13 cm −3 when fast ions were observed at the plasma edge due to parametric decay instability. We discuss the implication of these results with respect to LH application on a high power heated plasma.

Outgassing behavior during plasma disruptions in the JT-60 tokamak have been investigated using a newly developed fast pressure gauge. Large amounts of gases are released suddenly in two stages during a plasma disruption. The outgassing peaks coincide with the burst of Hα emission. The first gas release is ascribed to heat flux due to the loss of the plasma thermal energy. The second is caused by the heat flux due to the loss of plasma magnetic energy. Although this outgassing mechanism should be investigated, about three quarters of the released gases then return to the wall.

Remote radiative cooling can significantly reduce the divertor plate heat flux, which may provide a solution to one of the major problems of present and next-step machines. Strong remote radiative cooling of up to 50% of the total input power of 20 MW was observed in JT-60. JT-60 divertor plasmas are dense and cold (nediv ≈ 2.4 × 1020 m−3, Tediv ≈ 26 eV). Remote radiative cooling is a mixture of carbon and hydrogen radiation (+ CX). JT-60 divertor plasmas can be adequately modeled with a simple divertor model with radiative cooling.

This paper describes the upgraded JT-60. The objectives of this upgrade are to improve plasma performance with minimum modification and to obtain physical and engineering database for the design of...

In JT-60, MARFEs have been observed frequently in high- I p and high density limited discharges with neutral beam heating after replacement of the first wall from TiC coated molybdenum tiles to graphite ones. The MARFEs occur near the density limit of the JT-60 plasmas. The NB heating is effective to suppress the MARFE. The empirical scaling of the threshold density for MARFE onset taking account of the NB power is obtained, which is useful to predict the MARFE onset. The radiated power from the plasma with MARFE is about 90% of the absorbed power. Both stored energy and central electron temperatures do not change with the MARFE onset in spite of the such intense radiation loss. So the MARFE does not affect the core plasma. A simple model based on the thermal equilibrium in a peripheral plasma can explain the onset condition of the MARFE.

RF heating system with total power of 24 MW fed by 1 MW klystrons at around 2 GHz have been developed and operated successfully in the JT-60 tokamak. System performance has been demonstrated in a real plasma load. A super high power system has worked stably against changeable plasma load without output circulators. A maximum power input into plasmas reached 11 MW from three array launchers of four row by eight line waveguides, which corresponded to 17 MW generator power.

The energy confinement properties of JT-60 during Ohmic and neutral beam heated discharges are analysed on the basis of the heat pinch model proposed by Callen and Cordey. The feature of temperature profile consistency during variation of the density profile sets an important constraint for the heat pinch term. A heat pinch term proportional to neBp seems to be consistent with the experimental observation. A transport model is proposed which reproduces quite well the parameter dependence of the JT-60 experimental data. The formation of an edge temperature pedestal during the H-mode is observed with this model when the edge convective energy loss is significantly reduced.

The JT-60 LHRF heating system was designed with broad frequency capability to inject a long-pulse-width high-power RF wave into a plasma. The system uses 24 high-power klystrons, each with an output power rating of 1 MW. Five operating frequencies are available within the range of 1.74 GHz to 2.23 GHz. The frequency can be changed at a rate of 150 MHz/0.3 s. This feature provides both plasma heating and plasma current drive. It also permits frequency correlation with the plasma density and current profile. This paper describes both the high-speed frequency tuning system for the JT-60 klystrons and the results of fast frequency change tests of the LHRF over the range of 1.74 GHz to 2 GHz when RF power was launched from the same launcher during a plasma shot.

Particle transport in JT-60 was investigated with a simple perturbation method. From an analysis of the sawtooth induced density perturbations, it was found that the particle diffusivity has an inverse dependence on the electron density and that the measured diffusion coefficients correspond to the values obtained from the analysis of pellet induced perturbations. Also, from an analysis of the density and the heat pulse propagations, the ratio of the electron thermal diffusivity and the particle diffusivity, χe/D, was estimated to be about four. These experimental results were compared with the drift wave transport models.

Recent lower hybrid current drive (LHCD), and heating (LHH) experiments on JT-60 are reported. The current drive product of neRpIRF approximately 12.5*1019 m-2 MA was achieved at the LH power of approximately 4.5 MW, and the CD efficiency, the energy confinement, the global power balance and the heat load on divertor plates were investigated in high power LHCD plasmas. Nearly steady state H-mode discharges were found during LHCD with two different frequency injections. Sawtooth suppression in NB heated plasmas by LHCD have shown an improvement in confinement near the plasma center. Parametric instabilities in LH heating experiments were significantly reduced by increasing the plasma current, and the stored energy increased linearly with heating power of up to approximately 9 MW at ne approximately 7*1019 m-3 and Ip=2.75 MA. Parametric instabilities near the plasma edge in the ion heating regime were also reduced in peaked density plasmas produced by pellet injection and LH waves increased the central plasma pressure at ne(0) > 1.4*1020 m-3.

Module-type flat-field grazing-incidence spectrographs with holographic gratings and multichannel detectors for large TOKAMAK (JT-60) plasma diagnosis are developed. The spectrographs cover the different wavelength regions from 0.5–122 nm, and are set to measure impurity lines in the plasma every 20 ms with space resolution of 7 cm. The flat-field imaging properties with designed wavelength resolution were confirmed, and results of tokamak plasma measurements proved the value of these spectrographs for plasma diagnosis.

It was planned to move a stub rf terminator during rf injection for producing a good impedance matching of an antenna with a few plasma shots in JT-60 ICRF heating experiment. A choke stub was adopted to effect this operation. Rf leakage and high power performance were the major subject in the research and development of the choke stub. For the design of the choke plunger, a simple analytical model was developed to calculate the rf leakage of the plunger. The validity of the model was demonstrated by rf leakage measurements.

Major efforts in the recent JT-60 experiments have been concentrated on the improved confinement of plasmas with profile control and on the steady state operation study. Peaked density profiles were produced with the successive pellet injection. The energy confinement time was improved by 40% as large as that with the gas fuelled discharges. The fusion products n e(0)τET i(0)reached 1.2 × 1020m−3·s·keV, which was twice that of gas fuelled discharges. High-βp, plasmas were obtained in low-I p discharges with improved confinement and a high ion temperature T i, (0) of 12 keV. The bootstrap current reached 80% of the total plasma current at βp=3.2. The new concept of a steady-state tokamak power reactor has been proposed on the basis of this result. The maximum current drive efficiency ηCD of 3.4 × 1019m−2·MA/MW was obtained in the LH current drive experiments. Helium ash exhaust experiments using He-beam injection into H+ plasmas showed promising results for α-particle exhaust in a fusion power reactor.

The energy confinement in high power lower hybrid current driven (LHCD) plasmas has been studied in the JT-60 tokamak. At a plasma current of 1 MA, the diamagnetically estimated energy confinement time in LHCD plasmas has almost the same value as the confinement time in ohmically heated plasmas at e ~ 1.0 × 1019 m−3. The confinement time of high power LHCD plasmas (PLH ≤ 4.5 MW) is longer than that of neutral beam heated plasmas at the same power level. The energy confinement time depends on the LHCD power and on the electron density, but is independent of the plasma current, with α + β ~ −0.3.

In the JT-60 tokamak with nearly perpendicular neutral hydrogen beam injection of up to ~ 21 MW, some regimes of plasma species and beam directions have been observed in which the sawtooth periods are significantly extended. It is found that the use of helium as working gas together with co-injection of neutral beams can have an important influence on the occurrence of extended sawtooth periods. Beam induced stabilizing effects on sawteeth and internal kink modes are also discussed.

In JT-60, H-mode experiments with outer and lower divertors have been performed. In the outer divertor discharge, an H-mode similar to the modes observed in the lower/upper divertor discharges is obtained. Its threshold absorbed power and electron density are 16 MW and 1.8 × 1019 m−3. In the two schemes of combined heating with NB + ICRF and NB + LHRF, H-mode discharges are also obtained. Moreover, in the new configuration with the JT-60 lower divertor, H-mode phases with and without edge localized modes are obtained. The improvement in the energy confinement time in both divertor configurations is limited to values within 10%. The paper mainly presents the H-mode results of the outer divertor discharges. Also, typical results of the lower divertor discharges are shown for a comparison of the H-mode characteristics of the two configurations.

The {ital H} mode was achieved in limiter discharges of a tokamak with lower-hybrid current drive for the first time. Simultaneous application of rf powers at two different frequencies such as 1.74 and 2.23 GHz or 1.74 and 2.0 GHz appeared to be effective in the attainment of the {ital H} mode. The threshold lower-hybrid power was as low as Ohmic-heating power with hydrogen plasmas. A nearly steady-state edge-localized-mode-free {ital H} mode with a duration up to 3.3 sec was established without significant impurity accumulation.

The global energy confinement characteristics of neutral beam heated JT-60 discharges are presented. There is a difference in the dependence of the energy confinement time on the plasma current between limiter and divertor discharges. For limiter discharges, the energy confinement increases with plasma current up to 3.2 MA, whereas for divertor discharges this improvement saturates when the safety factor drops below 3, independent of the location of the X-point. The JT-60 L-mode results indicate that the deterioration in energy confinement for q < 3, which is also found in H-mode regimes of other devices, may be a universal characteristic of divertor discharges. Regarding the scaling with plasma size, it is shown that the global/incremental confinement time increases with plasma minor radius. For sufficiently large plasmas, however, the global/incremental confinement time is no longer a function of minor radius.

A strong ion tail with a temperature of ~ 60 keV was produced in JT-60 by lower hybrid wave acceleration of neutral beam injected fast ions. The incremental energy confinement time due to neutral beam heating of a lower hybrid heated plasma was improved by more than a factor of two and reached approximately 110 ms. The confinement of the tail ions is classical and the improvement of confinement can be fully explained by the contribution of the fast ions. The heating process of the beam ions is consistent with the predictions of the quasi-linear theory.

Operational experience and improvement of klystrons for the JT-60 LHRF heating system are reported. JT-60 klystron is the most powerful and wide band one forfusion plasma heating. Operational experience of these klystrons were accumulated after encountering many troubles to be removed and the stable operation is performed to repair the electron gun in the klystron. These are reported acompanying the data in detail.

Studies on plasma confinement, steady-state tokamak operations and boundary plasmas are main subjects in tokamaks relating to fusion plasmas. Emphases of recent JT-60 experiments from 1987 to the shutdown in October, 1989 were put on these studies. Investigations of L-mode improved modes such as pellet injection, high Ti mode, giant sawtooth by 3rd harmonics of ICRF etc., and H-mode were carried out as confinement studies. Current drive by LHCD made a large progress by using a new launcher. Bootstrap current reached 80% of plasma current, which has brought a new idea of steady-state tokamaks. Results on particle confinement, particle and heat control studies with divertor are also presented

A new gauge with a fast time response has been developed for measuring the vacuum pressure of a tokamak fusion device, JT-60. The pressure is measured during and after a hydrogen plasma discharge shot in order to investigate the mechanism of outgassing from the first wall, which is made of graphite. The evacuated gas quantity after a typical disruption is almost equal to or greater than the amount of gas input. After a typical plasma disruption, the amount of gas of 40 Pam3 is released within 50 ms. This is probably caused by the heat flux to the first wall due to both the plasma thermal and magnetic energies. It is observed that three-quarters of the gas released returns to the first wall. On the other hand, the evacuated quantity after normal plasma discharge is less than the gas input.

In a nuclear fusion reactor with carbon-based plasma facing material, tritium inventory inside the vacuum vessel is one of the critical problems because the reactor operation may be restricted by the tritium which piles up in the carbon tile. From this point of view, exhaust gas analysis was conducted in JT-60 operation and it was successfully demonstrated. Tritium was measured with an ion chamber and water bubbler. Exhausting gas composition and chemical form was measured with a micro gas-chromatograph and a residual gas analyzer. Initial results were obtained for deuterium plasma discharges, helium Taylor discharges and helium glow discharges. Removed tritium from the vacuum vessel in an operating day corresponded to about 8% of the generated tritium during the tokamak discharges in the day. The chemical forms of the removed tritium mainly consisted of elemental gas. Tritium removal rates were compared for glow discharges using H/sub 2/, He and Ar gases. Effectiveness of wall conditioning with H/sub 2/ was confirmed.

Operation of tokamaks has now reached few hundred device-years. Availability is a top level parameter vital for the efficient management of complex plants, like fusion devices, for decision making and to judge the quality of design, manufacturing and operation. There is no standard way to analyze the operating experience: different techniques are adopted, various levels of detail are reached and also the kind of analysis and corrective actions are different. Nevertheless most of the results for the various tokamaks are expressed in similar terms like number of plasma pulses, operating time and delays, experimental sessions lost. An overall analysis of JET, JT-60, DIII-D, ASDEX-U, Tore Supra, FTU and RFX operating experiences has been performed to find out main results, lessons learned and possible suggestions to improve availability of the devices themselves and of those under design or construction, like ITER. The main results are here presented and discussed

In the operation of tokamak such as ITER, it is important to understand the behavior of hydrogen isotopes and other elements in vacuum vessel for tritium inventory control and design of the plasma exhaust process. Particularly removal of carbon-tritium codeposited layer is considered as a crucial issue for ITER, because of its suspected tritium inventory. In JT-60U, exhaust gas was analyzed in the operation of discharge cleaning and tokamak discharge experiments, and the behavior of hydrogen isotopes and carbon were investigated. In discharge cleaning experiments, the degassing rate of hydrogen isotopes in GDC (glow discharge cleaning) was much larger than that in TDC (Taylor discharge cleaning) and ECRDC (electron cyclotron resonance discharge cleaning). Exhausted hydrocarbons were far less than hydrogen. During tokamak discharge experiments, various carbon compounds such as CD4, C2D2, C2D4, C2D6 and CO2 were detected. The maximum amount of exhausted carbon was several mg for each shot. There was a strong correlation between exhausted hydrocarbon and tritium. The exhausted hydrocarbon increased with the wall temperature

The modification of JT-60 is planned as a fully superconducting tokamak (JT-60SC). The mission of JT-60SC program is to establish scientific and technological bases for an advanced operation in an economically and environmentally attractive DEMO reactor and ITER. The research objectives are to accomplish high performance steady state operation with high beta and non-inductive full current drive, with high bootstrap current fraction, and demonstrate the plasma applicability of reduced-activation material for a plasma of break-even class relevant to the reactor plasma. Basic design of JT-60SC has been completed and the detailed design is under way. The engineering design for main components of JT-60SC is described.

Tokamak magnetic fusion research has progressed to an advanced project, i.e., ITER (International Thermonuclear Experimental Reactor). Therefore, existing tokamaks must explore operational scenarios necessary for a high-performance plasma while simultaneously suppressing any disruptive instabilities. To implement this investigation, various advanced real-time controls using processed profile data and all available actuators should be tested in the experiments. To satisfy these requirements at the large tokamak JT-60, we are upgrading its original real-time control system completely. Two criteria for the new system configuration have been adopted: (i) The distributed processors for the measurements, actuators, and supervisory controllers (SVCs) am linked through a "reflective memory (RM)" network for fast, real-time communication. The profile data in the RM modules are also updated in real time. (ii) For efficient parallel and pipeline processes performed by multiple controllers, synchronization is provided from two sources: external hardware clocks and software counters in the RMS, which are updated by the SVCs. For a control function, the following simple and flexible method has been employed to perform various experimental proposals: New test control methods are first registered in the supervisory controllers. In accordance with the preprogrammed waveforms that indicate which control method to use, the SVCs select the correct control method in real time. The upgraded configuration of the JT-60 plasma control system and its effectiveness in seeking advanced operational scenarios applicable to ITER are discussed in detail.

The first experiment on the Nb-Ti pulsed coils, DPC (demonstration poloidal coil)-U1 and -U2, was successfully performed. The coils have a cable-in-conduit conductor cooled by 4-K supercritical helium, and their windings have 1-m inner, and 2-m outer, diameters. Charging tests on the coils were performed in both DC and pulsed modes. In DC mode, the coils were charged individually to their design current of 30 kA, and attained a peak field of 5.2 T and a stored energy of 11 MJ. In pulsed mode, using the power supply of the JT-60, the designed pulsed field of 7 T/s were achieved by charging one of the two coils to its design current in 0.75 s. AC losses of the coil during the pulsed operation were confirmed to be quite small. Mechanical, thermal, and hydraulic properties were also analyzed on the basis of the experimental results. >

Tests of the X2242 were run in the JAERI (Japan Atomic Energy Research Institute) Tokamak-60 (JT-60) ICRF (ion cyclotron range of frequencies) system. The tests evaluated this newly developed tetrode's performance under conditions relevant to the planned CIT (Compact Ignition Tokamak) and JT-60U (JT-60 Upgrade) ICRF systems. The behavior of a large high-power tube at frequencies above 100 MHz was also studied. Complete instrumentation was provided, including an acoustic monitor that senses boiling at the tube anode to permit prompt diagnosis. High-power test results were 1.6 MW for 0.5 s at 131 MHz, 1.0 MW for 10 s at 131 MHz, and 1.5 MW for 6 s at 110 MHz. Performance was studied under load VSWR of 1.5 and 2 over a 0 to 360 degrees phase range. With a VSWR of 1.5, plate dissipation was 1.4 times higher than that of matched load. >

The JT-60 JAERI (Japan Atomic Energy Research Institute) Tokamak-60 neutral beam injector (NBI) has been successfully operated for three years, delivering a neutral beam power of 20 MW at a beam energy of 75 keV. The helium-beam injection into JT-60 was performed to investigate exhaustion of fusion alpha particles from the plasma. The helium-ion beam current extracted from ion sources was 40-45 A at an energy of 31 keV. The injection power was about 0.4 MW. In the helium-beam operation, the helium gas in the beamline chamber was pumped out by a condensed SF/sub 6/ gas cryosorption pump. The JT-60 neutral beam injector will be modified to inject a 120-keV deuterium beam into the JT-60 upgrade, which will start to operate at the beginning of 1991. >

Since the completion of the JAERI Tokamak-60 (JT-60) tokamak machine and the NBI (neutral beam injection) and RF heating systems, various improvements have been made and new auxiliary systems have been installed. The approximately 8000 shots since the first plasma have yielded valuable operation and engineering experiences. Development of a high-speed pellet injector and a multijunction-type LH launcher has contributed to obtaining useful experimental results. The JT-60 will be upgraded by replacing the vacuum vessel and the poloidal field coils for the study of plasmas near the break-even conditions. >

A pneumatic four-shot pellet injector had been in use for the Japan Atomic Energy Research Institute Tokamak-60 (JT-60), contributing to plasma studies in 1988. It could propel the pellets up to 1.6 km/s at a propellant gas pressure of 50 bar. IN 1989, a new gun assembly was installed in the upgraded system to provide higher performance and reliability. The supply pressure of the propellant gas is to be raised to 100 bar to obtain a higher pellet velocity of up to 2.3 km/s. The device is now in use in the JT-60 and is expected to contribute to further plasma studies. The features and performance of the device are outlined. >

In the JT-60 tokamak, the original outer single-null plasma containment vacuum vessel and poloidal field coils, will be exchanged with those of large D-shaped cross sections. The basic dimensions of the new vessel will allow plasmas of 6-MA current, 70 to 100 m/sup 3/ volume, and vertical elongations of 1.4 to 1.7. Installation of a large vessel is made possible at the cost of high toroidal ripple volume in the space occupied by the plasma and at the cost of high electromagnetic force acting on the coils and the vessel. A prime requirement for the modification is maximizing the plasma performance with minimum expense, while ensuring access to the plasma for heating and diagnostics. The plasma configuration is designed to allow the largest D-shape discharge with a lower single null, within the boundary of the original toroidal field coils. The D-shaped vessel is designed according to a new concept of Inconel 625 all-welded continuous chamber in a double skin construction. Rectangular tubes are sandwiched by three-dimensionally curved skins to form a totally everywhere rounded vacuum vessel. The gas is being changed to deuterium from hydrogen, which will allow neutral beams of 40 MW. The lower hybrid and ion-cyclotron systems will be provided with larger antennas. >

Fast pressure gauges were installed in the JT-60 (JAERI Tokamak-60) to monitor the pressure near a plasma. They are Penning configuration gauges. A Penning discharge in the cell is sustained by the magnetic field which confines the plasma. The gauges are controlled by a personal computer. The response time of the system is fast ( >

A multijunction-type launcher was installed on the JAERI (Japan Atomic Energy Research Institute) Tokamak-60 (JT-60). It is composed of 32 modules consisting of 96 secondary waveguides, in which E-plane junctions are used to divide each main waveguide into three secondary waveguides at its grill mouth. The waveguides are equipped with geometrical phase shifters yielding 70 degrees of phase shift in order to launch a peak parallel refractive index (N/sub /// peak) of 1-2.5 at 2 GHz. In the course of fabricating the multijunction launcher, the main technological problem was to fabricate multijunction modules with good RF properties (phase angle, power dividing ratio, etc.). A TIG (tungsten inert gas) welding method was adopted after a study of applicable fabrication methods. The phase shift and the power dividing ratio were verified to be 70 degrees +or-10 degrees and 1/3+or-1/15, respectively. The N/sub /// spectrum of the multijunction launcher calculated from measured values of phase shift was not so different from the design value. Using a test module, the authors also performed a high-power transmission test up to a power level of approximately 700 kW without breakdown. After installation of the multijunction launcher in the JT-60 tokamak, it was quickly conditioned up to a power level of 2.75 MW and maximum current drive efficiency eta /sub CD/ of 3.4*10/sup 19/ M/sup -2/A/W. Experimental results verify the adequacy of the launcher's method of fabrication and its satisfaction of requirements. >