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All-optical gas sensor consisting of graphene covered silicon microring resonator is proposed. Ammonia gas increases the Fermi level of graphene. The gas concentration can be detected by the resonant-wavelength shift by microring resonator.

Integrated optic ammonia gas sensor using graphene covered silicon microring resonator is proposed. Ammonia gas adsorption increases the Fermi energy of graphene. The gas concentration is detected by the resonant-wavelength shift of microring resonator. The NH3 gas concentration from 0.5 to 1000 ppm is shown to be measured by the resonant wavelength from 1551.75 to 1551.27 nm. By applying voltage to the graphene, the Fermi energy is controlled to adjust the sensitivity and the range of detectable concentration.

JT-60 data processing system (DPS) possesses three-level hierarchy. At the top level of hierarchy is JT-60 inter-shot processor (MSP-ISP), which is a mainframe computer, provides communication with the JT-60 supervisory control system and supervises the internal communication inside the DPS. The middle level of hierarchy has minicomputers and the bottom level of hierarchy has individual diagnostic subsystems, which consist of the CAMAC and VME modules. To meet the demand for advanced diagnostics, the DPS has been progressed in stages from a three-level hierarchy system, which was dependent on the processing power of the MSP-ISP, to a two-level hierarchy system, which is decentralized data processing system (New-DPS) by utilizing the UNIX-based workstations and network technology. At present, replacing the MSP-ISP with a UNIX-based workstation is under way as a final phase in building the New-DPS project. This replacement will be accomplished by October 2005. In addition, remote experiment in JT-60 is considered to be an important issue in extension of the collaborative research among the nuclear fusion community. Technical environments for this task have been developed. The participant from a remote site can set discharge parameters with assistance of the dedicated machine operating status-information tools.

A negative-ion-based neutral beam injection (NBI) system, in which large caesium-seeded negative-ion sources are utilized, has operated reliably in the Large Helical Device (LHD) since it was operational in 1998. The injection power of the hydrogen beam has been increased up to 13.1 MW with three injectors. In one injector with modified ion sources utilizing a multi-slotted grounded grid, the injection power reached 5.7 MW with an energy of 184 keV, which exceeds the designed value of 180 keV and 5 MW. Individual control of the local arc-discharge with independently divided arc and filament power supplies is effective in improving beam uniformity in a large negative ion source. The injection duration has been extended to 120 s, during which the LHD plasma is sustained by the NBI alone with a reduced power of 0.2–0.3 MW using one ion source with a cooled plasma grid. These results demonstrate that the injection performance of the negative-ion-based NBI is comparable to that of the conventional positive-ion-based NBI. The progress of the negative-NBI in the LHD is reviewed in detail from the point of view of the improvement of the negative ion sources.

The JFT-2M tokamak has been modified three times during the Advanced Material Tokamak Experiment (AMTEX) program to investigate the compatibility of the low-activation ferritic steel F82H with tokamak plasmas as structural material for future reactors. The ferritic steel plates/wall were installed inside and/or outside of the vacuum vessel to reduce the ripple of the toroidal magnetic field step by step through three modifications. This paper focuses on engineering aspects of these modifications: electromagnetic analysis to find a suitable way to attach these plates, installation procedure to keep small tolerances, and a three-dimensional magnetic field measurement device used to obtain information of the actual shape of the vacuum vessel used as a standard installation surface. To maintain good surface conditions of the ferritic steel plates/wall that rust easily, careful treatment was executed before the installation. To reduce oxygen impurities further, a boronization system with trimethyl boron, which is safe and easy to operate, was developed.

In JFT-2M, the toroidal magnetic field (TF) ripple was reduced by ferritic insert. Two kinds of ripple reduction were carried out. In the first case, ferritic steel was installed between the TF coil (TFC) and the vacuum vessel, just under the TFCs outside the vacuum vessel. In the second one, ferritic steel was installed inside the vacuum vessel covering almost the whole inside wall. The ripple was successfully reduced in both cases. The temperature increment on the first wall, which indicates the ripple-induced loss of fast ions, was measured by infrared television and was also reduced. The effect of the localized larger ripple was also investigated by attaching additional ferritic steel. A new version of the orbit-following Monte Carlo (OFMC) code was developed including the three-dimensional complex structure of the TF ripple and the nonaxisymmetric first-wall geometry. The experimental results and the new OFMC calculation were consistent.

Compatibility between plasma and reduced activation ferritic steel, which is the leading candidate for the structural material of a fusion demonstration reactor, has been investigated in the Advanced Material Tokamak EXperiment (AMTEX). Ferritic plates (FPs) were installed progressively in the JFT-2M tokamak. The effect of ferromagnetism on plasma production, control, confinement, and stability has been investigated. Impurity release behavior has also been investigated. Even when the inside vacuum vessel wall was fully covered with the FPs and the tokamak plasma was operated close to the wall, no deleterious effect was observed, and the normalized beta could be increased up to ~3.5. Thus, encouraging results are obtained for application of this material to the demonstration reactor.

A low-power test system using a vector network analyzer is constructed to evaluate the performance of phase correcting mirrors in a matching optics unit (MOU). The MOU is used in order to facilitate the good coupling of a millimeter wave beam from a gyrotron into corrugated waveguides. The newly developed diagnostic system allows us to make measurements not only of intensity-, but also of phase-profiles of millimeter wave beams and opens a new range of possibilities for designing and testing high-precision quasi-optical systems. Low-power tests are carried out to evaluate the performance of the newly designed MOU by reversely injecting a Gaussian beam from the output port of the MOU. The output beam from the input port of the MOU is measured with the analyzer. In this inverse process, the injected Gaussian beam should be converted to the real beam pattern radiated from the gyrotron. Using correlation analysis, the measured- and calculated-electric fields are compared in their amplitude- and phase-profiles. Good agreement between the measurement and the calculation is obtained. Therefore, it is judged that the phase-correcting mirrors in the MOU will work correctly in upcoming high-power tests.

Low activation ferritic steel plates were installed in the JFT-2M tokamak to fully cover the inside wall of the vacuum vessel (ferritic inside wall; FIW) as a simulation of the blanket wall of the fusion demonstration reactor. After the physics and engineering design concerns with the ripple reduction, the electro-magnetic force and so on, FIW was installed with a tolerance of a few mm. The ripple reduction was confirmed from the detailed measurement. Tokamak discharges were obtained without changing control system, as predicted theoretically. The reduction of ripple loss of fast ions was clearly demonstrated by both experiments and calculations. Compatibility of FIW with high normalized beta plasma up to 3.3 was demonstrated at relatively far wall position. Preliminary investigation showed that the slowdown of the growth rate occurred with closer wall position, which presumably corresponds to wall stabilization effect.

Data processing system for JT-60 is constructed by mini-computers, microcomputers and CAMACs for the control and the data acquisition of plasma diagnostics. However, it is difficult to apply this system to the remote diagnostics using the internet, because the old microcomputers used for over 15 years has no internet connectivity. Therefore, the data processing subsystem using mini-computers and microcomputers should be replaced, keeping the consistency with the main computer. The workstation based on UNIX-OS is a useful candidate for replacing those computers, since it is excellent in the network communication, the cost and the stability. We therefore the replaced microcomputers controlled and monitored the diagnostic CAMAC system, and executed its data acquisition. The new workstation-based system was simple system, which consisted of the serial highway driver module for controlling CAMAC and CPU module with SPARC CPU in VMEbus. By these replacements, the difficulty in the remote control has been solved by using the network and screen control of X Window. The maintainability is also improved and the function enhancement becomes easier. In this paper, basic design for replacing mini-computer and microcomputer with the UNIX workstation will be described.

Compatibility between the plasma and low activation ferritic steel, which is a candidate material for fusion demonstration reactors, has been investigated step by step in the JFT-2M tokamak. We have entered the third stage of the Advanced Material Tokamak EXperiment (AMTEX), where the inside of the vacuum vessel wall is completely covered with ferritic steel plates ferritic inside wall (FIW). The effects of a FIW on the plasma production, impurity release, the operation region, and H-mode characteristics have been investigated. No negative effect has been observed up to now. A high normalized beta plasma of βN ~ 3, having both an internal transport barrier and a steady H-mode edge was obtained. A remarkable reduction in ripple trapped loss from 0.26 MW m−2 (without ferritic steel) to less than 0.01 MW m−2 was demonstrated by the optimization of the thickness profile of FIW. A code to calculate fast ion losses, taking into account the full three-dimensional magnetic structure was developed, and values obtained using the code showed good agreement with experimental results. Thus, encouraging results are obtained for the use of this material in the demo-reactor.

To measure the electron temperature profile of tokamak plasma in a high confinement mode with edge localized mode (ELMy H-mode), a method of data analysis is presented on the fast-scanning Fourier transform spectrometer (FTS), and results in the JT-60U tokamak are described. In the ELMy H-mode, where the FTS interferogram of electron cyclotron emission (ECE) becomes incomplete due to nonthermal emission pulses of ELM, the spectrum of ECE is estimated by using the maximum entropy method (MEM) of nonparametric type. The optimization of the regularization parameter is studied in terms of adopting the generalized cross validation (GCV) from the linear regularization of Tikhonov–Phillips. A study is also performed on a method to remove the nonthermal emission pulses from interferogram data before the MEM analysis, and on an extension of minimum cross entropy in order to meet with the missing of interferogram values in the region of zero path difference. From the application to experimental data of many shots, it is found that the electron temperature profiles can be automatically measured using the FTS during the ELMy H-mode operation of the tokamak.

Energetic particle deposition to the wall due to toroidal magnetic field (TF) ripple was assessed for a 2 GW fusion power reactor. When the present allowance for the loss is applied, the alpha particle flux to the wall can be as high as 2×10 18 m −2 s −1 in the reactor, eroding tungsten by ∼20 μm per year. The peak particle fluence over a 2-year operation cycle can reach 10 26 m −2 , probably being larger than a critical fluence for blister formation. The result suggests that, for the steady-state tokamak fusion reactor, we should introduce a new design methodology of determining an acceptable level of TF ripple on the basis of particle fluence to the wall, instead of the present one based on a tolerable heat flux.

Compatibility of a low activation ferritic steel, F82H, with plasma has been investigated in the JFT-2M tokamak with three stages. The tokamak experiments in the third stage have just started, where the vacuum vessel is fully covered with the ferritic steel. Prior to the tokamak experiment, vacuum properties of F82H were investigated using a test-stand. The outgas rate is less than 6×10 −8 Pa m 3 /sm 2 after baking at 120 °C for 10 days. As is predicted from the results, the base pressure of 6×10 −6 Pa, which is the same as the previous level, was obtained in JFT-2M. In spite of the magnetic effect, the Taylor discharge was obtained without any difficulties and removal of oxygen was clearly observed. Tokamak discharges were obtained with the ferritic steel wall. The improved confinement mode (H-mode) was obtained with a neutral beam of 800 kW with a single-null divertor configuration. Within the preliminary results, the radiation loss did not increase due to the ferritic steel wall.

The compatibility of the low activation ferritic steel with a fusion plasma has been investigated in the JFT-2M tokamak. The program consists of three stages. In the first stage, the reduction of fast ion losses was well demonstrated by ferritic steel plates (FPs) outside the vacuum vessel (VV). In the second stage, 20% of the inner surface of the VV was covered by the FPs. The plasma control, stability, and impurity release were preliminary investigated. No deteriorative effect on the plasma was observed at least in the following conditions: partial covering of 20% and the normalized beta value less than 2.8. First boronization was applied to JFT-2M leading to a remarkable decrease of the oxygen impurity. After the boronization, plasmas with the highest normalized beta in JFT-2M were obtained. Thus encouraging results were gained for this stage. In the third stage, the VV was fully covered by FPs, where the ripple reduction and the plasma stability will be investigated as a full scale testing.

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...

Ripple loss reduction resulting in reduced losses of fast ions was demonstrated successfully at the JFT-2M tokamak by installing ferritic plates between the vacuum vessel and toroidal field coils. This paper describes the comparison between the experimental results and computation based on the orbit following a Monte Carlo code. They are in qualitatively good agreement in the cases where the toroidal distribution of the toroidal field ripple is almost periodic (e.g. ferritic plate installed resulting in a complex periodic structure of the magnetic ripple), but differ in the case of toroidally small average ripple exhibiting non-periodic large local ones (e.g. ferritic plate installed resulting in a more complex non-periodic structure of the magnetic ripple).

Research activities of the Japanese tokamaks JT-60U, JFT-2M, and TRIAM-1M are described. The recent JT-60 program is focused on the establishment of a scientific basis of advanced steady-state oper...

A folded waveguide (FWG) antenna was used in the ion cyclotron range of frequency (ICRF) on the large helical device (LHD) in the National Institute for Fusion Science. The FWG antenna is a waveguide antenna folded several times in order to make its size smaller. The FWG antenna in the LHD is designed so that slow waves are excited preferentially. It was used in the fourth experimental campaign in 2000-2001 for the purpose of plasma production and plasmas with an average electron density up to 3.0×1018 m-3 were obtained. This will be a high enough density for initial plasmas of neutral beam injection or ICRF to obtain plasmas with higher densities and temperatures. This is the first demonstration of the utility of an FWG antenna in magnetic confinement devices. Further investigations were made in order to understand the mechanism of plasma production. The maximum achievable plasma density increased with injection power and gas-puffing rate, and became saturated. The density became higher as the magnetic field strength was increased. Such experimental observations were explained by the wave accessibility conditions of a shear Alfven wave.

The JT-60 data processing system is a large computer complex, gradually being modernized by utilizing progressing computer and network technologies. There were three major changes in our system recently. A VME-based fast data acquisition system (FDS) has been developed to enlarge a transient mass-data storage system (TMDS). It can acquire the data of 6 MB per channel with 1 or 5 μs sampling. Now there are three FDSs with 29 channels. The TMDS, which was composed of a mini-computer and 61 channels of 6 MB memory modules with 5-μs sampling, has been replaced with a new system based on the technology of the VME-based fast data acquisition system. One of the outdated auxiliary crate controllers with micro-processor (ACM), which controls CAMAC modules in the crate on the serial highway, acquires data from them and transfers data to the main computer, has been replaced by a workstation with a VMEbus byte-serial highway driver. The combination of a workstation and a VMEbus byte-serial highway driver has been developed and used to replace many mini-computers in our system. A workstation with a digital linear tape (DLT) library has been introduced as final data storage. The conversion of a few terabytes of raw data tapes for a cartridge tape library (CTL), which is attached to a mainframe computer, Fujitsu GS8300, will be started by using data conversion software prepared to handle mainframe data on the workstation.

A 20-channel electron cyclotron emission (ECE) detection system for the grating polychromator in JT-60U has been developed to measure electron temperature and its perturbations. Indium antimonide (InSb) detectors aligned in a row are cooled by liquid helium held by a 50 cm-diam cryostat, in which a hold time of the liquid helium reaches 7 weeks. Twenty-channel preamplifiers are dc-coupled ones, which enable to measure low-frequency ECE. From the comparison between the new system and the existing 20-channel detector system, where a refrigerator for cooling InSb detectors and ac-coupled preamplifiers are used, it is found that the signal-to-noise ratio is improved, and that small perturbations can be detected. By using this system, the electron temperature and its perturbations near the internal transport barrier in a reversed shear discharge are measured in detail.

In the JFT-2M tokamak, discharges exhibiting good performance have been produced by counterNBI after boronization. An improved core confinement mode under an H-mode edge condition is sustained for -5~~E With H89p - 1.5 and pN - 2.0 at a line-averaged electron density of around 70%-80% of the Greenwald density limit. It is found that the q-profile changes from a monotonic one having qo ~ 1 to a zero or weak negative central magnetic shear configuration having I < qmin ~ qo < 3 during improved core confinement mode using counter-NBI under an H-mode edge condition. The central electron temperature very clearly increases when the sawtooth activity disappears, while the previous result without boronization exhibited a significant degradation of the energy confinement due to the accumulation of impurities. The mechanism governing the improved core confinement mode including the density peaking and/or ITB formation due to negative shear is presently unclear.

In the medium size tokamak JFT-2M, ferritic steel plates (FPs) were inserted in order to reduce the toroidal field ripple, which caused fast ion losses. A 2D infrared TV (IRTV) system with time and spatial resolutions appropriate for measuring the first wall temperature increment caused by the ripple ion losses was developed. An IR thermal-imaging camera was used in the present study; this device provided a field rate of 60 fields/sec and a detectable temperature range of 0-500oC with a resolution of 0.2oC. The PtSi detector, sensitive to 3˜5 µm IR radiation, was arrayed in 256 × 256 pixels. The optical system consisted of an IR-lens, a turning mirror, and a sapphire vacuum window with a distance from the camera position to the target wall of 3.5 m, resulting in spatial resolution of ˜3 mm. The target wall consisted of ˜9 cm × ˜15 cm carbon tiles. By using this system, local hot spots due to the ripple-trapped and banana drift losses were observed when fast ions were produced by tangential NBI (36 kV, ˜0.5 MW). The peak temperature increment reached ˜75oC and ˜150oC on the ripple-trapped and banana drift loss regions, respectively. In the optimized condition with FP insertion, the temperature increment was reduced to an almost negligible level for the ripple-trapped loss regions. The corresponding increment for the banana drift loss regions was also minimized. These IRTV data clearly demonstrate the efficacy of FPs for the reduction of fast ion losses.

Application testing of the low activation ferritic steel to plasma is in progress in the JFT-2M tokamak. The toroidal field ripple reduction with ferritic insertion between the vacuum vessel and the toroidal field coils (1st stage) was successfully demonstrated. So far, no deteriorating effects of ferritic boards (FBs) inside the vacuum vessel has been observed in the pre-testing on compatibility with plasma (2nd stage). The density limit was improved by more than a factor of 1.6 after boronization with inside FBs. Design and preparation works are in progress for the testing on compatibility with plasma (3rd stage), where inside wall of the vacuum vessel will be fully covered with ferritic steel.

Magnetic effect of ferritic steel board (FB) on the plasma was experimentally investigated in the JFT-2M tokamak. Clear deterioration by the error field was not observed even in the condition of toroidally unbalanced placement of FBs outside vacuum vessel (VV). Deteriorating effects were not observed with the FBs inside VV for the operational region and the control of plasma, although slight discrepancy in the relation of vertical-coil current and equilibrium was found. Compatibility of FBs with high normalized-beta plasma (β N up to 2.7) was demonstrated after boronization.

In order to investigate the effect of polarization change on measurements of electron temperature profile from electron cyclotron emission (ECE) in a tokamak plasma, we have calculated the spatial evolution of the electric field of ECE propagating in the plasma. In the case of almost on-axis propagation, the Boyd's prediction is a good approximation, that is, the electric vector rotates with magnetic field. On the other hand, along horizontal sight line far from the axis the Boyd's prediction has a discrepancy of 3% from the numerical results. The effect of the polarization change is important for the measurement of ECE imaging in large tokamaks.

An electron cyclotron emission (ECE) measurement system has been developed to obtain electron temperature with high time and spatial resolution for JT-60U plasmas. The system consists of three different diagnostic systems. These systems have different advantages and have been used complementarily. Several kinds of calibration are periodically carried out in the systems to secure reliable measurements. In order to reduce the transmission loss, a corrugated waveguide transmission line was installed and the reduction of the loss of ca. 3 dB was attained. As preliminary applications toward the operation of future fusion reactors, the feedback control of electron temperature profile and the automatic determination of the sawtooth inversion radius with an adaptive neural network were successfully demonstrated.

In order to test the effect of the toroidal field ripple on the fast ion losses, ferritic steel boards were inserted between the vacuum vessel and the toroidal field coil in the JFT-2M tokamak. The experimental and computational results show that the ripple amplitude is reduced from 2.2 to 1.1%. The ion losses are monitored from the increase in the wall temperature measured by the infrared TV. The region of the ripple trapped ion losses moves to outer side by about 6 cm and the increment of the wall temperature due to the ion loss of ripple trapped and banana drift is reduced. The ripple loss of the fast ions is reduced by ferritic steel insertion for the first time in the world. No deleterious effect of the ferritic insertion on plasma production and plasma control has been observed so far.

The JT-60 data processing system is a large computer complex and gradually modernized by utilizing progressing computer and network technology. There are two major changes in our system. A main computer of FACOM M-780 has been replaced with compatible GS8300 using state-of-art CMOS technology, which results in lower power and space usage with nearly the same performance. Now it can handle ∼500 MB of data per discharge. A gigabit ethernet switch with FDDI ports has been introduced to cope with the increase of handling data. The switch will connect a tera-byte (TB) data server at the bandwidth of a gigabit per second with the main computer and many data acquisition workstations. Other developments in our system are the realization of three workstation-based plans, the TB data server, the VME-based fast data acquisition system and a CICU. The TB data server is basically a UNIX workstation with ∼100 GB RAID disks and ∼900 GB MO auto-exchangers. The VME-based fast data acquisition system has been developed to enlarge the present TMDS. The CICU, which has a function of interfacing the main computer with the CAMAC system, has been replaced with the workstation-based system after the fine tuning.

To examine the effects of ferritic steel on error field, ripple reduction and plasma properties (especially for improved confinements), Advanced Material Tokamak EXperiment (AMTEX) is planned for JFT-2M. The AMTEX program consists of the test of toroidal field ripple reduction with ferritic inserts and that of the compatibility between ferritic steel and the tokamak plasma. To reduce fast ion losses due to toroidal field ripple, the ripple amplitude and its toroidal mode number have to be reduced. The guideline of the design for the reduction in ripple amplitude (δ) by inserting Ferritic Boards (FB) is as follows; the wider and thicker FB should be positioned far from the plasma and be located just under the toroidal field coil. Elimination of a central part of the FBs is effective in reducing a second harmonic ripple amplitude. Since the ferritic steel easily saturates by the toroidal magnetic field, the δ depends on the toroidal magnetic field. The JFT-2M Vacuum Vessel (VV) will be remodelled using ferritic steel for the plasma-matching test. Since the relative magnetic permeability of the ferritic steel is around two, the effect of the ferritic steel VV on plasma equilibrium is rather small. A simple port design can reduce any ripple amplitude of the fundamental mode but cannot reduce that of higher modes.

The JT-60 data processing system was originally a large computer complex system including a lot of micro-computers, several mini-computers, and a mainframe computer. Recently, several improvements have been made to the original system to modernize the system. Many sub-systems composed of aged mini-computers have been replaced with workstations by utilizing recent progress in computer and network technologies. The system can handle ∼300 MB of raw data per discharge, which is three times larger than the amount in the original system. These improvements have been applied to develop element technologies necessary to the remote participation in JT-60 experiments. A remote diagnostics control and monitoring system and a computer system for access to JT-60 data from the Internet are used together with video conferencing systems for a real-time communication. In 1996, the remote participation based on them was successfully demonstrated in collaboration with Japan Atomic Energy Research Institute, Los Alamos National Laboratory, and Princeton Plasma Physics Laboratory.

The importance of the relativistic frequency down-shift and the optical thickness is pointed out in the determination of the electron temperature profile from electron cyclotron emission (ECE) measurement in the medium electron temperature region. These effects cause a pseudo radial displacement (Δ r ) of the obtained electron temperature profile. The pseudo displacement depends on only three parameters (the electron temperature, T e , the optical depth, τ, and the major radius, R ). In order to correct the radial position in any tokamak, a scaling law of the displacement is derived as Δ r ( m)=0.0009 R (m) T e (keV) (1+40/(5τ)). This scaling law is applicable to the plasma with τ>5 where the error is less than 0.5% of the major radius.

In order to obtain the time evolution of the electron temperature in JT-60U plasmas, an electron cyclotron emission measurement system with fast time (of the order of microseconds) and high spatial (of the order of centimetres) resolution has been developed. The operation and data acquisition of the system and preliminary data analysis are made automatically. For precise measurement of the electron temperature profile it is important to correct the shift due to relativistic down-shift frequency in present-day tokamak plasmas.

The effects of relativistic down shift frequency on the determination of the electron temperature profile from electron cyclotron emission (ECE) in a large helical device(LHD) have been studied by a numerical calculation. since the value of toroidal magnetic field has maximum at the magnetic axis along the sight line in the LHD, the emissivity at the magnetic axis decreases due to the relativistic effect. We found an apparent drop in electron temperature around the magnetic axis due to the relativistic effect. The parametric dependences of the relativistic effect on electron temperature measurement have been studied. When the electron temperature is higher and/or optical thickness is thinner, the gap from the nonrelativistic electron cyclotron frequency is wider. In the case of Te(0)=5 keV and ne=1×1019 m−3, the frequency gap is about 0.5 GHz. Calculation shows that the electron temperature at the central region (ρ

Toroidicity-induced Alfven eigen (TAE) modes are observed during minority-ion cyclotron resonance heating (ICRH) in the JT-60U. The toroidal mode numbers of TAE modes are identified as 7, 8, 9, 10 and 11 from the Doppler shift in the TAE modes with scanning toroidal rotation at a plasma current of 3 MA. The toroidal mode number of TAE modes tends to increase during a giant sawtooth by ICRH with a decreasing safety factor for the central region. The TAE mode number increases with plasma current, so that nine TAE modes are observed sequentially during a giant sawtooth at a plasma current of 4 MA, where the maximum toroidal mode number is estimated to be at least 13. There are no Alfven continuum gaps for TAE modes in the safety-factor ranges of i-1/2n

A new electron cyclotron heating (ECH) system has been installed on the Heliotron-E helical device. The gyrotron output 106 GHz TE 12,2 mode is converted into a gaussian beam by the quasi-optical Vlasov convertor. The gaussian beam is coupled to the HE 11 waveguide mode, then transmitted by the corrugated waveguides. The transmitted beam is well focused and launched to the Heliotron-E device. Plasma has been produced and heated by this 106 GHz ECH. The second harmonic breakdown was investigated and compared with the fundamental harmonic breakdown. The improvement of the particle confinement was observed during 106 GHz second harmonic ECH. Compared with the 53 GHz fundamental ECH phase, the averaged electron density was increased by 20% without gas puffing. The outflux to the divertor plate was reduced.

An edge electron cyclotron heating (ECH) experiment has been performed on the Heliotron-E helical device using 35 GHz and 53 GHz gyrotrons. Transition phenomena have been observed during the 53 GHz edge ECH. The Hα emission and the outflux to the wall dropped, and then the core electron density increased by 20%–60%. The electron stored energy seems to be improved up to 20% after the transition. This phenomenon is found to depend strongly on the edge heating power, the target plasma density and the limiter position. A possible mechanism is discussed in terms of high energy electrons produced by the edge ECH.

Recent results on the minority ion second harmonic heating on JT-60U are presented. Maximum coupled power reached 6.4 MW. Good antenna-plasma coupling capability and a small fraction (less than 10%) of an incremental radiation loss to r.f. power are confirmed. Power absorption rate increases with increasing r.f. power and is saturated around unity at r.f. powers higher than 3 MW. The sawtooth stabilization by minority ion second harmonic heating was realized over a wide parameter range, i.e. Ip = 0.9 MA−4 MA, q95 = 2.3−8.6, ne = (1.3−5) × 1019m−3 and PIC ⩾ 2.2 MW. A figure of merit VP〈ne〉Ptot for efficiency of the sawtooth stabilization is about 50% higher than those in other devices where fundamental resonance minority ion heating is employed. The longest stable period reached 2.33 s. Attainable sawtooth-free periods scale with the resistive diffusion time. It was found that the energy confinement is further improved by 25% during the reheating phase after the giant sawtooth crash. The electron temperature profile became more peaked at the improved confinement phase. Those phenomena were observed only in low q discharges (q95 ⩽ 2.9).

Singular value decomposition (SVD) was applied to multichannel signals obtained by an electron cyclotron emission diagnostic system in the JT-60U tokamak. As a result, on grating polychromator signals, the equilibrium and perturbative terms of electron temperature profile were successfully separated, and the structure of a magnetic island produced by the tearing mode was revealed. By neglecting the components of small singular values, the noise component of the signals was removed to improve the signal-to-noise ratio while conserving well the pulse-like waveforms. The usefulness of the SVD-based spectral analysis is shown by a typical set of signal waveforms.

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.

Sufficient coupling of an ICRF antenna for high power experiments was obtained even for a wide gap between the separatrix and the antenna in JT-60U. The loading resistances for an out-of-phase mode are over 4 Omega for a gap of 13 cm between the separatrix and the Faraday shield over the wide range of electron density from 1 × 1019 to 5.5 × 1019 m-3. In particular, the loading resistances for an in-phase mode are about 5 Ω for a gap of 33 cm between the separatrix and the Faraday shield for the same plasma parameters. However, the heating response for the out-of-phase mode is much stronger than that for the in-phase mode

For compensation of the magnetically deflected beamlets, a new type of grid system has been applied to a large negative ion source of a negative-ion-based neutral beam injector in the Large Helical Device (LHD) fusion experimental machine. Corresponding to deviation of the negative ion beamlet trajectory due to the electron suppression magnetic field at the extraction grid, the exit aperture of the extraction grid is displaced to cancel the beamlet deflection by electrostatic lens effect. As a result, the beamlet deflection was suppressed less than 2 mrad in a wide range of energy of 110–180 keV, and the port-through efficiency of the injected beam was improved to 90%–95% from 70%–75% in the case of the previous compensation by the grounded grid aperture displacement. The injected beam power was increased to 3.5 MW with an energy of 165 keV using one injector. To improve beam uniformity of the large area beam, individual arc currents through filaments are controlled for uniform plasma production, with ind...

The effects of relativistic frequency broadening on the ECE measurements for a high temperature tenuous tokamak plasma are studied. Calculations of the radiation transfer are carried out assuming a spherically-symmetric relativistic Maxwellian distribution function for electrons. From the relativistic point of view, observation of the fundamental ordinary mode ECE along a major radius from the low field side is best for the measurement of the electron temperature profile. It is impossible to obtain the electron temperature profile using measurements from the high field side without information on the optical thickness along the propagation path.

In the data processing system for the JT-60 tokamak, a unique mass data acquisition system with fast sampling, a transient mass data storage system (TMDS), has been used since 1988. It is composed of a minicomputer and 61 channels of 4/6 MB memory modules with a sampling rate up to 200 kHz and about 300 MB of data are transferred to a main computer by using a special LAN developed by Fujitsu Ltd. TMDS can handle a large amount of data, but cannot be enlarged in its capability, such as CPU power or the number of channels. To solve the problems of TMDS, a new fast VME data acquisition system (FDS), has been developed. It can acquire 6 MB of data per channel with a sampling rate of 200 kHz or 1 MHz and consists of a workstation with VMEbus memory modules. Up to now there are three FDSs with 24 channels. The minicomputer of TMDS has been replaced with a new system based on the technology of FDS. To cope with mass data transfer to a data server, they are connected with a gigabit ethernet switch.

An additional beamline, BL5, equipped with four positive ion sources will be installed on Large Helical Device (LHD) in 2010. The performance of an ion source which generates 80 keV deuterium and 60 keV hydrogen beams was investigated. The structure of the ion source is based on that of a BL4 ion source on LHD. The main differences between the ion sources for the BL4 and BL5 are the acceleration voltages and the materials of plasma electrodes: copper and molybdenum, respectively. The molybdenum plasma electrode for BL5 has better performance than the copper plasma electrode of BL4. The integrated performance of the ion source for BL5 reached a value equivalent to approximately 58 A in the beam current of hydrogen positive ion at 60 keV in the beam energy.

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.

Since 1996, the remote participation in JT-60 experiments has been successfully conducted in collaboration with JAERI, Los Alamos National Laboratory and Princeton Plasma Physics Laboratory. The remote analysis system is a computer complex consisting of an analysis server and a data server for access to JT-60 data. Collaborators of the remote participation can analyze JT-60 data by using this system through the Internet. For the remote diagnostic system, an X Window control program has been customized to communicate via the overseas line using a UNIX workstation at the remote site and one in the JT-60 control room. It has a feature of real-time remote control of diagnostic equipment and remote access to CAMAC data. Both systems are used together with the ISDN-based video-conferencing system for real-time communication between the remote laboratory and the JT-60 control room.

A ripple reduction is made by fully covering ferritic inside wall (FIW) as simulation of the blanket wall. We pay attention on the avoidance of generation of higher harmonic ripple. The n-th higher harmonic mode becomes almost zero in the case that the toroidal width of ferritic steel corresponds to 2/spl pi//(n times number of toroidal field coils). Effective method to reduce toroidal field ripple with FIW is developed. The ripple reduction is confirmed by the magnetic field measurement and the reduction of ripple losses by FIW was successfully demonstrated.

On JFT-2M, to investigate whether low-activation ferritic steel (F82H) can be used as a structural material of the demonstration reactor, the Advanced Material Tokamak EXperiment (AMTEX) is being carried out step by step. The AMTEX program consists of three stages. In the 1st stage, the low-activation ferritic steel plates (FPs) were inserted between the vacuum vessel (VV) and the toroidal field coils (TFCs) to reduce the toroidal field ripple. In the 2nd stage, the pre-testing of compatibility with plasma was carried out by partially installing the FPs in the inside of VV. In the 3rd (final) stage, the testing of compatibility with plasma will be carried out by a full covering of the FPs on the inside of VV.

A modification of JFT-2M was done in order to reduce the toroidal magnetic field (TF) ripple, using ferritic boards (FBs) between the TF coil and vacuum vessel. The arrangements (thickness in minor radius direction, toroidal position and toroidal width) of the FBs are determined so as to minimize the ripple at each TF coil under the restriction of existing ports, diagnostics, and heating systems. The FBs are supported from TF coil cases. The support structures are designed taking into account all coil current patterns of the JFT-2M. The number of the boards can be changed depending on the TF strength. Results of an electromagnetic force test after installation show no deformation of the support structure. The computations about FBs show a reduction of the fundamental mode ripple from 2.1% to 0.8% at (R,Z)=(1.6 m,0.0 m) for a TF strength B/sub t/=2.2 T.