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Japan Domestic Agency (JADA) is responsible for six diagnostic systems in the ITER project. We have successfully developed a prototype instrumentation and control (I&C) system for the ITER divertor thermocouples system. The prototype I&C system includes a supervisory function, a sequencing management function, and a data acquisition function. The supervisory function, which was implemented using EPICS (Experimental Physics and Industrial Control System), manages internal operations for measurement such as health check of sensors. We developed a conversion tool from flowcharts to EPICS record source codes. The EPICS records generated by the conversion tool are used to trigger each operation step and to indicate the progress of the sequence. We also developed the sequencing management function that coordinates the execution of operation steps. Both the flowchart conversion tool and sequencing management function were developed to reduce designers’ mistakes and to ensure rapid development. The prototype I&C system for the DTC was tested. We confirmed that the I&C system performed satisfying the requirements. This validated design will be applied to other ITER diagnostic systems procured by JADA.

In ITER, it is important how the CODAC system conducts many plant systems including diagnostic systems. In order to establish necessary communications between the diagnostics systems and the CODAC system, Japan domestic agency (JADA) has proposed the new concept of supervisory system for the diagnostic system based on our experiences in operating plasma diagnostic systems. The supervisory system manages operation sequences, current state and configuration parameters for the measurement. JADA designed the supervisory system satisfying the requirements from both CODAC system and diagnostic systems. In our design, the tool which converts operational steps described as flowcharts into the EPICS (experimental physics and industrial control system) records source codes is introduced. This tool will ensure reduction of the system designers’ efforts. We designed a communication protocol to configure measurement parameters and proposed configuration parameter validation function. We also analyzed the management of the central/local control mode for the diagnostic systems. The function which selects the adequate limit values and consistency check algorithms in accordance with the conditions of the diagnostics system is proposed. JADA will develop a prototype of the supervisory system and validate the design in 2013.

Vertical displacement events (VDEs) and disruptions usually take place under intervention of vertical stability (VS) control and the vertical electromagnetic force induced on vacuum vessels is potentially influenced. This paper presents assessment of the force that arises from the VS control in ITER VDEs using a numerical simulation code DINA. The focus is on a possible malfunctioning of the ex-vessel VS control circuit: radial magnetic field is unintentionally applied to the direction of enhancing the vertical displacement further. Since this type of failure usually causes the largest forces (or halo currents) observed in the present experiments, this situation must be properly accommodated in the design of the ITER vacuum vessel. DINA analysis shows that although the ex-vessel VS control modifies radial field, it does not affect plasma motion and current quench behavior including halo current generation because the vacuum vessel shields the field created by the ex-vessel coils. Nevertheless, the VS control modifies the force on the vessel by directly acting on the eddy current carried by the conducting structures of the vessel. Although the worst case was explored in a range of plasma inductance and pattern of VS control in combination with the in-vessel VS control circuit, the result confirmed that the force is still within the design margin.

The harsh environment in the ITER device presents many diagnostic challenges. A proposed design engineered at the Japan Atomic Energy Agency (JAEA) is of an optical alignment system installed on mirrors behind double vacuum windows. Due to the harsh environment, a piezo-motor is used as an actuator and a sliding bearing is used as a pivot. The required torque of the actuator is estimated. Trial results indicate that the actuator as described fulfills requirements. A prototype of a key optical component, the micro-retro-reflector array, is produced through the electro-forming method. Measured reflectivity yields a sufficient signal for in situ calibration. A Cassegrain telescope prototype is designed and produced using simple spherical mirrors and achromatic lenses made of silica and CaF 2 . The measured focal length concurs with design objectives. Irradiation tests are necessary to verify the capabilities of these optical devices.

Quartz glasses of KU-1 and KS-4V are candidate window materials for the ITER optical diagnostics in UV and visible range. The temperature dependence of the radiation-induced transmission losses in UV range has been investigated under γ-ray irradiation for the KU-1 and KS-4V quartz glasses up to 10 MGy. The KU-1 and KS-4V quartz glass samples were irradiated at room temperature, 100, 200 and 300 °C. Dose rate was 0.46–1.1 MGy, and 2.6 Gy/s from 1.1 to 10.4 MGy. It was confirmed that there is no significant loss in wavelength range longer than 350 nm for both window materials. Transmission loss in KU-1 is larger than that in KS-4V under temperature below 200 °C. KU-1 has large temperature dependence. On the other hand, temperature dependence is not clear in KS-4V above 100 °C. Prominent recover of the transmission loss was not observed after irradiation for each sample. From the transmission loss point of view, KS-4V is better window material than KU-1 at temperature below 200 °C. KU-1 is available at 300 °C.

The technological challenges associated with preparing and installing diagnostic systems on ITER are summarized and the solutions that have been developed to overcome them are described. Areas where further developments are needed are highlighted.

Magnetic coils are indispensable for plasma position control in tokamaks. For ITER, non-inductive voltages generated under irradiation have to be reduced to well below 1 μV in order to support long pulse (3000 s) operation. In situ measurements of the differential voltage between the ends of two magnetic coils wound with mineral-insulated (MI) cable have been carried out at the JMTR fission reactor. The MI cables of the two magnetic coils had copper center conductor with diameter 0.5 and 0.8 mm, respectively, stainless steel outer sheath and MgO insulation. The measured differential voltage for the two MI cables increased with neutron fluence, reaching 4.5 and −0.7 μV in the two coils, at a fast neutron fluence of 1.26 × 1023 n/m2. The magnitude of the measured voltage can be explained by thermoelectric potentials, enabled mainly by non-uniform transmutation and displacement damages of the copper core of the cable.

Heat and particle control has been studied under the reactor-relevant high-power heating in the large tokamak of JT-60U with an open divertor and progressively a W-shaped pumped divertor. Heat and particle control is crucial for reduction in heat load onto the divertor plates, control of density in the main plasma, effective exhaust of helium ash, and reduction in impurity contamination. For the reduction of heat load, radiative divertor concept was developed based on understanding of heat and particle transport in scrape-off layer and divertor plasmas, which contributed to establishment of divertor concept in ITER. With argon injection, the total radiation loss power reached up to 80% of the net heating power with high confinement of HH{sub y2} {approx} 1, where HH{sub y2} is a confinement enhancement factor over the IPB98(y,2) ELMy H-mode scaling, at high density of 80% of the Greenwald density in the ELMy H-mode plasma. For the density control, the dependence of particle confinement on plasma parameters was systematically studied with two confinement times for center- and edge-fueled particles, which enabled discussion of density controllability. Core fueling using a high-field-side pellet injection extended the operation range of high confinement (HH{sub y2} {approx} 1) from 60 to 70% ofmore » the Greenwald density in the high {beta}{sub p} ELMy H-mode plasma. Efficient helium ash exhaust of {tau}{sub He}*/{tau}{sub E} = 2.8 was demonstrated in the ELMy H-mode plasma with the pumping from the private flux region, which is the same pumping geometry as that in ITER design. Reduction in Z{sub eff} by puff-and-pump scheme was demonstrated, and chemical sputtering yields were estimated with the consideration of not only methane but also heavier hydrocarbons. Their sputtering yields showed strong dependence on the wall temperature and weak dependence on the particle flux. The measured profiles of C II and C IV line intensities were well reproduced by the Monte Carlo impurity transport simulation code (IMPMC code). The estimation of sputtering yields and development of the simulation code enabled reliable predictions for impurity behavior in a fusion reactor.« less

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...

Irradiation tests on many diagnostic components have been carried out as a part of the R&D program of the ITER Engineering Design Activities (EDA). Five kinds of ITER round robin fibers were irradiated in JMTR and60Co γ-rays. Induced transmission loss of those fibers are much smaller than that of pure SiO2 core fiber. Especially, KU-H2G from Russian Federation and F-doped fibers from Japan have rather good radiation hardness even in the visible range. Those fibers might be available in the diagnostic port of ITER. The transmissivity of KU-1 quartz, which is a candidate of the window material for UV and visible spectroscopy in ITER, was measured under the irradiation of 14 MeV neutrons and γ-rays in the UV range (200 - 400 nm). Significant transmission loss was observed in the wavelength range of 200-300 nm. The Mica substrate bolometer was irradiated in JMTR during three irradiation cycles. Total neutron fluence was about 1×1024 n/m2 which was equivalent to 0.1 dpa. Significant increase in the meander resistance was observed, which was causedby the nuclear transmutation of gold into mercury. The use of gold meanders might be problematic in ITER.

As a part of the R&D program of the ITER Engineering Design Activities (EDA), we have carried out irradiation tests on a number of diagnostic components. In this paper the recent results of neutron irradiation tests on diagnostic windows, bolometer substrates and magnetic coils at JAERI are described. UV range transmission losses of a KU-1 quartz glass were measured during 14 MeV neutron irradiation at the Fusion Neutronics Source (FNS) with neutron fluences of up to 7.4×1019 n/m2. Significant transmission losses were observed in the wavelength of 200–300 nm. Mica substrates of bolometers were irradiated at the Japan Material Testing Reactor (JMTR) with neutron fluence between 1×1022 and 2.2×1023 n/m2. The size expansion was found to be only 0.13% at 2.2×1023 n/m2 and 150 °C, which gives us the good prospect of the bolometer in the ITER-FEAT application. Magnetic coils made with four kinds of Mineral Insulated (MI) cables have been irradiated in JMTR up to a fast neutron fluence of 1.1×1024 n/m2. Radiation Induced Electrical Motive Force (RIEMF) of several volts was observed between core inductor and sheath of the MI cable.

Transport coefficients of impurities in reversed shear plasmas are evaluated and compared with those in ELMy H-mode plasmas. The diffusivity is significantly reduced (∼1/10) at the internal transport barrier. An inward pinch velocity of −3 m/s is also observed at the ITB. The inward pinch velocity increases as charge number of impurity increases. Effects of Ar seeding to the main plasma are investigated in ELMy H-mode. Argon seeding maintain improved confinement HITER89=1.4, the clear edge pedestal, the radiation fraction of ∼80% and the divertor detachment up to 70% of Greenwald limit. Moreover, argon seeding reduces type I ELM frequency less than half, but the heat load of each ELM on the outer target plate does not change. Therefore, the reduction of energy losses due to ELMs probably causes the improvement of edge confinement in Ar seeded cases, though the radiated power inside the separatrix increases.

Recent development of the JT-60U charge exchange recombination spectroscopy (CXRS) diagnostic system is reported. The measurements of the radial profiles of the ion temperatures, the toroidal and poloidal rotation velocities, and the impurity density are based on CXR spectra of fully stripped carbon ions induced by neutral beam injection. Measurements are made simultaneously at 59 spatial points (23 toroidal and 36 poloidal) with a time resolution of 1/60 s throughout the full discharge pulse of 15 s. Considerable effort has been expended to ensure the automatic provision of reliable profile data, where the mechanical components of the system must cope with both a baked tokamak vessel (∼300 °C) and an expected disruption-induced acceleration of up to 60 g, where g denotes the gravitational acceleration, i.e., 9.8 m2/s.

Experimental results related to the divertor geometry such as divertor plasma detachment, neutral transport and plasma energy confinement, were compared in the open and W-shaped divertors. The ion flux near the outer strike point was larger than in the open divertor, and the electron temperature at the target, Tediv, was reduced. Divertor detachment and x-point MARFEs occurred at n e 10–20% lower than that for the open divertor. Although the leakage of neutrals from the divertor to the main chamber decreased, a neutral source in the main chamber due to an interaction of the outer scrape-off layer (SOL) plasma to the baffle plates became dominant above the baffle. Degradation in the enhancement factor of the energy confinement was observed similarly in the open and W-shaped divertors. The neutral density inside the separatrix was estimated to be a factor of 2–3 smaller, which did not affect the energy confinement.

The divertor geometry of JT-60U was modified from an open type to a w-shaped pumped configuration. This paper discusses behavior of carbon impurity in the w-shaped divertor. Carbon generation by chemical sputtering is important in deuterium plasmas. The reduction of carbon generation in helium discharges is consistent with this conclusion. CD-band intensities in the private flux region were smaller in the w-shaped divertor than in the open divertor, suggesting that the divertor dome is effective in decreasing chemically sputtered hydrocarbon molecules. The observed CD-band intensity was fairly reproduced by simulation using the IMPMC-code. When deuterium gas was puffed into the divertor, Zeff value in the main plasma became about 15% larger than with puffing from the main chamber. For reducing impurity in the main plasma, the gas puffing from the top of main plasma was effective.

Performance of inner leg pumping of JT-60U W-shaped divertor is presented. Inner leg pumping works well as the dome separates inner and outer divertor channels for neutral particles in the private flux region. In ELMy H-mode with NBI heating power of 18 MW, the operational regime where inner leg pumping is effective extends to the divertor radiation range of about 60% of SOL power. The fraction of inner particle flux pumped is estimated to be about 3%. It is qualitatively shown that a combination of main gas puff and inner leg pumping reduces carbon impurities in the main plasma, suggesting the existence of SOL flow effect on impurity reduction.

Emission of He I lines (668nm, 706 nm and 728nm) has been simultaneously observed in the JT-60U divertor plasma. From calculations using a collisional-radiative model, intensity ratios of these He I lines are considered to be useful to measure electron temperature and density. In order to evaluate the feasibility of the intensity ratio method for temperature and density measurements, the measured intensity ratios are compared with those calculated from the temperature and density measured with Langmuir probes. As a result, the measured intensity ratios are well reproduced by the calculations. The discrepancies in the ratios I (728nm)/I (706nm) and I (668 nm)/I (728nm) between the measurements and calculations are about 12%.

Collective Thomson scattering (CTS) diagnostic based on a pulsed CO 2 laser (wavelength 10.6 μm) has been developed to establish a diagnostic method of confined α-particles in burning plasmas. A high-repetition and high-energy transversely excited atmospheric (TEA) laser has been developed as a source of the CTS diagnostic. In order to obtain single-mode output, which is needed for CTS diagnostic, seed laser is injected into the cavity with unstable resonator. Pulse energy of 17 J with a repetition rate of 15 Hz has been achieved in a single-mode operation. This result gives a prospect for the CTS diagnostic on International Thermonuclear Experimental Reactor (ITER), which requires energy of 20 J with repetition rate of 40 Hz. Proof-of-principle test will be carried out in the JT-60U tokamak by using the newly developed laser. Preliminary consideration of the CTS diagnostic in the tandem mirror GAMMA 10 shows that axial profiles of ion temperature will be obtained using a circumferential collection mirror ofscattered power.

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.

A motional Stark effect polarimeter with five viewing points has been installed to obtain current profiles in the JT-60U tokamak. Offset angles caused by Faraday rotation were calibrated by rapidly moving a plasma, keeping the q profile unchanged. Polarized background light was observed and its effect was corrected by detecting the light which had a wavelength close to that of the beam emission. Current profiles in a current rampdown discharge were measured, and the q value at the inversion radius of the sawtooth oscillation agreed with the theoretical value (unity) within 10%.

In order to measure the line integrated electron density on the International Thermonuclear Experimental Reactor (ITER), a vibration-compensated five-channel CO2 laser interferometer/polarimeter system is being developed. The design of the interferometer/polarimeter has been improved from the previous design, which used a CO2 laser (wavelength, λ=10.6 μm) and a CO laser (λ=5.3 μm), on the basis of experience gained with a dual CO2 system (λ=10.6 and 9.27 μm) on JT−60U. The JT−60U system gives good reliability for both interferometry and polarimetry operation. The performance of the dual CO2 laser is expected to meet the requirements for measurements on ITER (accuracy; 1%, time resolution; 1 ms).

The performance of two array antennae of phased 2 × 2 loops for JT-60U is presented in the initial operation for higher harmonic ion cyclotron resonant frequency (ICRF) experiments. The array antenna is designed so as to have a sharp coupled power spectrum of n  peak ≈ 3 with Δn  ≈ 2.5 when out of phase with toroidally adjacent loop currents ((π, 0) phasing) at 116 MHz by separating the loops widely by 0.44 m. The maximum coupled power up to the present is 6.4 MW with (π, 0) phasing at Bt = 4 T, which corresponds to the power density of about 6.1 MW m−2. Anomalous antenna breakdown is observed during combined neutral beam injection (NBI) and ICRF heating for the particular magnetic field strength Bt = 3.0 T at which an ion cyclotron resonance layer of 7ωcD is located near the antennae. A fairly large antenna loading resistance is obtained, i.e. 4 Ω for (π, 0) phasing at 0.13 m of the distance δSF between the separatrix and the Faraday shield, and 5 Ω for in-phase ((0, 0) phasing) at very large distances of more than 0.33 m. results show the decrease in the loading resistance at the L to H transition is about 30%, from 5 to 3.7 Ω with (0, 0) phasing at δSF = 0.32 m. The ratio of the incremental radiation loss to ICRF power decreases with increasing ICRF power and is saturated at small values less than 10% at power levels more than 4 MW. The increase in the radiation loss seems to be mainly caused by the light impurities from the first wall.

The impurity behavior and the remote radiative cooling have been studied in the JT-60U divertor discharges with high power NB heating. NB at a power level of 20 MW was injected for 2 s, and no carbon bloom has been observed. For the first time, the measurement and the calculation for the carbon influx from the divertor plates are compared quantitatively. The carbon generation mechanism can be explained by the sputtering with deuterium (physical sputtering), oxygen and carbon. It is found that the contribution of the chemical sputtering with deuterium is small and that of the sputtering with oxygen is important. Remote radiative cooling can reduce the heat flux onto divertor plates. The dependence of the radiative cooling power on the electron density and the safety factor is presented. The favorable operation regime for discharges with high power auxiliary heating is discussed based on the experimental results.

In JT-60U, impurity profiles were measured with charge exchange recombination spectroscopy (CXRS) and impurity transport was analyzed. Density profiles of fully ionized carbon ions C6+ were obtained from neutral beam density profiles and excitation rate coefficients for CVI 529.0 nm (n=8-7). Peaked density profiles of C6+ in a typical NB heated deuterium plasma were measured. Peaking parameter Cν was not constant and was in the range of 0.5–1.0. OVII (1s2-2enp, 2.16 nm) and OVIII (1s-2p, 1.90 nm) were measured with a grazing incidence spectrometer viewing the plasma center vertically. Oxygen impurity transport was analyzed in the peripheral region from the ratio of OVIII/OVII.

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.

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.

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.

Many windows will be required on ITER for optical, infrared, and microwave diagnostics. The diagnostic transmission systems will be incorporate labyrinths in port shielding blocks and so the windows will not be located in high levels of neutron and gamma radiation. Nevertheless it is necessary for the window seals to be able to withstand pressure transients, temperature excursions and mechanical shocks which are significantly different from those experienced on present-day tokamaks. The objective of the RD> 200 °C, (4) Temperature ramp rate:> 20 °C/hr, (5) Temperature gradients across windows:3∼4 points for each surface of window, (6) Mechanical acceleration of structure:> 15g, duration 10ms, 1000cycles., (7) Optical property

The in situ calibration method for the impurity influx monitor (divertor) is experimentally examined. The total reflectance of the optical path from the focal point of the Cassegrain telescope to the first mirror is derived using a micro retroreflector array. An optical fiber with angled physical contact (APC) connectors reduces the return edge reflection. APC fibers and a multimode coupler increase the signal-to-noise ratio by about one order compared to that of triple-branched fibers and enable measurement of the wavelength dependence of the total reflectance of the optical system even after potential deterioration of mirror surfaces reduces reflectance.

Metallic mirrors will be used as plasma-viewing elements in all optical and laser diagnostic systems in ITER. In the harsh environment of ITER, the performance of mirrors will decrease mainly because of the erosion of their surfaces and deposition of impurities. The deterioration of the optical properties of diagnostic mirrors will directly affect the entire performance of the respective ITER diagnostics, possibly leading to their shutdown. Therefore, R&D on mirrors is of crucial importance for ITER diagnostics. There is a coordinated worldwide R&D programme supervised by the Specialists Working Group on first mirrors of the International Tokamak Physics Activity, Topical Group on Diagnostics. This paper provides an overview of new results in the field of first mirrors, covering the manufacturing of ITER mirror prototypes, investigations of mitigation of deposition and mirror cleaning and the predictive modelling of the mirror performance in ITER. The current status of research on beryllium deposition—a new critical area of mirror research—is given along with an outlook for future activities.

A collective Thomson scattering (CTS) diagnostic using carbon dioxide (CO2) laser is being developed to establish a diagnostic method of confined α-particles in burning plasmas. To realize the CTS diagnostic, a transversely excited atmospheric CO2 laser has been developed. To obtain single-mode output, which is needed for the CTS diagnostic, seed laser is injected to the cavity with an unstable resonator. Using this technique, output energy of 17 J at a repetition rate of 15 Hz has been achieved. This result gives a prospect for the CTS diagnostic on International Thermonuclear Experimental Reactor (ITER). Proof-of-principle test of the CTS diagnostic is being performed with the new laser system on JT-60U. A method to improve spatial and spectral resolutions, which is subject to be resolved for the CO2 laser CTS, is proposed using a resonance at the lower hybrid frequency. Calculation results show that the scattering power increases more than a factor of 10, and spatial and spectral resolution improve about 30% if the scattered wave vector is nearly perpendicular to the magnetic field.

A high-repetition and high-energy transversely excited atmospheric carbon dioxide (CO2) laser for a collective Thomson scattering (CTS) diagnostic has been developed to establish a diagnostic method of confined α particles in burning plasmas. To excite a single-transverse and single-longitudinal mode, a continuous wave seed laser was injected to an unstable resonator with a cavity length of ∼4.4m. Pulse energy of 10J with a repetition rate of 10Hz has been achieved in the single-mode operation. Pulse energies of 18J with a repetition rate of 10Hz and 36J with single shot operation have also been achieved in the multimode operation. These results give a prospect for the CTS diagnostic on International Thermonuclear Experimental Reactor (ITER), which requires single-mode energy of 20J with repetition rate of 40Hz.

A decaborane-based boronization system has been installed in the JT-60U tokamak in order to reduce the influx of impurities during plasma discharges. Boronization has been performed under a glow discharge using a helium-decaborane gas mixture. The properties of the boron films deposited through boronization and the effects of boronization on the tokamak discharges were investigated. It was found that the deposition of a boron layer with high purity was achieved with few impurities other than hydrogen through boronization, and that the present boronization deposited toroidally nonuniform boron film. It was also found that the decaborane-based boronization resulted in good plasma performance similar to that of conventional boronization.

Plasma-surface interactions during RF heating on JFT-2 tokamak were investigated. Light impurities (O, C) and also metal impurities (Mo, Fe, Ti) increased by about twice for 220 kW RF power, however, because impurities have been reduced to an allowable level by Ti gettering onto the wall and the limiter, it was not so deleterious on plasma confinement. The increase of metal impurities can be explained by the light impurity sputtering and the effect of high energy ions produced by RF heating seems to be minor for metal release. On condition of clean surface produced by Ti gettering, decrease of electron density was observed during RF heating. This phenomenon is discussed in connection with the behaviour of high energy ion.

ICRF heating of impurity-free tokamak plasmas is studied in DIVA. The frequency is fixed at 25 MHz, which is a second-harmonic cyclotron frequency of deuteron with a toroidal magnetic field of 16.4 kG. The toroidal field BT and the ratio of proton-to-deuteron density p are varied over ranges of 12 ? BT ? 19 kG and 2 p 40%, respectively. The most favourable ion heating is obtained at BT = 18 kG and p = 5 ~ 10%. Ion temperature increases from 200 to 300 eV were measured by charge-exchange neutral analysis and also by spectroscopy, where 100 kW RF net power were applied. The results are consistent with mode-conversion theory.

The radiation loss in DIVA is measured quantitatively by a pyroelectric detector calibrated by heavily contaminated resistive discharges. The results are: (1) radiation loss power due to low-Z impurities does not play an important role; (2) radiation loss power due to the pseudo-continuum from high-Z impurities affects the power balance of the confined plasma; (3) the time dependence of energy loss channels is clarified in the gross power balance, a fact which attests to the reliability of the radiation measurements carried out by the pyroelectric detector.

Detailed spectroscopic studies of impurities have been made on the hydrogen plasma in the wavelength region from 10 to 1300 A with the upgraded DIVA tokamak device with a divertor. The toroidal field is 20 kG and the peak plasma current is 40 kA. Oxygen is the dominant impurity with the density of about 2% of the electron density. The effective ionic charge estimated from the amount of low-Z impurities (oxygen and carbon) is 2.4 to 3 in agreement with the value derived from the plasma conductivity. The emission spectrum of many lines observed in the region from 42 to 200 A is due to gold as impurity. Its spatial distribution has a peaked profile at a shorter wavelength and a hollow profile at a longer wavelength. The radiation power losses due to low-Z impurities and gold are estimated to be 30–35 kW and 27 kW, respectively, and the total power loss corresponds to 41–44% of the Joule-input power.

Neutral-beam power of up to 1.2 MW injected into the plasma has produced a volume-averaged β of up to 2.6% and a central beta β0 of up to 7%, due to the thermal components in the JFT-2 tokamak. In these beam-dominated discharges, the magnetohydrodynamic behaviour was studied. Four types of internal oscillations were observed: i) enhanced sawtooth oscillations with long repetition time and large sawtooth amplitude; ii) round sawtooth oscillations and/or reduced sawtooth oscillations with short repetition time and small sawtooth amplitude; iii) high-frequency oscillations without sawtooth oscillations, and iv) high-frequency oscillations with sawtooth oscillations. The measured beta values are compared with the critical ones as found from high-n ballooning-mode analysis, and the relationship between MHD behaviour and beta values is also investigated.

Diagnostic instruments for measuring the JT-60 plasma are fully developed and ready to operate. Each instrument and its fundamental characteristics are described as well as some aspects of the plasma parameters. The new features in JT-60 diagnostics are that the diagnostics system which has its own data processing system separate from the machine control computer, and the utilities for each diagnostics instrument (vacuum, electricity, water, air, liquid nitrogen) constitute one system in the total diagnostics system.

Electron cyclotron heating (ECH) and pre-ionization using a 28 GHz 200 kW gyrotron were investigated in the JFT-2 tokamak. Three modes of the wave were launched from different types of launch system, i.e. the extraordinary mode from the high field side of the torus, the ordinary mode from the low field side, and TE 02 mode from the top to examine their heating characteristics. The oblique launch of the extraordinary mode gave the largest central heating in consistent with the theoretical prediction. It was shown that the polarization of the launched wave as well as the effect of the mode conversion at the upper hybrid resonance layer contribute to the heating mechanism.

A mode conversion theory of the fast magnetosonic wave is compared with the experimental results in a deuterium plasma with minority hydrogen. The direct measurements of toroidal wave length, toroidal damping length, and the increase of ion temperature show that parameter dependences of ion heating are well explained by the mode conversion theory including tunneling, reflection, and absorption of the fast wave at a two-ion hybrid resonance layer.

Parametric instabilities, which bring about undesirable peripheral heat deposition in lower-hybrid-wave (LHW) heating, are stabilized in the JFT-2 tokamak by an effective electron heating in the boundary plasma. The boundary electron temperature rises significantly by simultaneous injection of neutral beams into the plasma, which may be expected to improve the heating efficiency of LHW in the main plasma

Temporal variations of intensities of spectral lines in a normal-incidence vacuum ultra-violet region emitted from oxygen and carbon impurities of tokamak plasmas have been studied both experimentally and theoretically. Experimental results have been compared with theoretical ones calculated numerically by using a plasma model involving ionization, radiative and dielectronic recombinations, finite confinement times of impurity-ions, and influxes of neutral impurities from walls. The initial rise of electron temperature has been estimated in such a way that the peak-times for spectral lines in successive states of ionization coincide in both experimental and numerically-calculated results. The estimated electron temperatures are found to be consistent with those obtained from measurements of ruby-laser light scattering and soft X-ray spectra.

Doppler broadenings of impurity (O, N and C) lines in hydrogen discharges were observed with the JFT-2 and DIVA tokamak devices for the purpose of determining the spatial distribution of the ion temperature. A 1-meter Czerny-Turner type vacuum monochromator was used in higher orders for sufficient resolution. Impurity emissions are localized in shell-like regions in the minor cross-section. The radial distribution of the ion temperature was determined from the Doppler broadening of various impurity lines and from the localization of line intensities. The Doppler temperatures of the 0 +6 ion were found to be 256 eV for a 100 kA, 14 kG discharge with JFT-2 and 220 eV for a 28 kA, 20 kG discharge with DIVA.

Oxygen impurity has been reduced to about 1–2% of the electron density by titanium gettering onto the torus wall (~ 1 2 ) and the limiter. Radiation loss and effective ionic charge were decreased by a factor of ~2. As a result of reduced impurity influx, broader electron temperature profiles have been obtained. The energy confinement time and the scaling factor of the maximum electron density ( n e R/B t ) were improved by a factor of 1.6 with the titanium gettering. Limits on the density increase were investigated in connection with the radiation power. Mechanism limiting the density maximum is discussed.

A 3-m grazing-incidence monochromator was calibrated absolutely in the wavelength region from 40 to 1100 A by means of tokamak discharges. The monochromator was equipped with a platinum-coated ruled grating blazed at 146 A and a windowless Cu–Be Photomultiplier. Line-pairs used in the atomic-branching-ratio calibration were impurity lines emitted from oxygen and carbon ions, in various ionized states, in JFT-2 and DIVA hydrogen discharges. The effect due to self-absorption was negligibly small for the impurity resonance lines, because the densities of the impurity ions were very low. Errors in the calibration experiment were estimated to be typically within several tens of percent and were largely ascribable to the lack of accuracy in the values of transition probabilities. The absolute sensitivity of the monochromator determined through the present calibration was considered to be sufficiently reliable for the investigation of impurities in high-temperature tokamak plasmas.

Radiation from the Hα-line has been investigated in the hydrogen tokamak plasma obtained under a typical JFT-2a discharge. The profiles of the Hα-line were measured along both perpendicular and tangential paths to the toroidal magnetic field. The measured profiles can be interpreted in terms of Doppler broadening at the line wings due to hot hydrogen atoms originating from the resonance charge-exchange process and at the line center due to cold particles having energies of a few to several eV. The peak value of the Doppler temperature for the hot particle was approximately 50 eV. Equal temperatures were obtained from measurements along perpendicular and tangential directions to the toroidal magnatic field. Distributions in the minor cross-section of photon numbers of the Hα-line were also measured, from which the confinement time of charged particles was found to be 0.5 to 2.5 msec.

Deuterium and hydrogen energy spectra during ICRF heating as well as nH/nD were measured with the E||B type mass-separated neutral particle energy analyser. High-energy tails of deuterium and hydrogen were observed in the dilute-minority case. From Fokker-Planck calculations of the charge-exchange spectrum and power-balance analysis it is shown that second-harmonic cyclotron damping by the majority deuterium is the dominant mechanism of ion heating in the dilute-minority case.

Characteristics are presented of the energy confinement time resulting from neutral hydrogen beam heating of JT-60 tokamak plasmas with a plasma current, Ip, of 1-2 MA, a line averaged electron density, e, of 1.5-7 × 1019 m−3 and an absorbed beam power, Pabs, up to 20 MW. The plasma energy content follows an offset linear relation to the absorbed power. The incremental energy confinement time for thermal components, , is almost independent of Ip and e at 60 ms, while the incremental energy confinement time for the total plasma energy, , decreases with plasma density. The value of in JT-60 coincides with the prediction of the Shimomura-Odajima scaling of L-mode discharges.

A pulse of methane gas is injected into the DIVA tokamak. The major portion of the injected impurity is shielded by the outer edge plasma, i.e. only 20% of the injected carbon are introduced into the discharge. After radial penetration of the injected carbon has been accomplished, accumulation doubles and a new steady state is sustained over the period of proton confinement. The transient radial diffusion process of the carbon is compared with calculations using an impurity transport code. The behaviour of the carbon impurity in the whole minor radius is well described by assuming perfect re-cycling at the gold-plated shell surface and using neoclassical diffusion superposed by anomalous diffusion of protons. It is shown that spreading caused by anomalous diffusion is important. The calculated result indicates that the carbon ions are lost dominantly in a He-like state.