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Time-resolved measurements of the negative ion density and of other plasma parameters in a pure hydrogen Tandem source have been effected using the laser photodetachment technique. The study was made for hydrogen filling pressures in the range 7–11 mTorr and an arc current of 120 A. We found a strong time variation in negative ion density was present during the discharge pulse (0.5 s). The maximum negative ion density was observed at the beginning of the pulse, at the lowest filling pressure studied, followed by a rapid decrease. An improvement of the current distribution during the pulse was obtained by increasing the filling pressure. The electron density is highest at 7 mTorr and stays almost constant during the pulse. The increase in filling pressure leads to electron density values lower by a factor of 4 which are constant during the discharge pulse. The highest relative negative ion density during the pulse is obtained at 9 mTorr. The time variation of the negative ion density does not follow that o...

Accelerated electrons, which would lead to high thermal load of grids, have been suppressed in a high-current large hydrogen negative ion source. An extraction grid, with apertures shaped as the secondary electrons generated on the grid aperture surface would be shielded against the acceleration electric field, works well to prevent the secondary electrons from leaking to the acceleration gap, compared with a straight aperture extraction grid. Although the strong magnetic field at the extraction grid also lowers the electron leakage downstream, the aperture shaping of the extraction grid is more effective for the suppression of the accelerated electrons. The acceleration efficiency, defined by the ratio of the negative-ion current to the acceleration drain current, is improved to around 85%. There remains the accelerated electrons generated in the negative ion neutralization by collision with the residual neutral molecules during the acceleration. The direct interception of the accelerated negative ions w...

A high-power large negative ion source has been operated for a long pulse duration. A three-grid single-stage accelerator is used, where the extraction grid is shaped so that the secondary electrons generated on the extraction grid would be prevented from leaking into the acceleration gap. A stable long-pulse arc discharge with an arc power of 100 kW has been obtained over 15 s by balancing an individual arc current flowing through each filament. The cesium-seeded operation is not influenced by a temperature rise over 100 °C of the plasma grid during the long-pulse arc discharge. As a result, 330 kW (91 keV–3.6 A) of the negative ion beam was produced stably for 10 s from an area of 25 cm×26 cm, where the current density was 21 mA/cm2 and the negative ion power density was 1.9 kW/cm2. The neutralization efficiency of accelerated negative ions has been measured including the residual positive and negative ion ratios by the water calorimetry of the beam dumps. The result agrees well with the calculation result.

A negative-ion-based neutral beam injection (NBI) system is planned for plasma heating of the Large Helical Device (LHD). We have developed a negative ion source, which is 1/3 the scale of the source for the NBI. A magnetic filter held was generated by external permanent magnets to lower the electron temperature in a large-area bucket plasma source (35 cm/spl times/62 cm) for efficient H/sup -/ production. We investigated the magnetic field configuration and found a low electron temperature high density plasma ( 15 A: 1/3 current of LHD ion source). Based on the results, we are designing a negative ion source for the LHD.

A novel emittance meter is developed using the pepper‐pot method. Kapton foils are used to detect intensity distributions of small beamlets at the ‘‘image’’ plane of the pepperpot. Emittance of H− beams for the neutral beam injector of the large helical device has been measured. The normalized emittance (95%) of a 6 mA H− beam with an emission current density of about 10 mA/cm2 is ∼0.59 mm mrad. The present system is simple, and it eliminates many complexities of the existing schemes.

We report results of folded waveguide antenna experiments with high power, up to 1 MW, which include the dependence of plasma loading resistance on the plasma density and how much r.f. power can be delivered. We propose its application to the large helical device, which is a heliotron-torsatron device scheduled to produce its first plasma in 1998. The folded waveguide antenna launches an ion Bernstein wave to heat electrons via the Landau damping process. Preliminary experiments show a sharply defined wavenumber spectrum along the magnetic field line, which suggests core plasma electron heating. Direct electron heating is superior to fast wave minority heating because it avoids the orbit losses of deeply trapped high energy ions.

A large negative hydrogen ion source with a rod-type magnetic filter was investigated for the construction of a neutral beam injection (NBI) system for the large helical device (LHD). A 1 3 - scale device was constructed and investigated. Three main phenomena are observed on seeding a small amount of cesium into the ion source. These are enhancement of the hydrogen negative ion (H−) current, a reduction in the electron beam extracted with the H− beam, and a decrease in the operating gas pressure. The enhancement factor of the H− current is 3–4, and a total H− current of 8.5 A is obtained from 20 × 40 extraction apertures 9 mm in diameter. This current corresponds to a current density of 17 mA cm−2. A multi-wall structure forming a lattice, called an egg-box cell, is installed between the magnetic filter rod and the plasma grid. The electron current with this cell is half as low as the electron current without the cell, but the H− current is hardly affected. This phenomenon is observed when cesium is seeded in the source chamber. The electron temperature and its density increase by a factor of 2 on strengthening the cusp field. Experiments to obtain a higher H− current density were carried out using the improved are plasma. A hydrogen negative ion current of 54 mA cm−2 was extracted from a restricted aperture of 36 extraction holes. The H− current was extracted from 20 × 40 full apertures of the ion source with strong cusp field. A total H− current of 10.4 A was obtained from these apertures. The optimum filling pressure is less than 1.0 Pa.

A volume-averaged equilibrium β value 〈βeq〉 of 2.14% is achieved in a compact helical system using two neutral beam lines with balanced injection and intense wall conditioning with Ti gettering. This value is the highest β value realized so far in helical systems. Reheat mode, where the stored energy increases after turn-off of a strong gas puff, is employed in the experiment. Discharge conditions are as follows: Bt = 0.61 T; beam power through the port, 1.1 MW (coinjection) and 0.8 MW (counterinjection); line-averaged electron density n e = 6.5 × 10 13 cm −3 . Amplitudes of magnetic fluctuations integrated over the frequency range from 3 kHz to 100 kHz become saturated at 〈βeq〉 higher than 1%. Dominant coherent modes are m n = 2 1 and 1 1 when 〈βeq〉 is lower and higher respectively than 1%. Dependence of the energy confinement time τE on n e (up to 8 × 1013 cm−3) and Bt (from 0.6 to 1.8T) is also studied in this high β experiment. When the density increases τE degrades compared with the LHD scaling; the density dependence exhibits Bohm-like behaviour. On the contrary, τE scales as Bt≈0.75, which is rather close to the LHD scaling (gyro-Bohm-like behaviour).

A design for a high power negative-ion-based neutral beam system for the large helical device (LHD) is described. A total power of 20 MW with a pulse length of 10 s is required for neutral beam injection in the LHD. The characteristic of this system is its rather low energy (125 keV for hydrogen) for the use of negative ions, so a large amount of negative ions is needed. The basic design of the beamline is based on the grounded gas neutralizer. A large expansion chamber between the ion source and the neutralizer, and additional ion dumps are needed for the use of negative ions. The beam acclerating power supplies are similar to those of the positive ion system because of the similar beam energy. Some additional power supply is needed for the production of negative ions. The construction of the neutral beam injector will start in 1995.

A 1 3 - scale source was designed on the basis of previous multi-cusp H− sources developed at the National Institute for Fusion Science (NIFS). It has been developed for the negative-ion-based neutral beam injection system on the test stand in the large helical device (LHD). The ion source was immersed in a vacuum vessel, making the ion source simple and compact. A reduction in the stripping loss of H− ions is also expected. Three studies on the 1 3 - scale vacuum-immersed H− ion source (aiming at a 15 A × 125 kV beam) are described: design features for the vacuum-immersed source at high voltage, a Monte Carlo simulation for pumping gases in the accelerator to estimate the stripping loss of H− ions, and the experimental results of testing the plasma source and beam extraction and acceleration during beam conditioning. As a result of the simulation, the stripping loss was calculated to be about 26% for a reference design. A guiding principle for accelerator design was obtained from the viewpoint of enhanced pumping on transverse gas conductance and resulting reduction in the stripping loss. In experiments, a plasma density of more than 2.5 × 1012cm−3 was produced in the first chamber of the plasma source. During beam extraction, it was found that a beam-induced low density plasma (i.e. 105−107 cm−3) exists in the vacuum vessel. In order to produce a long pulse beam, this plasma must be shielded. Beam conditioning has been completed up to 1 A with beam energy of up to 85 keV for 0.1 s pulse duration. H− current scaling and engineering problems related to full scale performance are discussed.

A large r.f. driven negative hydrogen ion source has been constructed. R.f. negative ion sources have the potential of continuous and efficient operation for next-step neutral beam injection systems. The developed r.f. negative ion source is large for this fusion application. The r.f. plasma generator is a metal-walled multi-cusp bucket source with tent magnetic filter and has dimensions of 30 × 30 cm2 cross-section and 19 cm depth. An internal induction coil is used as an r.f. antenna. The negative ion accelerator, which consists of three grids, produces nine beamlets with an energy of 60 keV. The r.f. power amplifier system generates 2 MH – 30 kW for 1 s. The r.f. plasma was produced successfully under impedance matching condition. An ion current density of 150 mA cm−2 was obtained and the electron temperature was high, 5–10 eV, in the antenna vicinity. In the chamber center which is the negative ion production region, the electron temperature is still high, more than 3 eV. As a result, a negative ion current was not detected at an extraction energy of 1.2 keV.

A filament-type helium ion source has been developed for the purpose of alpha particle diagnostics and as a test beam for electrostatic quadrupole transport experiments. We have obtained preliminary results of He+ beam current, ∼60 mA/cm2. The normalized brightness of this beam is ∼1010 A/(m rad)2. The normalized emittance (90%) measured by multislit faraday cup emittance meter is ∼0.6 mm mrad below 100 V, 10 A arc discharge. The effects of plasma parameters in the source on the beam emittance and on the beam perveance are discussed.

Toroidal plasma rotation driven not by the momentum input but by the electric field induced by loss of ions has been observed for plasmas heated by perpendicular neutral beam injection. The rotation is found to increase in the direction opposite to the plasma current when the plasma ions are heated by ion cyclotron resonance frequency waves. The rotation driven by the electric field is quantitatively consistent with that inferred from the ambipolarity of particle fluxes.

The H-mode, an improved confinement regime, was attained recently in JIPP T-IIU high power heating experiments, in a limiter configuration without any shaping of the plasma cross-section. This H-mode is unusual because it was obtained with heating in the ion cyclotron range of frequencies (ICRF). It was also attained with combined ICRF and neutral beam injection (NBI) heating. The threshold power level obtained with ICRF alone is similar to or less than that obtained in the combined heating case. The dependence of the power threshold of the H-mode on various plasma parameters has been studied. It increases with the plasma current and is insensitive to the plasma density, and there is an optimum value of the toroidal field intensity. The power deposition profile for ICRF heating has been analysed with a ray tracing code and used to explain the observed dependence on the toroidal magnetic field. The paper also discusses a class of discharges with improved confinement observed in the same series of experiments. These discharges had a power level close to the H-mode threshold power and exhibited a marked improvement of confinement. They were, however, different from H-mode discharges in the time evolution of the profiles.

H− ion density in the extraction region of a volume-production-type ion source is measured, using the photodetachment method, and directly compared to H− ion beam extracted from the ion source in order to understand the effective production of H− ions. It is observed that there is qualitative similarity between H− density and extracted H− beam current, as expected. Characteristics of H− ion production under two values of magnetic strength are compared to study the effect of the filter field. It was found that the weaker magnetic field of 750 G cm is not sufficient for the filter field since high energy electrons can come into the extraction region across the magnetic field.

Microwave plasma source for the negative hydrogen ion production was constructed and tested. The plasma source consists of two chambers: a plasma production chamber and a plasma confinement chamber. The production chamber is placed into the strong axial magnetic field and the microwaves (2.45 GHz, up to 5 kW) are introduced through the quartz window along the magnetic field lines. It is found that the suppression of the fast electron loss to the window holder is important to improve the efficiency of hydrogen plasma production. The plasma density increases with the magnetic field strength in the plasma production chamber. For the microwave power of 4 kW the uniform plasma of 3×1012 cm−3 for argon and 3×1011 cm−3 for hydrogen is obtained in the area of 20×20 cm. Electron temperature in the plasma grid region is around 2 eV. The optimum gas pressure is around 6 mTorr.

An electron cyclotron resonance (ECR) plasma production method using electron cyclotron wave is presented. This method is not subjected to any cutoff-density scaling, and is capable of generating dense plasmas far beyond the critical density of an ordinary wave. It has been demonstrated that a plasma density more than 1013 cm−3 (argon) is obtained with a low frequency (2.45 GHz) and low power (11 kW) microwave. The experimental results and the application to plasma neutralizer are described. The experimental verification of superpermeability of Nb membrane in a plasma environment is in progress to develop a plasma neutralizer with Nb-membrane wall. The design parameters of Nb-membrane pump are presented.

A large rf‐driven hydrogen negative‐ion source has been developed for a neutral beam injection (NBI) system in the fusion application. The rf plasma generator is a metal‐walled multicusp bucket source, the dimensions of which are 30 cm×30 cm in cross section and 20 cm in depth. An induction coil antenna is fully immersed in the rf plasma generator, and is used for coupling the rf power into plasma. A strong external magnetic filter is generated in front of the plasma electrode for negative‐ion production, and a filter field strength of more than 1000 G cm is required to reduce the electron temperature in the extraction region as low as 1 eV. Precisely 5.5 mA of a negative‐ion current was obtained from a single aperture, corresponding to 4.1 mA/cm2 of current density. A small amount of cesium was supplied to the rf plasma generator from a cesium oven. As a result, a lowering of the operational gas pressure and improvement of the negative‐ion production efficiency were observed. The optimum bias voltage was...

The R&D program on a large current negative ion source has been carried out at the National Institute for Fusion Science for the Large Helical Device Project. So far a 1/3-scaled model ion source has been examined. By seeding a small amount of cesium in discharge, we obtained a total current of 8.5 A with an extracting energy of 40 keV. 1.2 A of high density negative ion beam was obtained from the restricted area of the source with the current density of 54 mA/cm/sup 2/. An engineering test of a vacuum-immersed source has also been performed.

A test stand for the negative-ion-based NBI (neutral beam injector) for the LHD, currently under construction, has neutral beam capabilities of 2.5 MW at a beam energy of 125 keV (H/sup 0/)/250 keV (D/sup 0/) and pulse duration of 10 s. Components being developed are as follows: (1) a large and high current negative ion source which is immersed in vacuum and composed of a compact modular ion source; (2) a beam dump to remove high heat flux up to 1.6 kW/cm/sup 2/; (3) a high-pumping-speed (450 m/sup 3//s) cryopump; (4) a 250-kV power supply system using a GTO (gate turn-off thyristor); and (5) a controller system to vary the beam energy within 0.5 s. >

The multiampere negative hydrogen ion source has been developed in National Institute for Fusion Science (NIFS). The ion source is a volume‐production type multicusp one with an extraction area of 25×25 cm2. It is found that high density negative hydrogen ions of more than 1 × 1012 cm−3 are produced in the center of the arc chamber with a double magnetic filter configuration. A supply of a small amount of cesium vapor into the arc chamber has greatly enhanced the H− ion current and reduced the operation pressure. The H− ion current of 3.3 A has been extracted from a Cs‐seeded plasma at the pressure of 0.9 Pa.

A new rf source producing a large and uniform plasma with magnetic field‐free geometry for neutral beam has been developed. The source consists of a small‐diameter multiple‐rf plasma generator and a bucket chamber as a plasma expansion section. The plasma density scaling, the spatial profile, and the ion beam extraction have been studied by using the single rf plasma generator. It was found that under the optimized conditions the plasma density and the profile in the expanded plasma were essentially ruled by qualities of the bucket chamber itself. The present rf source is promising for the application of large‐volume positive‐ion source or plasma processing.

A large rf-driven hydrogen negative ion source has been constructed for applying to the neutral beam injector of the next step fusion experimental devices. The negative ion source was optimized for plasma production and negative ion extraction. A high density plasma of 1.5 × 1012 cm−3 was generated in the plasma generator with dimensions of 30 × 30 × 20 cm3 at the rf input power of 18 kW. In the pure-volume operation the hydrogen negative ion current of 5.5 mA, corresponding to 5.0 (mA/cm2)/(W/cm3) was extracted at 13.1 mTorr of the gas pressure. Cesium effects which are observed in a filament-arc type negative ion source were also confirmed in the rf-driven negative ion source. The rf-transmission efficiency is improved by using a capacitively coupled DC isolator to the rf ion source.

Present status of the hydrogen negative ion source development is reported for an NBI system in the Large Helical Device. An external-filter type negative ion source has produced a 16.2 A of H - ion beam with the energy of 47 keV at the arc efficiency of 0.1 A/kW and at the operational gas pressure of 3.8 mTorr. The H - ion beam of 13.6 A was successfully accelerated to 125 keV with two-stage acceleration. A rod-filter type negative ion source has also produced a 16 A of H - ion beam with the current density of 45 mA/cm 2 . A vacuum-immersed negative ion source has demonstrated stable operation on condition that the ion source is shielded from the beam-induced plasma. For the future long-lifetime operation, an RF-driven negative ion source has been developed, where an induction coil antenna is placed inside a metal-walled magnetic multicusp source.

The 1/3‐scaled H− ion source for NBI‐system of Large Helical Device has been developed. By seeding cesium vapor into the multicusp plasma source, the H− ion current is enhanced 3 times larger than that in pure hydrogen operation. The reduction of electron current and operating gas pressure is also observed. Total H− ion current of 8.5 A has been obtained from 800 extraction holes of 9 mm in diameter arranging in a large area of 25×50 cm2. The ratio of electron to H− ion current can be reduced to 2 by using the EggBox Cell placed on the plasma grid. From 36 extraction holes, the H− ion current of 1.24 A has been obtained with the extraction voltage of 8.5 kV, which corresponds to the current density of 54 mA/cm2. The H− ion current density increases as an extraction voltage and an input arc power increase.

Negative hydrogen ions are extracted from a sheet plasma and extracted H− currents are optimized for several parameters; a location of the extraction electrodes, a bias voltage, a filling gas pressure and a discharge current. There is an optimum position of the electrodes near the plasma periphery and an optimum pressure for the H− production. The profiles of the electron density and temperature are measured and the relation between the profiles and the H− yield is investigated. It is found that the H− yield is optimized when the electron temperature in front of the extraction grids decreases below 1 eV and that this optimum condition can be achieved by changing the position of the electrodes and/or by changing the filling gas pressure.

Kapton foils can be used as a simple ion beam monitor. These foils act as a self-developing medium and quantitative information about the distribution of the striking beam can be obtained easily. The “burn” pattern on the film due to irradiation by the ion beam is analyzed to determine the beam distribution, here the beam power density distribution. Given the beam voltage this information can be translated to the ion current density distribution. The linearity of response of the Kapton foil with beam current density is shown. The threshold for response of the Kapton foil seems quite good; a clear response is observed with a foil of thickness=50 μm for H− beam current density of about 35 μA/cm2 at 88 kV.

A large rf-driven negative hydrogen ion source has been developed for a neutral beam injection (NBI) system in future experimental fusion devices. Dimensions of the rf plasma generator are 30×30×20 cm3. An induction coil antenna is fully immersed into the plasma generator. The rf-driven plasma with an electron density of 1.5×1012 cm-3 was generated at 28 kW rf power at a frequency of 2 MHz. The electron temperature in the extraction region decreases as the line-integrated magnetic filter field strength increases. An electron temperature in the extraction region of 1 eV, which is suitable for generating negative ions, was obtained at a line-integrated filter field strength of 1030 G cm. Negative hydrogen ions are extracted from a single aperture of 13 mm in diameter in a pure-volume operation. An extracted negative ion current of 5.5 mA was obtained at 13.1 mTorr gas pressure and 15 kW rf power. This negativeion current corresponds to a power efficiency of 5.0 ( mA/cm2)/( W/cm3), which is superior to that in filament-arc negative ion sources.

In the JIPP T-IIU tokamak, a high power ICRF heating experiment has been conducted, up to an extremely high power density (~2 MW·m−3), with a total RF power of PRF = 2 MW. Great attention has initially been paid to the problem of impurities, and it has been found that (a) the adoption of low Z materials for the limiter, (b) in situ carbon coating (i.e. carbonization) and (c) adequate gas puffing synchronized to the RF pulse are very effective in suppressing radiation loss. With these methods, a remarkable reduction in metal impurities (especially in iron impurity) has been achieved; the total radiation loss has been reduced to less than 30-40% of the input power. In these reduced radiation loss plasmas, the characteristics of ICRF heated plasmas have been studied intensively. With an increase in the ICRF heating power, a deterioration of the energy confinement time has been observed, indicating quantitative agreement with the Kaye-Goldston L-mode scaling. It is shown that the so-called profile consistency, which is the leading feature in neutral beam heated plasmas, also holds in ICRF heated plasma. It has been observed that the electron temperature profile only responds to the safety factor q(a) and does not change when the deposition profile is controlled by tailoring the k1 spectrum.

The effects of the ergodic magnetic limiter (EML) on edge modification, hydrogen recycling and impurity behavior are investigated in the HYBTOK-II tokamak. A plasma modification at the edge with the EML is confirmed by measuring the edge plasma parameters. An enhancement of recycling with the EML is demonstrated using computed tomography. Emissions from metal ions were reduced, while those from metal atoms and light impurities were increased with the EML. The reasons of this different effect on impurity behavior are discussed, based on an impurity control modeling in the EML, the magnetic field structure and a simple calculation on the penetration of neutral atoms into plasma.

A series of initial resonant helical divertor (RHD) experiments have been carried out in ohmically and auxiliary heated JIPP T-IIU plasmas. Disruptive and MHD instabilities make the interpretation of the RHD results difficult but an apparent increase in the energy confinement time is observed when the helical magnetic perturbation is applied. This may be due to the suppression of MHD activity or to a reduction in the edge convective heat losses. Magnetic island effects have been observed on the floating potential of a Langmuir probe array and energy scrape-off layer widths have been measured with and without helical perturbations during ICRF operation. Basic pump limiter data is presented including ion temperatures and C4+ impurity profiles. Energy confinement times are reported in ohmically and NBI heated discharges.

The confinement of ohmically heated stellarator plasma is studied experimentally. The energy confinement time ?E increases with the total rotational transform angle ?? in the case of ??/2? 4, while ?E decreases with ??/2? > 0.3, that is, q?(a) < 3-4, which is consistent with the Alcator scaling. When the rotational transform angle due to helical field ?h/2? is larger than 0.14, no major disruptions are observed even for q?(a) as low as 1.7 (?h/2?= 0.14, Ip ? 100 kA at Bt = 2.2 T), although strong MHD oscillations with low frequency (~1 kHz) appear when q?(a) < 2. Tangential injections of neutral beams are carried out for additional heating of stellarator and tokamak plasmas. With co-injection, the increase in ion temperature ?Ti at the center per injected power PN into the torus is 1.6 eV/kW in the stellarator and is 1.9 eV/kW in the tokamak configuration. In the case of counter-injection, ?Ti/PNis 1.6 eV/kW in both configurations. Lower hybrid heating (LHH) with the frequency of 0.8 GHz is applied. In both cases of stellarator and tokamak, ion heating of 0.8 eV/kW is observed with input power up to 160-180 kW. Electron cyclotron heating (ECH) by use of klystrons with frequency of 35.5 GHz is carried out. Preliminary results show heating of bulk electrons with the heating efficiency of 1.6 eV/kW. Current drive in steliarator plasma by lower hybrid wave is observed in the case of low density plasma (n?e ~5 × 1012 cm-3).

A 40 keV, 1.2 MW neutral beam injector for the Compact helical System (CHS) device has been constructed for the purpose of producing high-pressure plasma. The beam line consists of one ion source, and is characterized by a changeable injection angle for optimizing beam deposition and for investigating the confinement of high-energy ions. A technology for manufacturing cooling channels of extracting grids (both copper and molybdenum) is developed. The physical requirement on the neutral beam and the design features of the injector are presented. >

The energy confinement characteristics of ICRF-heated Tokamak plasmas are studied at high input power density approximately 2 MW m-3 volume averaged, on the JIPP T-IIU device (R=0.91 m, a=0.23 m). High electron and ion temperatures (Te approximately 2.5 keV, Ti approximately 2.0 keV, at each maximum) have been achieved by the operation at a plasma current IP of 280 kA, plasma line-averaged electron density ne of 7*1013 cm-3 and input power of 2 MW, with a suppression of total radiation loss (30-40% of the total input power) by a carbon coating on the vacuum vessel. The fraction of ICRF power absorbed by the plasma, alpha , is determined experimentally from the decay of the stored plasma energy just after the ICRF pulse is terminated. The value of alpha increases slightly with increasing electron density and decreases from 90 to 70% as the ICRF power is increased from 1 to 2 MW m-3 volume averaged. The global energy confinement time tau epsilon , defined by WP/(POH+ alpha Pr.f), decreases by a factor of 2 approximately 3 from that in Ohmic plasmas as the heating power increases up to 2 MW. It is found that the energy confinement time has a strong line-averaged electron density dependence as tau E varies as ne0.6, which is obtained by the use of the measured absorbed power, while the Kaye-Goldston scaling predicts tau E varies as ne0.26.

The influence of the filament on the discharge characteristics of the multicusp source (bucket source) is studied experimentally and theoretically. It is found that the magnetic field which is produced by the filament itself plays an important role to determine the trapped region of primary electrons by coupling with the magnetic multicusp field. Experimentally, arc efficiency varies drastically when only the direction of the filament current is changed. The measurements of the spatial profiles of the electron density and the temperature show that the plasma production occurs at the periphery of the chamber in the case of poor arc efficiency. Theoretical calculation also shows that in this case primary electrons are trapped in the local magnetic mirror of the multicusp field. The trapping condition depends on the magnetic field geometry around the filament, and, therefore, it is sensitive to the location of filament in the magnetic multicusp field as well as the strength and the direction of filament current.

炉心プラズマを核融合反応が持続する点火温度にまで加熱するためには, 高速中性粒子束あるいは高周波波動をプラズマに入射し, 中心部に伝送して, それが吸収され, プラズマのエネルギーとして蓄えられることが必要である。また, その運動量をプラズマ粒子に与えて非誘導駆動電流を発生し得れば, 定常トカマク炉の実現も想定される。これらのプラズマ加熱法及び電流駆動法の物理機構, 研究の概要について述べる。

A measurement of low frequency self-oscillating instability in a highly ionized magneto-plasma column is reported. Radial distribution of amplitude and phase velocity of the oscillation is observed by correlation technique. Radial profile of the plasma particle density consists of two regions each of which has a different radial scale length, n /(∂ n /∂ r ), and it is indicated that in each region, the oscillations with different frequency are localized. The oscillations have similar properties except for the difference in the frequency. All the properties of these oscillation can be described by the theory of resistive drift instability. A rational for the identifications of the oscillations is given.

PIG discharges with uniform magnetic field configuration and Divertor type field configuration are studied. Low frequency fluctuations in the plasma column are observed by a Langmuir probe and two photo-transisters. In the case of uniform field configuration, the frequencies of fluctuation agree well with the theory by Hoh and Simon. In the case of Divertor type field configuration, the amplitude of fluctuation is small compared with that for uniform field configuration. Plasma parameters (space potential, particle density and electron temperature) are measured for both magnetic field configurations. The most remarkable difference between two types of field configurations exists in radial electric field intensity. In a uniform configuration, the radial electric field E r is about 3–4 V/cm. Whereas in a Divertor type field configuration, E r is only 0.2 V/cm or less. This small E r results in a rather quiescent PIG discharge in a Divertor type field configuration.

The construction and the performance of a new type of plasma source, designed by applying a magnetic field similar to that in the divertor of the stellarator, are described. The plasma parameters are measured by Langmuir probes. A hollow plasma of He with a density of the order of 10 10 particles/cm 3 is produced along the separatrix line of force in this apparatus. The fluctuation of the plasma density is rather small. Since this new type plasma source has no electrode on the axis, it is possible to introduce it for some stationary torus machine.

Neutral beam injection heating is planned in JIPP T-II, a hybrid device of tokamak and stellarator. The injection power of 100 kW is expected for single beam line. The first injector has been installed. The design and the performance of JIPP T-II injector are described. The results of preliminary experiment to improve the source plasma are also given.

The heating systems for the new large helical device which has been designed as a joint effort among Japanese universities are reported. The conceptional design of the heating devices and basic hardware structures are discussed. ECH heating system (112 GH z ) with maximum power of around 10MW is proposed to produce and heat the net current-less plasma. For the further heatings, NBI (20 MW, 100 kV, H 0 ) and ICRF (9 MW, 30–90 MH z ) heating systems have been considered. High energy particle loss, which is a fairly serious problem in helical configurations (stellarator/heliotron), have been analyzed numerically. Monte-Carlo calculation shows that, for tangential injection, the beam energy over 100 kV is necessary to obtain the thermalization efficiency of above 60% inside the half radius. On the fast-wave (ICRF) heating, theoretical analysis of the wave field and velocity distribution function, which includes the loss cone effect, has been developed. In the case of 3 He minority heating, the orbit loss problem becomes small than in the case of proton minority heating. From these theoretical analysis and hardware considerations, the required thermalized power of around 20 MW can be obtained by ECH, NBI and ICRF heatings.

Neutral beam injection experiments are carried out in JIPP T-II, which is a hybrid device of stell arator and tokamak. Two neutral beam injectors are equipped tangentially in the direction of parallel (co-injection) and anti-parallel (counter-injection) to the plasma current. Neutral hydrogen beams of 22 keV, 60 kW (co-injection) and 30 kW (counter-injection) are injected into the ohmically heated stellarator plasma and also into the tokamak plasma. The bulk ion heating efficiency for co-injection is around 1.6 eV/kW in the case of stellarator and 1.9 eV/kW in the case of tokamak, while the heating efficiencies for counter-injection are about 1.6 eV/kW in both cases. The difference between the tokamak and the stellarator is considered to be caused by the enhanced orbit loss due to the helical ripples in the stellarator.

Recent experimental results in the Large Helical Device have indicated that a large pressure gradient can be formed beyond the stability criterion for the Mercier (high-n) mode. While the stability against an interchange mode is violated in the inward-shifted configuration due to an enhancement of the magnetic hill, the neoclassical transport and confinement of high-energy particle are, in contrast, improved by this inward shift. Mitigation of the unfavourable effects of MHD instability has led to a significant extension of the operational regime. Achievements of the stored energy of I MJ and the volume-averaged beta of 3% are representative results from this finding. A confinement enhancement factor above the international stellarator scaling ISS95 is also maintained around 1.5 towards a volume-averaged beta, (beta), of 3%. Configuration studies on confinement and MHD characteristics emphasize the superiority of the inward-shifted geometry to other geometries. The emergence of coherent modes appears to be consistent with the linear ideal MHD theory; however, the inward-shifted configuration has reduced heat transport in spite of a larger amplitude of magnetic fluctuation than the outward-shifted configuration. While neoclassical helical ripple transport becomes visible for the outward-shifted configuration in the collisionless regime, the inward-shifted configuration does not show any degradation of confinement deep in the collisionless regime (nu* < 0.1). The distinguished characteristics observed in the inward-shifted configuration help in creating a new perspective of MHD stability and related transport in net current-free plasmas. The first result of the pellet launching at different locations is also reported.

The azimuthal drift velocity of ions is measured with electrostatic ion wave propagating across the magnetic field. The relation between ion-neutral collision and the drift velocity across the magnetic field is investigated as a function of the magnetic field.

An explanation is presented for the existence of hydromagnetic waves in a partialiy ionized gas discharge that do not damp in a weak magnetic field and tend to have a phase velocity of the sound wave in a stronger field. Equation for a low frequency osciliation with smali amplitude in a plasma in a uniform magnetic field are used in the analysis. (C.E.S.)

Numerical calculations of the particle trajectory in a Divertor type plasma source show that the radius of the scattering center is proportional to (mass)1/4.

The role of the magnetic field in the behavior of magnetic multipole discharge is investigated experimentally by using a bucket type ion source with electromagnets. There exist two modes of discharge which transit each other in the critical current. The relation between the plasma density profile and the magnetic field is experimentally clarified by taking into account the filament position, that is, (1) the magnetic field strength near the filament must be below 80±20 G to get no peaked density profile, and (2) a region of a density uniformity better than 10% is obtained within 20±10 G contour. The dependence of the arc efficiency and the ion species composition on the magnetic field strength are studied and discussed.