Y. Fujiwara

Bio: Y. Fujiwara is an academic researcher from Japan Atomic Energy Research Institute. The author has contributed to research in topics: Beam (structure) & Ion beam. The author has an hindex of 9, co-authored 25 publications receiving 240 citations.

Negative hydrogen ion source for TOKAMAK neutral beam injector (invited)

Abstract: Intense negative ion source producing multimegawatt hydrogen/deuterium negative ion beams has been developed for the neutral beam injector (NBI) in TOKAMAK thermonuclear fusion machines. Negative ions are produced in a cesium seeded multi-cusp plasma generator via volume and surface processes, and accelerated with a multistage electrostatic accelerator. The negative ion source for JT-60U has produced 18.5 A/360 keV (6.7 MW) H− and 14.3 A/380 keV (5.4 MW) D− ion beams at average current densities of 11 mA/cm2 (H−) and 8.5 mA/cm2 (D−). A high energy negative ion source has been developed for the next generation TOKAMAK such as the International Thermonuclear Experimental Reactor (ITER). The source has demonstrated to accelerate negative ions up to 1 MeV, the energy required for ITER. Higher negative ion current density of more than 20 mA/cm2 was obtained in the ITER concept sources. It was confirmed that the consumption rate of cesium is small enough to operate the source for a half year in ITER-NBI without maintenance.

High power negative ion sources for fusion at the Japan Atomic Energy Research Institute (invited)

Abstract: Technologies producing high power negative ion beams have been highly developed over the years at Japan Atomic Energy Research Institute for use in neutral beam injectors for heating the thermonuclear fusion plasmas. At present, it is possible to produce multiampere H−/D− ion beams quasicontinuously at energies of more than a few hundred keV with a good beam optics of beamlet divergence of a few mrad. Based on these technologies, two research and development projects have been initiated; one is to develop a 22 A/500 keV/10 s D− ion source for the neutral beam injector for JT‐60U, and the other is to develop a 1 A/1 MeV/60 s H− ion source to demonstrate high current negative ion acceleration up to the energy of 1 MeV, the energy required for the neutral beam injector for the International Thermonuclear Experimental Reactor.

Development of a 500 keV, 22 A D− ion source for the neutral beam injector for JT‐60U

Abstract: The first results of the performance test of the large negative ion source for a JT‐60U negative‐ion‐based neutral beam injector (N‐NBI) are presented. The ion source consists of a cesium seeded multicusp plasma generator, where negative ions are produced via volume and surface processes, a 110 cm×45 cm multiaperture extractor, and a three‐stage electrostatic accelerator. After negative ion production and voltage holding tests in test stands, the ion source was installed in the N‐NBI system and the full power test began. Up to now, the ion source has produced 400 keV, 5.9 A (2.4 MW) D− ion beams, the world highest D− current and beam power, with a pulse duration of 0.1 s.

Negative ion sources for neutral beam injection into fusion machines

Abstract: Plasma heating and current drive in future machines for magnetically confined thermonuclear fusion will require neutral beam injectors based upon negative ion sources. Although the technology of filament arc ion sources is considered adequately mature for large scale injectors, such as for ITER, ion source efficiency and reliability need to be improved. The power loading of the source and accelerator are especially problematic. This is particularly important for long pulse (1000 s) operation. Investigations with the Kamaboko source, a small scale model of the ITER arc source, illustrate the importance of the conditions of the plasma grid for reducing electron loading of the accelerator.

Neutral beams for the International Thermonuclear Experimental Reactor

Abstract: Receiving higher emphasis on the neutral beam (NB) off-axis current drive, the NB system is being highlighted for the steady state operation of the International Thermonuclear Experimental Reactor (ITER). To fulfill the physics requirement of heating and current drive, the NB system delivers ∼50 MW of D0 beams at 1 MeV into the ITER plasmas. The NB injector was designed so as to minimize the axial length, to avoid cost impact on the building. It was estimated by nuclear analyses that the insulation gas around the beam source would cause radiation induced conductivity, which would result in a power dissipation of >100 kW in the gas itself. As a result the present design utilizes vacuum insulation around the beam source. Since the vacuum pressure inside/outside the beam source ranges 10−1–10−2 Pa, both gas (glow) and vacuum arc discharges are taken into account in the design.

Overview of the RF source development programme at IPP Garching

Abstract: The development of a large-area RF source for negative hydrogen ions, an official EFDA task agreement, is aiming at demonstrating ITER-relevant ion source parameters. This implies a current density of 200?A?m?2 accelerated D? ions at a source filling pressure of ?0.3?Pa and an electron-to-ion ratio of ?1 from an extraction area similar to the positive-ion based sources at JET and ASDEX Upgrade and for pulse lengths of up to 1?h. The work is progressing along three lines in parallel: (i) optimization of current densities at low pressure and electron/ion ratio, utilizing small extraction areas (<0.01?m2) and short pulses (<6?s), in this parameter range the ITER requirements are met or even exceeded; (ii) investigation on extended extraction areas (<0.03?m2) and pulse lengths of up to 3600?s and (iii) investigation of a size-scaling on a half-size ITER plasma source. Three different test beds are being used to carry out these investigations in parallel. An extensive diagnostic and modelling programme accompanies the activities. The paper discusses the recent achievements and the status in these three areas of development.

Internal transport barriers in tokamak plasmas

Abstract: Internal transport barriers in tokamak plasmas are explored in order to improve confinement and stability beyond the reference scenario, used for the ITER extrapolation, and to achieve higher bootstrap current fractions as an essential part of non-inductive current drive. Internal transport barriers are produced by modifications of the current profile using external heating and current drive effects, often combined with partial freezing of the initial skin current profile. Thus, formerly inaccessible ion temperatures and QDTeq values have been (transiently) achieved. The present paper reviews the state of the art of these techniques and their effects on plasma transport in view of optimizing the confinement properties. Implications and limits for possible steady state operations and extrapolation to burning plasmas are discussed.

Status of the ITER neutral beam injection system (invited)a)

Abstract: The ITER neutral beam injectors are the first injectors to be designed to operate under conditions and constraints similar to those that will be encountered with a fusion reactor. The injectors will use a single large ion source and accelerator that will produce 40A D− 1MeV beams for pulse lengths of up to 3600s. The accelerated ion beams will be neutralized in a gas (D2) neutralizer which is subdivided into four vertical channels to reduce the gas flow into the injectors that is needed to produce optimum target for neutralization. These injectors will have to operate in a hostile radiation environment and they will become highly radioactive due to the neutron flux from ITER. The design has been modified recently to have a rectangular vacuum vessel with a removable lid that allows vertical access to, and maintenance of, the beamline components, the incorporation of an absolute all metal valve at the exit of the injector, the choice of a rf driven ion source as the reference design of ion source, and to have a high voltage deck incorporating the various auxiliary power supplies in air rather that under high pressure SF6. A major development is that it has been agreed that a Neutral Beam Test Facility (NBTF) will be set up at Padua, Italy. The NBTF will consist of two test beds: one of which will be capable of operating a complete injector at full performance. The second will be an ion source test bed, which will be used for the development and testing, to full performance, of the large negative ion source.

Positive and negative ion sources for magnetic fusion

Abstract: The positive or negative ion sources which form the primary components of neutral beam injection systems used in controlled nuclear fusion using magnetic confinement have to meet simultaneously several demanding requirements. This paper describes the underlying physics of modern positive ion sources, which provide the required high proton fraction (>90%) and high current density (/spl ap/2 kA/m/sup 2/) at a low source pressure (0.4 Pa) with a high electrical efficiency and uniformity across the accelerator grids. The development of negative ion sources, which are required if high energy neutral beams are to be produced, is described, and the present understanding of the physics of negative ion production in sources is explained. The paper reports that negative ion sources have achieved many of the parameters required of sources for the neutral beam injectors of future fusion devices and reactors, >200 A/m/sup 2/ of D/sup -/ at low source pressure, <0.3 Pa, with a low co-extracted electron content. The development needed to meet all the requirements of future systems is briefly discussed.

Physics and engineering design of the accelerator and electron dump for SPIDER

Abstract: The ITER Neutral Beam Test Facility (PRIMA) is planned to be built at Consorzio RFX (Padova, Italy). PRIMA includes two experimental devices: a full size ion source with low voltage extraction called SPIDER and a full size neutral beam injector at full beam power called MITICA. SPIDER is the first experimental device to be built and operated, aiming at testing the extraction of a negative ion beam (made of H− and in a later stage D− ions) from an ITER size ion source. The main requirements of this experiment are a H−/D− extracted current density larger than 355/285 A m−2, an energy of 100 keV and a pulse duration of up to 3600 s.Several analytical and numerical codes have been used for the design optimization process, some of which are commercial codes, while some others were developed ad hoc. The codes are used to simulate the electrical fields (SLACCAD, BYPO, OPERA), the magnetic fields (OPERA, ANSYS, COMSOL, PERMAG), the beam aiming (OPERA, IRES), the pressure inside the accelerator (CONDUCT, STRIP), the stripping reactions and transmitted/dumped power (EAMCC), the operating temperature, stress and deformations (ALIGN, ANSYS) and the heat loads on the electron dump (ED) (EDAC, BACKSCAT).An integrated approach, taking into consideration at the same time physics and engineering aspects, has been adopted all along the design process. Particular care has been taken in investigating the many interactions between physics and engineering aspects of the experiment. According to the 'robust design' philosophy, a comprehensive set of sensitivity analyses was performed, in order to investigate the influence of the design choices on the most relevant operating parameters.The design of the SPIDER accelerator, here described, has been developed in order to satisfy with reasonable margin all the requirements given by ITER, from the physics and engineering points of view. In particular, a new approach to the compensation of unwanted beam deflections inside the accelerator and a new concept for the ED have been introduced.