大型ヘリカル装置用超伝導マグネットシステム

Stability and quench experiments of a Nb-Ti cable-in-conduit superconductor with bare strands have been carried out to determine the performance of the poloidal coil for the Large Helical Device (LHD). The conductor was formed into a double-pancake coil named IV-S and was mounted on TOKI-PF-a previously tested R&D coil. In excitation tests at 7.4-8.3 K, the IV-S coil reached the critical currents without premature quenches. Stability tests at 7.5 K indicated that the limiting current exceeds 20 kA, which is the nominal operating current of the LHD poloidal coils. These results demonstrated that the stability of the chosen conductor with bare strands is high enough for the LHD.< <ETX xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">&gt;</ETX>

Review of SC magnet technologies developed in LHD project

Large Helical Device (LHD) experiments are executed with the collaboration of many universities in Japan. Therefore, remote participation plays an important role. In this paper, the authors introduce the current remote participation facilities for these experiments. National Institute for Fusion Science (NIFS) remote participation facilities fall into three categories. The first, remote access, allows direct access to the experimental network. This is the most flexible way to use computer resources remotely. For this purpose, virtual private network (VPN) service is available for coresearchers. In addition, several laboratories are connected directly via SINET3. Once researchers connect to the network, they can use the computer as if they are at the NIFS. The next is remote data reference. Users can view experimental data in a Web browser. Also, they can use a browser to retrieve basic information about experiments that is stored in a relational database. The last is video services. Remote researchers can use a videoconference system to communicate with researchers at the NIFS, and they can use Web browsers to watch the main monitor image displayed in the control room. For security reasons, the entire network is protected by a firewall, and one-time password authentication is used to realize secure VPN access.

We have observed EUV spectra of xenon ions from the large helical device (LHD) at the National Institute for Fusion Science in Toki in the wavelength range of 10–17 nm using a high resolution SOXMOS spectrometer. A small quantity of xenon gas was injected into the large helical device. In some cases, the plasma evolution was stable and a steady discharge was obtained for several seconds, but sometimes the plasma underwent radiation collapse and rapid cooling and in this situation the EUV yield was significantly increased. Investigation of the spectra showed that during the heating phase and in a stable plasma, the emission was dominated by ions with open 4s and 4p subshells, while during radiation collapse, the spectra were dominated by lines from species with open 4d subshells. From a comparison of these spectra with theoretical data from atomic structure calculations and also with charge state specific data generated at Tokyo Metropolitan University, it was possible to make tentative assignments of the strongest lines arising from 4d–4f and 4p–4d transitions in Xe XVII and XVIII.

Simulation of the Large Helium Refrigeration Plant for LHD

The Large Helical Device (LHD) is a plasma physics experimental device with a magnetic stored energy of 960 MJ, consisting of two superconducting (SC) helical coils and six SC poloidal coils. The trial operation and the first plasma discharge of the Phase I project for LHD were completed on 31 March 1998 as initially planned. The second experimental campaign was conducted with additional heating using two NBI devices. The third campaign started in June 1999 and was completed in January 2000. Many plasma heating tests with a plasma field of up to 2.90 T were carried out. Major test results on the SC magnet system for LHD are as follows: (1) The LHD cryogenic system was successfully operated for 13 400 h and proved its high reliability. (2) A central field of 2.91 T at a radius of 3.60 m was achieved at an inner helical coil current of 11.08 kA, a middle helical coil current of 11.83 kA and an outer helical coil current of 12.02 kA. (3) All six poloidal coils were excited stably. (4) Nine flexible SC bus lines with a total length of 497 m were operated stably and safely.

Since 1989, National Institute for Fusion Science (NIFS) promotes academic approaches toward exploration of fusion science for steady-state helical reactor and realizes establishment of comprehensive understanding of toroidal plasmas as an inter-university research organization and a key center of worldwide fusion research. Three key projects, Large Helical Device (LHD) project, simulation science project, and fusion engineering project are organized for early realization of net current free fusion reactor. The LHD has been producing high-performance plasmas comparable to those of large tokamaks, and several new findings with regard to plasma physics have been obtained. The simulation science project contributes understanding and systemizing the physical mechanism of plasma confinement in fusion plasmas and exploring complexity science of a plasma for realization of the numerical test reactor. In the fusion engineering project, the design of the helical fusion reactor has been progressed based on the development of superconducting coils, the blanket, fusion materials and tritium handling.

Vertical profiles of edge impurity emissions have been measured in upper half region of elliptical plasmas at horizontally elongated plasma cross section in large helical device (LHD). The vertical profiles near upper O-point located just below helical coil are analyzed to study the plasma edge boundary of the ergodic layer consisting of stochastic magnetic field lines with connection lengths of 30 ≤ Lc ≤ 2000 m. As a result, C3+ ion emitting CIV spectrum is identified as the ion existing in the farthest edge of the ergodic layer. The peak position of CIV (312.4 Å: 1s23p 2P1/2,3/2-1s22s 2S1/2) vertical profile does not change at all in a wide temperature range of 150 ≤ Te(ρ = 1) ≤ 400 eV, whereas it moves inside the ergodic layer when Te(ρ = 1) is reduced below a threshold temperature, e.g., 130 eV at Rax = 3.75 m configuration. It is found that the C3+ ion exists at the boundary between ergodic layer and open magnetic filed layer at which the Lc distributes in lengths of 5 to 30 m. The result indicates that the edge boundary near the O-point in LHD is determined by a starting point of the open filed layer, where a tokamak-like steeper edge temperature gradient is formed, although the edge boundary is quite obscure at the X-point region. Any plasma does not exist between the edge boundary and the vacuum vessel. The CIV profile at the O-point is simulated using a three-dimensional edge transport code of EMC3-EIRENE in which the magnetic field structure in vacuum is used for the ergodic layer. A clear discrepancy of 8 mm is found in the peak positions of CIV between measurement and simulation for magnetic configurations with thick ergodic layer, i.e., Rax =3.90 m, while only a small discrepancy of 3 mm is observed for those with relatively thin ergodic layer, i.e., Rax = 3.75 m. It suggests that the discrepancy is caused by a modification of the magnetic filed due to the presence of plasma pressure.

Mechanical analysis and fabrication of the R&D forced-flow helical coil (TOKI-PF)

Enhancement of hydrogen isotope retention capacity for the impurity deposited tungsten by long-term plasma exposure in LHD

The Fusion Engineering Research Project (FERP) at the National Institute for Fusion Science (NIFS) is conducting conceptual design activities for the LHD-type helical fusion reactor FFHR-d1A. This paper newly defines two design options, ‘basic’ and ‘challenging.’ Conservative technologies, including those that will be demonstrated in ITER, are chosen in the basic option in which two helical coils are made of continuously wound cable-in-conduit superconductors of Nb3Sn strands, the divertor is composed of water-cooled tungsten monoblocks, and the blanket is composed of water-cooled ceramic breeders. In contrast, new ideas that would possibly be beneficial for making the reactor design more attractive are boldly included in the challenging option in which the helical coils are wound by connecting high-temperature REBCO superconductors using mechanical joints, the divertor is composed of a shower of molten tin jets, and the blanket is composed of molten salt FLiNaBe including Ti powers to increase hydrogen solubility. The main targets of the challenging option are early construction and easy maintenance of a large and three-dimensionally complicated helical structure, high thermal efficiency, and, in particular, realistic feasibility of the helical reactor.

Distributions of stress in helical coils with the geodesic winding based on virial theorem are analyzed theoretically and numerically. A force-balanced coil (FBC) is a multi-pole helical hybrid coil combining toroidal field (TF) coils and a solenoid helically wound on a torus. The combination reduces the net electromagnetic force in the direction of major radius by canceling the centering force due to the TF coil current and the hoop force due to the solenoid current. The FBC concept was extended using the virial theorem which shows that strength of magnetic field is restricted by working stress in the coils and their supporting structure. High-field coils should accordingly have same averaged principal stresses in all directions which is named the virial-limit condition. Recently, we made a model FBC which were neither impregnated with epoxy resin nor reinforced with stainless steel wires. Using acoustic emission measurements, we found that the wires vibrated in response to electromagnetic force changes. Since FBC winding is modulated to reduce the tilting force, the winding is slightly similar to but different from the shortest geodesic trajectory and has no tensile load. In order to reduce the vibration, the geodesic winding is expected to be effective. In this work, we analyze the effect of the winding modulations including the geodesic modulation for the stress distribution of helical windings.

A multichannel far-infrared laser interferometer system has been developed for the measurement of the spatial and temporal behaviors of the electron density on the Large Helical Device (LHD) at the National Institute for Fusion Science. Of several candidates for high power laser sources a 119 μm CH3OH laser was taken as a probing light. The optical configuration is of the Michelson interferometer type with 13 channels. The optical system of the interferometer is mounted on a massive stainless frame floated on three vibration isolating mounts. The resolution of the fringe counters is 1/100 fringes, and it corresponds to a line averaged density of 5.6×1016 m−3 at the central chord. Preliminary results have been obtained in the initial experiments on the LHD.

Magnetohydrodynamic (MHD) stability studies of low-mode-number free-boundary kink modes in a finite-β Large Helical Device (LHD) [A. Ilyoshi, et al. Fusion Technology 17, 169 (1990)] equilibrium series with self-consistent bootstrap current find some serious free-boundary modes. They indicate that care must be taken in the design of high-β experiments. Since the LHD configuration is flexible, with the possibility of introducing or modifying dipole and quadrapole fields, unbalancing the currents in the helical coils, operating with a high-temperature divertor, and changing the collisionality regime by working with different values of temperature and density, paths to high-β operation should exist. Comparison of the experimental results with these theoretical predictions for the studied equilibrium sequence will provide understanding of the MHD stability properties of LHD.

Research and development of superconductors and superconducting coils for the Large Helical Device