A. Iiyoshi

Design Study for the Large Helical Device

The Large Helical Device (LHD) is a Heliotron/torsatron-type superconducting helical confinement fusion device. The design study is described. The goal of the LHD is to demonstrate high energy confinement and high beta in a helical device, which are necessary steps toward a helical reactor system.

Initial physics achievements of large helical device experiments

The Large Helical Device (LHD) experiments [O. Motojima, et al., Proceedings, 16th Conference on Fusion Energy, Montreal, 1996 (International Atomic Energy Agency, Vienna, 1997), Vol. 3, p. 437] have started this year after a successful eight-year construction and test period of the fully superconducting facility. LHD investigates a variety of physics issues on large scale heliotron plasmas (R=3.9 m, a=0.6 m), which stimulates efforts to explore currentless and disruption-free steady plasmas under an optimized configuration. A magnetic field mapping has demonstrated the nested and healthy structure of magnetic surfaces, which indicates the successful completion of the physical design and the effectiveness of engineering quality control during the fabrication. Heating by 3 MW of neutral beam injection (NBI) has produced plasmas with a fusion triple product of 8×1018 keV m−3 s at a magnetic field of 1.5 T. An electron temperature of 1.5 keV and an ion temperature of 1.4 keV have been achieved. The maximum s...

Scalings of energy confinement and density limit in stellarator/heliotron devices

The paper presents a study of empirical scaling of energy confinement observed experimentally in stellarator/heliotron devices (Heliotron E, Wendelstein VII-A, L2, Heliotron DR) for plasmas heated by electron cyclotron heating and/or neutral beam injection. The proposed scaling of the gross energy confinement time is: , where P is the absorbed power (MW), n is the line average electron density (1020 m?3), B is the magnetic field strength on the plasma axis (T), a is the average minor radius (m) and R is the major radius (m). The empirical scaling of the density limit obtainable under the optimum condition is proposed to be: . These scalings for helical systems are compared with those in tokamaks. The energy confinement scaling has a similar power dependence as the L-mode scaling of tokamaks. The density limit scaling for helical systems seems to indicate an upper limit of the achievable density similar to that in many tokamaks. From the energy confinement time and the density limit , a beta limit can be derived: , which can be lower than the stability/equilibrium beta limit. Thus, from the viewpoint of designing a machine, the values of B, a and R should be selected with care because the dependence of the confinement time (or n?ET) and of the above beta limit on these values is different.

The large helical device (LHD) helical divertor

The Large Helical Device (LHD) now under construction is a heliotron/torsatron device with a closed divertor system. The edge LHD magnetic structure has been studied in detail. A peculiar feature of the configuration is the existence of edge surface layers, a complicated three dimensional magnetic structure which does not, however, seem to hamper the expected divertor functions. Two divertor operational modes are being considered for the LHD experiment-high density, cold radiative divertor operation as a safe heat removal scheme and high temperature divertor plasma operation. In the latter operation, a divertor plasma with a temperature of a few keV, generated by efficient pumping, is expected to lead to a significant improvement in core plasma confinement. Conceptual designs of the LHD divertor components are under way

Progress summary of LHD engineering design and construction

In March 1998, the LHD project finally completed its eight year construction schedule. LHD is a superconducting (SC) heliotron type device with R = 3.9 m, ap = 0.6 m and B = 3 T, which has simple and continuous large helical coils. The major mission of LHD is to demonstrate the high potential of currentless helical-toroidal plasmas, which are free from current disruption and have an intrinsic potential for steady state operation. After intensive physics design studies in the 1980s, the necessary programmes of SC engineering R&D was carried out, and as a result, LHD fabrication technologies were successfully developed. In this process, a significant database on fusion engineering has been established. Achievements have been made in various areas, such as the technologies of SC conductor development, SC coil fabrication, liquid He and supercritical He cryogenics, development of low temperature structural materials and welding, operation and control, and power supply systems and related SC coil protection schemes. They are integrated, and nowadays comprise a major part of the LHD relevant fusion technology area. These issues correspond to the technological database necessary for the next step of future reactor designs. In addition, this database could be increased with successful commissioning tests just after the completion of the LHD machine assembly phase, which consisted of a vacuum leak test, an LHe cooldown test and a coil current excitation test. These LHD relevant engineering developments are recapitulated and highlighted. To summarize the construction of LHD as an SC device, the critical design with NbTi SC material has been successfully accomplished by these R&D activities, which enable a new regime of fusion experiments to be entered.

The next large helical devices

Helical systems have the strong advantage of inherent steady‐state operation for fusion reactors. Two large helical devices with fully superconducting coil systems are presently under design and construction. One is the LHD (Large Helical Device) [Fusion Technol. 17, 169 (1990)] with major radius=3.9 m and magnetic field=3–4 T, that is under construction during 1990–1997 at NIFS (National Institute for Fusion Science), Nagoya/Toki, Japan; it features continuous helical coils and a clean helical divertor focusing on edge configuration optimization. The other one in the W7‐X (Wendelstein 7‐X) [in Plasma Physics and Controlled Fusion Nuclear Research, 1990, (International Atomic Energy Agency, Vienna, 1991), Vol. 3, p. 525] with major radius=5.5 m and magnetic field=3 T, that is under review at IPP (Max‐Planck Institute for Plasma Physics), Garching, Germany; it has adopted a modular coil system after elaborate optimization studies. These two programs are complementary in promoting world helical fusion resea...

Studies of currentless, high-beta plasma in the Heliotron E device

A currentless plasma with a volume-averaged beta value of 2% has been produced with neutral beam heating. Target plasmas were created by second harmonic resonance heating with electron cyclotron waves (150–350 kW and 53.2 GHz) at a magnetic field strength of 0.94 T. Neutral beam injection (23–30 keV and 1.3−2.6 MW) was used to heat the plasma further. MHD stable and unstable high-beta plasmas were observed. The Q-mode plasmas were produced with the help of intense neutral gas puffing. Properties of the MHD activity and confinement of high-beta plasmas are discussed and compared with theoretical studies.

Analysis of the plasma-wall interaction in the Heliotron E device

The plasma-wall interaction (PWI) of the currentless plasmas with temperature To, Tio ≤ 1.1 keV, density Ne = (2–10)× 1013/cm3, and volume-averaged beta value of β

Plasma confinement studies in LHD

≤ 2% was investigated. We have observed that PWI took place mainly where the divertor field line intersected the chamber wall (called divertor traces). Boundary plasmas were measured with electrostatic probes, which showed the presence of the divertor region with the parameters in the range of Ned = 1010–1011/cm3 and Ted = 10–50 eV. Surface analysis techniques (ESCA, AES, and RBS) were applied to analyze the surface probes (Si, graphite and stainless steel) and the test pieces (SiC, TiC, and stainless steel), which were irradiated by plasmas for short and long times respectively.

ICRF heating of currentless plasma in Heliotron E

The initial experiments on the Large Helical Device (LHD) have extended confinement studies on currentless plasmas to a large scale (R = 3.9 m, a = 0.6 m). Heating by NBI of 3 MW produced plasmas with a fusion triple product of 8 × 1018m-3keVs at a magnetic field strength of 1.5 T. An electron temperature of 1.5 keV and an ion temperature of 1.1 keV were achieved simultaneously at a line averaged electron density of 1.5 × 1019 m-3. The maximum stored energy reached 0.22 MJ with neither unexpected confinement deterioration nor visible MHD instabilities, which corresponds to β = 0.7%. Energy confinement times reached a maximum of 0.17 s. A favourable dependence of energy confinement time on density remains in the present power density (~40 kW/m3) and electron density (3 × 1019 m-3) regimes, unlike the L mode in tokamaks. Although power degradation and significant density dependence are similar to the conditions on existing medium sized helical devices, the absolute value is enhanced by up to about 50% from the International Stellarator Scaling 95. Temperatures of both electrons and ions as high as 200 eV were observed at the outermost flux surface, which indicates a qualitative jump in performance compared with that of helical devices to date. Spontaneously generated toroidal currents indicate agreement with the physical picture of neoclassical bootstrap currents. Change of magnetic configuration due to the finite β effect was well described by 3-D MHD equilibrium analysis. A density pump-out phenomenon was observed in hydrogen discharges, which was mitigated in helium discharges with high recycling.