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Various innovative divertor concepts have been developed to improve the LHD plasma performance. They are two divertor magnetic geometries (helical divertor configurations with and without n/m=1/1 island) and two operational scenarios (confinement improvement by generating high temperature divertor plasma and simultaneous achievement of radiative cooling and H-mode-like confinement improvement). In addition, technological development of new efficient hydrogen pumping schemes are being pursued for enhancing the divertor control capability. 16 refs., 4 figs.

In order to find out transport models applicable to transport simulation of next generation helical devices, we compare the thermal diffusivities predicted by some theoretical models with experimental data for present medium size helical devices (CHS (Compact Helical System), Heliotron E and W7AS (Wendelstein 7 AS)) with different magnetic configurations. We obtain the similar results in the three devices in spite of their difference of the magnetic configurations, except for the g-mode model. 9 refs., 2 figs.

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).

For the boron films made by a d.c. glow discharge for a gas mixture of diborane and helium, numerous surface properties were intestigated and the characteristics of the boron film as a plasma facing material in fusion devices were evaluated. The ability of the boron film to getter the oxygen was estimated as about 10 21 O atoms m -2 . The gettered oxygen in the boron film was observed in the surface to a depth of less than about 30 nm. The hydrogen concentration of the boron film made at room temperature was approximately 30%. The hydrogen in the film desorbed at about 573 K. This desorption temperature was considerably lower than that of graphite. These results directly indicate that the boron film coated on the graphite wall can largely reduce the particle recycling, in addition to the reduction in the oxygen impurity level in the plasma. After the oxygen gettering, the oxygen existed in the form of B—O bonds. It is also noted that the boron film absorbed both the oxygen and the carbon after exposure to the atmosphere.

The world largest superconducting fusion machine LHD (Large Helical Device) is under construction in Japan, aiming at steady-state operations. Its basic control system consists of UNIX computers, FDDI/Ethernet LANs, VME multiprocessors and VxWorks real-time OS. For flexible and reliable operations of the LHD machine a cooperative distributed system with more than 30 experimental equipments is controlled by the central computer and the main timing system, and is supervised by the main protective interlock system. Intelligent control systems, such as applications of fuzzy logic and neural networks, are planed to be adopted for flexible feedback controls of plasma configurations in addition to the classical PID control scheme. Design studies of the control system and related R&D programs with coil-plasma simulation systems are now being performed. The construction of the LHD control building on a new site will begin in 1995 after finishing the construction of the LHD Experimental Building, and the hardware construction of the LHD central control equipment will be started in 1996. The first plasma production by means of this control system is expected in 1997.

The Large Helical Device (LHD) is a non-tokamak type superconducting toroidal fusion apparatus. The LHD project contributes to the physics and technological developments exploring the currentless and steady-state plasmas, which have the high performance of temperature, density, confinement time and beta values being extrapolatable to the reactor condition. This paper reviews the major progress of the SC magnet technology attained from R&D programs, design and construction of LHD.

The frequency characteristic of AC losses in aluminum-stabilized superconducting conductors are investigated in order to elucidate their loss mechanism under changing transverse magnetic field. AC loss measurement is carried out for the short and straight sample conductor for the helical coil system of the Large Helical Device. The sample of about 50 cm in length is subjected to an AC ripple field with a frequency range from 0.05 to 337 Hz superimposed on the bias field. The experimental results show that the loss feature much depends both on direction of the transverse magnetic field and location of the stabilizer relative to the strand bundle. These results are quantitatively explained by the theoretical analysis. >

The stability and limiting current of a NbTi forced-cooled superconducting coil have been investigated experimentally and analytically at the subcooled condition below 4.5 K. The test coil was wound in a one layer solenoid from a cable-in-conduit type conductor of critical current of 5.1 kA at 4.2 K and 6.5 T, whose total length was 10 m. It was installed in the bore of a back-up superconducting magnet that provided field 8 T at the test coil. The stability margin of the coil as a function of normalized current (transport current/critical current) were measured. The results are compared with the formula proposed by Dresner and reported to predictions of stability margin and limiting current at subcooled condition below 4.5 K. >

Since a forced-flow superconducting coil consists of metallic conductors and organic insulation with relatively low mechanical strength, a knowledge of mechanical behavior and properties of the interlaminar insulation is of great importance. Shear fracture tests of interlaminar organic insulation under biaxial stresses were carried out at low temperature, by using test specimens made of stainless steel pieces bonded with impregnated organic insulation (glass-kapton-glass). Inclination of adhesive surface was varied to get combinations of the debonding stresses and to construct the critical curve of the debonding fracture. The test results show that the critical curve approximately agreed with the Mohr-Coulomb criterion and that shear fracture resistance of the bonding interface enhanced with the increase of compressive stress value. >

A DC 75 kA power supply system was constructed to test the SC (superconducting) R&D (research and development) cables and magnets for the Large Helical Device (LHD). It consists of three 25 kA unit banks. A unit bank has two double-star-rectifier connections with the inter-phase reactors. A digital feedback control method is applied to the automatic current regulation (ACR) in each unit bank. For shortening the dead time of the feedback process, a new algorithm of a digital phase controller for the ACR is investigated. A Bode diagram of the feedback process is directly measured. It is confirmed that the dead time of the feedback process is reduced to one sixth, and that the feedback gain of PID (proportional, integral and differential) compensation is improved by a factor of two from the original method. >

A flexible superconducting bus-line is proposed as an electrical feeder between the superconducting coils of the Large Helical Device (LHD) and the device's power supplies. The bus-line consists of superconducting cables and a cryogenic flexible transfer-line. A specially developed aluminum stabilized NbTi/Cu compacted strand cable satisfies requirements for large current capacity, high stability, high reliability and flexibility. A full-scale model with a length of 20 m was designed and constructed to investigate the feasibility and performance of the superconducting bus-line. Its fabrication, transportation, installation, cooling and excitation tests were successfully carried out. The bus-line was very stable and could be excited up to 40 kA (rated current is 30 kA) without a quench. The stability, current distribution and heat load were also measured. >

The Advanced Toroidal Facility (ATF) torsatron [Fusion Technol. 10, 179 (1986)] has completed experiments focusing on microwave scattering measurements of density fluctuations and transport studies utilizing the modulation of dimensionless parameters. Microwave scattering measurements of electron density fluctuations in the core of low‐collisionality electron cyclotron heated (ECH) plasmas show features that might be evidence of trapped electron instabilities. Starting from gyro‐Bohm scaling, the additional dependence of confinement on the dimensionless parameters ν* and β (collisionality and beta) has been investigated by modulating each of these parameters separately, revealing the additional favorable dependence, τE∝τgBν*−0.18β+0.3.

Excitation experiments for one of the R&D coils called TOKI-MC have been carried out to simulate the condition of excitation of helical coils for the Large Helical Device (LHD). TOKI-MC is a pool-cooled twisted solenoid coil with aluminum stabilized superconductors that were obliquely wound around an elliptical bobbin. In all nominal current excitations, quenches were initiated around 17 kA while the design current was 20 kA. There was no training effect. Signals of acoustic emission (AE) sensors indicated that the quenches were triggered by mechanical perturbations. It was found that the conductor of TOKI-MC moved freely without restriction due to the electromagnetic force, and that the winding moved reversibly. >

The module coil TOKI-MC is a twisted solenoid coil wound with an aluminum stabilized superconductor developed as an R&D program for the Large Helical Device (LHD). The TOKI-MC can simulate the conductor and winding structure cooled by pool boiling helium, the twisted winding and the large electromagnetic force of the helical coils for LHD. The TOKI-MC was designed as a cryostable coil at an operating current of 20 kA, but the coil quenched around 17 kA in excitation tests. The cause of quenches was thought to be the result of wire movement. Stability tests were also carried out and the measured recovery current was less than the designed value. The degradation of recovery current was due to the excess magnetoresistivity of the copper clad aluminum stabilizer. The stability of TOKI-MC was evaluated and compared with the data of short sample tests. >

The helical R&D coil TOKI-HB has been designed and fabricated to demonstrate the feasibility of constructing a pool-cooled superconducting helical coil extrapolatable to the Large Helical Device (LHD). In addition, the fabrication and tests of the coil will provide much needed engineering data on crucial materials and processes for the LHD. The major design parameters of the helical coil are: major radius 0.8 m, minor radius 0.2 m, helical pitch 3, coil number 1, maximum central field 0.75 T, maximum coil field 3 T, superconductor NbTi, operating current 8.93 kA and stored magnetic energy 1.9 MJ. This system consists of a helical coil wound on a hollow torus, liquid-nitrogen shields, a vacuum vessel and a control dewar. Major items of measurement plans are cool-down characteristics, cryogenic stability, stress due to the electromagnetic force, and accuracy of the magnetic field. This paper will summarize the design concept and the fabrication process of this system.

For development of superconducting conductors and coils, the National Institute for Fusion Science prepared a new superconducting test facility which is a combination of an electro-mechanical testing device and a large cryogenic cooling system at the new Cryogenics and Superconductivity Building at the Toki site. The cryogenic cooling system has a capacity of 250 liters of liquid helium per hour or 500 W at 4.5 K, and can also supply supercritical helium up to 50 g/s. A dc power supply generates 75 kA current at 21 V. A split-type superconducting coil generates a magnetic field of 9 T of 100 mm in width. The mechanical testing machine has a capacity of 10 MN at liquid-helium temperature. The following R & D experiments for the Large Helical Device have been done using the facility: critical current and stability measurement of helical coil conductors of which the nominal current is 21 kA at 7 T, an elasticity measurement of a helical coil, and stability measurement of a cable-in-conduit conductor.

The composite superconductor called Design-M for the TOKI-MC has been designed as part of a chain of research and development programs. The conductor must be conductive to the requirements of electrical and mechanical properties. The designed conductor can be divided into three main parts: a five-layered and compacted superconducting cable, four copper-clad 5-nines pure aluminum stabilizers and a half-hard copper sheath. These parts are soldered together. The surface of the copper sheath is coated with copper oxide to assist heat transfer toward liquid helium. The operating current is designed to be 20 kA in a magnetic field of 8 T at a temperature of 4.2 K. The conductor has been manufactured in total lengths of about 900 m, and approximately 2000 kg in weight. Technical improvements have been made and production facilities for assembly have been set up for mass-scale production. Electrical and mechanical properties have been investigated.

A calculation of the electromagnetic forces was carried out to design the supporting structure and the coil package. LHD has one pair of helical coils and three pairs of poloidal coils, all of which are superconducting coils. Because of the large electromagnetic force, the helical coils are fixed to a cylindrical supporting structure that is called the shell support. From the comparison between three types of methods to support the poloidal coils is found that the most suitable is that with the poloidal coils fixed firmly to the shell support in all directions. When considering the rigidity of the poloidal coils, the maximum stress of the shell support is smallest with this type. The forces on the electrical insulators of the helical coils are estimated by calculating the electromagnetic forces on each conductor and summing them. The maximum loads on the insulators between turns and between layers are 159 and 116 MPa, respectively. They occur in the center area of the coil cross section.

The Large Helical Device (LHD) is a heliotron/torsatron-type confinement device (B = 4 T, R = 3.9 m) equipped with a helical divertor. In the LHD configuration, the plasma region is shifted inwards approximately a third of the plasma minor radius relative to the center of the two pairing helical coils, thus making the distance between the coil center and the edge plasma small, 328 mm on the small major radius side of the torus. Within this space, many components must be installed, such as the superconducting helical coil (163), its coil can (45), the thermal shield (30), the vacuum gap (35), the plasma vacuum vessel (15), the first wall (25) (the numbers in the parentheses are allocated radial space in mm for the respective components). Effective edge plasma control by the divertor requires a space of more than 15 mm between the plasma and the first wall. To meet the above space requirement, the thickness of the helical coil is designed to be as small as possible, yet the coil current density and maximum field strength are within the limits of reliable coil operation.

A forced-flow NbTi superconducting helical coil was fabricated as research and development of the helical coil for the Large Helical Device (LHD), corresponding to a 14 scale model of the LHD. A computer-controlled helical-winding machine employing the roller bending method with twisting capability was also developed. The maximum assembly error of the conductors decreased to about 0.7 mm by optimal control of the winding curvature and torsion of the conductors. Three-dimensional structural analysis of the helical coil and the support structure was carried out to assess the mechanical integrity of the coil and support structure and to understand their mechanical behavior under the electromagnetic force. Analytical results show that the coil support structure can keep the maximum deformation of the conductors less than .4 mm, which corresponds to the required field accuracy of 5 × 10−4.

To investigate properties of superconducting R&D conductors and coils for the Large Helical Device (LHD), a dc 75 kA power supply system which consists of three 25 kA unit banks has been designed and constructed. One unit bank has two double-star-rectifier connections with the inter-phase reactors. The algorithm of the digital feedback control system is installed in each unit bank. In this algorithm, each 6 phase is operated independently on the process of averaging, proportioning and integrating. The experiments of several short samples and one of the R&D coils have been done under the maximum ramp rate of the power supply, 999 A/s per unit. It is confirmed that the measured current stability is less than ±0.75%, and the deviation between a set value and an output value of the load current is less than ±1% under the short circuit test.

The TOKI-MC (called “Module Coil” before), a superconducting solenoid, is one of the R&D coils for the Large Helical Device (LHD). The coil was made of one of the candidate conductors for the helical coils on its actual size. Being rated at 20 kA operating current and 11 MJ stored energy, the coil was designed to simulate almost all major parameters of the helical coils, i.e., conductor torsional rate, electromagnetic force of 10 MN/m, operating current, magnetic field of 7.7 T, current density in the winding of 40 A/mm2, and pool-boiling cooling method. The conductors had 19 mm square cross sections and are a composite of NbTiCu compacted cable, aluminum stabilizer, and copper mechanical sheath. The conductors were wound with a maximum torsional rate of 36°/m on a three-dimensionally shaped stainless-steel winding form. The coil forming was operated in twist-and-wind way piece by piece. Seven conductors were wound in layers with 0.4 m inner diameter, 0.7 m outer diameter, and 0.4 m height. The finished winding was canned into a stainless-steel coil case, having a mass of 5300 kg altogether.

Irregular magnetic fields resonate with the rational surface of the magnetic confinement systems, form magnetic islands and ergodic layers, and destruct the plasma confinement. To avoid this confinement destruction the requirement of an accuracy of 10 −4 in the magnetic field is adopted as the magnetic-accuracy design criterion for the LHD machine. Following this criterion the width of the undesirable magnetic island is kept less than one tenth of the plasma radius. The irregular magnetic field from the superconducting (SC) helical and poloidal coils is produced by winding irregularity, installing irregularity, cooling-down deformations and electromagnetic deformations. The local irregularities such as feeders, layer connections, adjacent-conductor connections of the coils also produce an error field. The eddy currents on the supporting shell structure of SC coils, the cryostat, etc. are also evaluated. All irregular effects are analyzed using Fourier decomposition and field mapping methods for the LHD design, and it is confirmed that the present design of the superconducting coil system satisfies the design criterion for these field irregularities.

Heat loads under various operating conditions of the Large Helical Device (LHD) are analyzed, the particular requirements for the refrigeration design are described first, then the concept “switching heat exchanger” of a new large helium-liquefier refrigerator is discussed as a candidate which can provide mechanically and thermodynamically stable operation of the cold box at the time-varying heat load of the LHD.

The Large Helical Device (LHD) is designed to have six pairs of superconducting coils. In the first phase of the project, these will be energized by six steady-power supplies. These power supplies are thyristor types, which generate positive and negative voltages and large currents (up to 30 kA). Each of them has a smoothing filter which reduces the voltage ripple to less than 0.2 V, and a protection circuit to transfer rectifier current to a dump resistor and drop it rapidly during coil quenches. These six steady power supplies are powered from the commercial line. In the second phase of the project, three more power supplies powered by a motor generator (120 MVA) or superconducting energy storage will be added to the poloidal coil circuits. These will be used in experiments which require rapid changes in the poloidal coil currents during a plasma discharge of 5 s. A protection circuit is now under development, composed of disconnectors, a vacuum circuit breaker, a power fuse and diodes. It has worked well up to 21 kA, and will be tested up to 30 kA.

A vacuum system for the LHD cryostat has been preliminary designed to guarantee reasonable operation of the system. The required working pressure for the cryostat had to be determined from the heat transfer. In order to complete the engineering design of the system, it is urgent to generate the outgassing data of the thermal insulation materials in the cryostat at low temperatures and establish a realistic operational scenario including conditioning and maintenance phases.

A forced-flow NbTi superconducting poloidal coil (TOKI-PF) using a cable-in-conduit conductor (CICC) was fabricated as an R&D project for the LHD coils. The main objective of TOKI-PF is to gather engineering data for the superconducting poloidal field coil. The major parameters are as follows: inner and outer radii 0.6 and 0.82 m, respectively, operating current 25.6 kA, stored energy 1.04 MJ, conductor NbTi CICC sized 10.0 × 22.5 mm. The mechanical coil configuration and nominal current of 25.6 kA simulate those of the LHD poloidal coils. Useful technical data were obtained and the feasibility of fabricating the LHD poloidal coils was confirmed. In this paper, preliminary test results will also be described.

We have been developing aluminum-stabilized superconductors for the helical coils of the Large Helical Device (LHD). Two types of conductors, Design-M and KISO-4B, were selected to proceed to the full-scale short sample tests. The test facility for full-scale conductors consists of 100 kA vapor-cooled current leads, 9 T background field split coils, a cryostat and a 75 kA dc power supply. Since both conductors have multi-layered strand cables without transposition, it is feared that an uneven current distribution between strands may cause degradation of the critical current. To measure the current distribution inside the conductor, a new method using a pick up coil was introduced. Using this experimental setup, we have measured the critical current, the current distribution inside the conductor and the stability of both full-scale conductors.

A centralized control system necessary for experimental fusion machines is discussed. A two-layered configuration where a number of complete local systems are controlled by a central computer, is appropriate as control system for the Large Helical Device. All the local systems can be uniformly operated with global cooperation by a few operators through general-purpose consoles as the central computer. Fast control, which cannot be supported by the uniform control, is aided by a precise but pre-programmable timing generator and an interlock for emergency response. Special safety requirements during a coil quench are studied under the condition that the enormous stored energy must be released in a short time after the quench.

The Large Helical Device (LHD) will be operated under various loads, such as strong electromagnetic forces, atmospheric pressure, and thermal stress. The outer vessel of the cryostat has a toroidal shape and is the largest component of the LHD. The designed maximum pressure of the vessel is 0.1 MPa inward and 0.01 MPa outward. Since the cross section of the outer vessel is not circular, the presence of unbalanced atmospheric pressure requires carefully designed studies of the rigidity of the cryostat. The poloidal cross section of the vessel is designed as a bell-shape, of which the bottom is a flat plate of 150 mm and the top has a major radius of 4 m, the minor radius is 2 m, with a thickness of 50 mm, by the finite element method using ANSYS. In this case, the maximum deformation appears at the center of the bottom flat plate; its value is 1.7 mm when stainless steel is used. The poloidal cross section of the outer vessel of the cryostat is desirable to have a symmetric circular shape which increases the mechanical stiffness and reduces the unbalanced magnetic field due to the eddy current during experiments. However, from an engineering point of view, the bottom is required to have the shape of a flat plate to have working space for inner components. We analyzed the stress field and deformation of the outer vessel under atmospheric pressure by using a finite element analysis code to pursue the optimized shape and thickness of the body.

It was discovered several decades ago that wall cleaning procedures, primarily to remove the light impurities C and O, were an important part of preparing fusion devices for plasma operation. Since that time, conditioning procedures making use of low‐temperature plasmas in different gasses have become more elaborate, and have also found application in other fields. However, there are still many questions remaining with regard to the cleaning efficiency, and the most appropriate method for reducing impurity concentrations in the shortest time. In this paper, a short review is given, outlining the current status of removing oxygen from metal surfaces by the application of low‐temperature plasmas.

Thermonuclear ignition and subsequent burn are key physics for achieving laser fusion. In fast ignition, a highly compressed fusion fuel generated with multiple ns‐laser beams is rapidly heated with a large energy, ps‐laser pulse in prior to core disassembly. This scheme has a high potential to achieve ignition and burn since driver energy required for high fusion gain is predicted to be about one tenth of that needed for the central ignition scheme. In Japan, Fast Ignition Realization Experiment (FIREX) project has been started to clarify the physics of energy transport and deposition in the core plasma and to demonstrate fuel temperature of above 5 keV. After the success, FIREX‐I will be followed by the second phase of the project (FIREX‐II) to demonstrate ignition and burn. LFEX laser, designed to deliver a laser pulse of 10 kJ in 10 ps, are operational and the first phase of FIREX experiments has been stated. A new target is proposed to attain dense compression of fuel and improve laser‐core coupling efficiency by adopting double‐cone structure, a low‐density inner liner, low‐Z outer coating, and Br‐doped fuel shell. In this paper, present status and near term prospects of the FIREX‐I project will be reported together with activities on target designing, laser development, and plasma diagnostics.

A self-consistent analytical solution for net erosion of a divertor plate which is set perpendicular to magnetic field lines is presented. The primary flux profile of hydrogen and impurities except redepositing particles is externally given as well as the return ratio of sputtered atoms to the plate. In the direction along the divertor trace, all conditions are uniform. The ionization mean free path is assumed constant to simplify equations. The analytical solution is compared with net erosion experiments carried out in compact helical system (CHS) by exposing a graphite target to a neutral beam heated plasma column introduced perpendicularly to the target along the magnetic field lines through a 2 cm slit opened on a graphite limiter. After exposure to 98 discharges, the target surface is analyzed with Rutherford backscattering method. Deposition profiles of Ti and O impurities are very well explained with the analytical predictions.

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. However, it does not seem to hamper the expected divertor functions. As a confinement improvement scheme in LHD, we have proposed a high temperature divertor plasma operation in which a divertor plasma with a temperature of a few keV, generated by efficient pumping, leads to the confinement improvement. Conceptual designs of the LHD divertor components are under way.

Using the island divertors (IDs) of W7-AS and W7-X and the helical divertor (HD) of LHD as examples, the paper presents a comparative divertor transport study for three typical helical devices of different machine sizes following two distinct divertor concepts, aiming at identifying common physics issues/effects for mutual validation and combined studies. Based on EMC3/EIRENE simulations supported by experimental results, the paper first reviews and compares the essential transport features of the W7-AS ID and the LHD HD in order to build a base and framework for a predictive study of W7-X. The fundamental role of low-order magnetic islands in both divertor concepts is emphasized. Preliminary EMC3/EIRENE simulation results for W7-X are presented and discussed with respect to W7-AS and LHD in order to show how the individual field and divertor topologies affect the divertor transport and performance. For instance, a high recycling regime, which is absent from W7-AS and LHD, is predicted to exist for W7-X. The paper focuses on identifying and understanding the role of divertors for high density plasma operations in helical devices. In this regard, special attention is paid to investigating the divertor function for controlling intrinsic impurities. Impurity transport behaviour and wall-sputtering processes of CX-neutrals are studied under different divertor plasma conditions. A divertor retention effect on intrinsic impurities at high SOL collisonalities is predicted for all the three devices. The required SOL plasma conditions and the underlying mechanisms are analysed in detail. Numerical results are discussed in conjunction with the experimental observations for high density divertor plasmas in W7-AS and LHD. Different SOL transport regimes are numerically identified for the standard divertor configuration of W7-X and the possible consequences on high density plasmas are assessed. All the EMC3-EIRENE simulations presented in this paper are based on vacuum fields and comparisons with local diagnostics are made for low-ß plasmas.

A pool-cooled NbTi superconducting helical coil system (named TOKI-HB) has been fabricated as part of one of the research and development programs for the Large Helical Device (LHD). The main purpose of the construction of this device is to build an entire pool-cooled superconducting helical coil system which is large enough to be extrapolable to LHD. The helical coil has a major radius of 0.8 m and a minor radius of 0.2 m with a helical pitch number of 3. An operating current of 8.9 kA produces a maximum magnetic field of 0.75 T at the geometrical center of the poloidal cross section and 3 T at the coil surface. Basic design concepts of the TOKI-HB device are described together with some issues of the fabrication.

Two forced-flow cooled NbTi superconducting coils (TOKI-TF, PF) have been designed and fabricated. The helical coil (TOKI-TF) is a 1/4-scale model of the Large Helical Device (LHD). It has a major radius of 0.9 m, a minor radius of 0.25 m, and a pitch number of 4. Nominal current and maximum field were designed to be 8 kA and 2.8 T, respectively. Another coil (TOKI-PF) was fabricated for the demonstration of LHD poloidal field coils. It consists of two double pancakes with an inner radius of 0.6 m and an outer radius of 0.82 m. The nominal current of 25 kA simulates that of LHD poloidal field coils. Cable-in-conduit-type conductors were used for the both coils. The test facility was also constructed with a vacuum vessel, a liquid nitrogen shield, 30-kA power leads, a heat exchanger, and cryogenic supports. Design concepts and details are presented.

The superconducting helical coils of the Large Helical Device (LHD) require superconducting conductors with large current capacities (from 20 kA to 30 kA) and high current densities (55 A/mm/sup 2/ at 8 T). An NbTi superconductor/bin with pool boiling is being used because of the large electromagnetic force and the complicated helical windings. Several conductors are designed to show how the difference of the position of pure aluminum in the conductors affects the stability and the mechanical properties. Scaled-down R&D conductors with operational currents from 7 kA to 10 kA were made on an experimental basis. The superconducting characteristics, stability, and mechanical properties of these scaled-down conductors were tested. The design and the test results concerning the superconducting characteristics are described.

A force-cooled superconducting coil has been studied for the Large Helical Device (LHD), which will be a main experimental apparatus of the Japanese National Institute for Fusion Science (NIFS). To achieve the main requirements for the LHD, a large-current, high-current-density NbTi cable-in-conduit-type conductor of 10 kA and 47 A/mm/sup 2/ at 8 T was developed. The test coil was wound in the form of a one-layer solenoid and tested at an operating current of 10 kA in a superposed field of about 7 T, at a hydraulic condition of 1.0 MPa, 4.5 K, and 10 g/s supercritical helium. The stability, normal propagation velocity, and pressure rise during a quench for the high-current-density case were also studied.

A twisted solenoid coil (TOKI-MC) has been designed and fabricated in order to study the mechanical properties of the Large Helical Device (LHD). One of the most important R&D items of the LHD is the mechanical behavior of the helical coils under a large electromagnetic force. The TOKI-MC was wound obliquely on the 3D-machined elliptical bobbin with a maximum torsional rate of 36deg/m at the innermost conductor. The maximum field in the coil is 7.7 T with an operating current of 20 kA, an average current density of 40 A/mm^2, and a stored energy of 11 MJ. The TOKI-MC can simulate the electromagnetic force, conductor torsional rate, magnetic field, operating current, and current density of the LHD superconducting helical coils. The design and test results of the conductor and the design and fabrication of the coil are described

The superconducting coil design for the Large Helical Device, an interuniversity magnetic confined fusion research device, is described. The major and minor radii of the helical coil are 3.9 m and 0.975 m, respectively. The helical coil is made of a pool-cooled conductor, and the field experienced by the conductor is 7.2 T at phase I and 9.6 T at phase II. Three pairs of poloidal coils are also superconducting. The coils will be made of NbTi conductor.

Helical coils of the Large Helical Device (LHD) are large scale superconducting magnets for heliotron plasma experiments. The helical coils had been cooled by saturated helium at 4.4 K, 120 kPa until 2005. An upgrade of the cooling system was carried out in 2006 in order to improve the cryogenic stability of the helical coils and then it has been possible to supply the coils with subcooled helium at 3.2 K, 120 kPa. A designed mass flow of the supplied subcooled helium is 50 g/s. The subcooled helium is generated at a heat exchanger in a saturated helium bath. A series of two centrifugal cold compressors with gas foil bearing is utilized to lower the helium pressure in the bath. The supplied helium temperature is regulated by rotational speed of the cold compressors and power of a heater in the bath. The mass flow of the supplied helium is also controlled manually by a supply valve and its surplus is evaporated by ten heaters at the outlet above the coils. In the present study, the performance of the cooli...

The pace of the ITER project in St Paul-lez-Durance, France is accelerating rapidly into its peak construction phase. With the completion of the B2 slab in August 2014, which will support about 400 000 metric tons of the tokamak complex structures and components, the construction is advancing on a daily basis. Magnet, vacuum vessel, cryostat, thermal shield, first wall and divertor structures are under construction or in prototype phase in the ITER member states of China, Europe, India, Japan, Korea, Russia, and the United States. Each of these member states has its own domestic agency (DA) to manage their procurements of components for ITER. Plant systems engineering is being transformed to fully integrate the tokamak and its auxiliary systems in preparation for the assembly and operations phase. CODAC, diagnostics, and the three main heating and current drive systems are also progressing, including the construction of the neutral beam test facility building in Padua, Italy. The conceptual design of the Chinese test blanket module system for ITER has been completed and those of the EU are well under way. Significant progress has been made addressing several outstanding physics issues including disruption load characterization, prediction, avoidance, and mitigation, first wall and divertor shaping, edge pedestal and SOL plasma stability, fuelling and plasma behaviour during confinement transients and W impurity transport. Further development of the ITER Research Plan has included a definition of the required plant configuration for 1st plasma and subsequent phases of ITER operation as well as the major plasma commissioning activities and the needs of the accompanying R&D program to ITER construction by the ITER parties.

The anomalous particle transport in a tokamak core is believed to be linked to the advection of magnetically trapped electrons alone, owing to the passing electrons maintaining a thermal equilibrium along the field lines. Surprisingly, in nonlinear numerical studies, the radial flux of passing electrons rivals that of the trapped ones. The strong interaction of passing electrons and electric fluctuations is mediated by long tails of the modes along the magnetic field, which are generated by the passing electrons in the first place.

A new slow transition process for energy transport in magnetically confined plasmas is reported. The slow transition is characterized by the change between two metastable transport conditions characterized by a weak and a strong electron temperature (${T}_{e}$) dependence of normalized heat flux. These two branches are found to merge at the critical ${T}_{e}$ gradient. In metastable transport, the derivative of normalized heat flux to the ${T}_{e}$ gradient, $\ensuremath{\partial}({Q}_{e}/{n}_{e})/\ensuremath{\partial}(\ensuremath{-}\ensuremath{\nabla}{T}_{e})$, is positive, while it becomes negative during the transition phase. The time for the transition increases as the normalized ${T}_{e}$ gradient is increased and exceeds the transport time scale characterized by the global energy confinement time.

The Large Helical Device (LHD) is the largest heliotron‐type super‐conducting fusion experimental device. One of the features of the heliotron configuration is its unique edge magnetic field topology. There exist an intrinsic stochastic layer just outside of the last closed flux surface (LCFS), residual islands embedded in the stochastic layer, whisker structures, laminar layers and intrinsic divertor structure (helical divertor). That contrasts to ‘onion‐skin’ like magnetic field structure in poloidal divertor tokamaks scrape‐off layer (SOL). The edge field line structure can be characterized by Kolmogorov length and field line connection length from wall to the other wall. In the stochastic layer, Kolmogorov length is much longer than connection length. Typical connection length of field lines in the stochastic layer and laminar layers are several kilometers and below several tens of meters, respectively, and these structures co‐exist. Therefore, the radial profile of field lines connection length becom...

Measurement systems for magnetic flux and magnetic field are being developed in the large helical device (LHD). Specification of the measurement systems is described. Careful attention should be paid to the set-up, especially thermal insulation to superconducting coils. Physical interpretation of magnetic flux as well as local magnetic field measured in finite-β plasmas has been conducted with three-dimensional magnetic field analysis code. This calculation indicates that non-linear changes in magnetic flux as a function of 〈β〉 should be considered for precise estimates.

The movable limiter, the head plate of which is formed with a boron carbide plate hot-pressed in vacuum, is used for studying plasma impurity removal and the possibility of boronization during discharge cleaning in Uragan-3M torsatron. Measurements of current-voltage characteristics, boron and oxygen spectroscopies, residual-gas mass-spectrometry were carried out for different biased and unbiased limiter positions. It is shown that at a negative limiter bias during pulsed discharges cleaning the boron spectral signal from the plasma increases due to arcing. At the same time the oxygen spectral line intensity decreases and it probably means desorbed oxygen and water effective trapping by boronized surfaces and the limiter B 4 C-plate, since the main plasma parameters do not change at this moment. The estimations of the limiter application for impurity pumping and solid target boronization shows that the negatively biased movable B 4 C-limiter can be used to improve pulsed discharge cleaning in Uragan-3M torsatron.

Diagnostics in infrared and far-infrared range for Heliotron E are described; (a) FIR interferometer for measuring electron density profile and (b) ECE for electron temperature profile as routine work, and (c) Fraunhofer diffraction method with a CO/sub 2/ laser for density fluctuation and (d) Thomson scattering with a D/sub z/O laser (/spl lambda/ = 385 /spl mu/m) for Ion temperature, as new methods.

Boronization at near room temperature has been performed in plasma processing teststand (PPT) by using a 5% diborane gases B 2 H 6 in He on electrically floating or cathode potential Al samples under various DC glow discharge conditions as a function of power or total gas pressure. The hydrogen concentration was analyzed by using elastic recoil detection method (ERD) and a new modified normalizing technique with Rutherford back scattering (RBS). Results showed that a high growth rate of film formation and floating surface were effective in reducing hydrogen concentration in B films. This result was in good agreement with earlier measurements of H with flash filament (FF) desorption method. In particular it was presumed from the experiments that the H B ratio was reduced by decreasing energetic ions but increasing radicals for B film formation.