Table of contents

High resolution spectral analysis of lithium plasma formed by single and double laser ablation has been undertaken to understand the plume-laser interaction, especially at the early stages of the plasma plume. In order to identify different atomic processes in evolving plasma, time resolved spectral emission studies at different inter-pulse delays have been performed for ionic and neutral lithium lines emitting from different levels. Along with the enhancement in emission intensity, a large line broadening and spectral shift, especially in the case of excited state transition Li I 610.3 nm have been observed in the presence of the second pulse. This broadening and shift gradually decrease with increasing time delay. Another interesting feature is the appearance of a multi-component structure in the ionic line at 548.4 nm and these components change conversely into a single structure at the later stages of the plasma. The multi-component structures are correlated with the presence of different velocity (temperature) distributions in non-LTE conditions. Atomic analyses by computing photon emissivity coefficients with an ADAS code have been used to identify the above processes.

Two pairs of high-frequency magnetic probes were installed in the Large Helical Device (LHD). During the injection of a perpendicular neutral beam, ion cyclotron emissions (ICEs) with the fundamental frequency corresponding to the ion cyclotron frequency at the plasma edge were detected, which are the same type of ICE as measured with the former spare ion cyclotron range of frequencies (ICRF) heating antennas. This type of ICE was further investigated with regard to the phase and intensity of signals. Another type of ICE was found in the LHD, and these ICEs were synchronized with bursts of toroidicity induced Alfvén eigenmodes (TAE) and the rise of intensity of lost ion flux. Therefore the source of these ICEs was thought to be the particles transferred from the core to the outer region of plasma by the TAE bursts. The frequency of ICEs induced by the TAE bursts increases linearly with the magnetic field strength, since the ion cyclotron frequency increases with the magnetic field strength.

A fast video camera is verified to be a powerful tool for observation of filaments/blobs near the last closed flux surface (LCFS). In order to extract the fluctuation component from the raw data of the fast camera, a pre-processing technique, sliding time window averaging subtraction (STWAS) has been developed to remove the background of slowly varying emission from the bulk plasma. By using this pre-processing technique, the fast camera data are analyzed. A method to identify the filaments in the pre-processed image is also discussed.

A 16-channel electron cyclotron emission (ECE) radiometer has been employed to observe the (m, n) = (2, 1) magnetic island structure on HT-7 tokamak, where m and n represent the poloidal and toroidal mode number respectively. The results indicate that the island width is about 7 cm when the magnetic island is saturated during the m/n = 2/1 mode. The location of resonance surface can be determined by plotting the contour of ECE relative fluctuation. This method could be applied to the HT-7 and EAST campaigns in the future for the research of neoclassical tearing modes (NTMs).

Perturbative experiments on electron heat transport have been successfully conducted on the HL-2A tokamak. The pulse propagation of the electron temperature is induced by the supersonic molecular beam injection (SMBI), which has characteristics of good localization and deep deposition. A model based on the electron heat transport in cylindrical geometry has been applied to reconstruct the measured amplitude and phase profiles of the electron temperature perturbation. The results show that the heat transport is significantly reduced near the pedestal region of the H-mode plasma. In the "profile stiffness/resilience" region, similar heat diffusivities have been observed in L-mode and H-mode plasmas, which verifies the gradient-driven transport physics in tokamaks.

An infrared camera (IR) has been put into operation in the Experimental Advanced Superconducting Tokamak (EAST), which is used to measure the temperature distribution on the surface of lower divertor target plates. With a finite difference method, the heat flux onto the divertor target plates is calculated from the surface temperature profile. The high confinement mode (H-mode) with type-III edge localized modes (ELMs) has been obtained with about 1 MW lower-hybrid wave power on the EAST in the autumn experiment in 2010. The analyzed H-mode discharges were lower single null X-point diverted discharges with a density range of < n_e > (1 ∼ 4) × 10¹⁹ m⁻³. The surface temperature of the inner target plate increases with heating power. The peak temperature on the surface of target plates is lower than 200°C with about 2.4 MW heating power. Comparison among the heat flux profiles occurring in different phases in the same discharge has been performed. It indicates that the heat flux profile obviously changes from the ohmic phase to the H-mode phase, and the full width at half maximum (FWHM) of the heat flux profile is the narrowest during the ELM-free H-phase. On the outer target plate, the peak heat flux exceeds 2 MW/m² during the ELMy H-mode phase, whereas it is only about 0.8 MW/m² during the ELM-free phase in the same discharge.

Impurity accumulation is studied for neutral beam-heated discharges after hydrogen multi-pellet injection in Large Helical Device (LHD). Iron density profiles are derived from radial profiles of EUV line emissions of FeXV-XXIV with the help of the collisional-radiative model. A peaked density profile of Fe²³⁺ is simulated by using one-dimensional impurity transport code. The result indicates a large inward velocity of −6 m/s at the impurity accumulation phase. However, the discharge is not entirely affected by the impurity accumulation, since the concentration of iron impurity, estimated to be 3.3 × 10⁻⁵ to the electron density, is considerably small. On the other hand, a flat profile is observed for the carbon density of C⁶⁺, which is derived from the Z_eff profile, indicating a small inward velocity of −1 m/s. These results suggest atomic number dependence in the impurity accumulation of LHD, which is similar to the tokamak result.

Prediction of volt-second consumption has been done by the tokamak simulation code (TSC), which includes the whole plasma discharge of HL-2M conception design. It covers the volt-second consumptions at the entire current ramp-up phase and the plasma flattop phase. More important, the sensitivities of volt-second consumption with respect to the current ramp-up time and the impurity concentration have been studied, respectively.

A new high repetition rate Nd:YAG Thomson scattering system is developed for the Heliotron J helical device. A main purpose of installing the new system is the temporal evolution measurement of a plasma profile for improved confinement physics such as the edge transport barrier (H-mode) or the internal transport barrier of the helical plasma. The system has 25 spatial points with ∼10 mm resolution. Two high repetition Nd:YAG lasers (> 550 mJ@50 Hz) realize the measurement of the time evolution of the plasma profile with ∼10 ms time intervals. Scattered light is collected by a large concave mirror (D = 800 mm, f/2.25) with a solid angle of ∼100 mstr and transferred to interference filter polychromators by optical fiber bundles in a staircase form. The signal is amplified by newly designed fast preamplifiers with DC and AC output, which reduces the low frequency background noise. The signals are digitized with a multi-event QDC, fast gated integrators. The data acquisition is performed by a VME-based system operated by the CINOS.

A new spherical torus called VEST (Versatile Experiment Spherical Torus) is designed, constructed and successfully commissioned at Seoul National University. A unique design feature of the VEST is two partial solenoid coils installed at both vertical ends of a center stack, which can provide sufficient magnetic fluxes to initiate tokamak plasmas while keeping a low aspect ratio configuration in the central region. According to initial double null merging start-up scenario using the partial solenoid coils, appropriate power supplies for driving a toroidal field coil, outer poloidal field coils, and the partial solenoid coils are fabricated and successfully commissioned. For reliable start-up, a pre-ionization system with two cost-effective homemade magnetron power supplies is also prepared. In addition, magnetic and spectroscopic diagnostics with appropriate data acquisition and control systems are well prepared for initial operation of the device. The VEST is ready for tokamak plasma operation by completing and commissioning most of the designed components.

In an effort to simulate the dynamic behavior of a non-ferromagnetic conducting structure with consideration of the magnetic damping effect, a finite element code is developed, which is based on the reduced vector potential (A_r) method, the step-by-step integration algorithm and a time-partitioned strategy. An additional term is introduced to the conventional governing equations of eddy current problems to take into account the velocity-induced electric field corresponding to the magnetic damping effect. The TEAM-16 benchmark problem is simulated using the proposed method in conjunction with the commercial code ANSYS. The simulation results indicate that the proposed method has better simulation accuracy, especially in the presence of a high-intensity external magnetic field.

This paper describes a conceptual design study for the circuit configuration of the Error Field Correction Coil (EFCC) power supply (PS) to maximize the expected performance with reasonable cost in JT-60SA. The EFCC consists of eighteen sector coils installed inside the vacuum vessel, six in the toroidal direction and three in the poloidal direction, each one rated for 30 kA-turn. As a result, star point connection is proposed for each group of six EFCC coils installed cyclically in the toroidal direction for decoupling with poloidal field coils. In addition, a six phase inverter which is capable of controlling each phase current was chosen as PS topology to ensure higher flexibility of operation with reasonable cost.

An Ion Cyclotron Range of Frequency (ICRF) system with a radio frequency (RF) power of 4 × 1.5 MW was developed for the Experimental Advanced Superconducting Tokamak (EAST). High RF power transmitters were designed as a part of the research and development (R&D) for an ICRF system with long pulse operation at megawatt levels in a frequency range of 25 MHz to 70 MHz. Studies presented in this paper cover the following parts of the high power transmitter: the three staged high power amplifier, which is composed of a 5 kW wideband solid state amplifier, a 100 kW tetrode drive stage amplifier and a 1.5 MW tetrode final stage amplifier, and the DC high voltage power supply (HVPS). Based on engineering design and static examinations, the RF transmitters were tested using a matched dummy load where an RF output power of 1.5 MW was achieved. The transmitters provide 6 MW RF power in primary phase and will reach a level up to 12 MW after a later upgrade. The transmitters performed successfully in stable operations in EAST and HT-7 devices. Up to 1.8 MW of RF power was injected into plasmas in EAST ICRF heating experiments during the 2010 autumn campaign and plasma performance was greatly improved.

Electric potential near a wall for plasma with the surface produced negative ions with magnetic field increasing toward a wall is investigated analytically. The potential profile is derived analytically by using a plasma-sheath equation, where negative ions produced on the plasma grid (PG) surface are considered in addition to positive ions and electrons. The potential profile depends on the amount and the temperature of the surface produced negative ions and the profile of the magnetic field. The negative potential peak is formed in the sheath region near the PG surface for the case of strong surface production of negative ions or low temperature negative ions. As the increase rate of the magnetic field near the wall becomes large, the negative potential peak becomes small.

For a robust design of vacuum vessel of HL-2M, the electromagnetic (EM) loads have to be understood clearly. In this paper, some crucial transient events, such as plasma major disruptions (MDs), vertical displacement events (VDEs), fast discharge of toroidal field (TF) coils, have been investigated to evaluate the eddy currents and EM forces on vacuum vessel and in-vessel components. The results show that the eddy currents depend strongly on the current decay time, and the maximum toroidal eddy current flowing in the whole vessel can reach up to 2.4 MA during MDs that is close to the plasma current. Large symmetric radial forces and a net vertical force on vessel shells could be caused by these transient events. Combination of eddy currents in in-vessel components and toroidal field could twist the copper plates and other internal parts, however, if these plates are supported and connected carefully, the twist moments will not have a big effect on the vessel shells and vessel support.

For safe operation with active water cooling plasma facing components (PFCs) to handle a large input power over a long pulse discharge, some design optimization, R&D and maintenance were accomplished to improve the in-vessel components. For the purpose of large plasma current (1 MA) operation, the previous separated top and bottom passive stabilizers in the low field were electrical connected to stabilize plasma in the case of vertical displace events (VDEs). The design and experiments are described in this paper

Tungsten was exposed to pure Ar or Ne plasmas over 1550 K at several incident ion energies. Even under the irradiation condition that the tungsten nanostructure is formed by He plasma irradiation, holes/bubbles and fiberform nanostructures were not formed on the surface by exposure to Ar or Ne plasmas. In addition, the results from energy dispersive X-ray spectroscopy supported the facts that Ar and Ne did not remain in the sample. We will discuss the reason for the differences in the damage to the tungsten surface exposed to noble gas plasmas.

A conceptual design review of the ITER gas injection system (GIS) function, safety, operation, and maintenance has recently been successfully completed. The GIS design can now continue to the preliminary design stage. This paper gives an overall description of the requirements and implementation at the concept design level. The designs of the sub-systems according to its breakdown structure are discussed against the corresponding requirements.

Tritium breeder and neutron multiplier as functional materials play an important role not only in ITER test blanket module (TBM) but also in fusion reactor. The paper describes the status of the fabrication of the two materials in Southwestern Institute of Physics (SWIP). Li₄SiO₄ pebbles were fabricated by melt-spraying method. Most of the pebbles with the diameter of 1.0 mm are well spherically shaped. The properties of the pebbles have been investigated. The results show that the pebbles produced by this method have a high density of 93% TD (theoretical density). It was also found that the open/closed porosity will be decreased after thermal treatment, but the average crush load will be increased to 7 N. The rotating electrode process (REP) has been adopted to produce beryllium pebble for impurity control and mass production. The pebbles with the diameter of 1.0 mm were produced by REP. The beryllium pebbles produced by REP look almost perfectly spherical with a very smooth surface and a high density of 98% TD. The test results indicate that REP method has excellent prospects for the fabrication of beryllium pebbles and the attractiveness of their properties.

QUEST has a divertor configuration with a high and a negative n-index, and the problem of plasma vertical position instability control in QUEST is still under extensive study for achieving high efficiency plasma. The instability we considered is that the toroidal plasma moves either up or down in the vacuum chamber until it meets the vessel wall and is extinguished. The actively controlled coils (HCU and HCL) outside the vacuum vessel are serially connected in feedback with a measurement of the plasma vertical position to provide stabilizing control. In this work, a robust controller is employed by using the loop synthesis method, and provides robust stability over a wide range of n-index. Moreover, the gain of the robust controller is lower than that of a typical proportional derivative (PD) controller in the operational frequency range; it indicates that the robust controller needs less power consumption than the PD controller does.

A new 300 MVA/1350 MJ motor generator (MG) will be built to feed all of the poloidal field power supplies (PFPS) and auxiliary heating power supplies of the HL-2M tokamak. The MG has a vertical-shaft salient pole 6-phase synchronous generator and a coaxial 8500 kW induction motor. The Ohmic heating power supply (OHPS) consisting of 4-quadrant DC pulsed convertor is the one with the highest parameters among the PFPS. Therefore, the match between the generator and the OHPS is very important. The matching study with Matlab/Simulink is described in this paper. The simulation results show that the subtransient reactance of the generator is closely related to the inversion operation of the OHPS. By setting various subtransient reactance in the simulation generator model and considering the cost reduction, the optimized parameters are obtained as x''_d = 0.405 p.u. at 100 Hz for the generator. The models built in the simulation can be used as an important tool for studying the dynamic characteristics and the control strategy of other HL-2M PFPSes.

In this paper, surface modification of the strut dowel used in ITER PF support is reported. Different ions (nitrogen/titanium) with different doses are implanted into the surface of strut dowel. The result of Auger Electron Spectroscopy (AES) indicates that nitrogen can be implanted more deeply than titanium under the implantation condition of 60 kV accelerating voltage and a dose of 8 × 10¹⁷/cm² nitrogen. Surface Micro Hardness (SMH) and wear resistance are improved remarkably. Further SEM observation shows that there are no obvious scratches and damages after wear test.

Neutral beam injection (NBI) is recognized as one of the most effective means for plasma heating. A 100 s long pulse neutral beam with 30 keV beam energy, 10 A beam current and a 100 s long pulse modulating neutral beam with 50 keV beam energy, 16 A beam current were achieved in the EAST neutral beam injector on the test-stand. The preliminary results suggest that EAST-NBI system initially possess the ability of long pulse beam extraction.