Table of contents

A diamond-shaped shock wave was created in a helium arcjet plasma. Visible/ultraviolet emission spectroscopy was used to investigate the condition for the formation of stable shocks and to determine characteristics of the plasma. Dependence of the position of the shock front on the gas pressure in the expansion region was investigated. It was found that the shock wave arises from the collision of plasma particles and residual neutral atoms in that region. Continuum and line spectra of neutral helium were measured, from which the electron temperatures were derived. The electron density was deduced from the Inglis-Teller limit of the He I 2p³P-3d³D series. The temperature and density were found to have almost constant values of 0.2 eV and 8.5×10¹³ cm⁻³, respectively, across the shock front.

To realize an excitation of electron Bernstein waves (EBW) via mode conversion from X-mode waves injected from the high magnetic field side (HFS), new inner-vessel mirrors were installed close to a helical coil in the large helical device (LHD). 77 GHz electron cyclotron (EC) wave beams injected from an existing EC-wave injection system toward the new mirror are reflected on the mirror so that the beams are injected to plasmas from HFS. Evident increases in the electron temperature at the plasma core region and the plasma stored energy were observed by the HFS beam injection to the plasmas with the line-average electron density of 7.5×10¹⁹ m⁻³, which is slightly higher than the plasma cut-off density of 77 GHz EC-waves, 7.35×10¹⁹ m⁻³. The heating efficiency evaluated from the changes in the time derivative of the plasma stored energy reached ∼70%. Although so far it is not clear which is the main cause of the heating effect, the mode-converted EBW or the X-mode wave itself injected from the HFS, an effective heating of high-density plasma over the plasma cut-off of EC-wave was successfully demonstrated.

An attempt was made to improve the spatio-temporal resolution of the tangential X-ray crystal spectrometer (XCS) on the Experimental Advanced Superconducting Tokamak (EAST) by evaluating experimentally the applicability of a novel X-ray photon detection technology for measuring the satellite spectra of Ar XVII with a high counting rate. High-resolution experimental data on the profiles of ion temperature and plasma rotation velocity facilitate the studies of the mechanisms underlining important physical phenomena, such as plasma heating, L-H transition and momentum transport. Based on silicon diode array and single-photon counting technology, a relatively small area (83.8×33.5 mm²) two-dimensional detector was successfully installed and tested in the recent EAST campaign. X-ray photon counting rate higher than 20 MHz was observed for the first time, and high quality satellite spectra were recorded for ion temperature and plasma rotation measurement, indicating that the new technology is suitable for the next-step high-resolution XCS on EAST, and the deployment of a detector array with a much larger X-ray sensing area is planned for better plasma coverage.

On HL-2A, the ion temperature profile can be roughly estimated by analyzing the energy spectrum of the neutral particles that escape from core plasma through 11 different spatial channels. However, this profile deviates from the real radial distribution because of the averaging and attenuation effects. The DEGAS code is used to estimate these effects and reconstruct a more reasonable ion temperature profile by simulating these neutral spectra and making them conform to the measurements. In this paper, critical parameters that affect the escaped neutral spectra are enumerated, and an example of a reconstruction of the ion temperature profile is presented based on experimental data.

A space-resolved EUV spectrometer for measuring the one-dimensional distribution of impurity line emissions in Large Helical Device (LHD) has been upgraded to measure two-dimensional distributions of impurity line emissions with an extension of working wavelength range to 30∼650Å. The two-dimensional measurement is performed by scanning the observation chord horizontally. A rectangular plasma region of 520 × 700 mm² in vertical and horizontal sizes can be observed during a single horizontal scan. The horizontal scan requires a time duration of 5 s at least. The spatial resolution is 10 mm in the vertical direction when a spatial-resolution slit of 0.2 mm in width is adopted. Although a spatial resolution in the toroidal direction is 75 mm, it is a function of CCD exposure time and horizontal scanning speed. Two-dimensional distribution of EUV line emissions from several impurities has been successfully observed for the first time from steady discharges in LHD. In this paper two-dimensional distributions of He II (303.78 Å), C V (40.27 Å), C VI (33.73 Å) and Fe XX (132.85 Å) located at different radial positions are presented with simple analysis on the magnetic field structure of LHD.

A normal incidence vacuum ultraviolet (VUV) and a grazing incidence extreme ultraviolet (EUV) spectrometers have been developed for the edge and core impurity measurements in HL-2A tokamak. The VUV and the EUV spectrometers cover wavelength ranges of 300∼3200 Å and 50∼500 Å, respectively. A spatial resolution of 2 mm has been achieved for the VUV spectrometer when a space-resolved slit 50 μm in width is used. The space-resolved slit is placed between the entrance slit and the grating of the spectrometer. The spectral resolutions of 0.15 Å for the VUV spectrometer in the wavelength coverage of the concave 1200 grooves/mm grating and of 0.22 Å for the EUV spectrometer at λ=200 Å with a flat-field laminar-type holographic grating are obtained. The sensitivity of the VUV spectrometer was calibrated in situ with the plasma bremsstrahlung radiation. The experimental results from both spectrometers are presented, especially the line intensity radial profiles measured by the VUV spectrometer.

The efficiency of energetic ion confinement is reduced in a tokamak plasma by the non-axisymmetric field, namely the ripple field. The ripple field is produced by a finite number of toroidal field coils. It is affected by the non-axisymmetric finite beta effect. The three-dimensional MHD equilibrium calculation code VMEC is used to analyze the non-axisymmetric finite beta effect in a ripple tokamak. In the VMEC code, the flux coordinates are used, so the calculation region is limited to the area of plasma. To calculate the orbit outside the plasma, we develop a field calculation code, which is based on the Biot-Savart law. The details of the method and results are described in this paper.

We have observed reduction of the runaway electron synchrotron radiation, hard X-ray (HXR) intensity, and HXR energy after applying 110 GHz 2nd harmonic electron cyclotron resonant heating (ECRH) during runaway electron (RE) discharges at low density with startup runaway electrons. However, we did not see a significant reduction of X-rays from 170 GHz 2nd harmonic ECRH at a higher field. A recently installed IR TV camera was used to observe the forward cone of synchrotron radiation from high energy REs in KSTAR. We have observed changes to the synchrotron images and reduction of the HXR by application of resonant magnetic perturbations (RMP) from in-vessel control coils (IVCC) installed inside KSTAR in the n = 1 configuration.

Lithiumization of the vacuum vessel wall of the Aditya tokamak using a lithium rod exposed to glow discharge cleaning plasma has been done to understand its effect on plasma performance. After the Li-coating, an increment of ∼100 eV in plasma electron temperature has been observed in most of the discharges compared to discharges without Li coating, and the shot reproducibility is considerably improved. Detailed studies of impurity behaviour and hydrogen recycling are made in the Li coated discharges by observing spectral lines of hydrogen, carbon, and oxygen in the visible region using optical fiber, an interference filter, and PMT based systems. A large reduction in O I signal (up to ∼40% to 50%) and a 20% to 30% decrease of H_α signal indicate significant reduction of wall recycling. Furthermore, VUV emissions from O V and Fe XV monitored by a grazing incidence monochromator also show the reduction. Lower Fe XV emission indicates the declined impurity penetration to the core plasma in the Li coated discharges. Significant increase of the particle and energy confinement times and the reduction of Z_eff of the plasma certainly indicate the improved plasma parameters in the Aditya tokamak after lithium wall conditioning.

Microhollow cathode discharge (MHCD) plasmas were generated in gas mixtures containing water vapor at pressures of up to 100 kPa of He or 20 kPa of air. The cathode diameter was 1.0 mm with a length of 2.0 mm. The electrical characteristics showed an abnormal glow mode. Spectroscopic measurements were carried out to examine the plasma and radicals. An analysis of the spectral profile of Hα at 656.3 nm enabled a derivation of the electron densities, namely 2×10¹⁴ cm⁻³ (at 10 kPa) and 6×10¹⁴ cm⁻³ (at 4 kPa) for the helium and air atmospheres, respectively, in the negative glow region. By comparing the observed OH radical spectra with those calculated by the simulation code LIFBASE, the gas temperature was deduced to be 900 K for 4 kPa of He at a discharge current of 50 mA.

Reduced activation ferritic/martensitic steel (RAFM) is recognized as the primary candidate structural material for ITER's test blanket module (TBM). To provide a material and property database for the design and fabrication of the Chinese helium cooled ceramic breeding TBM (CN HCCB TBM), a type of RAFM steel named CLF-1 was developed and characterized at the Southwestern Institute of Physics (SWIP), China. In this paper, the R&D status of CLF-1 steel and the technical issues in using CLF-1 steel to manufacture CN HCCB TBM were reviewed, including the steel manufacture and different welding technologies. Several kinds of property data have been obtained for its application to the design of the ITER TBM.

Refurbishment of steady state tokamak (SST-1) primarily focused at addressing the issues and bottle-necks involving various subsystems of SST-1 as observed during earlier commissioning attempts, have progressed significantly. Under the refurbishment spectrum, all joints in the superconducting magnet system have been re-fabricated as low DC leak tight joint resistances, all toroidal field (TF) magnets have been equipped with 5 K radiation shields on the inner side and successfully tested for their rated parameters in cold under nominal currents, all vessel sectors and modules have been baked and tested under representative conditions, supporting helium and nitrogen cryogenic facilities have been made > 99% reliable in various envisaged operating scenarios of SST-1. The reassemblies of the critical subsystems of the SST-1 machine shell have progressed aggressively and are nearing completion. Auxiliaries such as the baking facility for the vacuum vessel and first wall components, current leads assembly distributions, synchronized timing system, reliable data acquisition and plasma control systems as well as essential diagnostics have also been readied towards the first plasma. A detailed engineering validation of the assembled SST-1 machine shell including field error measurements has been planned prior to first plasma attempts.

The ITER neutron shielding blocks are located between the outer shell and the inner shell of the vacuum vessel to provide neutron shielding. Considering the combined loads acting on the shielding blocks during ITER plasma operation, the structure of the shielding blocks must be evaluated. Using the finite element method with ANSYS analysis software, static structural analysis is performed, including elastic analysis and limit analysis for one typical shielding block. The evaluated results based on RCC-MR code show that the structure of this shielding block can meet the design requirement.

A superconducting coil system is actually complicated by the distributed parameters, e.g. the distributed mutual inductance among turns and the distributed capacitance between adjacent conductors. In this paper, such a complicated system was modeled with a reasonably simplified circuit network with lumped parameters. Then, a detailed circuit analysis was conducted to evaluate the possible voltage transient in the coil circuit. As a result, an appropriate (minimum) snubber capacitance for the Switching Network Unit, which is a fast high voltage generation circuit in JT-60SA, was obtained.

An experimental Paschen test setup has been established to analyze the quality of ITER current lead (CL) insulation and extend the research on Paschen's law under various conditions. Insulation problems can destroy a machine if a Paschen discharge is triggered by an insulation defect that is caused by faulty manufacturing, electromagnetic force, and thermal stress load with a certain degree of vacuum helium or pipe leakage. The results show that the CL insulation mock-up worked well under normal temperature and pressure. Besides, the mock-up also worked well in helium conditions and at 80 K temperature at different pressures. One area of CL insulation was severely destroyed when the 80 K test was conducted after 5 thermal cycles, resulting in Paschen discharge phenomenon. The breakdown voltage is maintained at a relatively low level under different pressure conditions; the change of breakdown voltage was mainly due to the change of pressure, and such change was in line with the Paschen law.

Steady State Tokamak (SST-1) vacuum vessel baking as well as baking of the first wall components of SST-1 are essential to plasma physics experiments. Under a refurbishment spectrum of SST-1, the nitrogen gas heating and supply system has been fully refurbished. The SST-1 vacuum vessel consists of ultra-high vacuum (UHV) compatible eight modules and eight sectors. Rectangular baking channels are embedded on each of them. Similarly, the SST-1 plasma facing components (PFC) are comprised of modular graphite diverters and movable graphite based limiters. The nitrogen gas heating and supply system would bake the plasma facing components at 350°C and the SST-1 vacuum vessel at 150°C over an extended duration so as to remove water vapour and other absorbed gases. An efficient PLC based baking facility has been developed and implemented for monitoring and control purposes. This paper presents functional and operational aspects of a SST-1 nitrogen gas heating and supply system. Some of the experimental results obtained during the baking of SST-1 vacuum modules and sectors are also presented here.

One of the serious concerns for tungsten materials in fusion devices is the radiation defects caused by helium plasma irradiation since helium is a fusion product. The fiber-formed nanostructure is thought to have a possible weakness against the plasma heat flux on the plasma-facing component and also may destroy the reflectivity of optical mirrors. In this paper an interesting method for the recovery of such tungsten surfaces is shown. The recovery process depends on the grade and manufacturing process of tungsten materials.

The 14 MeV neutrons produced in the D-T fusion reactions have the potential of breeding Uranium-233 fissile fuel from fertile material Thorium-232. In order to estimate the amount of U-233 produced, experiments are carried out by irradiating thorium dioxide pellets with neutrons produced from a 14 MeV neutron generator. The objective of the present work is to measure the reaction rates of ²³²Th + ¹n → ²³³Th → ²³³Pa → ²³³U in different pellet thicknesses to study the self-shielding effects and adopt a procedure for correction. An appropriate assembly consisting of high-density polyethylene is designed and fabricated to slow down the high-energy neutrons, in which Thorium pellets are irradiated. The amount of fissile fuel (²³³U) produced is estimated by measuring the 312 keV gammas emitted by Protactinium-233 (half-life of 27 days). A calibrated High Purity Germanium (HPGe) detector is used to measure the gamma ray spectrum. The amount of ²³³U produced by Th²³² (n, γ) is calculated using MCNP code. The self-shielding effect is evaluated by calculating the reaction rates for different foil thickness. MCNP calculation results are compared with the experimental values and appropriate correction factors are estimated for self-shielding of neutrons and absorption of gamma rays.

When one wants to simply estimate tritium breeding ratio (TBR), the TBR may be reduced from a "local" TBR for the breeding zones of a blanket module by multiplying the breeder coverage (= the surface area of effective breeding region / the surface area of the first wall around plasma). When blanket modules are arranged, the gap between neighboring modules and the frames of the modules are regarded as non-breeding zones. On the other hand, neutrons scattered in the non-breeding zones can enter breeding zones, contributing to tritium production. This means that the estimation method mentioned above tends to underestimate TBR. In order to assess the scattering effect quantitatively, we carried out a three-dimensional Monte Carlo N-particle transport MCNP-5 calculation. It was found from the calculation that there is little decrease in TBR for gaps less than 4 cm when the blanket thickness is 70 cm. The result indicates that such a wide allowance of the gap will facilitate access of remote handling equipment for the replacement of blanket modules and improve access of diagnostics.

A novel method for calculating the magnetic stiffness matrix was proposed for the numerical analysis of the magneto-elastic stability of complicated current-carrying structures aiming for application in the magneto-elastic behavior of the tokamak system. A code based on the proposed method was developed and applied to the numerical analysis of two typical current-carrying structures. The good consistency of the numerical and analytical results validated the proposed method and the related numerical code.

Vacuum insulation of −1 MV is a common issue for the HV bushing and the accelerator for the ITER neutral beam injector (NBI). The HV bushing as an insulating feedthrough has a five-stage structure and each stage consists of double-layered insulators. To sustain −1 MV in vacuum, reduction of electric field at several triple points existing around the double-layered insulators is a critical issue. To reduce electric field simultaneously at these points, three types of stress ring have been developed. In a voltage holding test of a full-scale mockup equipped with these stress rings, 120% of rated voltage was sustained and the voltage holding capability required in ITER was verified. In the MeV accelerator, whose target is the acceleration of a H⁻ ion beam of 1 MeV, 200 A/m², the gap between the grid support was extended to suppress breakdowns triggered by electric field concentration at the edge and corner of the grid support. This modification improved the voltage holding capability in vacuum, and the MeV accelerator succeeded in sustaining −1 MV stably. Furthermore, it appeared that the H⁻ ions beam was deflected and a part of the beam was intercepted at the acceleration grid. This causes high heat load on the grids and breakdowns during beam acceleration. To suppress the direct interception, a new grid was designed with proper aperture displacement based on a three dimensional beam trajectory analysis. As a result, 980 keV, 185 A/m² H⁻ ion beam acceleration has been demonstrated, which is close to the ITER requirement.

This paper describes an asymmetric control method for the firing angle and a start/stop timing shift control of four thyristor converters called "Booster PS" to minimize the reactive power fluctuation during plasma initiation in JT-60SA. From the simulation using the "PSCAD/EMTDC" code, it is found that these control methods can drastically reduce the reactive power induced by the four units of the "Booster PS". In addition, the voltage fluctuation of the motor-generator connected to the "Booster PS" is expected to be suppressed. This can also contribute to achieve stable control of the JT-60SA magnet power supplies.

JT-60SA is a fully superconducting coil tokamak upgraded from the JT-60U. This paper focused on the integrity of the top lid of cryostat in JT-60SA. The design requirement for the cryostat in normal operations is to achieve vacuum insulation of 10⁻³ Pa, and the top flange of the top lid is lightly welded onto its body flange. The weld is tensile-loaded by bending deformation of the top lid due to vacuum pressure of external 0.1 MPa. This weld integrity is evaluated with tensile-load reduction, which results in clamp reinforcement. The structural integrity of the top lid is validated.

ITER magnet gravity support (GS) has been redesigned as a structure of pre-assembled multi-flexible plates instead of the original welded structure. In the past several years, engineering tests of the new structure have been proposed. A prototype engineering test platform is being developed. In order to apply the loads/load combinations onto the test mock-up, seven hydraulic bolt tensioners in three directions have been applied to simulate various loads (forces and moments), through which the deformation of bolts, flexible plates and clamp blocks, the stress distribution in the flexible plates, the friction between the contact surface, etc. can be monitored/tested. The measurement and control system includes seven sets of synchronization controller, a 16-channel strain gauge, 25 sets of displacement sensors, etc. Principles of EDC220 digital controller and development of multi-channel control software are also demonstrated.

This paper presents the evolution of the design of cold mass support for the ITER magnet feeder system. The glass fibers in the cylinder and the flanges of the normal G10 support are discontinuous in the preliminary design. The heat load of this support from the analysis is only 4.86 W. However, the mechanical test of the prototype showed that it can only endure 9 kN lateral force, which is significantly less than the required 20 kN. So, the configuration of the glass fibers in the cylinders and flanges of this G10 support are modified by changing it to a continuous and knitted type to reinforce the support, and then a new improved prototype is manufactured and tested. It could endure 15 kN lateral forces this time, but still not meet the required 20 kN. Finally, the SS316LN material is chosen for the cold mass supports. The analysis results show that it is safe under 20 kN lateral forces with the heat load increased to 14.8 W. Considering the practical application, the requirements of strength is of primary importance. So, this SS316LN cold mass support is acceptable for the ITER magnet feeder system. On the other hand, the design idea of using continuous and knitted glass fibers to reinforce the strength of a G10 support is a good reference for the case with a lower heat load and not too high Lorentz force.