Design of the power system for dynamic resonant magnetic perturbation coils on the J-TEXT tokamak

A set of in-vessel saddle coils has been installed on J-TEXT tokamak. They are proposed for further researches on controlling tearing modes and driving plasma rotation by static and dynamic resonant magnetic perturbations (RMPs). The saddle coils will be energized by DC with the amplitude up to 10 kA, or AC with maximum amplitude up to 5 kA within the frequency range of 1–5 kHz. At DC mode two antiparallel 6-pulse phase thyristor rectifiers are chosen to obtain bidirectional current, while at AC mode an AC–DC–AC converter including a series resonant inverter can generate current of various amplitudes and frequencies. The paper presents the design of the power supply system, based on the definition of the power supply requirements and the feasibility of implementation of the topology and control strategy. Some simulation and experimental results are given in the end.     

The upgrade of KSTAR timing system to support long-pulse operation and high-speed data acquisition

Since the first campaign of KSTAR in 2008, the home-made timing system had run for the synchronized operation of tokamak. The timing board which featured PMC-form factor, giga-bit optical communication, home-made protocol, multi-triggering capability, using GPS time and being integrated to EPICS (Experimental Physics and Industrial Control System), had advantages of compactness, modularity, platform independency and full functionality for the synchronized tokamak operation. However, there was deficiency in timing accuracy resulting from the engagement of software in realization of timing function and timing jitter due to poor isolation in output ports. Moreover, new requirements were on the rise as the plasma pulse length was getting longer and diagnostics operating at the higher frequency were newly installed.In order to meet new requirements and overcome the problems, the new timing board has been developed. As a result, the performance is remarkably enhanced: timing accuracy less than 5 ns, jitter less than 100 ps, 8 configurable multi-triggering sections, provision of maximum 100 MHz sampling clock. The KSTAR timing system upgraded by introducing the new timing board is participating in the 2011 campaign after calibration and consolidating the established timing system.This paper describes design, development and commissioning results of the new KSTAR timing system.     

Introduction to resonant magnetic perturbation coils of the J-TEXT Tokamak

To investigate the interactions between both the static and rotating resonant magnetic perturbations (RMP) and the tokamak plasma, two sets of coils, namely static RMP (SRMP) and dynamic RMP (DRMP), are constructed on the J-TEXT tokamak. SRMP is reconstructed from TEXT-U and mainly produces static m/n = 1/1, 2/1 and 3/1 resonant perturbation field, where m and n are the poloidal and toroidal mode numbers, respectively. DRMP, newly designed and installed inside the vacuum vessel, can generate pure 2/1 RMP. DRMP is also designed to operate in the AC mode and can produce rotating 2/1 RMP which will be used to study the tearing mode control. Due to the effect of the eddy current in the vacuum vessel wall, the amplitudes of the 2/1 component will be attenuated to about 1/3.6 of the DC value when the operation frequency is larger than 500 Hz. However, DRMP can still provide sufficient large rotating 2/1 perturbation for tearing mode related studies.     

Design of FPGA based high-speed data acquisition and real-time data processing system on J-TEXT tokamak

Tokamak experiment requires high-speed data acquisition and processing systems. In traditional data acquisition system, the sampling rate, channel numbers and processing speed are limited by bus throughput and CPU speed. This paper presents a data acquisition and processing system based on FPGA. The data can be processed in real-time before it is passed to the CPU. It provides processing ability for more channels with higher sampling rates than the traditional data acquisition system while ensuring deterministic real-time performance. A working prototype is developed for the newly built polarimeter–interferometer diagnostic system on the Joint Texas Experimental Tokamak (J-TEXT). It provides 16 channels with 120 MHz maximum sampling rate and 16 bit resolution. The onboard FPGA is able to calculate the plasma electron density and Faraday rotation angel. A RAID 5 storage device is adopted providing 700 MB/s read–write speed to buffer the data to the hard disk continuously for better performance.     

Proto-CIRCUS tilted-coil tokamak–torsatron hybrid: Design and construction

We present the field-line modeling, design, and construction of a prototype circular-coil tokamak–torsatron hybrid called Proto-CIRCUS. The device has a major radius R = 16 cm and minor radius a < 5 cm. The six “toroidal field” coils are planar as in a tokamak, but they are tilted. This, combined with induced or driven plasma current, is expected to generate rotational transform, as seen in field-line tracing and equilibrium calculations. The device is expected to operate at lower plasma current than a tokamak of comparable size and magnetic field, which might have interesting implications for disruptions and steady-state operation. Additionally, the toroidal magnetic ripple is less pronounced than in an equivalent tokamak in which the coils are not tilted. The tilted coils are interlocked, resulting in a relatively low aspect ratio, and can be moved, both radially and in tilt angle, between discharges. This capability will be exploited for detailed comparisons between calculations and field-line mapping measurements. Such comparisons will reveal whether this relatively simple concept can generate the expected rotational transform.     

Overview of the COMPASS CODAC system

This paper presents an overview of the Control, Data Acquisition, and Communication system (CODAC) at the COMPASS tokamak: the hardware set-up, software implementation, and communication tools are described.The diagnostics and the data acquisition are tailored for high spatial and temporal resolution required by the COMPASS physics programme, which aims namely at studies of the plasma edge, pedestal, and Scrape-off-Layer (SOL). Studies of instabilities and turbulence are also an integral part of the programme. Therefore, the data acquisition consists of more than 1000 channels, sampled at rates from 500 kS/s up to 2 GS/s.Presently, the feedback system controls the plasma position and shape, plasma current, and density and it includes 32 analogue input channels as well as 1 digital input/output channel and 8 analogue outputs. The feedback control runs within the Multi-threaded Application Real-Time executor (MARTe) framework with two threads, a 500 μs cycle to control slow systems and a 50 μs cycle to control the fast feedback power supplies for plasma position control.In this paper, special attention is paid to the links between the systems, to the hardware and software connections, and to the communication. The hardware part is described, the software framework is addressed, and the particular implementation – the dedicated software modules, communication protocols, and links to the database are described.     

Lithium technologies for edge plasma control

We have investigated two new modes of operation been in T-10 limiter tokamak experiments with a novel rotary feeder of lithium dust. Quasi steady-state mode I and pulse mode II of dust delivery were realized in both OH and OH + ECRH disruption free plasmas at the lithium flow rate up to 2 × 10²¹ atoms/s. A higher flow rate in mode II with injection rate of ～5 × 10²¹ atoms/s caused a series of minor disruptions, which was completed by discharge termination after the major disruption. The observed decreases of bolometer and D_β signals, with increase of the electron density during the lithium dust injection, reveal the effects of the first wall conditioning. The lithium technology may provide inherent safety pathway for major disruption mitigation in a tokamak reactor, which requires demonstration in contemporary tokamak experiments.     

Conceptual design of 5 MW-NBI injector for HL-2 M tokamak

The HL-2 M tokamak is now under construction in Southwestern Institute of Physics in China. As one of the main auxiliary heating systems for HL-2 M tokamak, a new NBI beam line with 5 MW NBI power, 42° injection angle, based on 4 sets of 80 kV/45 A/5 s bucket ion sources with geometrical beam focus, is conceptually designed with geometrical calculation and engineering simulations. The preliminary structure and layout of key components including ion sources, neutralizers, ion dumps, deflection magnet, beam edge scraper, long pulse calorimeter target, short pulse calorimeter target, injection port and beam drift duct are determined. The magnetic shielding of the stray field of HL-2 M tokamak is analyzed. Beam power transmission efficiency is calculated with geometrical algorithm. The ratio of neutral beam injection power to ion beam power is as high as 48%.     

Pulsed repetition rate nanosecond laser heating and ablation of the tokamak graphite tile deposited layers

Laser heating and ablation of the plasma-facing surface of a graphite tile from TEXTOR tokamak that was covered by a deposited carbon layer has been studied. Laser heating measurements were performed with a pulsed nanosecond Nd-YAG laser (2nd harmonic, 10 kHz repetition rate, 100 ns pulse duration). Surface temperature measurements were recorded with a home-made pyrometer having a response time of 15 μs (t_99%). The experimental results are simulated with an analytical model of laser heating of a surface covered by a deposited layer and heated repeatedly by laser pulses. The comparison between experimental and theoretical data of the observed temperature excursions enables us to assess the deposited carbon layer physical parameters (thermal conductivity, porosity, etc.) if the thermal and optical properties of the graphite substrate are known. Laser ablation measurements were performed with two pulsed nanosecond Nd-YAG lasers (20 Hz and 10 kHz repetition rate with 5 ns and 100 ns pulse duration, respectively). For a plasma-facing graphite surface covered by a thick (～30–50 μm) deposited carbon layer, the ablation threshold is 4.5 ± 1 kJ/m² regardless of the pulse duration. The obtained ablation threshold is significantly lower than the one measured for a virgin tokamak graphite sample. The comparison of the experimental results and theoretical data demonstrated that the laser ablation mechanisms for tokamak graphite and thick carbon layers deposited on plasma-facing surface are different.     

An initial study of demountable high-temperature superconducting toroidal field magnets for the Vulcan tokamak conceptual design

Recent developments have made it possible to consider high-temperature superconductor (HTS) for the design of tokamak toroidal field (TF) magnet systems, potentially influencing the overall design and maintenance scheme of magnetic fusion energy devices. Initial assessments of the engineering challenges and cryogenic-dependent cost and parameters of a demountable, HTS TF magnet system have been carried out using the Vulcan tokamak conceptual design (R = 1.2 m, a = 0.3 m, B₀ = 7 T) as a baseline. Jointed at the midplane to allow vertical removal of the primary vacuum vessel and routine maintenance of core components, structural D-shaped steel support cases provide cryogenic cooling for internally routed YBCO superconducting cables. The cables are constructed by layering ～50 μm thick commercially available YBCO tape, and the interlocking steel support cases self align during assembly to form internal resistive joints between YBCO cables. It is found that designing the TF magnet system for operation between 10 K and 20 K minimizes the total capital and operating cost. Since YBCO is radiation-sensitive, Monte Carlo simulation is used to study advanced shielding materials compatible with the small size of Vulcan. An adequate shield is determined to be 10 cm of zirconium borohydride, which reduces the nuclear heating of the TF coils by a factor of 11.5 and increases the YBCO tape lifetime from two calendar years in the unshielded case to 42 calendar years in the shielded case. Although this initial study presents a plausible conceptual design, future engineering work will be required to develop realistic design solutions for the TF joints, support structure, and cryogenic system.     

RF calculation and thermal-stress analysis of the ECRH launcher on EAST tokamak

EAST is a medium sized superconducting tokamak with major radius R = 1.8 m, minor radius a = 0.45 m, plasma current I_p ≤ 1 MA, toroidal field B_T ≤ 3.5 T and expected plasma pulse length up to 1000 s. An electron cyclotron resonance heating (ECRH) launcher for four-beam injection is being installed on EAST tokamak. Four electron cyclotron wave beams which are generated from four sets of 140 GHz/1 MW/1000 s gyrotrons will be injected into the plasma by the spherical focusing mirrors and plane mobile mirrors. The focusing mirrors are spherical to focus Gaussian beams after reflection. Four plane mobile mirrors independently steer continuously in the poloidal and toroidal direction controlled by motors. With the suitable distance between mirrors and appropriate focal length of focusing mirror, the beam radius in the resonance layer of plasma is 31.145 mm. The heat from plasma radiation and metal losses is loaded on the mobile mirror. In order to decrease the temperature and thermal stress, the inner equivalent diameter of water channels is 8 mm and the suggested water velocity is 4 m/s.     

10 years of engineering and physics achievements by the Large Helical Device project

This article reviews 10 years of engineering and physics achievements by the Large Helical Device (LHD) project with emphasis on the latest results. The LHD is the largest magnetic confinement device among diversified helical systems and employs the world's largest superconducting coils. The cryogenic system has been operated for 50,000 h in total without any serious trouble and routinely provides a confining magnetic field up to 2.96 T in steady state. The heating capability to date is 23 MW of NBI, 2.9 MW of ICRF and 2.1 MW of ECH. Negative-ion-based ion sources with the accelerating voltage of 180 keV are used for a tangential NBI with the power of 16 MW. The ICRF system has full steady-state operational capability with 1.6 MW. In these 10 years, operational experience as well as a physics database have been accumulated and the advantages of stable and steady-state features have been demonstrated by the combination of advanced engineering and the intrinsic physical advantage of helical systems in LHD. Highlighted physical achievements are high beta (5% at the magnetic field of 0.425 T), high density (1.1 × 10²¹ m^?3 at the central temperature of 0.4 keV), high ion temperature (T_i of 5.2 keV at 1.5 × 10¹⁹ m^?3), and steady-state operation (3200 s with 490 kW). These physical parameters have elucidated the potential of net-current free helical plasmas for an attractive fusion reactor. It also should be pointed out that a major part of these engineering and physics achievements is complementary to the tokamak approach and even contributes directly to ITER.     

EAST ion cyclotron resonance heating system for long pulse operation

Radio frequency (RF) power in the ion cyclotron range of frequencies (ICRF) is one of the primary auxiliary heating techniques for Experimental Advanced Superconducting Tokamak (EAST). The ICRF system for EAST has been developed to support long-pulse high-β advanced tokamak fusion physics experiments. The ICRF system is capable of delivering 12 MW 1000-s RF power to the plasma through two antennas. The phasing between current straps of the antennas can be adjusted to optimize the RF power spectrum. The main technical features of the ICRF system are described. Each of the 8 ICRF transmitters has been successfully tested to 1.5 MW for a wide range of frequency (25–70 MHz) on a dummy load. Part of the ICRF system was in operation during the EAST 2012 spring experimental campaign and a maximum power of 800 kW (at 27 MHz) lasting for 30 s has been coupled for long pulse H mode operation.     

Mechanical design for modification of a neutral beam for off-axis injection

DIII-D is planning to implement off-axis neutral beam current drive by neutral beam injection through a midplane port at angles up to 15° from horizontal. To accommodate the beam-line tilting, the following modifications are planned: (1) move the beam line away from the tokamak by 0.39 m to allow for a 0.68 m inside diameter welded bellows of necessary length to provide 15° of vertical motion between the vessel port and the beam line; (2) reduce the vertical height of the injected beam from 0.48 m to 0.43 m to provide clearance for the inclined beam as it passes through the length of the vessel port; (3) add a linkage system between the front of the beam line and the tokamak to restrain the NB against the vacuum loading from the bellows while maintaining zero roll about the axis of the beam line as it is moved about a virtual pivot axis; (4) add a forward and two rear vertical actuators for raising and lowering the beam line (These actuators require coordinated position control to rotate the NB about a virtual pivot axis.); (5) incorporate lateral restraint to comply with seismic requirements.     

Transmutation characteristics of minor actinides in a low aspect ratio tokamak fusion reactor

The transmutation characteristics of minor actinides in the transmutation reactor of a low aspect ratio (LAR) tokamak are investigated. One-dimensional neutron transport and burn-up calculations coupled with a tokamak systems analysis were performed to determine optimal system parameters. The dependence of the transmutation characteristics, including the neutron multiplication factor, produced power, and the transmutation rate, on the aspect ratio A in the range of 1.5–2.0 was examined. By adding Pu239 to the transmutation blanket as a neutron multiplication material, it was shown that a single transmutation reactor producing a fusion power of 150 MW_th can destroy minor actinides contained in the spent fuels for more than 38 units of 1 GW_e pressurized water reactors (PWRs) while producing a power in the range of 1.8–6.8 GW_th.     

Engineering design for the HL-2M tokamak components

The HL-2A tokamak will be modified into HL-2M. The Bt at the plasma center (major radius R = 1.78 m) is 2.2 T, the minor radius is 0.65 m. The plasma current I_P of HL-2M will reach up to 2.5 MA, the elongation and triangularity is more than 1.8 and more than 0.5, respectively. The vacuum vessel torus consists of 20 sectors with “D” shaped cross-section and double wall structure. 20 toroidal field coil bundles comprise 140 turns which are designed with demountable joints, the poloidal field coils system consists of 25 coils. The engineering design and calculation for field coil system, vacuum vessel, support structure, etc. are finished, many key issues for manufacture process have been discussed with industry and the fabrication of main components of HL-2M tokamak will be carried out in factories.     

Cluster storage for COMPASS tokamak

The COMPASS tokamak is expected to produce several gigabytes of data per shot in near future. A new storage system is needed to accommodate and access all the data. It should be scalable, fault-tolerant, and parallel. It should not be based on proprietary solutions to maintain independence from hardware and software manufacturers and preferably it should be built on inexpensive commodity hardware. One of the promising distributed parallel fault-tolerant file systems, GlusterFS, was selected for testing. The aim of the work was to make initial tests of a particular small GlusterFS setup to confirm its aptitude for the COMPASS storage system. Aggregated reading throughput from multiple NFS clients was one of the most important figures that were benchmarked, it scaled well with the number of clients, starting just above 60 MiB/s¹ for 1 client and going slightly over 300 MiB/s for 6 clients.     

Feasibility of a multi-purpose demonstration neutron source based on a compact superconducting spherical tokamak

Tokamak neutron sources would allow near term applications of fusion such as fusion–fission hybrid reactors, elimination of nuclear wastes, production of radio-isotopes for nuclear medicine, material testing and tritium production. The generation of neutrons with fusion plasmas does not require energetic efficiency; thus, nowadays tokamak technologies would be sufficient for such purposes. This paper presents some key technical details of a compact (～1.8 m³ of plasma) superconducting spherical tokamak neutron source (STNS), which aims to demonstrate the capabilities of such a device for the different possible applications already mentioned. The T-11 transport model was implemented in ASTRA for 1.5 D simulations of heat and particle transport in the STNS core plasma. According to the model predictions, total neutron production rates of the order of ～10¹⁵ s^?1 and ～10¹³ s^?1 can be achieved with deuterium/tritium and deuterium/deuterium respectively, with 9 MW of heating power, 1.4 T of toroidal magnetic field and 1.5 MA of plasma current. Engineering estimates indicate that such scenario could be maintained during ～20 s and repeated every ～5 min. The viability of most of tokamak neutron source applications could be demonstrated with a few of these cycles and around ～100 cycles would be required in the worst cases.     

The lower hybrid current drive system for steady-state operation of the Vulcan tokamak conceptual design

The steady-state current drive system for the Vulcan tokamak concept has been designed, taking into account requirements of high field, small size, and high operational wall temperature (B₀ = 7 T, R₀ = 1.2 m, T_wall > 800 K). This lower hybrid current drive system allows steady-state operation by utilizing high field side launch, high RF source frequency (8 GHz), and dedicated current drive ports. An iterative MHD and current drive solver is used to determine the ideal launching spectra and location to assure strong single pass absorption. It is found that with nominal Vulcan operational parameters (n_e ≈ 4 × 10²⁰ m^?3, T_e ≈ 2.8 keV, I_p = 1.7 MA, P_LHCD = 19.8 MW) bootstrap currents of ～70% and lower hybrid current drive efficiencies of 1.16 × 10¹⁹ A W m^?2 could be achieved. The optimized solution yielded advanced tokamak profiles with q values on-axis above 2. A conceptual design of the system is presented, which takes into account space, power, cooling, and launched spectrum requirements. The system is found to be compatible with the vacuum vessel design and requires cooling power of <1 MW per waveguide bundle.