Design, fabrication and testing of UHV compatible high power RF devices for lower hybrid current drive system on SST-1 tokamak

Maintaining plasma current under steady state conditions is one of the most important pre-requisites for a tokamak-based reactor. Lower hybrid current drive (LHCD) system aims to drive tokamak plasma current by means of RF power. The LHCD system on SST-1 tokamak is based on two 500 kW, CW klystrons operating at 3.7 GHz. A waveguide transmission line transmits power from source to the antenna. A phased array waveguide antenna is used to couple power to the plasma. The antenna side of the transmission line is placed inside the tokamak vacuum vessel. The design and fabrication of this In-Vessel system has to satisfy the demands of high power RF as well as ultra high vacuum (UHV) compatibility. This paper describes some of the critical UHV compatible In-Vessel RF devices, their design, fabrication, and test results.     

Numerical Simulation of Non-Inductive Current Driven Scenario in EAST Using Neutral Beam Injection

For achieving the scientific mission of long pulse and high performance operation,experimental advanced superconducting tokamak(EAST) applies fully superconducting magnet technology and is equiped with high power auxiliary heating system.Besides RF(Radio Frequency) wave heating,neutral beam injection(NBI) is an effective heating and current drive method in fusion research.NBCD(Neutral Beam Current Drive) as a viable non-inductive current drive source plays an important role in quasi-steady state operating scenario for tokamak.The non-inductive current driven scenario in EAST only by NBI is predicted using the TSC/NUBEAM code.At the condition of low plasma current and moderate plasma density,neutral beam injection heats the plasma effectively and NBCD plus bootstrap current accounts for a large proportion among the total plasma current for the flattop time.     

DIII-D research program

Summary The DIII-D tokamak is uniquely positioned to contribute to the development of Magnetic Fusion Energy over the next decade. Recent stability and confinement improvements resulting from current profile control and the discovery of the VH-mode stress the importance of non-inductive current drive. The DIII-D program plan calls for the implementation of high rf systems for such profile control; localized heating and current drive for improved to tokamak performance and advancing divertor research under current drive conditions. Funding levels will determine the potential impact of this $400 M research facility by determining the pace for implementation of rf and divertor upgrades as well as determining the amount of experimental operating time. A strong DIII-D program can effectively address the R&D issues of next generation tokamaks and allow the large number of DIII-D collaborators to explore new ideas to advance the tokamak to a commercially attractive energy option.     

Analysis of high β regimes for DEMO

Stability limit calculations are presented for a range of tokamak power plant equilibria. The current drive requirements to sustain the optimised equilibrium profiles are confirmed by a transport code and the plasma shape is obtained from free-boundary equilibrium calculations. A pressure pedestal is included according to empirical scaling and ballooning mode stability limits. A terative optimisation of the profiles is undertaken to improve the baseline profiles in order to achieve the highest possible plasma performance and most favourable magnetohydrodynamic stability within conservative assumptions in order to increase confidence in the availability and control of the plasma. This results in a fully noninductive baseline operating scenario for a tokamak power plant design which has a broad low-shear q-profile which is meant to complement previous advanced tokamak design studies.     

Optimization of an in-vessel visible inspection system for a long-pulse operation in KSTAR

To monitor the global formation of shaped plasmas, motion, and damage to the internal structures of the vacuum vessel, an in-vessel visible inspection system has been developed and operated on the Korean superconducting tokamak advanced research (KSTAR) device. The system contributed much to research progress on KSTAR such as the plasma start-up, plasma wall interactions, edge-localized modes, and disruptions. Moreover the need to perform inspections became important with high plasma power operation because of the increased frequency of first wall damage following off-normal events. Therefore the system is being improved from its original concept, and its final goal is operation during steady-state operation of the tokamak. The system consists of three fast visible cameras and two light-emitting diode illuminators. They are designed to be controlled fully from the control room to provide inspection capability at any time during hostile operating conditions. In this paper, we describe the upgrade of the system and recent results of the visible inspection system with the images of the KSTAR discharges for the last four years. Finally, we discuss the technical issues for a long pulse steady-state operation.     

Poloidal Field Control for the HT—7U Superconducting Tokamak

Controlling the poloidal field(PF) in the HT-7U superconducting tokamak is critical to the realization of the mission of advanced tokamak research.Plasma start-up,plasma position,shape,current control and plasma shape reconstruction have been performed as a part of its design process.The PF coils have been designed to produce a wide range of plasmas,Plasma start-up can be achieved for multiple conditions.Fast controlling coils for plasma position inside the vacuum vessel are sued for controloling the plasma vertical position on a short timescale.The PF coils control the plasma current and shape on a slower timescale,VXI(VME bus extensions for Instrumentation)Bus system and DSP(Digital Signal Processor is a basic unit of the feedback control system),the response time of which is about(2-4)ms.The basic unit of this system ,the shape-controlling algorithms of a few critical points on plasma boundary and real-time equilibrium fitting(RTEFIT)will be described in this paper.     

The ITER divertor pumping system,design evolution,simplification and performance

The ITER tokamak will be fuelled at a time averaged rate of up to 200 Pam³ s^?1 requiring neutralised gas in the divertor to be pumped to balance the fuelling and remove the fusion helium and other impurities in the exhaust. This is achieved on ITER using large bespoke cryo-sorption pumps. In this paper design evolution of the ITER divertor pumping system is outlined from the 1998 configuration to the current design. Details of the new, 6 direct pump, system design which will be used in the build of ITER are given. The operating modes of the new system configuration for different plasma scenarios are described and the performance of the new system is analysed and compared with previous baselines.     

Multi-Inputs/Multi-Outputs control of plasma current and loop voltage on Tore Supra

During a tokamak discharge, several control modes may have to be run in sequence in order to perform the control of the different discharge phases. The transitions between these control modes are not always easy to handle because in most cases the coupling between the controlled plasma quantities is not taken into account in each control mode design process. This paper presents a new Multi-Inputs/Multi-Outputs (MIMO) controller applied on Tore Supra to control both plasma current and flux variations through the central solenoid voltage and the lower hybrid current drive (LHCD) system power. It deals with the transition from a loop voltage floating mode to a loop voltage control mode. The controller, synthesized and tuned using a model-based approach, has been validated in simulation before its successful implementation on Tore Supra experiments.     

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.     

Upgrade of plasma density feedback control systemin HT-7 tokamak

The HT-7 is a superconducting tokamak in China used to make researches on the controlled nuclear fusion as a national project for the fusion research. The plasma density feedback control subsystem is the one of the subsystems of the distributed control system in HT-7 tokamak (HT7DCS). The main function of the subsystem is to control the plasma density on real-time. For this reason, the real-time capability and good stability are the most significant factors, which will influence the control results. Since the former plasma density feedback control system (FPDFCS) based on Windows operation system could not fulfill such requirements well, a new subsystem has to be developed. The paper describes the upgrade of the plasma density feedback control system (UPDFCS), based on the dual operation system (Windows and Linux), in detail.     

Plasma Shape and Current Control Simulation of HT-7U Tokamak

This paper describoes the discharge simulation of HT-7U tokamak plasma equilibrium and plasma current by solving MHD equations and surface average transport equations using an equilibrium evolution code. The simulated result shows the evolution of plasma parameter versus time .The simulated result can play an important role in the design of the plasma equilibrium and control system of a tokamak.     

Paralleling Inverters Study of EAST Vertical Stability Coil Power Supply

Experimental advanced superconducting tokamak vertical stability (VS) coil power supply is a large capacity single phase inverter power supply. To meet the requirement of large current and fast response, multi-inverters in parallel is presented, which based on carrier phase-shifted modulation technology. In parallel inverter system, the disperse circuit parameters and phase-shift carriers between parallel inverter units will cause circulating current, which contains fundamental component and a large number of harmonic components. In this paper, the model of circulating current is analyzed when VS coil power supply is working in voltage given mode, and an instantaneous current sharing control strategy is proposed based on the combination of current sharing inductor and instantaneous circulating current feedback control. Parallel inverter units are connected together through current sharing inductors which can change the impedance characteristic of the circulating impedance and well restrain the high-frequency circulating current. Then, the real part of the circulating impedance will be increased and the ability to restrain the low-frequency circulating current will be advanced by introducing virtual resistance, which is realized in the instantaneous circulating current feedback routine. The designations of the current sharing inductor and the virtual resistance are provided. The results of simulation and experiment verify that this current-sharing strategy is available and efficient.     

中国环流器二号A(HL-2A)装置物理设计

文章是关于中国环流器二号A(HL-2A)装置物理设计的总结报告,包括以下几方面的内容:分析计算等离子体截面变形及由截面拉长引起的垂直不稳定性,提出对HL-2A极向磁场线圈电流和控制系统的要求;研究通过中性束注入加热(NBI)和低混杂波电流驱动(LHCD)实现等离子体剖面控制,模拟并设计HL-2A的高性能的运行模式;分析HL-2A先进约束位形(RS位形)下的磁流体力学不稳定性,为实现高性能模式稳态运行的等离子体控制指出方向;同时,利用数值模拟分析HL-2A偏滤器等离子体性能,为偏滤器的改进提供依据。     

Management of complex data flows in the ASDEX Upgrade plasma control system

Establishing adequate technical and physical boundary conditions for a sustained nuclear fusion reaction is a challenging task. Phased feedback control and monitoring for heating, fuelling and magnetic shaping is mandatory, especially for fusion devices aiming at high performance plasmas. Technical and physical interrelations require close collaboration of many components in sequential as well as in parallel processing flows. Moreover, handling of asynchronous, off-normal events has become a key element of modern plasma performance optimisation and machine protection recipes.The manifoldness of plasma states and events, the variety of plant system operation states and the diversity in diagnostic data sampling rates can hardly be mastered with a rigid control scheme. Rather, an adaptive system topology in combination with sophisticated synchronisation and process scheduling mechanisms is suited for such an environment. Moreover, the system is subject to real-time control constraints: response times must be deterministic and adequately short.Therefore, the experimental tokamak device ASDEX Upgrade employs a discharge control system DCS, whose core has been designed to meet these requirements. In the paper we will compare the scheduling schemes for the parallelised realisation of a control workflow and show the advantage of a data-driven workflow over a managed workflow. The data-driven workflow as used in DCS is based on signals connecting process outputs and inputs. These are implemented as real-time streams of data samples. Consequently, real-time signal management forms the foundation of DCS. The paper explains the principal features such as tagged samples, signal groups, algorithmic blocks and processes as well as scheduling schemes which allow DCS control applications to be defined as self-contained modular building blocks glued together by a software framework.By virtue of this sound foundation, DCS is a mature but still evolving system for reliable, distributed control of an entire tokamak device coordinating and monitoring 20 diagnostic systems, 14 magnetic power supplies, 5 heating systems with a total power of more than 25 MW, 8 gas fuelling channels, a pellet injector and a killer gas gun.     

Design,Construction, and Installation of Movable Electrode Biasing System on Tokamak

The first results of the movable electrode biasing experiments performed on the IR-T1 tokamak are presented. For this purpose, a movable electrode biasing system was designed, constructed, and installed on the IR-T1 tokamak, and then the positive voltage applied to an electrode inserted inside the tokamak limiter and the plasma current, poloidal and radial components of the magnetic fields, loop voltage, and diamagnetic flux in the absence and presence of the biased electrode were measured. Results compared and discussed.     

A small,novel, remote in-vessel inspection system for the Alcator C-Mod tokamak

A small robust system has been constructed for in-situ visual inspection of the Alcator C-Mod tokamak. The system consists of a small, light, wide-angle high definition camera and LED package housed in a nacelle on the end of thin, rigid, 3.5 m long support pole. The nacelle has two actuated degrees of freedom allowing the camera to observe nearly 4π steradians. The support pole has a specific slight curve that allows it to pass to either side of the center column of the tokamak to observe the entirety of the vessel interior, while still fitting through the small aspect ratio Alcator C-Mod vacuum port structure. The support pole and camera can enter the vessel through any horizontal vacuum port with an inner diameter greater than 4 cm, thus a dedicated port is not required. The inspection is typically undertaken during maintenance periods when the vessel is filled with a noble gas near atmospheric pressure thus minimizing the influx of water vapor and the concomitant loss of wall conditioning. The system is operated manually, producing photos and video which are reviewed in near real-time. Nearly the entire vessel, including the plasma facing components, can be carefully inspected in 3–5 h. The system provides improved characterization of the interior components and surfaces of the tokamak with a modest engineering and operational effort. Information gathered from the system has identified damage to plasma facing components that were interfering with tokamak operation, as well as damage to mechanical components which were redesigned during the remainder of the campaign, thereby enhancing program planning.     

Rview of pellet fueling

The injection of frozen pellets composed of the isotopes of hydrogen has become the leading candidate for refueling fusion power reactors based on the tokamak concept. This lofty position has been reached partly as a result of efforts to find an attractive solution to the perplexing problem of depositing atoms of fuel deep within the magnetically confined, hot plasma, and because of some recent experimental successes. To some extent, the relative merits of this technique will depend upon the distance that the cryogenic pellet will penetrate such a plasma, and the early exploratory research has addressed this problem on both theoretical and experimental fronts. The conclusion from the theoretical effort is that a protective blanket consisting of hydrogenic gas or cold plasma will envelope the pellet and partially shield the surface from the intense plasma heat flux. The blanket prolongs pellet lifetime, but penetration to the plasma center might require pellet injection velocities in excess of 10 km/s. The need for central penetration has not yet been established either theoretically or experimentally. The experiments performed to date have verified the existence of a shielding mechanism in general, and pellet ablation models that incorporate neutral gas shielding in particular are in adequate agreement with the experiments. Magnetic shielding effects are expected to contribute to, but not dominate, self-shielding in the higher plasma temperature regimes of the future. The tokamak plasma has demonstrated a surprising resilience even to massive density perturbations caused by the large refueling pellets used in present experiments. The characteristic discharge behavior is qualitatively not unlike that observed with gas puffing; but, for the first time, central plasma fueling has been studied, and this does not appear to be superior to refueling by partial pellet penetration. If relatively large pellets containing a significant fraction of the total plasma charge are acceptable in the present resistive plasma regimes, then it can be argued that they should have little impact on the gross stability of a hot thermonuclear tokamak plasma. Large pellets are preferable from the standpoint of attaining deep penetration, and this has important implications for the technology of pellet injection. The interesting velocity regime of 1 km/s has already been achieved with simple gun-type devices and this should be adequate for near-term tokamak experiments. Further improvements are anticipated, but the 10 km/s and above regime is uncertain; and, if current theory and experiments extrapolate to the future, such velocities might be desirable but unnecessary.     

Studies on the requirements of the power supply system for the resistive-wall-mode control in JT-60SA

The JT-60SA satellite tokamak will be built in Naka, Japan. One of the main aims of this machine is to achieve steady-state high-beta plasmas. To reach this result, passive stabilizing plate (SP) and resistive wall modes (RWM) active control system based on 18 in-vessel coils will be installed. In the present design, these coils are placed on the plasma side of the SP, behind the first wall. This solution maximizes the efficiency in producing fast magnetic fields into the plasma by minimizing the shielding effect of the passive structures. Then, if the power supply (PS) and the control system have sufficient dynamic performance, it is possible to control the RWM with very low magnetic fields. This allows minimizing the Ampere-turns and the power requested to control the RWM. Conversely, the very fast dynamics required represents one of the main issues for the design of the RWM control system. This paper, after having recalled the main specification data for the RWM control system deriving from the physics studies, describes the analyses performed to complete the set of requirements necessary for the PS design. The characterization of coils and feeders is shown and the voltage necessary to produce the required current and bandwidth is quantified. Possible connections among PS and coils are analyzed in order to achieve the highest possible flexibility in controlling the RWM with a reduced set of independent PS. Finally, considerations on reasonable voltage margins to cope with load uncertainties are given.