Synergetic Effects of Runaway and Disruption Induced by VDE on the First Wall Damage in HL-2A

The plasma facing component in HL-2A has been damaged seriously after disruption,and for this reason its operation is suspended for maintenance.The experimental phenomena and plasma configurations,calculated by the current filament code(CF-code) using the plasma parameters measured by diagnostics and the signals of the magnetic probes,confirm that the first wall is damaged by the synergetic effects of runaway electrons and disruption induced by a vertical displacement event(VDE).When the plasma column is displaced upward/downward,the strong runaway electrons normally hit the baffle plate of the MP 3 or MP 1 coil in the upper and lower divertor during the disruption,causing the baffle plates to be holed and wrinkled by the energetic runaway current,and water(for cooling or heating the baffle plates) to leak into the vacuum vessel.Another disastrous consequence is that bellows underlying the baffle plate and outside the coil of MP 3 for connecting two segments of the jacket casing pipe are punctured by arcing.The arc may be part of the halo current that forms a complete circuit.The experimental phenomena are indirect but compelling evidence for the existence of a halo current during the disruption and VDE,though the halo current has not been measured by the diagnostics in the HL-2A tokamak.     

Physical Engineering Test and First Divertor Plasma Configuration in EAST

Physical engineering capability on the superconducting magnetic system of EAST was tested and first divertor plasma configuration in EAST was obtained. The extrapolation of the safety limit has verified the reliability of the system for long pulse operation. A stably controlled diverted plasmas configuration with an elongation n in excess of 1.8 and plasma current of up to 500 kA, by using the （copper） internal coils to control the vertical displacement instability was obtained by an optimized plasma control algorithm. Highly shaped plasma at various configurations, which almost covers all designed configurations for EAST, was generated stably. A number of operational issues, such as plasma initiation, ramp up and configuration control with constraints of superconducting coils, were successfully investigated. All of the results obtained proved both the capability of the superconducting poloidal magnets for operation under steady-state condition and effectiveness of the plasma control algorithm for EAST.     

Numerical Simulation Approaches to Evaluating the Electromagnetic Loads on the EAST Vacuum Vessel

Numerical simulation approaches are developed to compute the electromagnetic forces on the EAST vacuum vessel during major disruptions and vertical displacement events, with the halo current also considered. The finite element model built with ANSYS includes the vacuum vessel, the plasma facing components and their support structure, and the toroidal and poloidal field coils. The numerical methods are explained to convince of its validity. The eddy current induced by the magnetic flux variation and the conducting current caused by the halo current are also presented for discussion. The electromagnetic forces resulting from the numerical simulation are proven to be useful for structure design optimization. Similar methods can be applied in the upgrades of the EAST device.     

Predictions for EAST Divertor Performance

A detailed study of the divertor performance in EAST has been performed for both its double null （DN） and single null （SN） configurations. The results of application of the SOLPS （B2-Eirene） code package to the analysis of the EAST divertor are summarized. In this work, we concentrate on the effects of increased geometrical closure and of magnetic topology variation on the scrape-off layer （SOL） and divertor plasma behavior. The results of numerical predictions for the EAST divertor operational window are also described in this paper. A simple Core-SOL- Divertor （C-S-D） model was applied to investigate the possibility of extending plasma operational space of low hybrid current drive （LHCD） experiments for EAST.     

Main Directions and Recent Test Modeling Results of Lithium Capillary-Pore Systems as Plasma Facing Components

At present the most promising principal solution of the divertor problem appears to be the use of liquid metals and primarily of lithium Capillary-Pore Systems (CPS) as of plasma facing materials. A solid CPS filled with liquid lithium will have a high resistance to surface and volume damage because of neutron radiation effects, melting, splashing and thermal stressinduced cracking in steady state and during plasma transitions to provide the normal operation of divertor target plates and first-wall protecting elements. These materials will not be the sources of impurities inducing an increase of Zeef and they will not be collected as dust in the divertor area and in ducts. Experiments with lithium CPS under simulating conditions of plasma disruption on a hydrogen plasma accelerator MK-200 [-(10 - 15) MJ/m^2, - 50 μs] have been performed. The formation of a shielding layer of lithium plasma and the high stability of these systems have been shown. The new lithium limiter tests on an up-graded T-11M tokamak (plasma current up to 100 kA, pulse length -0.3 s) have been performed. Sorption and desorption of plasma-forming gas, lithium emission into discharge, lithium erosion, deposited power of the limiter are investigated in these experiments. The first results of experiments are presented.     

Design Features and Commissioning of the Versatile Experiment Spherical Torus (VEST) at Seoul National University

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 preionization 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.     

Plasma production and preliminary results from the ADITYA Upgrade tokamak

The Ohmically heated circular limiter tokamak ADITYA (R0 =75 cm,a =25 cm) has been upgraded to a tokamak named the ADITYA Upgrade (ADITYA-U) with an open divertor configuration with divertor plates.The main goal of ADITYA-U is to carry out dedicated experiments relevant for bigger fusion machines including ITER,such as the generation and control of runaway electrons,disruption prediction,and mitigation studies,along with an improvement in confinement with shaped plasma.The ADITYA tokamak was dismantled and the assembly of ADITYA-U was completed in March 2016.Integration of subsystems like data acquisition and remote operation along with plasma production and preliminary plasma characterization of ADITYA-U plasmas are presented in this paper.     

Predictive Modeling of EAST Divertor Target Power Loading with Varying Chemical Sputtering Conditions

The Experimental Advanced Superconducting Tokamak （EAST） is being built in China to achieve high power and long pulse operation for studies of reactor-relevant issues under steady-state conditions. A major concern for EAST is the power handling capability of the divertor target plates, which is a critical issue for future high-powered steady-state tokamaks, such as ITER. Detailed modeling using B2/EIRENE code package and the most recent chemical sputtering data shows that the presence of strong chemical sputtering at the main chamber wall leads to strong carbon radiation in the periphery of the confined plasma, significantly reducing the heat fluxes to the target plates and facilitating plasma detachment at a lower density desired for lower hybrid current drive in EAST, with only a slight increase in Zeff at the edge. The target heat load can be further reduced by operating with a double-null divertor configuration, which also leads to a significant reduction in the edge Zeff. However, the code predicts that the double-null operation would result in a strong divertor asymmetry in target power loading, favoring the outside targets.     

Progress Toward Steady-State Operation on Tore Supra

Important technological and physics issues related to steady-state operation required for next step are being examined on Tore Supra,after a major upgrade of internal components in order to increase the heat extraction capability to 25 MW for 1000 s.Here,we show first experimental results,where all the plasma facing components were actively cooled during pulses exceeding four minutes,with reactor-relevant heat load.New physics was observed in non-inductively driven plasmas,including a stationary peaked radial profile of the plasma density generated by an anomalous inward pinch;and a regime characterized by sinusoidal oscillations of central electron temperature,governed by non-linear coupling between heat transport and plasma current analogous to a predator-prey mechanism.     

Design of Controller for New EAST Fast Control Power Supply

The effectiveness of the magnetic confinement of plasma can be improved by elongat- ing the plasma cross-section in tokamak devices. But elongated plasma has vertical displacement instability, so a feedback control system is needed to restrain the plasma＇s vertical displacement. A fast control power supply is needed to excite the active feedback coils, which produces a magnetic field to control the plasma＇s displacement. With the development of EAST, the fast control power supply needs to keep on enhancing the fast response and output current. The structure of a new power supply is introduced in this paper. The method of multiple inverters paralleled with the current sharing reactor is presented to meet the need for large current and fast control. According to the design demands of the EAST fast control power supply, the adjuster of the current close loop is applied to the inverter, which can advance its ability to restrain the loop current in low frequency and DC output. The result of the experiment confirms the validity of the proposed scheme and control strategy.     

A Lower Rigid Support Structure for the HT—7U Vacuum Vessel

Vacuum vessel of the HT-7U is a fully welded toroidal structure with a noncircularcross-section nested in the bore of the TF coils. According to the requirement of the physicsdesign, sixteen horizontal ports on outboard mid-plane and thirty-two vertical ports on the topand bottom are designed for diagnostics, plasma heating, current driving, vacuum pumping andgas puffing. Bellows on these port necks are used for flexible components to absorb the relativedisplacement in radial and vertical directions due to external load, thermal expansion or contrac-tion and assembly tolerance, and also used for isolation of mechanical vibration. For the supportsystem of vacuum vessel it should be not only strong enough to withstand forces acting on thevessel interior components and the vessel itself due to the dead weight and electromagnetic inter-actions during plasma disruption, but also sufficiently flexible to be suited to thermal expansionduring baking. In order to solve this contradiction a new kind of low rigid s     

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.     

Capability Assessment of the Equilibrium Field System in KTX

Radial equilibrium of the KTX plasma column is maintained by the vertical field which is produced by the equilibrium field coils.The equilibrium is also affected by the eddy current,which is generated by the coupling of copper shell,plasma and poloidal field coils.An equivalent circuit model is developed to analyze the dynamic performance of equilibrium field coils,without auxiliary power input to equilibrium field coils and passive conductors.Considering the coupling of poloidal field coils,copper shell and plasma,the evolution of spatial distribution of the eddy current density on the copper shell is estimated by finite element to analyze the effect of shell to balance.The simulation results show that the copper shell and equilibrium field coils can provide enough vertical field to balance 1 MA plasma current in phase 1 of a KTX discharge.Auxiliary power supply on the EQ coils is necessary to control the horizontal displacement of KTX due to the finite resistance effect of the shell.     

Stress and Thermal Analysis of the In-Vessel RMP Coils in HL-2M

A set of in-vessel resonant magnetic perturbation(RMP) coils for MHD instability suppression is proposed for the design of a HL-2M tokamak.Each coil is to be fed with a current of up to 5 kA,operated in a frequency range from DC to about 1 kHz.Stainless steel(SS) jacketed mineral insulated cables are proposed for the conductor of the coils.In-vessel coils must withstand large electromagnetic(EM) and thermal loads.The support,insulation and vacuum sealing in a very limited space are crucial issues for engineering design.Hence finite element calculations are performed to verify the design,optimize the support by minimizing stress caused by EM forces on the coil conductors and work out the temperature rise occurring on the coil in diferent working conditions,the corresponding thermal stress caused by the thermal expansion of materials is evaluated to be allowable.The techniques to develop the in-vessel RMP coils,such as support,insulation and cooling,are discussed.     

Present Status of the KSTAR Superconducting Magnet System Development

The mission of Korea Superconducting Tokamak Advanced Research (KSTAR) project is to develop an advanced steady-state superconducting tokamak for establishing a scientific and technological basis for an attractive fusion reactor. Because one of the KSTAR mission is to achieve a steady-state operation, the use of superconducting coils is an obvious choice for the magnet system. The KSTAR superconducting magnet system consists of 16 Toroidal Field (TF) coils and 14 Poloidal Field (PF) coils. Internally-cooled Cable-In-Conduit Conductors (CICC) are put into use in both the TF and PF coil systems. The TF coil system provides a field of 3.5 T at the plasma center and the PF coil system is able to provide a flux swing of 17 V-sec. The major achievement in KSTAR magnet-system development includes the development of CICC,the development of a full-size TF model coil, the development of a coil system for background magnetic-field generation , the construction of a large-scale superconducting magnet and CICC test facility. TF and PF coils are in the stage of fabrication to pave the way for the scheduled completion of KSTAR by the end of 2006.     

Electromagnetic Modeling of the Passive Stabilization Loop at EAST

A passive stabilization loop (PSL) has been designed and manufactured in order to enhance the control of vertical instability and accommodate the new stage for high-performance plasma at EAST.Eddy currents are induced by vertical displacement events (VDEs) and disrup-tion,which can produce a magnetic field to control the vertical instability of the plasma in a short timescale.A finite element model is created and meshed using ANSYS software.Based on the simulation of plasma VDEs and disruption,the distribution and decay curve of the eddy currents on the PSL are obtained.The largest eddy current is 200kA and the stress is 68MPa at the outer current bridge,which is the weakest point of the PSL because of the eddy currents and the magnetic fields.The analysis results provide the supporting data for the structural design.     

Investigation of Divertor Detachment in EAST by ＇Two-Point＇ Model

Divertor plasma detachment offers one of the most promising operating modes for fusion devices because of low target power loading. In this article a ＇two-point＇ model is used to investigate the formation of detachment and explore the route to detachment in EAST, in order to find an ideal operation window. The simulation results show that impurity radiation and ionneutral friction are the main causes of divertor plasma detachment at the target plates. Raising the safety factor and reducing the upstream power density provide effective means to achieve the detachment due to the increased radiation power fraction. Puffing Ar and Ne impurities and raising the safety factor can bring the upstream high plasma temperature region （above 100 eV） and the low target plasma temperature region （below 10 eV） close to each other in terms of the separatrix density. But it is difficult to find a common operating region which satisfies both conditions. High recycling and detached regimes provides an ideal operation window because of the steady upstream condition and low target power load.     

Preliminary study of divertor particle exhaust in the EAST superconducting tokamak

The particle exhaust of the upper tungsten and lower carbon divertors in EAST has been preliminarily studied during the 2016 experimental campaign. The density decay time during terminating gas puffing has been employed as a key parameter to evaluate the divertor particle exhaust performance. Comparative plasma discharges have been carried out on the particle exhaust performance between two toroidal field directions in the upper single null and lower single null divertor configurations. This work has enhanced the understanding of the effects of the in–out asymmetry and divertor geometry on the efficiency of the divertor particle exhaust. In addition, the sensitivity of the particle exhaust capability on different strike point locations has been analyzed. The experimental results are expected to provide important information on the future upgrade of EAST bottom divertor and facilitate the realization of longer pulse operation.     

Investigation of turbulence rotation in the SOL and plasma edge of W7-X for different magnetic configurations

The W7-X stellarator is optimized with respect to neoclassical transport. Therefore turbulent transport plays an important role. It is equipped with an inertial cooled graphite divertor which intersects the island chain at the plasma edge depending on the magnetic configuration. Additional control coils and the plasma current modify the iota profile at the plasma edge and shift the position of the island chain. To monitor the effects on the poloidal propagation velocity in the scrape-off layer(SOL) and the plasma edge, an O-mode Poloidal Correlation Reflectometer(PCR) is used which simultaneously monitors the propagation of low-k turbulence. Operating in the density range of 0.6?×?10~(19) m~(-3) to2?×?10~(19) m~(-3) it covers a large part of the SOL and the plasma edge and allows for the experimental determination in the last closed flux surface(LCFS) and the associated shear layer in low to middensity discharges. In this paper it is shown that the propagation in the shear layer and its vicinity is determined best, when based on an elliptical model. Different magnetic configurations with magnetic edge topology of five independent islands for ι?=?1 and six linked islands for ι?=?0.81 are investigated. Also the effects of the plasma current and additional control coils on the edge magnetic topology are studied. The coherence spectra of antenna pairs for different poloidal separations is investigated. Using a decomposition method for the measured coherence spectra the characterization of turbulence spectra is possible with respect to e.g. broad band turbulence and quasi coherent modes.A strong reduction of the broad band turbulence is observed in the vicinity of the LCFS which is evidence for the suppression of low-k turbulence at the shear layer.     

Electromagnetic–thermal–structural coupling analysis of the ITER edge localized mode coil with flexible supports

In a fusion reactor, the edge localized mode(ELM) coil has a mitigating effect on the ELMs of the plasma. The coil is placed close to the plasma between the vacuum vessel and the blanket to reduce its design power and improve its mitigating ability. The coil works in a high-temperature,high-nuclear-heat and high-magnetic-field environment. Due to the existence of outer superconducting coils, the coil is subjected to an alternating electromagnetic force induced by its own alternating current and the outer magnetic field. The design goal for the ELM coil is to maintain its structural integrity in the multi-physical field. Taking as an example the middle ELM coil(with flexible supports) of ITER(the International Thermonuclear Fusion Reactor), an electromagnetic–thermal–structural coupling analysis is carried out using ANSYS. The results show that the flexible supports help the three-layer casing meet the static and fatigue design requirements. The structural design of the middle ELM coil is reasonable and feasible. The work described in this paper provides the theoretical basis and method for ELM coil design.