Mitigation of edge localised modes with magnetic perturbations in ASDEX Upgrade

The first stage of a significant enhancement of the ASDEX Upgrade experiment with in-vessel coils for non-axisymmetric magnetic perturbations is now operational. First experiments have shown that ELM mitigation can be achieved using various perturbation field configurations with toroidal mode numbers n = 1, 2, 4. The main access criteria is the plasma edge pedestal density to exceed a threshold, which takes the lowest value of about 60% of the Greenwald density for resonant |n| = 1 perturbations. In H-mode plasmas, mode locking or error field-induced magnetic islands are generally not observed. Due to the low local shear of the plasma magnetic field in the vicinity of the perturbation coils around the outboard midplane, the magnetic perturbation is resonant simultaneously on several rational surfaces. It is hypothesised that the existence of image currents on these surfaces ensures good shielding of the error field in the confined plasma.     

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.     

Fabrication and installation of KSTAR in-vessel control coils

The in-vessel control coil (IVCC) system, which has been designed for dedication of various active feedback plasma control functions, successfully fabricated and installed in the vacuum vessel of the Korea Superconducting Tokamak Advanced Research (KSTAR). The IVCC system consists of sixteen segmented coils that were independently fabricated outside the vacuum vessel and installed without any inside welding or brazing joints. The segmented coil system has several advantages such as eliminating possibility of cooling water leakage at the welded or brazed joints, simplification in fabrication and installation, and easy repair and maintenance of the coil system. Each segment contains eight oxygen-free high conductive coppers, which are grouped to four pairs, called as sections. Consequently, a segmented coil forms four sections for position control, field error correction (FEC), and resistive wall mode (RWM) control in accordance with electrical connection outside the cryostat. The eight conductors (or four sections) with internal coolant holes were enclosed in a rectangular welded jacket made of stainless steel 316LN and electrically insulated from the conductors by epoxy/glass composite layers. This coil system was commissioned up to 5 kA (30 kA-turns) for 5 s to achieve tentative use for the fast vertical plasma position control in the 2010 campaign of the KSTAR. This paper describes the several remarkable results in the fabrication and installation of the IVCC as well as commissioning results.     

Initial plasma start-up using partial solenoid coils in Versatile Experiment Spherical Torus (VEST)

Initial plasma start-up experiments based on ohmic discharge using partial solenoid coils located at both vertical ends of a center stack have been carried out in Versatile Experiment Spherical Torus (VEST) at Seoul National University. Ohmic discharges with the help of microwave pre-ionization have been performed according to the pre-programed start-up scenario which was experimentally verified by a series of vacuum field measurements using an internal magnetic probe array. A plasma current of around 0.4 kA has been achieved by ohmic discharge using partial solenoid coils, under the toroidal magnetic field of 0.1 T. The vacuum field calculation and fast camera image have revealed that the small plasma current even with significant amount of loop voltage up to 9.7 V is attributed to the imbalance of poloidal field for equilibrium. Modification of the start-up scenario and upgrade of power supplies are proposed to be carried out in order to achieve higher plasma current in the future experiments.     

Manufacturing of the Wendelstein 7-X divertor and wall protection

The in-vessel components of Wendelstein 7-X (W7-X) with a total surface of 265 m² comprise the divertor and the wall protection. The high heat flux (HHF) and lower heat flux (LHF) target, the baffle, the end plates closing the divertor chamber, a cryo vacuum pump (CVP) and a control coil form one divertor unit. Steel panels and the graphite heat shield protect the wall, including the ports. The HHF target elements, the steel panels and the control coils are manufactured by industry. The remaining components will be manufactured by the Max-Planck-Institute für Plasmaphysik (IPP) at its Garching workshops. For all components the final acceptance tests will be performed by IPP. This paper summarizes the main aspects for manufacturing, the preceding development and qualification tests as well as the final acceptance tests for the in-vessel components.     

Effect of an additional ferromagnetic material on the toroidal magnetic field ripple in the ITER

In the ITER tokamak, the toroidal magnetic field (TF) ripple is estimated with TF coils only, with the installation of ferromagnetic inserts (FIs), and with test blanket modules (TBMs) by using a 2-D code for easy and fast calculation. We assessed the effects of the thickness of the FIs on the TF ripple in order to optimize the FI. And we analyzed how the TBMs distort the TF, and calculated the TF ripple for various amounts of a ferromagnetic material and the positions of the TBMs. Even in the case of moving the TBMs outward up to 60-cm, and reducing the ferromagnetic material to 52%, the TF ripple is not decreased below 0.38%. So we had to adopt ripple correction coils. With a 52% reduced amount of the ferromagnetic material in a TBM, we could reduce the TF ripple to 0.28% at a coil current of 100 kA turn per each coil. And with an outward recess of the TBM up to 60 cm, we could reduce the TF ripple to 0.23% at a coil current of 250 kA turn per each coil. As a combined approach, if we reduce the amount of a ferromagnetic material in a TBM to 30%, and recess the TBM to 15 cm, we can efficiently obtain the TF ripple of 0.25% at a coil current of 150 kA turn per each coil.     

A Systematic study of modular coil characteristics for 2-field periods quasi-axisymmetric stellarator QAS-LA

Modular coil characteristics of a 2-field periods quasi-axisymmetric stellarator QAS-LA configuration with an aspect ratio A_p = 3, magnetic pressure ∼4% and rotational transform ι ∼ 0.15 per field period supplied by its own shaping have been detailed studied. In addition, the characteristics of modular coils for QAS-LA were compared with those of an intermediate QA configuration QAS-LAx and a tokamak based on the same center magnet field B₀, aspect ratio and number of coils. As expected, the B_max/B₀, force F and overturning moment M, increase with the increased complexity of the coil shape. The relationships between the modular coils’ parameters (such as radius curvature ρ, distance from coil to coil Δ_c–c and the cross-section of coils) and the electromagnetic characteristics have been systematically summarized. The approximate formula for the maximum magnetic field in the coil body as functions of modular coil parameters (Δ_c–c, ρ) was derived for a simple two wire system which will be useful when optimizations of coil properties are called for.     

A web application for poloidal field analysis on HL-2M

Recently, many web tools [1], [2], [3] in fusion society have been designed and demonstrated, which has been proved to be powerful and convenient to fusion researchers. Many physicists and engineers need a tool to compute the poloidal magnetic field for some purposes (for example, the calibration of magnetic probes for EFIT, the field null structure analysis for control, the design of some plasma diagnostic systems), so to develop a powerful and convenient web application for the calculation of magnetic field and magnetic flux produced by PF coils is very important. In this paper, a web application tool for poloidal field analysis on HL-2M with a totally original framework is presented. This web application is full of dynamic and interactive interface, and can run in any popular browser (IE, safari, firefox, opera), on any hardware (smart phone, PC, ipad, Mac) and operating system (ios, android, windows, linux, Mac OS). No any plugins is needed. The three layers (jQuery + PHP + Matlab) of this framework are introduced. The front top client layer is developed by jQuery code. The middle layer, which plays a role of a bridge to connect the server and client through socket communication, is developed by PHP code. The behind server layer is developed by Matlab, which compute the magnetic field or magnetic flux through a Special Function called Complete Elliptic Integral, and returns the results in the client favorite way, either by table or by JPG image. The field null structure and the vertical and radial field structure calculated by this tool are introduced with details. The idea to design a web tool with jQuery + PHP + Matlab framework may apply to other machines. The address for this application is http://dp.swip.ac.cn/hl2m.     

Gravitational flow of a thin film of liquid metal in a strong magnetic field

The influence of a poloidal magnetic field of the spherical Tokamak on super thin (h ≈ 0.1 mm) film flow of liquid metal driven by gravity over the surface of the cooled divertor plate is addressed. The experimental setup developed at the Institute of Physics, University of Latvia (IPUL) is described, which makes it possible to drive and visualize such liquid metal flows in the solenoid of the superconducting magnet “Magdalena”. As applied to the above setup, the magnetic field effect on the operation of the capillary system of liquid metal flow distribution (CSFD) is evaluated by using molten metal (lithium or eutectic InGaSn alloy) with a very small linear flowrate q ≤ 1 mm²/s, spread uniformly across the substrate. The magnetic field effect on the main parameters of the fully developed film flow is estimated for the above-mentioned liquid metals.An approximation technique has been proposed to calculate the development of the gravitational film flow. A non-linear differential second order equation has been derived, which describes the variation of the film flow thickness over the substrate length versus the flowrate q, magnetic field B and the substrate sloping α.Results of InGaSn film flow observations in a strong (B = 4 T) poloidal magnetic field are presented. Analysis of the video records evidences of experimental realization of a stable stationary film flow at width-uniform supply of InGaSn.     

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.     

AC loss measurements on Cable in Conduit Conductor with random-changing magnetic fields

Beside the generation of the poloidal component of the field, the main function of the poloidal field coils in a tokamak is the control of the shape and the position of the plasma, according to the chosen plasma scenario. A plasma scenario, namely a sequence of plasma shapes, is obtained and controlled by varying the current in the PF coils. The control currents create magnetic fields having complex trends, being almost random waveforms with frequencies in the range between 0.1 and 10 Hz. As a consequence, AC loss is generated in the superconducting coils exposed to those signals, and the feasibility of a plasma scenario is strictly related to the ability to withstand and remove the heat coming from the AC loss.In order to study what the behavior of the loss is in random magnetic fields, namely similar to the control fields, a SULTAN sample is tested under two kinds of random field signals. The first signal is obtained by summing several harmonic frequency components, in the range between 0.2 and 6 Hz, having random amplitude. The second waveform is generated by a random function generator and it has a much broader spectrum of frequencies. The tests are carried out by varying also the maximum amplitude of the signals. The results are here discussed and compared to the results of the single frequency AC loss tests, and a correlation between them is studied.     

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.     

Comparative assessment of application of low melting metals with capillary pore systems in a tokamak

Capillary-pore systems (CPS) with liquid metals are considered as advanced plasma facing material for application in DEMO-type fusion reactor. The estimation of opportunity of liquid Li, Ga and Sn application is carried out on the basis of its physical, chemical and technological properties, and with respect to prospective design of the tokamak in-vessel elements and technology.It has been shown that Li now is the most attractive and most investigated liquid metal for fusion devices application with CPS. The temperature limit for normal operation is about 550 °C and determined by appropriate Li flux to plasma due to evaporation. Wide range of structural materials is appropriate for Li based in-vessel elements.Ga and Sn are very corrosive and embrittlement inducing metals. As a result the temperature limit of these application is determined by compatibility with structural materials of CPS and in-vessel element. Only W can be used with Ga and Sn up to 500 °C. Moreover these metals have lower thermal properties comparing to Li.Surface temperature analysis for possible in-vessel element design (1 mm thick of porous W based CPS) has shown the similar power flux limit ∼21 MW/m² for Li, Ga and Sn application at normal operation. Taking into account the latent heat of vaporization and screening effect with re-radiation the CPS with Li has a priority at ELM and disruption conditions.     

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.     

Radiation shielding mass saving for the magnet coils of the VISTA spacecraft

Radiation shielding structure of a design concept with inertial fusion energy propulsion for manned or heavy cargo deep space missions beyond earth orbit has been investigated. Fusion power deposited in the inertial confined fuel pellet debris delivers the rocket propulsion with the help of a magnetic nozzle. The nuclear heating in the super conducting magnet coils determines the radiation shielding mass of the spacecraft. It was possible to achieve considerable mass saving with respect to a recent design work, coupled with higher design limits for coil heating (up to 5 mW/cm³). The neutron and γ-ray penetration into the coils is calculated using the S_N methods with a high angular resolution in (r–z) geometry in S₁₆P₃ approximation by dividing the solid space angle in 160 sectors. Total peak nuclear heat generation density in the coils is calculated as 3.143 mW/cm³ by a fusion power of 17 500 MW. Peak neutron heating density is 1.469 mW/cm³ and peak γ-ray heating density is 1.674 mW/cm³. However, volume averaged heat generation in the coils is much lower, namely 74, 163 and 337 μW/cm³ for neutron, γ-ray and total nuclear heating, respectively. The net mass of the radiation shielding for the magnet coils is 200 tonne by a total mass of 6000 tonne of the space craft.     

Power dissipation and energy transfer during testing of the ITER toroidal field model coil

The testing of the ITER toroidal field model coil (TFMC) in the background field of the EURATOM-LCT coil took place in autumn 2002 at the TOSKA facility of the Forschungszentrum Karlsruhe in the framework of the ITER R&D programme. The maximum currents in the two coils, in combined operation, were 16 kA in the LCT coil and 80 kA in the TFMC, respectively. The heat load of both coils, including the eddy current losses in the passive structures and the joule losses due to the joint resistances, was removed by a secondary loop of forced flow supercritical He. About 2% of the stored energy was transferred to the cryogenic system after all the safety discharges of both coils together. Most of the energy (about 98%) was extracted and transferred to the dump resistors of both coils, located outside the vacuum vessel. A computer code, based on the full inductance and resistance matrices, has been developed with SIMULINK™. After validation with experimental data the code has been used to perform circuit analysis and to evaluate the power dissipation and energy transferred to the cryogenic plant and to the external power circuits.     

Concept of curved magnetically guided liquid lithium target without a back plate for IFMIF

Curved magnetically guided lithium target (MGLT) without a back plate was newly proposed in light of simplified structure, easy maintenance and enhanced availability and performance for international fusion materials irradiation facility (IFMIF). It can replace conventional lithium target with a curved material back plate under the most severe condition on neutron irradiation. Magnetic field suited for the curved MGLT is produced in combination of a couple of radiation-proof resistive coils and reduced activation ferritic/martensitic steel (F82H) parts (yokes, ducts/nozzles and high flux test module (HFTM)). Shape of the magnetic field becomes curved automatically in the target region by setting HFTM closely to MGLT. Characteristics of the lithium flow on MGLT was analyzed in detail by two dimensional equations of motion with the magnetic field calculated by the Poisson Superfish code. The necessary magnetic flux density at the target region was found to be about 0.5 T to fulfill the IFMIF target conditions, i.e., lithium flow speed of 15 m/s, curvature radius of 1–1.6 m and flow thickness of 0.025 m. A narrow gap (a few mm) between MGLT and HFTM could be controlled by adjusting the coil current. Future subjects for further development of this concept were identified.     

The procurement and testing of the stainless steel in-vessel panels of the Wendelstein 7-X Stellarator

320 In-vessel water cooled stainless steel panels, poloidal closure plates and pumping gap panels, covering an area of approximately 100 m², are used in Wendelstein7-X to protect the plasma vessel. The panels are manufactured at Deggendorf, Germany by MAN Diesel & Turbo SE. The panels consist of a laser welded sandwich of stainless steel plates together with a labyrinth of cooling channels and have a complicated geometry to fit the plasma vessel of Wendelstein 7-X. The hydraulic and mechanical stability requirements whilst maintaining the tight tolerances for the shape of the components are very demanding. The panels are designed to operate at up to an average heat load of 100 kW/m² and a maximum heat load of 200 kW/m² with a water velocity of approximately 2 m s^?1. High heat flux testing of an un-cooled panel at a time averaged load of 200 kW/m² for 10 s were successfully performed to support the start up phase of Wendelstein 7-X operation. Extensive testing both during manufacture and after delivery to IPP-Garching demonstrates the suitability of the delivered panels for their purpose.     

Influence of construction errors on Wendelstein 7-X magnetic configurations

Wendelstein 7-X, currently under construction at the Max-Planck-Institut für Plasmaphysik in Greifswald, Germany, is a modular advanced stellarator, combining the modular coil concept with optimised properties of the plasma. The magnet system of the machine consists of 50 non-planar and 20 planar superconducting coils which are arranged in five identical modules, forming a toroidal five-fold symmetric system. The majority of operational magnetic configurations will have rotational transform ι/2π = 1 at the boundary. Such configurations are very sensitive to symmetry breaking perturbations, which are the consequence of imprecisely manufactured coils or assembly errors. To date, all 70 coils have been fabricated, and the first two half-modules of the machine have been assembled. The comparative analysis of manufactured winding packs and estimates of the corresponding level of magnetic field perturbation are presented. The dependency of the error fields on the coil assembly sequence is considered, as well as the impact of the first assembly errors. The influence of different construction uncertainties is discussed, and measures to minimise the magnetic field perturbation are suggested.     

Design of JET ELM control coils for operation at 350 °C

A study has confirmed the feasibility of designing, fabricating and installing resonant magnetic field perturbation (RMP) coils in JET¹ with the objective of controlling edge localized modes (ELM). A system of two rows of in-vessel coils, above the machine midplane, has been chosen as it not only can investigate the physics of and achieve the empirical criteria for ELM suppression, but also permits variation of the spectra allowing for comparison with other experiments. These coils present several engineering challenges. Conditions in JET necessitate the installation of these coils via remote handling, which will impose weight, dimensional and logistical limitations. And while the encased coils are designed to be conventionally wound and bonded, they will not have the usual benefit of active cooling. Accordingly, coil temperatures are expected to reach 350 °C during bakeout as well as during plasma operations. These elevated temperatures are beyond the safe operating limits of conventional OFHC copper and the epoxies that bond and insulate the turns of typical coils. This has necessitated the use of an alternative copper alloy conductor C18150 (CuCrZr). More importantly, an alternative to epoxy had to be found. An R&D program was initiated to find the best available insulating and bonding material. The search included polyimides and ceramic polymers. The scope and status of this R&D program, as well as the critical engineering issues encountered to date are reviewed and discussed.