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.     

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.     

Revisiting the poloidal magnetic confinement

This note proposes a closed poloidal magnetic configuration with an in-vessel coil system held by shielded supports. A dipole field is bounded by external coils and constrained into a hollow torus aiming at uniform intensity. In the horizontal mid-plane region the external coils and the dipole outer coils are broken in four arcs and bridged by couple of straight branches. Arcs and straight branches build a set of four side coils. In the clearance between their straight branches four tunnels in the poloidal magnetic field are achieved, to pass the supports and the feeders of the in-vessel coil system.A poloidal machine with a plasma thick as those of present large experiments is outlined. The dipole radius is 5.4 m, the plasma about it has a constant poloidal cross-section about 40 m², a volume about 1300 m³ and a minimum thickness 1 m in the outboard. The magnetic field ranges from 1.4 to 1.8 T.     

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.     

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.     

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.     

JT-60SA Toroidal Field Coils test cryostat development

Within the Broader Approach Agreement, Fusion for Energy will deliver to the Japanese Atomic Energy Association, amongst other components, the 18 Toroidal Field Coils (TFCs) for the superconducting Tokamak JT-60SA [1]. These coils will be individually tested at cryogenic temperatures and at the nominal current in a test cryostat. This cryostat is provided as an in-kind contribution by Belgium and is being developed jointly with CEA-Saclay/France.The vessel is large, oval shaped with an overall length of 11 m, a width of 7.2 m and a height of 6.5 m. To reduce the heat load to the coils the cryostat is covered by LN₂ cooled thermal shields. In addition to the cryostat, three test frames for the coils, the valve box vessel and the insulation vacuum system are also provided by Belgium. The Belgian contribution is design, manufacturing, assembly and test of the vacuum chamber, thermal shield and test frames by the Belgian company Ateliers de la Meuse (ALM), with the support of Centre Spatial de Liège (CSL). The TF coil test facility is assembled and the coil tests are performed by CEA/Saclay.The Belgian contribution, namely the design, manufacturing, assembly and test of the vacuum vessel, the thermal shields, and the test frames as well as of the vacuum pumping system are described in the presentation.     

Configuration space control for Wendelstein 7-X

The Wendelstein 7-X stellarator (W7-X) is a superconducting fusion experiment, presently under construction at the Greifswald branch of the Max-Planck-Institut für Plasmaphysik. W7-X is a device with high geometrical complexity due to the close packing of the components in the cryostat and their complex 3D shape e.g. of the superconducting coils. The tasks of configuration space control are to ensure that all these components do not collide with each other under a set of defined configurations, i.e. at the time of assembly, at 4 K or for various coil currents. To fulfill these tasks sophisticated tools and procedures were developed and implemented within the realm of a newly founded division that focuses on design, configuration control and configuration management.     

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.     

Analysis of edge magnetic field line structure in ITER due to in-vessel ELM control coils

In this work we evaluated the ITER ELM coils design based on two metrics: the Chirikov vacuum magnetic island overlap parameter, and the vacuum Field Line Loss Fraction. The study was performed for a range of current amplitudes for three different n = 4 waveforms: square, cosine and sine. The results indicated that ITER ELM coils are designed with a high level of flexibility to accommodate different operation scenarios (H-mode and Steady State) with different values of q₉₅ and q-profiles. The magnetic island overlap analysis showed that ITER ELM coils are capable of matching the DIII-D I-coil spectrum. The Field Line Loss analysis showed that edge vacuum stochastization might be achieved that is similar or greater than in DIII-D. Fault analysis of the coils indicated that ITER ELM coils are robust and show good characteristics even with 11% of dead coils.     

Design and test program of a simplified divertor dummy coil structure for the WEST project

In order to fully validate actively cooled tungsten plasma facing components (industrial fabrication, operation with long plasma duration), the implementation of a tungsten axisymmetric divertor structure in the tokamak Tore-Supra is studied. With this major upgrade, so-called WEST (Tungsten Environment in Steady state Tokamak), Tore-Supra will be able to address the problematic of long plasma discharges with a metallic divertor target.To do so, it is planned to install two symmetric divertor coils inside the vacuum vessel. This assembly, called divertor structure, is made up of two stainless steel casings containing a copper winding pack cooled by a pressurized hot water circuit (up to 180 °C, 4 MPa) and is designed to perform steady state plasma operation (up to 1000 s).The divertor structure will be a complex assembly ring of 4 m diameter representing a total weight of around 20 tons. The technical challenge of this component will be the implementation of angular sectors inside the vacuum vessel environment (TIG welding of the coil casing, induction brazing and electrical insulation of the copper winding). Moreover, this complex assembly must sustain harsh environmental conditions in terms of ultra high vacuum conditions, electromagnetical loads and electrical isolation (13 kV ground voltage) under high temperature.In order to fully validate the assembly and the performance of this complex component, the production of a scale one dummy coil is in progress.The paper will illustrate, the technical developments performed in order to finalize the design for the call for tender for fabrication. The progress and the first results of the simplified dummy coils will be also addressed.     

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.     

Completion of designing and manufacturing of the coil support structure of W7-X

In February 2000, the project called coil support structure for the Wendelstein 7-X fusion machine was started. Since October 2009 the full production of this big (80 tons) and complex component is now completed and delivered at IPP Greifswald. The W7-X coil system consists of 20 planar and 50 non-planar coils. They are supported by a pentagonal 10 m diameter, 2.5 m high called coil support structure (CSS). The CSS is divided into five modules and each module consists of two equal half modules around the radial axis. Currently, the five modules were successfully assembled with the coils meeting the tight manufacturing tolerances. Designing, structural calculation, raw material procurement, welding & soldering technologies, milling, drilling, accurate machining, helium cooling pipe forming, laser metrology, ultra sonic cleaning and vacuum test are some of the key points used all along this successful manufacturing process. The lessons learned in the large scale production of this difficult kind of support structure will be presented as relevant experience for the realization of similar systems for future fusion devices, such as ITER.     

Simulation of runaway electron generation and diffusion during major disruptions in the HL-2A tokamak

The generation and diffusion of runaway electrons (REs) during major disruptions in the HL-2A tokamak has been studied numerically. The diffusion caused by the magnetic perturbation is especially addressed. The simulation results show that the strong magnetic perturbation (δB/B ∼ 1.0 × 10⁻³) can cause a significant loss of REs due to the radial diffusion and restrain the RE avalanche effectively. The results also indicate that the REs are generated initially in the plasma core during disruptions, and that the toroidal electric field does not exhibit a centrally hollow phenomenon. In addition, it is found that the toroidal effects have little impact on the generation of RE and the evolution of toroidal electric field.     

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.     

Overview of magnetic control in ITER

ITER is targeting Q = 10 with 500 MW of fusion power. To meet this target, the plasma needs to be controlled and shaped for a period of hundreds of seconds, avoiding contact with internal components, and acting against instabilities that could result in the loss of control of the plasma and in its disruptive termination.Axisymmetric magnetic control is a well-understood area being the basic control for any tokamak device. ITER adds more stringent constraints to the control primarily due to machine protection and engineering limits. The limits on the actuators by means of the maximum current and voltage at the coils and the few hundred ms time response of the vacuum vessel requires optimization of the control strategies and the validation of the capabilities of the machine in controlling the designed scenarios.Scenarios have been optimized with realistic control strategies able to guarantee robust control against plasma behavior and engineering limits due to recent changes in the ITER design. Technological issues such as performance changes associated with the optimization of the final design of the central solenoid, control of fast transitions like H to L mode to avoid plasma-wall contact, and optimization of the plasma ramp-down have been modeled to demonstrate the successful operability of ITER and compatibility with the latest refinements in the magnetic system design.Validation and optimization of the scenarios refining the operational space available for ITER and associated control strategies will be proposed. The present capabilities of magnetic control will be assessed and the remaining critical aspects that still need to be refined will be presented. The paper will also demonstrate the capabilities of the diagnostic system for magnetic control as a basic element for control. In fact, the noisy environment (affecting primarily vertical stability), the non-axisymmetric elements in the machine structure (affecting the accuracy of the identification of the plasma boundary), and the strong component of eddy current at the start-up (resulting in a poor S/N ratio for plasma reconstruction for I_p < 2 MA requiring a robust plasma control) make the ITER magnetic diagnostic system a demanding part of the magnetic control and investment protection systems. Finally the paper will illustrate the identified roles of magnetic control in the PCS (plasma control system) as formally defined in the recent first step of the design and development of the system.     

Design of the dummy coil for magnet power supply

It is necessary to test it on a dummy coil, before using a magnet power supply (MPS) to energize a Poloidal Field (PF) coil in the Korea Superconducting Tokamak Advanced Research (KSTAR) device. The dummy coil should accept the same large current from the MPS as the PF coil and be within the capability of the utilities located at the KSTAR site. Therefore a coil design based on the characteristics of the MPS and other restrictive conditions needed to be made. There are three requirements to be met in the design: an electrical requirement, a structural requirement, and a water cooling requirement. The electrical requirement was that the coil should have an inductance of 40 mH. For the structural requirement, the material should be non magnetic. The coil support structure and water cooling manifold were made of SUS 304. The water cooling requirement was that there should be sufficient flow rate so that the temperature rise ΔT should not exceed 12 °C for operation at 12.5 kA for 5 min. Square cross-section hollow conductor with dimensions of 38.1 mm × 38.1 mm was used with a 25.4 mm center hole for cooling water. However, as a result of tests, it was found that the electrical and structural requirements were satisfied but that the water cooling was over designed. It is imperative that the verification will be redone for a test with 12.5 kA for 5 min.     

Fabrication and installation of equilibrium field coils for the JT-60SA

Recently, fabrication of the first superconducting coil in JT-60SA tokamak (EF4 coil) was finished. EF4 coil has ten double pancake (DP) coils. All DP coils were stacked up to form the winding pack. In order to check the manufacturing error of DP coils, their circularities (in-plane ellipticities) were evaluated for all DP coils. Positions of conductors for each DP coil were measured before curing process. Error bars of the current centers, which were used for the index for DP coil's circularity, ranged between 1.1 and 2.5 mm. During stacking the DP coils, the positions of these coils were optimized in order to cancel the error of circulation of the winding pack. As the result, the manufacturing error of the radial current center was achieved 0.6 mm for the winding pack. This value was an order of magnitude smaller than the required manufacturing error of EF4 coil.     

Updating the design of the feeder components for the ITER magnet system

The ITER superconducting magnet system generates an average heat load of 23 kW at 4 K to the cryoplant, from nuclear and thermal radiation, conduction and electromagnetic heating, and requires current supplies 10–68 kA to 48 individual coils. The helium flow to remove this heat, consisting of supercritical helium at pressures up to 1.0 MPa and temperature between 4.3 and 4.7 K, is distributed to the coils and structures through 30 separate feeder lines. The feeders also contain the electrical supplies to the coil, helium supply pipes and the instrumentation lines, and are integrated with the current lead transitions to room temperature. The components consist of the in-cryostat feeders, the cryostat feedthroughs and the coil terminal boxes (CTBs). This paper discusses the functional requirements on the feeder system and presents the latest design concept and parameters of the feeder components.     

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.