In-vessel Coils Research Articles

The DC-link of the inverter system BUSSARD for ASDEX Upgrade in vessel saddle coils

For the nuclear fusion experiment ASDEX Upgrade (AUG) an inverter system to individually feed the 16 in-vessel saddle coils was built. After an operating phase with four inverters that supplied groups of up to 4 coils connected in series, now the full system is working. The common DC-link is fed by an existing current-mode thyristor rectifier.The idea was to use the thyristor rectifier in voltage control mode. Therefore the control system and the power stack had to be adjusted. This publication describes the modifications made to the thyristor rectifier in order to feed a DC-link and stabilize its voltage under all load conditions. Also the integrated safety systems are shown.

An optimized upper divertor with divertor-coils to study enhanced divertor configurations in ASDEX Upgrade

A new upper divertor with in-vessel coils to investigate ‘advanced’ magnetic configurations will be installed in ASDEX Upgrade. ‘Advanced’ magnetic configurations are part of a risk mitigation strategy should the conventional divertor configuration approach not be applicable for future fusion reactors. The concept of the new upper divertor, Div-IIo, consists of an outer divertor with embedded divertor coils, an inner divertor and a cryo pump with 15m3/s effective pumping speed in the private flux region. The new divertor coils will be designed for 50kAt and can be operated with 5s plateau phase. This paper presents the design concept of the divertor and coil support structure, the coils and the cryo pump. In addition the forces to the divertor structure due to normal operation and during disruptions are discussed.

Merging compression start-up predictions for ST40

Merging compression is an efficient and robust method for plasma start-up which can achieve both high plasma current and high temperatures at fusion relevant densities. It involves the formation of plasma around two in-vessel poloidal field coils followed by a magnetic reconnection event after which the plasma can be compressed. In this paper we extrapolate from START/MAST experimental data to predict the performance of ST40 (currently under construction). In our analysis we found that induced eddy currents within passive elements greatly affected the scenario.

DTT device: Conceptual design of the superconducting magnet system

In the European Fusion Roadmap, one of the main challenges to be faced is the risk mitigation related to the impossibility of directly extrapolate to DEMO the divertor solution adopted in ITER, due to the very large loads expected. Thus, a satellite experimental facility oriented toward the exploration of robust divertor solutions for power and particles exhaust and to the study of plasma-material interaction scaled to long pulse operation, is currently being designed. Clearly, design requirements for this experiment are quite challenging, to account for the extreme operation conditions, which shall be as representative as possible of the DEMO ones, but in a much smaller device and at lower costs. A feasibility assessment has been carried out for the fully superconducting magnet system of the compact Divertor Tokamak Test (DTT) facility project. The overall magnet system is based on NbTi and Nb3Sn Cable-in-Conduit Conductors, and it adopts some of the most recent developments in this field. It consists of 18 Toroidal Field (TF), 6 Poloidal Field (PF) and 6 Central Solenoid (CS) module coils. In order to cope with the machine requirements such as plasma major and minor radii, magnetic field on plasma axis, plasma current, and inductive flux requirement, the Nb3Sn TF coil is characterized by a peak field of 11.4T on the conductor, operating at 46.3kA; the Nb3Sn CS modules are characterized by a peak field of about 13T, with a conductor operating current of 23kA; the PF coils are wound using NbTi conductors operating at a maximum peak field of 4.0T, with operating currents in the range 21kA to 25kA, depending on the PF coil. Profiting of the compact machine size, and thus of relatively short conductor lengths, the TF coil winding pack is conceived as layer wound and made of two distinct sections, a low- and a high-field one, employing different superconductor cross-sections, and electrically connected through an embedded “ENEA-type” joint.The main features of the magnet system are described here; the results of mechanical, electrical and thermo-hydraulic analyses, which are discussed here, indicate that the proposed design fulfills all the required criteria. In addition, a brief description of the In-Vessel coils is given, though they are not superconducting, for the sake of completeness.

Nonlinear burn control using in-vessel coils and isotopic fueling in ITER

In future burning tokamaks, active control of the plasma temperature and density will be necessary to produce a determined amount of fusion power. Conventional actuation methods, such as auxiliary power modulation and fueling rate modulation, are usually employed to tackle this so-called burn control problem. Fueling rate modulation can be used not only to directly control the plasma density but also to indirectly control the plasma energy by means of isotopic fuel tailoring. Moreover, based on recent experiments, the in-vessel coil system arises as a possible actuation method to decrease the plasma energy by generating non-axisymmetric fields in the plasma that may reduce the energy confinement time. In this work, a nonlinear burn controller that integrates the three mentioned actuators is proposed. The controller performance is tested via simulations for an ITER-like scenario.

Proposal of an alternative upper divertor in ASDEX Upgrade supported by EMC3-EIRENE simulations

We discuss the benefits of installing a pair of in-vessel coils with currents |Ifx| ≲ 50 kAt in the upper divertor of ASDEX Upgrade (AUG) to study a series of ‘alternative’ divertor configurations, like the Snowflake (SF) and the X-divertor (XD), that are currently considered as alternative solutions for the power exhaust problem. The possibility of operating the standard lower single-null (SN) and double-null (DN) would be preserved. Potential effects to reduce the peak parallel- and/or perpendicular heat flux are predicted from a simple geometrical-diffusive model as well as by numerical EMC3-EIRENE simulations for pure deuterium attached conditions with spatially constant diffusion coefficients. Beyond that a series of other potential transport- and radiation related heat flux mitigation effects are identified and could be studied experimentally with the modified upper divertor in the high-power divertor Tokamak AUG.

Heat dissipation research on the water-cooling channel of HL-2M in-vessel coils

HL-2M in-vessel coils are positioned in high vacuum circumstance, and they will generate joule heat when they carry 15kA electrical current, but joule heat is very difficult to dissipate in vacuum, so a hollow cable with 8mm inner diameter is design as water-cooling channel for heat convection. By using the methods of the theoretical derivation, together with CFD numeric simulation method and the experiment of the heat transfer, the water channel of HL-2M in-vessel coils has been studied, and the temperature of HL-2M in-vessel coils under different cooling water flow rates is obtained and acceptable. Simultaneously, the external cooling water supply system parameters for the water-cooling channel of the coils are estimated. Three methods’ results are in good agreement; the theoretical model is verified and could be popularized for predicting the temperature rise of HL-2M in-vessel coils.

Engineering design and integration of in-vessel single turn segmental coil in vacuum vessel of SST-1

SST-1 tokamak is having the error field due to unsymmetrical positioning of Toroidal field coils which push the plasma to inner side from its major radius of 1100 mm. hence it is required to install the In-vessel Coil (PF6) at a location of 1350 mm radius and elevation of 350 mm above and below the mid plane of the toroidal field coils. The In-Vessel coil was decided to make in eight segments for futuristic use, to control the individual localized error field correction by supplying the different current. A single turn, eight segments, copper conductor with 18 mm diameter with GFRP insulation and in housed in SS304 L casing to carry 8000 A current for 10 s was designed and installed in vacuum vessel of SST-1. This paper will present the design drivers, material selection, advantages and constraints of the in-vessel coils, its conceptual and engineering design, CAD models, finite element analysis using ANSYS, its fabrication, quality assurance/control and assembly/integration aspects inside vacuum vessel of SST-1.

The DTT device: Poloidal field coil assessment for alternative plasma configurations

This paper firstly illustrates the objectives, the figure of merits and the specifications considered in the design of the equilibrium configurations in the Divertor Tokamak Test (DTT) facility. The reference single null scenario is detailed. The range of alternative plasma shapes and current capabilities are then discussed. A number of in-vessel coils can be used to locally modify the magnetic configuration in the divertor region. Finally, a comparison of costs and benefits of the various configurations is given, with particular reference to the power exhaust issues.

A current-pulsed power supply with rapid rising and falling edges for magnetic perturbation coils on the J-TEXT tokamak

This study presents the design and principle of a current-pulsed power supply (CPPS) for the tearing mode (TM) feedback control of the J-TEXT tokamak. CPPS is a new method of stabilizing large magnetic islands and accelerating mode rotation through the use of modulated magnetic perturbation. In this application, continuous magnetic perturbation pulse trains with frequency of 1kHz to kHz, amplitude of 0.25G, and duty ratio of 20%–50% are required generating via in-vessel magnetic coils. A modified topology based on buck chopper is raised to satisfy the demands of inductive load. This modified topology is characterized by high frequency, rapid rising and falling edges, and large amplitude of current pulses. Appropriate RCD snubber circuit is applied to protect the Insulated Gate Bipolar Transistor (IGBT) switch device. Equipment with peak current that reaches 1kA, frequency that ranges from 1kHz to 3kHz, and rising and falling time within 100μs was constructed and applied to physical experiment.

Investigation of instabilities and rotation alteration in high beta KSTAR plasmas

H-mode plasma operation of the Korea Superconducting Tokamak Advanced Research (KSTAR) device has been expanded to significantly surpass the ideal MHD no-wall beta limit. Plasmas with high normalized beta, βN, up to 4.3 have been achieved with reduced plasma internal inductance, li, to near 0.7, exceeding the computed n = 1 ideal no-wall limit by a factor of 1.6. Pulse lengths at maximum βN were extended to longer pulses by new, more rapid control. The stability of the observed m/n = 2/1 tearing mode that limited the achieved high βN is computed by the M3D-C1 code, and the effect of sheared toroidal rotation to tearing stability is examined. As a method to affect the mode stability in high βN plasmas, the non-resonant alteration of the rotation profile by non-axisymmetric magnetic fields has been used, enabling a study of the underlying neoclassical toroidal viscosity (NTV) physics and stability dependence on rotation. Non-axisymmetric field spectra were applied using in-vessel control coils (IVCCs) with varied n = 2 field configurations to alter the plasma toroidal rotation profile in high beta H-mode plasmas and to analyze their effects on the rotation. The rotation profile was significantly altered with rotation reduced by more than 60% without tearing activity or mode locking. To investigate the physical characteristics and scaling of the measured rotation braking by NTV, changes in the rotation profile are analytically examined in steady state. The expected NTV scaling with the square of the normalized applied field perturbation agrees with the measured profile change δB2.1–2.3. The NTV is also found to scale as Ti2.1–2.4, in general agreement with the low collisionality “1/ν” regime scaling of the NTV theory (TNTV-(1/ν) ∝ Ti2.5).

Suppression of edge localized mode crashes by multi-spectral non-axisymmetric fields in KSTAR

Among various edge localized mode (ELM) crash control methods, only non-axisymmetric magnetic perturbations (NAMPs) yield complete suppression of ELM crashes beyond their mitigation, and thus attract more attention than others. No other devices except KSTAR, DIII-D, and recently EAST have successfully achieved complete suppression with NAMPs. The underlying physics mechanisms of these successful ELM crash suppressions in a non-axisymmetric field environment, however, still remain uncertain. In this work, we investigate the ELM crash suppression characteristics of the KSTAR ELMy H-mode discharges in a controlled multi-spectral field environment, created by both middle reference and top/bottom proxy in-vessel control coils. Interestingly, the attempts have produced a set of contradictory findings, one expected (ELM crash suppression enhancement with the addition of n = 1 to the n = 2 field at relatively low heating discharges) and another unexpected (ELM crash suppression degradation at relatively high heating discharges) from the earlier findings in DIII-D. This contradiction indicates the dependence of the ELM crash suppression characteristics on the heating level and the associated kink-like plasma responses. Preliminary linear resistive MHD plasma response simulation shows the unexpected suppression performance degradation to be likely caused by the dominance of kink-like plasma responses over the island gap-filling effects.

Impact of resistive MHD plasma response on perturbation field sidebands

Single fluid linear simulations of a KSTAR RMP ELM suppressed discharge with the M3D-C1 resistive magnetohydrodynamic code have been performed for the first time. The simulations show that the application of the n = 1 perturbation using the KSTAR in-vessel control coils (IVCC), which apply modest levels of n = 3 sidebands (~20% of the n = 1), leads to levels of n = 3 sideband that are comparable to the n = 1 when plasma response is included. This is due to the reduced level of screening of the rational-surface-resonant n = 3 component relative to the rational-surface-resonant n = 1 component. The n = 3 sidebands could play a similar role in ELM suppression on KSTAR as the toroidal sidebands (n = 1, 2, 4) in DIII-D n = 3 ELM suppression with missing I-coil segments (Paz Soldan et al 2014 Nucl. Fusion 54 073013). This result may help to explain the uniqueness of ELM suppression with n = 1 perturbations in KSTAR since the effective perturbation is a mixed n = 1/n = 3 perturbation similar to n = 3 ELM suppression in DIII-D.

Superconducting Magnet Systems in Tokamak Engineering and Optimization

The aim of this paper is to study the interplay between plasma shaping and magnet systems in superconducting tokamaks. In particular, we study the optimization of the layout of the superconducting toroidal field (TF) coil system as a function of plasma elongation (κ xp ) and aspect ratio (A). By excluding the use of in-vessel control coils, we proceed by maximizing plasma elongation as a function of the aspect ratio to keep plasma operation safe with respect to vertical displacement events. In this respect, design criteria coming from tokamak operational experience as well as other designs are used as guidance. The layout of the superconducting magnet systems is then optimized to meet the key requirements of reference TF on the plasma axis and maximum TF ripple at the plasma boundary, volt-seconds availability, and plasma shaping while satisfying the constraints of conductor temperature margins, peak stresses in the TF coil central vault, and hoop stress in the central solenoid.

Experimental and numerical study of error fields in the CNT stellarator

Sources of error fields were indirectly inferred in a stellarator by reconciling computed and numerical flux surfaces. Sources considered so far include the displacements and tilts of the four circular coils featured in the simple CNT stellarator. The flux surfaces were measured by means of an electron beam and fluorescent rod, and were computed by means of a Biot–Savart field-line tracing code. If the ideal coil locations and orientations are used in the computation, agreement with measurements is poor. Discrepancies are ascribed to errors in the positioning and orientation of the in-vessel interlocked coils. To that end, an iterative numerical method was developed. A Newton–Raphson algorithm searches for the coils’ displacements and tilts that minimize the discrepancy between the measured and computed flux surfaces. This method was verified by misplacing and tilting the coils in a numerical model of CNT, calculating the flux surfaces that they generated, and testing the algorithm’s ability to deduce the coils’ displacements and tilts. Subsequently, the numerical method was applied to the experimental data, arriving at a set of coil displacements whose resulting field errors exhibited significantly improved agreement with the experimental results.

Radial Compression to Transiently Achieve Higher Beta Poloidal in a Tokamak

An increase in the magnetic strength of the vertical field will, in general, compress and heat plasmas, which thus allows high temperature, high density, and high beta to be reached. In this paper, according to the scaling laws for the magnetic compression, the scaling range of both the major-radius compression and the minor-radius compression is figured out. Then, the comparison between the major-radius compression and the minor-radius compression is carried out. It demonstrates that the minor-radius compression can obtain higher ${\beta }_{p}$ of plasma. Consequently, we try to significantly enhance discharge parameters by the minor-radius compression by inserting the sets of the vertical-field coils in the experimental advanced superconducting tokamak (EAST) vacuum chamber. Afterward, the minor-radius compression based on EAST shot #34128 is deduced with ${C}_{a}=1.5$ and ${C}_{R}=0.925$ , which increases ${\beta }_{p}$ by 76.19%. It notes that the minor-radius compression is a promising way to effectively improve ${\beta }_{p}$ in a tokamak. Finally, the preliminary design of the minor-radius compression, which includes the control system, the in-vessel vertical-field coils, and its fast-control power supply, is described.

Equilibrium drives of the low and high field side n = 2 plasma response and impact on global confinement

The nature of the multi-modal n = 2 plasma response and its impact on global confinement is studied as a function of the axisymmetric equilibrium pressure, edge safety factor, collisionality, and L-versus H-mode conditions. Varying the relative phase () between upper and lower in-vessel coils demonstrates that different n = 2 poloidal spectra preferentially excite different plasma responses. These different plasma response modes are preferentially detected on the tokamak high-field side (HFS) or low-field side (LFS) midplanes, have different radial extents, couple differently to the resonant surfaces, and have variable impacts on edge stability and global confinement. In all equilibrium conditions studied, the observed confinement degradation shares the same dependence as the coupling to the resonant surfaces given by both ideal (IPEC) and resistive (MARS-F) MHD computation. Varying the edge safety factor shifts the equilibrium field-line pitch and thus the dependence of both the global confinement and the n = 2 magnetic response. As edge safety factor is varied, modeling finds that the HFS response (but not the LFS response), the resonant surface coupling, and the edge displacements near the X-point all share the same dependence. The LFS response magnitude is strongly sensitive to the core pressure and is insensitive to the collisionality and edge safety factor. This indicates that the LFS measurements are primarily sensitive to a pressure-driven kink-ballooning mode that couples to the core plasma. MHD modeling accurately reproduces these (and indeed all) LFS experimental trends and supports this interpretation. In contrast to the LFS, the HFS magnetic response and correlated global confinement impact is unchanged with plasma pressure, but is strongly reduced in high collisionality conditions in both H- and L-mode. This experimentally suggests the bootstrap current drives the HFS response through the kink-peeling mode drive, though surprisingly weak or no dependence on the bootstrap current is seen in modeling. Instead, modeling is revealed to be very sensitive to the details of the edge current profile and equilibrium truncation. Holding truncation fixed, most HFS experimental trends are not captured, thus demonstrating a stark contrast between the robustness of the HFS experimental results and the sensitivity of its computation.

An Optimal Real-Time Controller for Vertical Plasma Stabilization

Modern Tokamaks have evolved from the initial axisymmetric circular plasma shape to an elongated axisymmetric plasma shape that improves the energy confinement time and the triple product, which is a generally used figure of merit for the conditions needed for fusion reactor performance. However, the elongated plasma cross section introduces a vertical instability that demands a real-time feedback control loop to stabilize the plasma vertical position and velocity. At the Tokamak \`a Configuration Variable (TCV) in-vessel poloidal field coils driven by fast switching power supplies are used to stabilize highly elongated plasmas. TCV plasma experiments have used a PID algorithm based controller to correct the plasma vertical position. In late 2013 experiments a new optimal real-time controller was tested improving the stability of the plasma. This contribution describes the new optimal real-time controller developed. The choice of the model that describes the plasma response to the actuators is discussed. The high order model that is initially implemented demands the application of a mathematical order reduction and the validation of the new reduced model. The lower order model is used to derive the time optimal control law. A new method for the construction of the switching curves of a bang-bang controller is presented that is based on the state-space trajectories that optimize the time to target of the system. A closed loop controller simulation tool was developed to test different possible algorithms and the results were used to improve the controller parameters. The final control algorithm and its implementation are described and preliminary experimental results are discussed.

Electrical design of the BUSSARD inverter system for ASDEX upgrade saddle coils

A set of 16 in-vessel saddle coils is installed in the ASDEX Upgrade (AUG) nuclear fusion experiment for mitigation of edge localized modes (ELM) and feedback control of resistive wall modes (RWM). The coils were driven by DC current only during previous campaigns. Now, a new inverter system “BUSSARD” (German abbr. for “Bayerischer Umrichter, schnell schaltend für AUGs rasche Drehfelder”, translated: “bavarian fast switching inverter for AUG's fast rotating fields”) is built for the experiment. A four-phase system has been assembled to simultaneously operate up to 4 groups of coils consisting of up to 4 serial-connected coils each. The maximum current is 1.3kA with a ripple in the range of 7% and the frequency is variable between DC and approx. 100Hz. The switching frequency is variable between approximately 3 and 10kHz. As a first application, rotating fields are generated. The system can be enhanced in two stages to 16-phase operation with a bandwidth of 500Hz and a 24 phase system with a bandwidth of up to 3kHz.

In-vessel coils for magnetic error field correction in JT-60SA

We have designed in-vessel coils for a correction of magnetic error fields in JT-60SA. In order to design the in-vessel coils, namely, error field correction coils (EFCCs), error fields from several sources such as manufacturing and assembly errors of superconducting coils are calculated by Monte-Carlo approach. Required EFCC currents to correct error fields are evaluated by a least square method. Additionally, by the field line tracing, it is found that resonant magnetic perturbations (RMPs) by the EFCC enable to produce stochastic magnetic structures at the edge region, that is applicable to an ELM control.