In-vessel Coils Research Articles

Fault Analysis of In-Vessel Compression Coils in HFRC

Fault Analysis of In-Vessel Compression Coils in HFRC

Low-Resistance Joint for the ITER In-Vessel Coils

Low-Resistance Joint for the ITER In-Vessel Coils

Origin of the up/down poloidal asymmetric dependence of ELM control in ITER-like RMP configuration in KSTAR

Recent edge-localized modes (ELM) control experiments via resonant magnetic perturbation (RMP) in KSTAR have shown a strong up/down poloidal asymmetric coupling dependence. Specifically, in lower single null (LSN) plasmas at q95 ≳ 5, the lower two-row (middle/bottom) RMPs among ITER-like three-row (top/middle/bottom) in-vessel control coils (IVCC) in KSTAR were more effective in suppressing ELM-crashes than the upper two-row (top/middle) RMPs. In contrast, at q 95 ∼ 4, the upper two-row RMPs turned out to be more effective than the lower counterpart. Since the ITER baseline scenario is planned to operate at q 95 ∼ 3, the understanding of the origin of such up/down poloidal asymmetric coupling dependence, as well as the prediction about ITER-relevant conditions at a lower q 95, would be quite important and potentially impactful to the RMP ELM control in ITER. A linear, resistive, single-fluid MHD code MARS-F has been utilized to address and model the up/down poloidal asymmetric RMP coupling dependence. Specifically, based on two contrasting exemplary discharges with up/down poloidal (i) asymmetric at q 95 ∼ 4 and (ii) symmetric behavior at q 95 ∼ 5, among tens of otherwise similar discharges, a systematic MARS-F modeling has been thoroughly conducted. As a result, the plasma response investigation suggests that the X-point displacement (ξ X), rather than any other figures of merit, would be a directly relevant metric for the up/down poloidal asymmetric coupling in RMP-driven, ELM-crash suppression in KSTAR. Based on a sensitivity study of the edge safety factor in MARS-F modeling, the ξ X variation follows the same quantitative trend as observed in experiments. However, no or little plasma pressure dependence has been found, though ξ X increases with plasma pressure. At the ITER-relevant low q 95 ∼ 3 in a scaled KSTAR equilibrium, such modeling predicts the upper two-row RMPs would be more favorable in suppressing the ELM-crashes than the lower counterpart.

Control of elongated plasmas in superconductive tokamaks in the absence of in-vessel coils

The roadmap for the commissioning and first operations of superconductive tokamaks envisages the possibility of running discharges with fairly elongated plasmas before the complete installation of the in-vessel components, including vertical stabilization coils, or any other specific sets of coils to be used for the magnetic control of fast transients. In the absence of dedicated actuators, the magnetic control system shall perform the essential fast control actions by using the out-vessel superconductive coils, if needed. These are typically less efficient in reacting to fast transients, due to the shielding effect of the vessel and imply a coupling with other control tasks relying on the same actuators, such as plasma current, position, and shape control. Hence, effective actuator-sharing strategies must be put in place. This paper presents an architecture and a possible control strategy that is able to cope with vertically unstable elongated plasmas subject to fast varying disturbances, in the absence of dedicated in-vessel coils. The architecture exploits a model-based actuator-sharing approach to effectively accomplish the main magnetic control objectives while minimizing the cross-couplings among the various tasks. The effectiveness of the approach is demonstrated by means of nonlinear simulations of realistic JT-60SA scenarios. In particular, an isoflux plasma shape controller is integrated with plasma current control and vertical stabilization. The proposed control approach proves to control vertical displacement events and plasma deformations due to fast variations of poloidal beta with satisfactory performance.

Assembly procedure and tooling development plan for in-situ winding of ITER in-vessel vertical stability coils

The ITER in-vessel Vertical Stabilization (VS) coils, composed by the upper and lower VS coils, are designed to provide fast vertical stabilization of the plasma through a fast-current noise profile to compensate small plasma oscillations and to recover vertical displacement events.The VS coils are made by 4 continuous turns of Mineral Insulated Conductor (MIC) joined by Inconel ® 625 welded brackets, installed on rails on the inner wall of ITER Vacuum Vessel (VV). Thus, the coil turns winding/forming and the entire coils assembly have to be performed inside ITER VV, forcing the assembly procedure and tooling design to cope with the very limited space and strict constraints of ITER VV environment. The proposed procedure foresees that MIC conductor is un-spooled in the Neutral Beam (NB) port area and inserted into the ITER VV through a NB equatorial port, then a set of tooling is installed on a reinforced staging platform to wind, form and seal the conductor turns and finally to weld brackets under pre-compression. For this purpose, the tooling design including the reinforcement staging has been developed to meet the required tooling function, configuration optimization, assembly procedure and environmental constraints. The dedicated tools to be used for the assembly procedure of the VS coils are categorized into ex-vessel tools for unspooling and cleaning, in-vessel tools for winding and bump forming, bracket assembly tools for pre-compression and welding, termination and sealing tools, ancillary lifting tools and reinforcement staging platforms. This paper provides the overall assembly procedure and tooling development for the in-situ winding of VS coils.

HIL Investigation of a Single-Phase Inverter for a Tokamak Non-Axisymmetric In-Vessel Coil Power Supply

HIL Investigation of a Single-Phase Inverter for a Tokamak Non-Axisymmetric In-Vessel Coil Power Supply

Impact of increasing plasma-wall gap on plasma response to RMP fields in ITER

The impact of increasing plasma-wall gap on controlling edge-localized modes (ELMs) is numerically evaluated for the ITER 5 MA/2.65 T H-mode scenarios with full tungsten wall, based on the MARS-F computed plasma response to the applied n= 3–5 (n is the toroidal mode number) resonant magnetic perturbation (RMP) fields. Three new scenarios, referred to as standard, clearance and outergap, are considered assuming different plasma-wall gap sizes over a range on which vertically stability can be maintained by in-vessel coils in ITER. The latter are shown to have both direct and indirect effects on the plasma response and hence ELM control in ITER. The indirect and also primary influence occurs via change of the equilibrium edge safety factor q95 , which decreases with increasing the plasma-wall gap (at fixed plasma current and toroidal field), leading to a multi-peaking structure in the plasma response as measured by the plasma displacement near the X-point or the edge-localized resonant radial magnetic field perturbation. The direct, albeit secondary effect, is the reduction of local peak amplitudes with increasing the plasma-wall gap thus weakening the RMP field efficiency for ELM control with a given current in the control coils. A slight reduction of the plasma current, from 5 MA to 4.77, 4.92 and 4.65 MA for the standard, clearance and outergap scenarios, respectively, is found to be sufficient to access the q 95 window for the best ELM control with the n= 3 RMP. The n= 4 coil current configuration with the n= 5 sideband is also found favorable for ELM control in ITER, by producing RMP fields with mixed toroidal spectra compared to n = 3.

DEMO in-vessel equatorial coils for power-exhaust and fast plasma control

This paper addresses the issue of magnetic control of the DEMO device through the equatorial in-vessel coils. While vertical stabilization and radial control are typical applications of equatorial coils in plasma control, the cumbersome compatibility of in-vessel divertor coils with the DEMO divertor structures and remote handling draws attention to the possibility of using the in-vessel equatorial coils also for active plasma wobbling, as an alternative to traditional strike-point sweeping. In this work, the technical specifications of the DEMO equatorial in-vessel coils and the design requirements of the vertical stabilization and radial control systems are assessed. Finally, an optimized strategy for efficient power-exhaust control through active wobbling in DEMO is proposed.

Gaussian process-based quasi-coherent noise suppression in magnetic confinement devices with superconductors

Gaussian process (GP)-based technique suppressing quasi-coherent noises, i.e. structured noises, is developed which is more effective than conventional denoising techniques such as using frequency-domain filters. Superconducting devices like KSTAR, EAST, JT-60SA and ITER require separate sets of normal conducting magnetic coils inside the tokamak vacuum vessels to achieve a prompt control of fusion-grade plasmas in response to various fast and abrupt plasma activities such as vertical displacement events. Hence, these in-vessel control coils are typically operated with high-frequency switching power supplies which generate quasi-coherent noises. Semi-conductor based bolometers in KSTAR, for instance, are vulnerable to the quasi-coherent noise that makes a tomographic reconstruction for the 2D poloidal radiation map with the noise-contaminated signals flawed. By modeling the quasi-coherent properties of the noise as multivariate Gaussian distribution and generating the kernel function for the GP solely based on the measurements, the proposed method is able to suppress the noise whose performance is superior to the conventional filtering schemes. The method not only suggests an estimate of the denoised signal but also informs the consistent (with the measurements) uncertainty of the estimate at a level smaller than the standard deviation of the quasi-coherent noise. Performance of the method is confirmed with synthetic data containing the quasi-coherent noises, and it is applied to the measured data obtained by the KSTAR bolometers.

Research and development of copper joint for ITER In-Vessel Coils

The ITER In-Vessel Coils (IVCs) are crucial components of the Tokamak fusion reactor. They are located within the vacuum vessel (VV) behind the blanket shield modules and are responsible for compensating for fast perturbations of the plasma. There are a total of 176 joints used to connect the IVCs, feeders and feedthrough. The joint structure consists of a central copper conductor, a mineral (MgO) insulation layer, and an outer stainless-steel jacket. Given the inaccessibility of the IVCs once the blanket modules are installed, the highest level of quality is required for the joints. One key item is the welding of the copper conductor which has to be done inside the VV. To address the high thermal conductivity of copper conductor and the challenging constraints working inside the VV, a preheating method by the arc of the welding head has been developed. Macro- and micro-structure examination were performed on the welding cross sections, and tensile tests were carried out to check the strength of the copper joint, which demonstrated that arc preheating does not have a significant impact on the heat-affected zone. After more than 40 welding attempts and several design updates, the copper joint has been successfully developed and can be repeatedly and successfully welded.

Design of new resonant magnetic perturbation coils on the J-TEXT tokamak

The resonant magnetic perturbation (RMP) system is a powerful auxiliary system on tokamaks. On the J-TEXT tokamak, a set of new in-vessel coils is designed to enhance the amplitude of the RMP. The new coils are designed to be two-turn saddle coils. These two-turn saddle coils have been optimized in terms of their structure, support, and protection components to overcome the limitations of the narrow in-vessel space, resulting in a compact coil module that can be accommodated in the vessel. To verify the feasibility of this design, an electromagnetic simulation is performed to investigate the electrical parameters and the generated field of the coils. A multi-field coupled simulation is performed to investigate the capacity of heat dissipation. As a result of these efforts, the new RMP coils have been successfully installed on the J-TEXT tokamak. It has significantly enhanced the RMP amplitude and been widely applied in experiments.

Structural optimization design and numerical analysis of in-vessel magnetic compression coils

An FRC (field reversed configuration) device HUST-FRC (HFRC) is under construction at Huazhong University of Science and Technology, and the magnetic compression experiments will be investigated. 22 compression coils installed inside the metal vacuum chamber are designed with a peak current of about 110 kA rising within 50 μs. The coils are subjected to large electromagnetic forces during the compression process, which will challenge the structural stability of the coils. Therefore, it is necessary to estimate the structural stability of the compression coils during the compression progress. The compression process involves complex physical problems such as the coupling of circuit-electromagnetic-thermal-mechanical and non-linear contacts between adjacent coil components. It is also a challenge to calculate the coil current, electromagnetic load, and structure stress. To solve these problems, a joint analysis method is developed based on a two-dimensional (2-D) circuit-electromagnetic-thermal model and a three-dimensional (3-D) electromagnetic-mechanical model. The submodeling method is employed in mechanical simulations to improve the efficiency of coil optimization design. The electromagnetic and mechanical results confirm that the design of the compression coils is valid and feasible.

Analysis and design of in-vessel magnetic compression coil system for HFRC

A field-reversed configuration (FRC) experiment facility HUST-FRC (HFRC) is being built at Huazhong University of Science and Technology, where the magnetic compression of FRC will be investigated. To generate a pulsed compression magnetic field, a compression coil system is under development, consisting of in-vessel compression coils, a pulsed power supply, and feedthroughs. Energized by the pulsed power supply, a large current peaking ∼110 kA within 50 μs is transferred to in-vessel compression coils across the feedthroughs. This article presents the analysis and design of the compression coil system. At first, field ripple optimizations are performed to reduce the axial and toroidal ripples within 5%, resulting in an optimized coil arrangement and an improved coil conductor structure. In addition, the power supply circuits are quite complex due to the strong electromagnetic couplings among the coils and between each coil and the metal chamber wall. To rapidly and accurately design the electrical parameters of the pulsed power supply for achieving the target coil currents, a decoupled optimization method based on a circuit-electromagnetic model and a decoupled equivalent circuit model is proposed. The errors of the coil currents and the magnetic field are within 1%. Lastly, a feedthrough design and an optimized feeder scheme are proposed to achieve the structural stability of the feedthroughs since they are subjected to large electromagnetic forces.

Conceptual design of the DTT in-vessel equatorial coils

For the next generation fusion devices, the presence of in-vessel coils is fundamental to deal with elongated plasmas and alternative configurations, highly performant, yet critically sensitive to the vertical unstable mode. A pair of equatorial in-vessel coils connected in anti-series is sufficient to deal with the vertical control problem. In case of independent equatorial coils, they can also be used to produce a fast radial control action able to preserve plasma facing components during fast plasma transients. The main criticality of independent in-vessel coils is related to the induced current during disruptive events, making the coils extremely vulnerable due to overcurrents and mechanical stress. In this paper, we present the solution proposed for the conceptual design of the in-vessel equatorial coils for the Divertor Tokamak Test fusion device. This solution is able to guarantee radial and vertical control performance and robustness in case of disruption. Simulations are provided to show the efficiency of the design in case of disruptions and the effectiveness of the closed-loop control actions.

Compaction technology development for ITER In-vessel coil joints

Two types of ITER In-vessel coils (IVCs) are installed inside the vacuum vessel inner wall, behind blanket shield modules (BSMs), and near the plasma, namely: edge localized mode (ELM) and vertical stabilization (VS) coils. Overall, 27 ELM coils are distributed around the vacuum vessel inner wall, while one VS coil each is mounted above and below the plasma. Moreover, 176 IVC joints are used for joining the IVC coils, feeders, and feedthroughs, while 144 and 32 joints are used for joining the 27 ELM and two VS coils, respectively.The joints, which constitute a stainless steel (SS) sleeve, MgO insulation, and copper tube, show comparable performance to that of the conductor and can thus withstand severe conditions. Furthermore, the minimum available space for joint manufacturing is limited to 60 mm, which is a major obstacle. In joint manufacturing, two adjacent copper conductors are first welded via tungsten inert gas (TIG) welding, then the MgO insulation and sleeve will be applied. To maintain the insulation performance, the MgO insulation and SS sleeve should be compacted. Since compaction is one of the most important processes for joint manufacturing, this paper elucidates the development of compaction technology for IVC joints.

Conceptual design of in-vessel divertor coils in DTT

The Divertor Tokamak Test (DTT) facility will be equipped with in-vessel divertor coils able to locally modify the flux surfaces in the divertor region. When the first DTT divertor had not been selected, four in-vessel divertor coils were considered, with 10 turns each, fed by independent 4-quadrant SCR (thyristor) 0.5 kV - 5 kA - power supplies. This configuration was able to meet the design criteria in terms of requested performance and operational constraints, but one of the coils is not compatible with the geometry of the first DTT divertor, and another coil, scarcely efficient, can be removed. This paper discusses the design criteria, illustrates the possible revised layout, and illustrates the performance that can be achieved.

Estimation of the error field due to winding manufacturing and assembly tolerances of the DTT SC magnet system

The Divertor Tokamak Test facility (DTT) currently under construction in Frascati, Italy, is an experimental fusion reactor part of the EUROfusion roadmap to fusion energy. Due to their non-ideal winding distribution and assembly and manufacturing tolerances, real magnets are responsible for what is referred to as Error Fields (EFs). In some cases, EFs above a certain level can be unacceptable for achieving the sought performance of the fusion reaction and the implementation of dedicated in-vessel correction coils may become a necessary requirement. This work describes an approach to verify the magnet winding disposition by computing the EFs caused by the non-ideal shape of the DTT superconducting (SC) magnets. The results here presented are a contribution to the other activities aimed at deriving and verifying the requirements for the in-vessel correction coils.

Controlled movement of lower strike points by applying a divertor coil on the EAST tokamak

A new in-vessel water-cooled coil has been placed below the lower tungsten-covered divertor in EAST, aiming at moving the strike points in the divertor and spreading the heat flow on the targets. To achieve controlled movement of the strike points by the plasma control system (PCS) when applying the in-vessel coil, the plasma equilibrium configurations are predicted by the tokamak equilibrium (TEQ) code. The divertor coil current is involved in these calculations. The shape control parameters in the PCS are adjusted according to the predesigned equilibria from the TEQ code. The strike point movement over a wide range has been achieved in the experiment while keeping the main plasma stable. Thus, this method for divertor strike point movement with the assistance of the divertor coil is feasible.

Varistor development for in-vessel magnetic field coils in nuclear fusion devices

Varistor development for in-vessel magnetic field coils in nuclear fusion devices

An optical-input Maximum Likelihood Estimation feedback system demonstrated on tokamak horizontal equilibrium control

A readily parallelized Maximum Likelihood Estimation (MLE) algorithm with linear computational complexity is demonstrated in real time using only measurements from an extreme ultraviolet (EUV) diagnostic to control the horizontal position of a tokamak plasma. A set of trial emissivity profiles are parameterized by the control quantity of interest (Rm), and the MLE is identified from the profile which minimizes the signal reconstruction residual. The algorithm depends on an empirically determined likelihood function with exponential form. EUV emission (λ≈ 15 eV-1 keV) is captured in a poloidal plane by four 16-channel AXUV diodes mounted at different poloidal angles with radial and angular resolution sufficient to discern plasma equilibrium evolution in HBT-EP. Calculations of the plasma major radius by the system are consistent within diagnostic uncertainty for the majority of the discharge with those of: a weighted average of vertical soft X-ray or EUV chords, magnetic sensors, and an equilibrium reconstruction. The feedback system corrects for a horizontal displacement of the major radius equal to 20% of the plasma minor radius by adjusting the vertical field produced from 40 in-vessel control coils in real time. The MLE calculation is performed on a GPU in a 15μs cycle, with similar performance in this application to a simple weighted average of vertical chords. Results demonstrate horizontal position control using magnetic actuators and an optical observer.