Stabilization Coils Research Articles

Implications of vertical stability control on the SPARC tokamak

To achieve its performance goals, SPARC plans to operate in equilibrium configurations with a strong elongation of κareal∼1.75 , which in turn will destabilize the n = 0 vertical instability. However, SPARC also features a relatively thick conducting wall that is designed to withstand disruption forces, leading to lower vertical instability growth rates than usually encountered. In this work, we use the TokSyS framework to survey families of accessible shapes near the SPARC baseline configuration, finding maximum growth rates in the range of γ≲100 s−1. The addition of steel vertical stability plates has only a modest ( ∼25% ) effect on reducing the vertical growth rate and almost no effect on the plasma controllability when the full vertical stability system is taken into account, providing flexibility in the plate conductivity in the SPARC design. Analysis of the maximum controllable displacement on SPARC is used to inform the power supply voltage and current limit requirements needed to control an initial vertical displacement of 5% of the minor radius. From the expected spectra of plasma disturbances and diagnostic noise, requirements for filter latency and vertical stability coil heating tolerances are also obtained. Small modifications to the outboard limiter location are suggested to allow for an unmitigated vertical disturbance as large as 5% of the minor radius without allowing the plasma to become limited. Further, investigations with the 3D COMSOL code reveal that strategic inclusion of insulating structures within the VSC supports are needed to maintain sufficient magnetic response. The workflows presented here help to establish a model for the integrated predictive design for future devices by coupling engineering decisions with physics needs.

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.

Assessment of vertical stability for negative triangularity pilot plants

Negative triangularity (NT) tokamak configurations may be more susceptible to magneto-hydrodynamic instability, posing challenges for recent reactor designs centered around their favorable properties, such as improved confinement and operation free of edge-localized modes. In this work, we assess the vertical stability of plasmas with NT shaping and develop potential reactor solutions. When coupled with a conformal wall, NT equilibria are confirmed to be less vertically stable than equivalent positive triangularity (PT) configurations. Unlike PT, their vertical stability is degraded at higher poloidal beta. Furthermore, improvements in vertical stability at low aspect ratio do not translate to the NT geometry. NT equilibria are stabilized in PT vacuum vessels due to the increased proximity of the plasma and the wall on the outboard side, but this scenario is found to be undesirable due to reduced vertical gaps which give less spatial margin for control recovery. Instead, we demonstrate that informed positioning of passively conducting plates can lead to improved vertical stability in NT configurations on par with stability metrics expected in PT scenarios. An optimal setup for passive plates in highly elongated NT devices is presented, where plates on the outboard side of the device reduce vertical instability growth rates to 16% of their baseline value. For lower target elongations, integration of passive stabilizers with divertor concepts can lead to significant improvements in vertical stability. Plates on the inboard side of the device are also uniquely enabled in NT geometries, providing opportunity for spatial separation of vertical stability coils and passive stabilizers.

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.

A Mutually Coupled Stabilization Coil Configuration for Electromagnetic Levitation Melting With Frequency Consideration

A Mutually Coupled Stabilization Coil Configuration for Electromagnetic Levitation Melting With Frequency Consideration

A data-driven Vertical Stabilization system for the ITER tokamak based on Dynamic Mode Decomposition

In this article, we propose an online method to estimate an approximate linear model of a vertically unstable tokamak plasma to consequently adapt the parameters of a Vertical Stabilization controller. The identification procedure is based on the Dynamic Mode Decomposition with control approach, while the tuning procedure takes advantage of linear control theory to impose the desired crossing frequency and gain margins. The proposed technique is aimed at controlling ITER elongated plasmas using the VS3 stabilization coils, located inside the vessel. The effectiveness of the method is proven by means of numerical simulations carried out with the CREATE-NL+ free boundary evolutionary code, also considering the presence of realistic measurement noise levels.

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.

Final Design of ITER In-Vessel Coils and Manufacturing of In-Vessel Coil Conductor

Following a decision made at the ITER Council in November 2013, two types of in-vessel coils (IVCs), namely, edge-localized mode (ELM) coils to mitigate ELMs and vertical stability (VS) coils to provide vertical stabilization of the plasma, have been incorporated in the ITER baseline design. The in-vessel environment is severe, characterized by large transient electromagnetic fields, high radiation flux, and high temperature. To withstand this environment and provide the required functionality, a “mineral insulated conductor” (MIC) technology has been selected for the IVC conductor. It consists of an axially water-cooled copper conductor surrounded by magnesium oxide insulation and a stainless steel jacket. A major advantage of the coil design is the choice of the same conductor material and dimensions for both VS and ELM coils, long conductor length which eliminates the need for any internal joints, and Cu welding for the joints between coils and feeders. In situ winding of the VS coils is asking for the development of a creative solution for the unspooling, straightening, precise winding tools, bending, forming, and metrology processes in a tight and congested environment. Mock-ups for design verification and manufacturing feasibility have been produced, in particular, for the bracket manufacture and to perform thermal and mechanical fatigue testing on an ELM control coil winding pack assembly. Installation trials of a half-size ELM control coil have been performed to verify the feasibility of the assembly procedure. This article will give an overview of the IVC design status and the progress of ITER IVC conductor manufacturing.

Feasibility Study of ITER In-Vessel Coils Bracket Manufacture and Integration

Feasibility study of ITER In-vessel coils bracket manufacture and integration had been developed in Institute of Plasma Physics, Chinese Academy of Sciences. The ITER In-Vessel Coil system is comprised of Edge-Localized Mode (ELM) and Vertical Stabilization (VS) coils. The ELM coils are used to mitigate the Edge Localized Modes and the VS coils are used to provide Vertical Stabilization of the plasma. Designed bracket for IVC coils is a kind of building block type three or four stacked components with arcuate groove matching with round conductor. This paper describes structure design, manufacture and integration process of the ELM and VS bracket. R&D of bracket weld and assembly sequence optimization are carried out to determine the welding and assembly process. At last, three brackets are integrated in ASIPP.

The Mechanical Performance evaluation of Vertical Stability Coil under Electromagnetic-structure Coupling Analyses

The vertical stability coil is a set of active feedback control coil that is used to deal with the vertical instability of plasma. To improve the response performance, the coil is mounted in the vacuum vessel, which denotes the coil-body will suffer from large electromagnetic force from the plasma current and poloidal field coils. Besides the current flowing in the feeder is nearly perpendicular with magnetic field originated from toroidal coil. It implies large electromagnetic force will be generated on the feeder. In order to withstand the impact from the electromagnetic force, a series of reinforce compo-nents are designed and installed on the coil. It is necessary to verify whether or not the coil conductor and auxiliary components could successfully bear the shock of large electromagnetic force. A three-dimensional magnetic field model is built to accurately calculate the magnetic field and electromag-netic force. Corresponding to the magnetic field calculation model, a more detailed me-chanical anal-ysis model is created to launch the electromagnetic-structural coupling analysis. Based on the stress analysis results, the local structure of the coil is optimized to decrease the peak stress. The updated model is reanalyzed and stress linearization is exerted to extract the different kinds of stress on the coil components. Finally, the stress is evaluated based on ASME analytical design. The evaluation result is helpful to guide the further design optimization.

Structural integrity assessment for ITER lower VS coil feedthrough

The ITER vertical stabilization coil system is designed to provide fast plasma stabilization in all scenarios with elongated plasma cross-section. It consists of two water-cooled coils located in upper and lower portions of the vacuum vessel (VV), with feedthroughs responsible for the transfer of electricity and cooling water across the vacuum boundary of VV. Depending on location, the feedthroughs can be divided into upper port feedthroughs (UPR feedthroughs) and lower port feedthroughs (LWR feedthroughs), each family sharing the same design. Both designs are subject to demanding operation scheme, a severe working environment and the presence of interfacing components, which impose important thermal, electromagnetic and seismic loads on the components. In particular, the LWR feedthrough, object of present paper, is more impacted by seismic loads due to the more flexible structure compared to the UPR feedthrough. During the worst seismic event, it is subjected to a vertical acceleration of 28.8 m s−2, acting on an effective mass of 249.9 kg at its first order modal mode. Furthermore, the current circulating in the conductor of the LWR feedthrough, impose significant linear forces, up to 27.7 kN m−1 during normal operation and 30.3 kN m−1 during categories I & II abnormal plasma events, respectively. Finally, the effect of differential thermal expansions of interfacing structures imposes a relative displacement between the supports up to 32.9 mm in X (radial) direction and up to 14.67 mm in Z (vertical) direction, during baking. Being part of the vacuum boundary, the feedthrough must assure confinement function under all normal and accidental loading conditions, with proper structural integrity assessment by applying analysis approach and design criteria foreseen by the applicable nuclear codes. The paper details the load types and categories, the analysis methodologies, the structural design criteria, and the assessment results in views of static stress and fatigue for the LWR feedthrough. This study could be a comprehensive reference for the structural integrity assessment of other similar components in fusion devices.

The Forming Parameters Study Based on ITER Prototypical Vertical Stability Coil Small Bending Arcs

The vertical stability (VS) coil is a set of circle-shaped multilayer in-vessel coil that is arranged in the vacuum vessel. For International Thermonuclear Experimental Reactor (ITER) prototypical VS coil, to avoid the interference with other in-vessel components some small bending arcs are designed. They are the R600 S-shaped bending arc with the angle of 9.2°, the R600 W-shaped bending arc with the angle of 20.3°, 38.1°, 20.3°, and the R215 S-shaped bending arc with the angle of 53.6°. The forming parameters of these arcs are studied in the article. Given the fact of conductor rebound after bending, the key issues of forming the arcs are to determine the suitable wheel mold radius and rotation angle. With regard to the determination of wheel mold radius, the forming deformation states of conductor under different wheel mold radius are simulated. Then the conductor displacement data after rebound are extracted to acquire the conductor inner and outer arcs fitting equations and the corresponding conductor radii are obtained simultaneously. If the calculated conductor central radius is different with the target value, the wheel mold radius will be adjusted and new simulation is employed. The reiteration work is terminated until the target radius is obtained. Subsequently, the wheel mold radius obtained in the data fitting process is referenced to fabricate the matching wheel mold. And the forming experiment is launched to verify the feasibility of conductor bending based on the above wheel mold radius. As to the determination of wheel mold rotation angle, the bending simulation and data fitting are carried out first. Then the process is repeated continuously until the target bending angle is obtained. Finally, the obtained wheel mold rotation angle is assigned as the preliminary bending angle in the forming experiment. If the forming results fail to satisfy the angle requirement, the preliminary bending angle is adjusted slightly and the forming experiment is performed again until the target bending angle is achieved. In addition, wrinkles are found during the forming process for the R215 S-shaped arc. To deal with the wrinkle issue a series of forming technology are attempted, while nearly all of them seem to be not feasible. Finally, bushing is installed at the outer surface of conductor jacket to reinforce its strength and resist the buckling impact from magnesium oxide (MgO). The forming experiments indicate bushing technology successfully eliminates the large wrinkles on the jacket.

The Study of Induction Brazing for Vertical Stability Coil Copper Conductor

The vertical stability coil will be used in China Fusion Engineering Test Reactor (CFETR) to improve its control capability on plasma. In view of the severe operation environment and high reliable requirement, it is a challenge for the coil to keep reliability during the whole lifetime, considering the prominent advantages of induction brazing, which is proposed for the coil joint connection. According to the electromagnetic-thermal coupling theory, the induction brazing simulation is carried out under different induction coil parameters. And a set of relatively reasonable induction coil parameter is obtained based on the average temperature and standard deviation of copper sample. In order to verify the brazing performance of the determined induction coil, the brazing experiment is launched with different joint structures and the brazing inspection is performed subsequently. Although it seems some joint design have the advantage of minimizing the damage on the outer magnesium oxide insulation layer, while they fail to satisfy the strength requirement. Based on the lessons, a new joint with thin external sleeve is designed and a series of performance tests are employed. It indicates that there is no large temperature gradient in the joint radial direction and the joint strength is prominently improved.

The Thermal and Structural Analysis of Vertical Stability Coil

The vertical stability coil is a new set of saddle shaped non-superconducting coil designed for the purpose of improving the control capability of plasma vertical movement. To avoid the electromagnetic shielding and en-hance the response performance, the vertical stability coil is installed in the inner wall of vacuum vessel. The subsequent disadvantage accompanying the benefit is that the coil is under severe neutron radiation. Besides the neutron radiation the coil will also encounter the ohmic heat once it is energized. The temperature rising of vertical stability coil is not allowed to beyond the specific threshold to guarantee the reliability of the coil com-ponents. Thus, the ohmic heat and nuclear heat calculation methods are presented and the detail temperature field is analysed by using ANSYS to check whether or not the coil can bear the thermal load. In addition, the thermal load will result in the thermal stress. To verify whether the thermal stress will lead to the structural damage, the thermal-structural coupling analysis is launched and the stress is evaluated based on ASME ana-lytical design. The analysis results will provide guidance for the local structural optimization of vertical stabil-ity coil.

Challenges and status of the ITER IVC conductor qualification process in China

The coils inside the ITER vacuum vessel are part of the overall plasma control system that ensures and maintains stable plasma operations. Two vertical stability coils(VS)will provide fast vertical stabilization of the plasma, while an array of 27 edge localized mode (ELM) coils will create resonant magnetic perturbations to magnetically control the plasma exterior in order to suppress potentially harmful power deposition on plasma-facing components. For magnet coils operating inside of the vacuum vessel, conventional insulation schemes are not an option. After a decade of development and intensive collaboration with industry and research institutes, a copper mineral insulated conductor (MIC) in stainless steel jacket will be used as for its ability to withstand large transient electromagnetic forces, high radiation flux, and high temperature.Complex technologies have been developed by ASIPP for the series of production of IVC conductors, in terms of conductor initial dimension design, MgO block assembly and elaboration of Cu pipe insertion, compaction and winding processes. China has finished all qualification phases, especially for 100-meter unit length conductor manufacturing. The first 100-meter unit length of copper tube is successfully manufactured in China. All conductor samples required for qualification process have successfully passed the electrical tests with good performance.The challenges and status of the production of ITER IVC conductors in China are described in this paper.

MCNP Models for Tokamak Cooling Water System Equipment Evaluations

Heat generated in the ITER fusion reactor is deposited in the tokamak vacuum vessel, in-vessel components, and in the components of the neutral beam injector during plasma operations and during subsequent decay of activation products. This heat is managed by the tokamak cooling water system (TCWS). The stainless steel material in the integrated loop of blanket edge-localized mode vertical stabilization coils and divertor (IBED) components (e.g., piping, heat exchangers (HXs), and pumps) contains activation sources because of its exposure primarily to neutron radiation from the decay of 17N, which is a short-lived radionuclide produced by neutron capture reactions with oxygen nuclei in the IBED primary heat transfer system (PHTS) cooling water during plasma operations. A detailed geometry model of the IBED stainless steel components and neutron radiation sources is required for an accurate assessment of the gamma activation sources on level 3 of the tokamak building. In the baseline design, each of the eight IBED PHTS cooling trains has two shell-and-tube heat exchangers (HXs) connected in series. Because these HXs are very large and contain a large amount of radioactive water, the possibility of using compact HXs of the welded shell-and-plate type is under investigation. This paper presents two Monte Carlo N-Particle (MCNP) TCWS geometry models, one model for each HX type, along with the associated piping. These models were obtained by automatic geometry conversion from TCWS computer-aided design models. The TCWS geometry models and neutron source definitions were incorporated into a baseline MCNP model of the Tokamak Complex.

Coaxial Tilt Damping Coil With Additional Active Actuation Capabilities

This paper presents a tilt stabilization coil inside an exemplary high-speed bearingless disk drive. The coil is based on the passive, electrodynamic “Null-Flux” principle, where there are no coil currents and, thus, no losses when the rotor is nondeflected during rotation. The proposed conductor structure is then used as a means of active actuation on the rotor by energizing the coil actively for producing tilt torque. Modifying the structure geometrically into multiple, coaxial conductor loops which are energized with currents in the equal or opposing directions then allows producing additional axial forces and tilt torque with multiple degrees of freedom, depending on the coil design. Different connection possibilities allow the combination of active and passive characteristics. Simulation and measurement at prototype drives separately demonstrate the functioning of both, the passive stabilization and the active actuation principles.

Research on Nondestructive Examination of Bracket Welds of ITER In-Vessel Coils (IVC)

The ITER in-vessel coils are comprised of edge-localized mode (ELM) and vertical stabilization (VS) coils. There are 27 ELM coils and 2 VS coils. All the ELM and VS coils are supported by the clamped brackets bolted to the rails which are welded to the vacuum vessel wall. The in-vessel environment is severe, characterized by large transient electromagnetic fields, high radiation flux, and high temperature. Therefore, an appropriate nondestructive examination approach should be established and qualified. The difficulty in the development of the bracket weld nondestructive examination is the complex structure which brings space limit and various reflect echoes. Phased array ultrasonic test (PAUT) was studied and developed for nondestructive examination. In order to obtain the appropriate test parameters, CIVA simulation software was utilized to visualize the acoustic field distribution and coverage. The sensitivity of the PAUT method was also demonstrated by using machined defect provided by ITER Organization. Experiment results shows that all of reference defects can be detected with peak amplitude higher than 80% of FSH for both first leg and second leg.

A Maximum Current Control Method for Three-Phase PWM Rectifier for the ITER In-Vessel Vertical Stability Coil Power Supply

The required peak current and peak voltage of ITER in-vessel vertical stability (VS) coil power supply are up to 80 kA and 2.4 kV, so VS coil power supply needs a pulse width modulation (PWM) rectifier to achieve high power factor operation under the highly transient power demand of the value of 144 MW. A new maximum current control method for three-phase voltage source PWM rectifier based on its mathematical model in $d$ – $q$ frame and improved by dual-closed-loop control method has been presented. This control method meets the fast power charging demand and the dc-side voltage stabilization demand of power supply, and achieves a unity or high power factor operation; its controller is easier to be designed and implemented than dual-closed-loop control at charging mode. The advantages and feasibility of this control method have been verified by simulation and experiment.

On the control system preparation for ELM pacing with vertical kicks experiments at TCV

High confinement plasma regimes (H-mode) are the scenarios with the highest fusion rates and the operating regime foreseen for the ITER scenario with QDT = 10. The steep gradients of density and temperature drive magnetohydrodynamic (MHD) edge localized modes (ELMs). Although ELMs are beneficial for controlling particle exhaust and plasma impurities, the excessive energy they deposit on the plasma facing components may lead to erosion or melting, reducing the components life cycle. By controlling the ELM frequency in an ELMy H-mode plasma it may be possible to reduce the energy expelled per ELM, mitigating damage to the tokamak parts.One technique for mitigating these risks was first implemented successfully in TCV Ohmic H-mode plasmas, using the so called ELM pacing with vertical kicks [1]. TCV vertical stabilization coils located inside the vessel proved to be well suited to apply the necessary perturbations to induce vertical displacements of the plasma. This method was then reproduced on AUG [2,3] and later on JET [4] with significant improvements.During 2016 a set of experiments were planned to further study the physics of ELM pacing taking advantage of the new neutral beam heating (NBH) system installed on TCV [5]. This submission reports on the related preparation of the TCV digital vertical stabilization control system [6] actuating the in-vessel coils. A number of tools were deployed to drive a fast radial-field variation and preliminary experimental validation was carried out. The tools and simulations, as well as preliminary experiments aiming at evaluating the controller's accuracy are reported.