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In magnetic fusion devices, error field (EF) sources, spurious magnetic field perturbations, need to be identified and corrected for safe and stable (disruption-free) tokamak operation. Within Work Package Tokamak Exploitation RT04, a series of studies have been carried out to test the portability of the novel non-disruptive method, designed and tested in DIII-D (Paz-Soldan et al 2022 Nucl. Fusion 62 126007), and to perform an assessment of model-based EF control strategies towards their applicability in ITER. In this paper, the lessons learned, the physical mechanism behind the magnetic island healing, which relies on enhanced viscous torque that acts against the static electro-magnetic torque, and the main control achievements are reported, together with the first design of the asynchronous EF correction current/density controller for ITER.

The National Spherical Torus Experiment (NSTX) has undergone a major upgrade, and the NSTX Upgrade (NSTX-U) Project was completed in the summer of 2015. NSTX-U first plasma was subsequently achieved, diagnostic and control systems have been commissioned, the H-mode accessed, magnetic error fields identified and mitigated, and the first physics research campaign carried out. During ten run weeks of operation, NSTX-U surpassed NSTX record pulse-durations and toroidal fields (TF), and high-performance ∼1 MA H-mode plasmas comparable to the best of NSTX have been sustained near and slightly above the n = 1 no-wall stability limit and with H-mode confinement multiplier H98y,2 above 1. Transport and turbulence studies in L-mode plasmas have identified the coexistence of at least two ion-gyro-scale turbulent micro-instabilities near the same radial location but propagating in opposite (i.e. ion and electron diamagnetic) directions. These modes have the characteristics of ion-temperature gradient and micro-tearing modes, respectively, and the role of these modes in contributing to thermal transport is under active investigation. The new second more tangential neutral beam injection was observed to significantly modify the stability of two types of Alfven eigenmodes. Improvements in offline disruption forecasting were made in the areas of identification of rotating MHD modes and other macroscopic instabilities using the disruption event characterization and forecasting code. Lastly, the materials analysis and particle probe was utilized on NSTX-U for the first time and enabled assessments of the correlation between boronized wall conditions and plasma performance. These and other highlights from the first run campaign of NSTX-U are described.

The aim of this article is to describe and model a sawtooth crash in RFX-mod2 and compute the eddy currents induced in the stabilizing shell surrounding the plasma for two alternative gap concepts and the subsequent error field. The sawtooth crash displacement of the plasma column is represented by a vacuum model coupled with a suitable magneto-quasi-static (MQS) volume integral (VI) formulation. [ABSTRACT FROM AUTHOR]

A measurement-based numerical reconstruction technique is proposed to control winding geometry of the poloidal coils manufactured for ITER. A detailed coil model is parameterized in terms of the magnetic characteristics. Deviations from the specified coil shape are evaluated in a comparison of reconstructed “ideal” field and data of magnetic measurements. The technique has been validated in the course of quality inspection of the dummy double pancake similar to the pancakes of the ITER PF1 coil. Experimental results are presented. [ABSTRACT FROM AUTHOR]

Anadvanced computationmodel of the Poloidal Field coil #1 (PF1) of ITER tokamak is developed that involves a detailed description of the winding geometry. The model describes all basic current-carrying components, including turn joggles and inter-layer joggles of the coil. The model has been validated in comparative simulations with the simplified models and analytical calculations based on Ampere's law. A high computational accuracy of 7 · 10-7 is demonstrated. The error field produced by PF1 coil winding was estimated to be an order of magnitude below the threshold level. The model is planned to be applied for the quality control of the manufactured coil. The approach utilized in the PF1 coil model is quite generalized and flexible that makes it applicable for the modeling of other coils of the ITER magnet system. [ABSTRACT FROM AUTHOR]

After several years of successful RFX-mod operations, both in RFP and Tokamak configurations, an upgrade of the machine has been conceived. In particular, the vacuum vessel might be removed with the aim of improving passive MHD control and enabling plasma rotation at higher currents. In this paper, a detailed analysis of the new magnetic front-end is presented, with particular emphasis on the computation of the magnetic field errors generated at the poloidal gaps during transient phases of the discharge. At this purpose, a non-linear MHD equilibrium code, MAXFEA, has been used to provide the input data for the 3D electromagnetic analyses in the time domain carried out with the CAFE code. [ABSTRACT FROM AUTHOR]

The Divertor Tokamak Test (DTT) facility, construction starting at Frascati, Italy, is designed to test different solutions for divertor in view of DEMO. A preliminary analysis of the error fields (EFs) assumed a simplified model of rigid and independent displacements and rotations. A methodology based on the first order truncated Taylor expansion has been applied, linking the displacement parameters and the EFs within the required accuracies. A system of in-vessel copper coils has been designed to counteract EFs and the ampere-turns necessary to force them back within the request limits has been calculated. Here, the details of the analysis have been provided. [ABSTRACT FROM AUTHOR]

Discrepancies in magnetic field maps produced by confinement coils in thermonuclear fusion reactors may drive plasma to lose stability and must be carefully controlled during the whole time evolution of each shot using suitable correction coils. Even when kept below safety thresholds, these error fields (EFs) may alter the geometry of magnetic flux lines, defining the plasma geometry. This paper, using high accuracy 3-D magnetic field computations for confinement coils, addresses the issue of evaluating the effect of EFs on the plasma boundary shape during the shot, modeled as a sequence of equilibrium configurations. In particular, a procedure that can compute the shape perturbations due to given deformations of the coils has been set up and used to carry out an analysis of relationship between the EF and shape perturbations during the time evolution of International Tokamak Experimental Reactor programmed scenario. [ABSTRACT FROM AUTHOR]

Discrepancies between design and actual magnetic field maps in tokamaks are unavoidable, and are associated to a number of causes, e.g. manufacturing and assembly tolerances on magnets, presence of feeders and joints, non-symmetric iron parts. Such error fields may drive plasma to loss of stability, and must be carefully controlled using suitable correction coils. Anyway, even when kept below safety threshold, error fields may alter the behavior of plasma. The present paper, using as example the error fields induced by tolerances in toroidal field coils, quantifies their effect on the plasma boundary shape in equilibrium configurations. In particular, a procedure able to compute the shape perturbations due to given deformations of the coils has been set up and used to carry out a thorough statistical analysis of the error field-shape perturbations relationship. [ABSTRACT FROM AUTHOR]

Abstract: Error fields in tokamaks are small departures from the exact axisymmetry of the ideal magnetic field configuration. Their reduction below a threshold value by the error field correction coils is essential since sufficiently large static error fields lead to discharge disruption. The error fields are originated not only by magnets fabrication and installation tolerances, by the joints and by the busbars, but also by the presence of ferromagnetic elements. It was shown that superconducting joints, feeders and busbars play a secondary effect; however in order to estimate of the importance of each possible error field source, rough evaluations can be very useful because it can provide an order of magnitude of the correspondent effect and, therefore, a ranking in the request for in depth analysis. The paper proposes a two steps procedure. The first step aims to get the approximate magnetization state of ferromagnetic parts; the second aims to estimate the full 3D error field over the whole volume using equivalent sources for magnetic masses and taking advantage from well assessed approximate closed form expressions, well suited for the far distance effects. [Copyright &y& Elsevier]

Geometrical deformations and assembly errors in the ITER Toroidal Field (TF) coils will lead to magnetic field perturbations, which could degrade plasma confinement and eventually lead to disruption. Extensive computational studies of the influence of coil deformations and assembly errors on plasma behavior have given the basis for definition of the geometric tolerance of the Current Center Line (CCL) of the winding pack of the TF coil. This paper describes an analysis method to establish the feasibility to measure the magnetic CCL locus of the final winding pack (WP) with accuracy better than 1 mm. The proposed method is based on arrays of gradient coils accurately mounted with respect to the WP fiducial marks and datum surfaces. The magnetic measurements will be performed at defined locations around the WP perimeter to characterize accurately the CCL locus. The analysis emphases the robustness and sensitivity of the method versus the measurement location and the TF coil 3D geometrical deformation. The analysis and proposed measurement techniques will be described in detail. [ABSTRACT FROM PUBLISHER]

Abstract: Error fields in Tokamaks are small departures of the exact axisymmetry of the ideal magnetic field configuration. Their reduction beyond a threshold value by the error field correction coils is essential since sufficiently large static error fields lead to discharge disruption. The error fields are originated not only by coils fabrication and installation alignment tolerances, joints and busbars but also due to the presence of ferromagnetic elements. The start of plasma current flattop with relatively low plasma density is considered as a critical state of the 15MA scenario for the onset of locked modes causing disruptions. A figure of merit, the ‘3-mode’ criterion, based on the lowest error field harmonics [(1,1); (2,1); (3,1)] was chosen to assess the error fields expected in ITER. Analysis performed last years in an independent way by CREATE (EU) and Efremov Inst. (RF) groups has allowed a deep understanding of the error fields induced by all the possible sources. Both groups were successfully benchmarked with the estimation of the error fields induced by a given deformed shape of a TF coil obtaining a perfect match of the results up to 2 orders of magnitude smaller than the defined threshold of B 3-mode/B to <5×10−5. Three different sets of independent variables based on a 3D rigid body movement of the coils have been provided for the 6 CS modules, 18 TF coils and 6 PF coils tolerances, which have allowed a clear understanding of the weight of each of the variables in the induced error fields. The results obtained in 2008 with a realistic set of magnets tolerances concluded that the system of correction coils provides effective suppression of the error fields with margin to correct possible impact of other sources. In particular, it has been shown that superconducting joints, feeders and busbars play a secondary effect; however the radial position of the TF coils and the tilt and radial shift of the CS stack would have a relevant influence. The ensuing recent sets of variables studied aimed at deepening the understanding of the operational limits. The present paper addresses the impact of the fabrication tolerances and installation misalignments of the magnet system; the impact of ferromagnetic inserts and busbars as well as the design of the correction coils will be thoroughly covered elsewhere (Amoskov et al., in press ). [Copyright &y& Elsevier]

The Correction Coils system (CC) within ITER, is intended to reduce the range of magnetic error fields created by assembly or geometrical imperfections of the other coils used to confine, heat, and shape the plasma. The proposed magnet system consists of three sets of 6 coils each, located at the top (TCC), side (SCC) and bottom (BCC) of the Tokamak device and uses a NbTi cable-in-conduit superconducting conductor (CICC) operating at 4.2 K. The ITER Organization (IO) and the Institute of Plasma Physics at the Chinese Academy of Sciences (ASIPP) are jointly preparing the definition of the technical specifications and the upcoming qualification program for the Correction Coils. The proposed design consists of a one in hand conductor winding without internal joint inserted in a structural casing which reacts the electromagnetic loads. The development of major items such as terminal joints, casing manufacture, and vacuum impregnation system, is an essential phase before the series production which will take place at the premises of the supplier. This paper discusses the key technologies on CC coils and future plans for short sample prototypes fabrication. [ABSTRACT FROM AUTHOR]

Abstract: A set of 24 saddle loops is used on the Tokamak à Configuration Variable (TCV) to measure the radial magnetic flux at different toroidal and vertical positions. The new system is calibrated together with the standard magnetic diagnostics on TCV. Based on the results of this calibration, the effective current in the poloidal field coils and their position is computed. These corrections are then used to compute the distribution of the error field inside the vacuum vessel for a typical TCV discharge. Since the saddle loops measure the magnetic flux at different toroidal positions, the non-axisymmetric error field is also estimated and correlated to a shift or a tilt of the poloidal field coils. [ABSTRACT FROM AUTHOR]

Perturbations of the axial symmetry of the poloidal magnetic field in the order of 10-4 of the toroidal magnetic field (5.3 T) might induce the occurrence of magnetic islands which could degrade plasma confinement and eventually lead to disruption. A weighted parameter of the lowest harmonics, the '3-mode' index, is used as the figure of merit of ITER confinement capability with respect to possible magnet geometrical deformations and assembly errors. Three sets of correction coils will ensure that error fields are kept below the Three Mode Error Index (TMEI) during operation. Fabrication tolerances and assembly misalignments of the TF coils beyond specified values might become a relevant contributor to possible error fields in the magnetic configuration. An accurate determination of the coil current center line (CCL) is essential as the reference for survey operations during fabrication and final installation phases. The CCL will be defined geometrically as the barycenter of the conductor grooves of the winding pack, which will consist of seven, stacked D-shaped radial plates. However, it also seems feasible to reconstruct the CCL of the winding pack from magnetic measurements. A current of 1 kA circulating in the winding pack at room temperature will generate a magnetic field gradient of ~0.07 mT/mm at 500 mm from the winding pack center. With an array of probes operated in flip mode and accurately mounted with respect to the WP fiducial marks the CCL can be determined with accuracy < +/-1 mm. The paper will evaluate the accuracy of both geometric and magnetic procedures for current center line definition. [ABSTRACT FROM AUTHOR]

Error fields produced by the ferromagnetic rebar of the building of the ITER tokamak complex have been studied. It is shown that the '3-mode' error field produced by the global asymmetry of the building is about 2×10-6 of the toroidal magnetic field. Analysis of error fields caused by local asymmetry of the ferromagnetic elements was performed for the one nearest to the tokamak ferromagnetic structure, bioshield. It was shown that the bioshield produces an error field less than 10-5 of the toroidal magnetic field. Such error fields are acceptable. [ABSTRACT FROM AUTHOR]

The error fields produced by localized steel objects in the ITER tokamak building have been studied in this paper. The critical mass of the objects, producing in the plasma region the '3-mode' error field of 0.1 Unit (10-6 of the toroidal magnetic field), is estimated for a set of object positions. It is shown that the critical mass is scaled with distance as R7, if R≥30 m. [ABSTRACT FROM AUTHOR]

Error fields in tokamak magnet systems are primarily associated with displacements of toroidal field and poloidal field coils, arising during their manufacture and assembly. The Fourier analysis of error fields on a rational magnetic flux surface, anticipated in the ITER machine, exhibited the lower-order modes ((m, n)=(1,1), (2,1), (3,1)) of error fields as the most crucial ones. A statistical study using the Monte Carlo simulations has demonstrated that an averaged three-mode error field from the superposition of coil displacements can be two to three times higher than the acceptable level. The estimates obtained have been used to assess the capability of the correction coil system. [ABSTRACT FROM AUTHOR]

An analysis has been carried out to study error fields anticipated in the ITER machine. The error fields were simulated with the use of a computer code, PRORCODE, developed at the Efremov Institute. The Fourier decomposition of the error fields was performed on equilibrium rational magnetic surfaces inside the plasma filament to evaluate lower modes ( m , n ) = (1,1), (2,1), (3,1). Major sources of error field associated with misalignments of the magnet system in tokamaks are discussed. [ABSTRACT FROM AUTHOR]

Non-axisymmetric (error) fields in tokamaks lead to a number of instabilities including so-called locked modes [J.T. Scoville, R.J. La Haye, Nucl. Fusion 43 (4) (2003) 250–257] and resistive wall modes (RWM) [A.M. Garofab, R.J. La Haye, J.T. Scoville, Nucl. Fusion 42 (11) (2002) 1335–1339] and subsequent loss of confinement. They can also cause errors in magnetic measurements made by point probes near the plasma edge, error in measurements made by magnetic field sensitive diagnostics, and they violate the assumption of axisymmetry in the analysis of data. Most notably, the sources of these error fields include shifts and tilts in the coil positions from ideal, coil leads, and nearby ferromagnetic materials excited by the coils. New measurements have been made of the n=1 coil-related field errors in the DIII–D plasma chamber. These measurements indicate that the errors due to the plasma shaping coil system are smaller than previously reported and no additional sources of anomalous fields were identified. Thus they fail to support the suggestion of an additional significant error field suggested by locked mode and RWM experiments. [Copyright &y& Elsevier]

The paper assesses the n = 1 "overlap" error fields produced by localized steel objects (the object size is much less than the distance from the object to the plasma) placed at different levels of the ITER Tokamak building. The value of the error field proportionally depends on the mass of steel object. This paper studies the effect of the shape of steel objects on error field. For evaluations, three shapes of steel objects are considered allowing analytical calculation of their magnetization: a steel ball and a thin steel cylinder oriented either perpendicularly or parallel to the external magnetic field. The paper demonstrates a strong effect on the error field of the object shape and orientation. [ABSTRACT FROM AUTHOR]