High amplitude energetic particle geodesic acoustic mode (EGAM) oscillations driven by Neutral Beam (NBI) injected in the direction counter to the toroidal plasma current are observed in the TCV tokamak. The modes appear at frequencies close to the geodesic acoustic mode (GAM) frequency, corresponding to the mode radii calculated without plasma elongation corrections. The spatial structure of the EGAM density oscillation is analyzed using multichannel soft x-ray (SXR) and broadband light emission diagnostics. The analysis of spatiotemporal emission data shows a non-rotating structure of the density oscillations. The non-rotating mode discrimination is the additional new capability of the multichannel spatiotemporal SXR data analysis technique in TCV. We present the discrimination method of assessing the standing character of the EGAM wave. The structure of EGAM density oscillations in TCV is consistent with the theoretical GAM poloidal structure, namely a m= 1 standing wave with density oscillations amplitude proportional to sin(theta), where the poloidal angle theta is measured from the equatorial plane. The poloidal structure of magnetic the field oscillations is analyzed using a poloidal array of magnetic coils. The structure of the EGAM-induced magnetic field oscillations in TCV is a m = 2 standing wave. Time evolution of EGAMs suggests that a nonlinear EGAM chirping is observed. The chirping depends on the EGAM radial location, which varies as the NBI deposition is varied through a vertical shift of the plasma magnetic axis. The chirping disappears at the plasma periphery. A fast periodic radial shift of the EGAMs radial location is also observed to occur during the single chirp.

Tokamak a configuration variable (TCV), recently celebrating 30 years of near-continual operation, continues in its missions to advance outstanding key physics and operational scenario issues for ITER and the design of future power plants such as DEMO. The main machine heating systems and operational changes are first described. Then follow five sections: plasma scenarios. ITER Base-Line (IBL) discharges, triangularity studies together with X3 heating and N2 seeding. Edge localised mode suppression, with a high radiation region near the X-point is reported with N-2 injection with and without divertor baffles in a snowflake configuration. Negative triangularity (NT) discharges attained record, albeit transient, beta(N) similar to 3 with lower turbulence, higher low-Z impurity transport, vertical stability and density limits and core transport better than the IBL. Positive triangularity L-Mode linear and saturated ohmic confinement confinement saturation, often-correlated with intrinsic toroidal rotation reversals, was probed for D, H and He working gases. H-mode confinement and pedestal studies were extended to low collisionality with electron cyclotron heating obtaining steady state electron iternal transport barrier with neutral beam heating (NBH), and NBH driven H-mode configurations with off-axis co-electron cyclotron current drive. Fast particle physics. The physics of disruptions, runaway electrons and fast ions (FIs) was developed using near-full current conversion at disruption with recombination thresholds characterised for impurity species (Ne, Ar, Kr). Different flushing gases (D2, H2) and pathways to trigger a benign disruption were explored. The 55 kV NBH II generated a rich Alfvenic spectrum modulating the FI fas ion loss detector signal. NT configurations showed less toroidal Alfven excitation activity preferentially affecting higher FI pitch angles. Scrape-off layer and edge physics. gas puff imaging systems characterised turbulent plasma ejection for several advanced divertor configurations, including NT. Combined diagnostic array divertor state analysis in detachment conditions was compared to modelling revealing an importance for molecular processes. Divertor physics. Internal gas baffles diversified to include shorter/longer structures on the high and/or low field side to probe compressive efficiency. Divertor studies concentrated upon mitigating target power, facilitating detachment and increasing the radiated power fraction employing alternative divertor geometries, optimised X-point radiator regimes and long-legged configurations. Smaller-than-expected improvements with total flux expansion were better modelled when including parallel flows. Peak outer target heat flux reduction was achieved (>50%) for high flux-expansion geometries, maintaining core performance (H-98 > 1). A reduction in target heat loads and facilitated detachment access at lower core densities is reported. Real-time control. TCV's real-time control upgrades employed MIMO gas injector control of stable, robust, partial detachment and plasma beta feedback control avoiding neoclassical tearing modes with plasma confinement changes. Machine-learning enhancements include trajectory tracking disruption proximity and avoidance as well as a first-of-its-kind reinforcement learning-based controller for the plasma equilibrium trained entirely on a free-boundary simulator. Finally, a short description of TCV's immediate future plans will be given.

An overview is presented of the progress since 2021 in the construction and scientific programme preparation of the Divertor Tokamak Test (DTT) facility. Licensing for building construction has been granted at the end of 2021. Licensing for Cat. A radiologic source has been also granted in 2022. The construction of the toroidal field magnet system is progressing. The prototype of the 170 GHz gyrotron has been produced and it is now under test on the FALCON facility. The design of the vacuum vessel, the poloidal field coils and the civil infrastructures has been completed. The shape of the first DTT divertor has been agreed with EUROfusion to test different plasma and exhaust scenarios: single null, double null, X-divertor and negative triangularity plasmas. A detailed research plan is being elaborated with the involvement of the EUROfusion laboratories.

In 2021 JET exploited its unique capabilities to operate with T and D-T fuel with an ITER-like Be/W wall (JET-ILW). This second major JET D-T campaign (DTE2), after DTE1 in 1997, represented the culmination of a series of JET enhancements-new fusion diagnostics, new T injection capabilities, refurbishment of the T plant, increased auxiliary heating, in-vessel calibration of 14 MeV neutron yield monitors-as well as significant advances in plasma theory and modelling in the fusion community. DTE2 was complemented by a sequence of isotope physics campaigns encompassing operation in pure tritium at high T-NBI power. Carefully conducted for safe operation with tritium, the new T and D-T experiments used 1 kg of T (vs 100 g in DTE1), yielding the most fusion reactor relevant D-T plasmas to date and expanding our understanding of isotopes and D-T mixture physics. Furthermore, since the JET T and DTE2 campaigns occurred almost 25 years after the last major D-T tokamak experiment, it was also a strategic goal of the European fusion programme to refresh operational experience of a nuclear tokamak to prepare staff for ITER operation. The key physics results of the JET T and DTE2 experiments, carried out within the EUROfusion JET1 work package, are reported in this paper. Progress in the technological exploitation of JET D-T operations, development and validation of nuclear codes, neutronic tools and techniques for ITER operations carried out by EUROfusion (started within the Horizon 2020 Framework Programme and continuing under the Horizon Europe FP) are reported in (Litaudon et al Nucl. Fusion accepted), while JET experience on T and D-T operations is presented in (King et al Nucl. Fusion submitted).

Within the 9th European Framework programme, since 2021 EUROfusion is operating five tokamaks under the auspices of a single Task Force called ‘Tokamak Exploitation’. The goal is to benefit from the complementary capabilities of each machine in a coordinated way and help in developing a scientific output scalable to future largre machines. The programme of this Task Force ensures that ASDEX Upgrade, MAST-U, TCV, WEST and JET (since 2022) work together to achieve the objectives of Missions 1 and 2 of the EUROfusion Roadmap: i) demonstrate plasma scenarios that increase the success margin of ITER and satisfy the requirements of DEMO and, ii) demonstrate an integrated approach that can handle the large power leaving ITER and DEMO plasmas. The Tokamak Exploitation task force has therefore organized experiments on these two missions with the goal to strengthen the physics and operational basis for the ITER baseline scenario and for exploiting the recent plasma exhaust enhancements in all four devices (PEX: Plasma EXhaust) for exploring the solution for handling heat and particle exhaust in ITER and develop the conceptual solutions for DEMO. The ITER Baseline scenario has been developed in a similar way in ASDEX Upgrade, TCV and JET. Key risks for ITER such as disruptions and run-aways have been also investigated in TCV, ASDEX Upgrade and JET. Experiments have explored successfully different divertor configurations (standard, super-X, snowflakes) in MAST-U and TCV and studied tungsten melting in WEST and ASDEX Upgrade. The input from the smaller devices to JET has also been proven successful to set-up novel control schemes on disruption avoidance and detachment.

The Tokamak a Configuration variable (TCV) is equipped with two neutral beam injection (NBI) systems delivering up to 1.2 MW each for pulse durations of up to 2 s. The first system (NBI1), designed for an injection energy in the range of 25 to 30 keV has been operational since 2016. The existing concrete neutron shielding of the experimental hall proved insufficient for fully protecting human accessible areas, limiting the number of daily plasma pulses using NBI1. The recently commissioned second system (NBI2) is designed for injection synergies in the range 50 to 60 keV. Both systems are tangentially oriented in opposite directions in order to permit experiments with low or no net torque. Calculations with the TRANSP and ORBIS heating codes show that neutron rates from deuterium-deuterium fusion reactions may be as high as 10(14) n/s, up to 10 times higher than with the lower energy beam only. This is due both to the similar to five times larger beam-plasma neutron rates from the higher energy beam and to an exceptionally high contribution from beam-beam reactions between the opposing beams. The radiation protection policy at the Swiss Plasma Center is that all staff members be considered as members of the general public, limiting the daily personal dose to 4 mu Sv. This is also the maximum admissible daily dose in any publicly accessible zone, whether occupied or not. Currently, with only the lower energy beam, this limit can be attained in the control room adjacent to the device hall after only five NBI pulses out of a possible 30 daily pulses. To allow for exploitation of the two beams at full specifications, the source side of the existing barite concrete walls of the 15 x 20 x 8 m large TCV hall will be covered with 20-cm-thick polythene (PE) cladding and a ceiling made of 35-cm-thick PE will be added. The total mass of PE will be 200 tons. The usage of PE at this scale for neutron shielding is unprecedented at any fusion research facility.

An innovative and very detailed end-to-end system modelling tool has been developed and applied to test on simulated data the actual measurement capabilities of any generic high frequency (HF) magnetic diagnostic systems. The main goal of this rather complex tool is to obtain estimates of the intrinsic measurement uncertainties and then assess the actual vs. intended system measurement performance for correctly detecting individual components in the frequency spectrum of HF magnetic instabilities in the plasma. This has paramount consequences not solely for off-line analyses but also, and more importantly, for any real-time application where, as an example, the mode frequency, amplitude and {toroidal, poloidal} mode numbers are used to determine whether, and which, corrective actions need to be taken to stabilize the discharge. The algorithm has been applied to some of the various ITER HF magnetic diagnostic systems, most notably the AJ (LTCC-1D sensors) system as currently designed, hence providing specific confidence levels and error bounds for detecting the modes highlighted in the ITER measurement specifications. Additional analyses have been performed for the TCV and JET HF magnetic diagnostics, providing further constraints on the results obtained with these systems.

Inductive magnetic sensors are needed for tokamak operation to provide the low-frequency (LF) measurements leading to the equilibrium reconstruction and to monitor the higher frequency (HF) instabilities; the HF magnetic sensors are often also used as a back-up to the LF ones. For the HF inductive magnetic sensors (fluctuation measurements), we need to minimize the self-inductance ( L SELF ) provided that the effective area ( N A EFF ) remains sufficiently large, typically requiring L SELF < 100 μH and N A EFF ∼ 0.01 m 2 . For the LF inductive magnetic sensors (equilibrium reconstruction), the only physics-based design criteria are that of maximizing N A EFF > 0.10 m 2 , essentially independently of the resulting L SELF . Due to these rather different measurement specifications, it is quite the common case that different sets of LF and HF inductive magnetic sensors are used, which significantly complicates the R&D activities and the ensuing manufacturing processes. Starting in 2007, our group at the Swiss Plasma Centre at the Ecole Polytechnique Fédérale de Lausanne (EPFL) has originally developed the Low-Temperature Co-fired Ceramic (LTCC) technology for producing inductive magnetic sensors, this technology being exceptionally suitable for operating temperatures up to ∼1'000C in very harsh environmental conditions. The similarly named High-Temperature (HTCC) technology uses the same processes but different materials, and it is then suitable for operating temperatures up to ∼1'600C. ITER will have about ∼250 such LTCC sensors, of EPFL design and prototyped at the EPFL, but manufactured by a commercial entity to technical specifications lower than those achieved at the EPFL in terms of track width, track separation and overall manufacturing yields. While state-of-the-art in 2007, the LTCC and HTCC technologies are now at least 20 years old and new processes have been developed for commercial applications, essentially based on different photolithography (PL) techniques. Recent, and already industrialized, advances in microfabrication using PL techniques offer the possibility to create more compact and optimized designs starting from current industrial standards, therefore extremely facilitating the very costly lab-to-fab step of production, namely all the activities that lead from the ideas in the lab to the actual (industrialized) fabrication. Our goal is to continue to push the frontier of magnetic sensors, bring a commercially viable design that can be used in tomorrow's scientific projects, such as nuclear fusion and astronomy/astrophysics (for instance, in the new generations of miniaturized satellites). The main advantage of the PL techniques is that a much smaller track width ( dd 1) can be achieved, down from the routine dd 1 = 100 μm of the EPFL LTCC sensors (for ITER: dd 1 = 400 μm) to dd1 < 10 μm. A smaller dd1 allows to pack more planar winding loops (m) enclosing a larger area over a smaller geometrical surface. Therefore, PL coils can have a lower self-inductance L SELF ∝ m 2 while having the same effective area N A EFF ∝ m compared to coils manufactured by other more common technologies. Therefore, with PL techniques a similar design could be used for both HF and LF applications, the difference simply being the number of stacked-up layers (n) used to make-up the entire sensor. In this paper we will present the first developments towards the production of inductive magnetic sensors using PL techniques. Whilst providing the possibility of designing better performing and more compact sensors, the introduction of PL techniques in the manufacturing processes for inductive magnetic sensors has uncovered new limitations and obstacles, such as the required vertical track thickness that is poorly suited for the existing deposition techniques, the need for stacking up multiple wafers and the connection between the sensor and the in-vessel cabling.

A well-known, previously only 1D, algorithm using the Sparse Representation of Signals and an iterative Block Coordinate Descent method (the SparSpec-1D algorithm) has been further developed and tested in a 2D spatial domain to obtain the toroidal and poloidal periodicities of magnetic fluctuations in a tokamak. The tests are performed essentially using simulated data, because we know what the answer must be , and therefore it is straightforward to verify the accuracy of the algorithm. Two more examples using actual data from the JET and TCV tokamaks are considered to test the algorithm in real-life experiments; a further example using simulated data constructed from nominal test cases for the forthcoming ITER tokamak is also considered. The CPU run-time and the precision of the SparSpec-2D algorithm are studied as function of different analysis parameters. The stability of the algorithm is also tested via the introduction of random errors in the input signal. We find that the spatial-2D version of the baseline SparSpec-1D algorithm accurately finds the modes in the 2D toroidal and poloidal space, provided the set of magnetic sensors used for the analysis do not have a (quasi-)ignorable coordinate. The number of probes and their position are the key parameters that must be optimized for finding correct solutions. The main difficulty, as for the baseline SparSpec-1D algorithm, lies in dealing correctly with the intrinsic measurement uncertainties associated to the input magnetic fluctuation data, particularly the phase error, and this has been already separately reported in a companion work. However, the required CPU run-time for SparSpec-2D is significantly longer than that needed for 2 × SparSpec-1D, and thus SparSpec-2D is effectively suitable for use only when the 2 × 1D analyses cannot provide accurate results, which is the case when the set of measurements does not have an ignorable coordinate.

Internal kinks are a common magneto hydro-dynamic (MHD) instability observed in tokamak operation when the q profile in the plasma core is close to unity. This MHD instability impacts both the transport of the bulk plasma (current, particle and energy transport) and minority species, such as fast ions. In tokamak a configuration variable (TCV) (R (0)/a = 0.88 m/0.25 m) the fast ion population is generated in the plasma by neutral beam tangential injection of energies up to 28 keV. TCV features 16 active shaping coils permitting a great flexibility in plasma shape, including negative triangularity (delta) configurations that show surprisingly high confinement. This study focuses on the transport of fast ions induced by sawteeth, by comparing two triangularity cases and simulation results with experimental data. Comparison of two equilibria with opposite delta shows that the fast ion drifts are larger for delta < 0. Furthermore, the sawtooth-induced transport in this case is larger than delta > 0 in similar conditions. Comparison with experimental data confirms the dominance of the modification of thermal kinetic profiles following the sawtooth crash in explaining drops in the neutron rates and fast ion D-alpha signals. Additional fast ion diffusion, however, improves the interpretation of the experimental data. For delta < 0, the amplitude of the perturbation better representing the experimental data is larger. Finally, an exploratory study for 50 keV particles (soon available in TCV) shows that the situation does not worsen for such particles.

The tokamak à configuration variable (TCV) continues to leverage its unique shaping capabilities, flexible heating systems and modern control system to address critical issues in preparation for ITER and a fusion power plant. For the 2019–20 campaign its configurational flexibility has been enhanced with the installation of removable divertor gas baffles, its diagnostic capabilities with an extensive set of upgrades and its heating systems with new dual frequency gyrotrons. The gas baffles reduce coupling between the divertor and the main chamber and allow for detailed investigations on the role of fuelling in general and, together with upgraded boundary diagnostics, test divertor and edge models in particular. The increased heating capabilities broaden the operational regime to include T e / T i ∼ 1 and have stimulated refocussing studies from L-mode to H-mode across a range of research topics. ITER baseline parameters were reached in type-I ELMy H-modes and alternative regimes with ‘small’ (or no) ELMs explored. Most prominently, negative triangularity was investigated in detail and confirmed as an attractive scenario with H-mode level core confinement but an L-mode edge. Emphasis was also placed on control, where an increased number of observers, actuators and control solutions became available and are now integrated into a generic control framework as will be needed in future devices. The quantity and quality of results of the 2019–20 TCV campaign are a testament to its successful integration within the European research effort alongside a vibrant domestic programme and international collaborations.

This paper presents a dedicated study of plasma-antenna (PA) coupling with the Alfven Eigenmode Active Diagnostic (AEAD) in JET. Stable AEs and their resonant frequencies f, damping rates gamma < 0, and toroidal mode numbers n are measured for various PA separations and limiter versus X-point magnetic configurations. Two stable AEs are observed to be resonantly excited at distinct low and high frequencies in limiter plasmas. The values of f and n do not vary with PA separation. However, |gamma| increases with PA separation for the low-f, but not high-f, mode, yet this may be due to slightly different edge conditions. The high-f AE is detected throughout the transition from limiter to X-point configuration, though its damping rate increases; the low-f mode, on the other hand, becomes unidentifiable. The linear, resistive MHD code CASTOR is used to simulate the frequency scan of an AEAD-like external antenna. For the limiter pulses, the high-f mode is determined to be an n = 0 GAE, while the low-f mode is likely an n = 2 TAE. During the transition from limiter to X-point configuration, CASTOR indicates that n = 1 and 2 EAEs are excited in the edge gap. These results extend previous experimental studies in JET and Alcator C-Mod; validate the computational work performed by Dvornova et al (2020 Phys. Plasmas 27 012507); and provide guidance for the optimization of PA coupling in upcoming JET energetic particle experiments, for which the AEAD will aim to identify the contribution of alpha particles to AE drive during the DT campaign.

A common problem in many complex physical systems is the determination of pulsation modes from irregularly sampled time-series, and there is a wealth of signal processing techniques that are being applied to post-pulse and real-time data analysis in such complex systems. The aim of this report is studying the problem of detecting in real-time spatial periodicities in the spectrum of magnetic fluctuations in tokamaks, for which optimization of the algorithm run-time is essential. The main tool used hereafter will be the SparSpec algorithm, initially devised for astrophysical purposes and already applied to the analysis of magnetic fluctuation data in various tokamaks, both currently or previously operating (JET, TCV, Alcator C-mod) and under construction (ITER, DTT). For JET, the baseline version of the SparSpec algorithm, dubbed SS-H2, already regularly runs in real-time on a 1 ms clock for detecting Toroidal Alfvén Eigenmodes using synchronously-detected magnetic perturbation. It was noted that the solution is only slowly changing in time as the background plasma typically also slowly evolves. Therefore, as a specifically real-time acceleration tool, we will focus on the use of a memory with relaxation scheme, whereby solutions obtained at previous time points are used to provide weighted input constraints for the solution at the current time point. Use of the measurement uncertainties to weight the data, the spectral window and the ensuing penalization criterion (dubbed the SS-V5ν0 algorithm) is reported in a companion work. The behaviour of the SparSpec algorithm under a variety of simulated circumstances, and one actual test case from the JET tokamak, is analysed and appropriate conditions for the convergence of the memory-penalised solutions are derived. The tuning of the input parameters is discussed based on the results of our simulations. It is found that the implementation of SparSpec using such a memory with relaxation scheme is quite a complex procedure, and only when correctly optimized the results are superior, both in terms of the speed and the accuracy of the calculations, to those obtained with the SS-H2 and SS-V5ν0 versions of the SparSpec algorithm.

Within the European development of a future tokamak demonstration fusion power plant (DEMO) [1] the pre-conceptual studies on the plasma diagnostic and control (D&C) system are progressing to prepare the basis for reliable plasma operation at high overall performance [2]. A variety of plasma diagnostics will be employed on DEMO together with advanced control techniques in order to provide an accurate knowledge of the plasma state, which is needed to maintain plasma operation within the allowed physical and technical limits. The integration of diagnostic front-end components has to cope with strong adverse effects arising from neutron and gamma irradiation, heat loads, impinging particles and forces. In this environment, the quality of measurements can only be ensured for longer periods by using robust diagnostic components, mounting them in sufficiently protected (retracted) locations, and any maintenance can only be performed via remote handling. Major open issues are the durability of magnetic measurements in the presence of irradiation induced effects and the feasibility of detachment control under DEMO conditions. In parallel to diagnostic developments, the details of the main control issues are being formulated and investigated by quantitative plasma control simulations. To obtain the envisaged performance DEMO operates close to some physics limits where even small disturbances, if not properly controlled, can trigger major variations of the plasma parameters. Equilibrium control requires high control power and can drive the poloidal field coil system to its operational limits. Within this paper, we will provide an overview on the current status of the ongoing D&C developments for the European DEMO concept.

The Tokamak à Configuration Variable (TCV) tokamak is in the midst of an upgrade to further its capability to investigate conventional and alternative divertor configurations. To that end, modular and removable gas baffles have been installed to decrease the coupling between the divertor and the plasma core. The baffles primarily seek to suppress the transit of recycling neutrals to closed flux surfaces. A first experimental campaign with the gas baffles has shown that the baffled divertor remains compatible with a wide range of configurations including snowflake and super- X divertors. Plasma density ramp experiments reveal an increase of the neutral pressure in the divertor by up to a factor ×5 compared to the unbaffled divertor and thereby qualitatively confirm simulations with the SOLPS-ITER code that were used to guide the baffle design. Together with a range of new and upgraded divertor diagnostics, the baffled TCV divertor is now used to validate divertor models for ITER and next step devices with particular emphasis on geometric variations.

A common problem in many complex physical systems is the determination of pulsation modes from irregularly sampled time-series, and there is a wealth of signal processing techniques that are being applied to post-pulse and real-time data analysis in such complex systems. The aim of this report is studying the problem of detecting discrete spatial periodicities in the spectrum of magnetic fluctuations in tokamaks, for which the optimization of the algorithm performance is essential, particularly when multiple sensors are used with different measurement uncertainties, and some of the processed output signals are then used in real-time for discharge control. The main tool used hereafter will be the SparSpec algorithm, initially devised for astrophysical purposes and already applied to the analysis of magnetic fluctuations in various tokamaks. In its baseline version, dubbed SS-H2, the SparSpec algorithm runs in currently or previously operating tokamaks (JET, TCV and Alcator C-mod), and is foreseen to be deployed for data analysis in tokamak under construction (ITER, DTT). For JET, SS-H2 regularly runs also in real-time on a 1ms clock for detecting Alfvén Eigenmodes using synchronously-measured magnetic perturbations. On JET and TCV, it was noted that often a reduced set of sensors had to be used as the measurement uncertainties were not the same for all available sensors, somewhat deteriorating the overall performance of the algorithm. Hence, as part of a major update of the SparSpec algorithm, specifically intended for accelerating the real-time performance, use of the measurement uncertainties to weight the data, the spectral window and the ensuing penalization criterion was introduced. The behaviour of this new version of the SparSpec algorithm under a variety of simulated circumstances is analysed. It is found that the implementation of SparSpec using such error weighting produces superior results to those obtained with SS-H2, both in terms of the speed and the accuracy of the calculations. A test on actual data from the JET tokamak also shows a clear improvement in the performance of the algorithm.

The effect of plasma shaping on scrape-off layer (SOL) plasma turbulence is investigated through a rigorous validation exercise. Two- and three-dimensional simulations of the SOL plasma dynamics in three TCV limited discharges are carried out with the GBS code. These discharges realize an almost circular magnetic equilibrium, an elongated equilibrium, and an elongated equilibrium with negative triangularity. For the three plasma discharges, three simulations are performed, considering (i) a three-dimensional model with an explicit dependence on elongation, triangularity, and the inverse aspect ratio; (ii) a circular three-dimensional model in the infinite aspect ratio limit; and (iii) a two-dimensional model, which assumes a circular magnetic equilibrium in the infinite aspect ratio limit, cold ions, and interchange driven turbulence in the sheath limited regime. Ten validation observables common to simulations and experimental measurements from a reciprocating probe located at the TCV outer midplane are identified, and the agreement between experimental and numerical results relative to each observable is evaluated. The composite metric introduced by Ricci et al. [Phys. Plasmas 18, 032109 (2011)] is then used to assess the overall agreement between simulations and experimental measurements. It is found that the shaping model implemented in GBS improves the description of SOL plasma turbulence, taking into account the impact of elongation and triangularity.

Innovative high-frequency magnetic sensors have been designed and manufactured in-house for installation on the Tokamak à Configuration Variable (TCV), which are now routinely operational during the TCV experimental campaigns. These sensors combine the Low Temperature Co-fired Ceramic (LTCC) and the classical thick-film technologies and are in various aspects similar to the majority of the in-vessel inductive magnetic sensors foreseen for ITER (around 450 out of the 505 currently being procured are of the LTCC-1D type). The TCV LTCC-3D magnetic sensors provide measurements in the frequency range up to 1 MHz of the perturbations to the wall-aligned toroidal (δBTOR), vertical (δBVER), and radial (δBRAD) magnetic field components. Knowledge of the equilibrium at the last closed flux-surface allows us to then obtain the field-aligned parallel (δBPAR ∼ δBTOR), poloidal (δBPOL), and normal (δBNOR) components, the latter being in most cases rather different from the vertical and radial components, respectively. The main design principles were aimed at increasing the effective area and reducing the self-inductance of the sensor in each of the three measurement axes, which are centered at the same position on each sensor, while reducing the mutual and parasitic coupling between them by optimizing the on-board wiring. The physics requirements are set by the installation of two high-power/high-energy neutral beam injection systems on TCV, i.e., studying fast ions physics, coherent instabilities, and turbulence in the (super-)Alfvénic frequency range. In this paper, we report the manufacturing, installation, and commissioning work for these high-frequency LTCC-3D magnetic sensors and conclude with an overview of illustrative experimental results obtained with this system. The LTCC-3D data provide new insights into the δBPOL coherent (eigenmodes, up to ∼400 kHz) and in-coherent background turbulent fluctuations in the higher frequency range up to ∼1 MHz, which were not previously available with the TCV Mirnov sensors. Furthermore, the LTCC-3D δBPOL measurements allow us to cross-check the data obtained with the standard Mirnov coils and have led to the identification of largeelectromagnetic (EM) noise pick-up for the Mirnov data acquisition (DAQ). When the sources of EM noise pick-up on the Mirnov DAQ are removed, the LTCC-3D data for δBPOL are in good overall agreement, i.e., within the expected measurement uncertainties, with those obtained with the standard Mirnov sensors located at the same poloidal position in the frequency range where the respective data acquisition overlap, routinely up to 125 kHz and up to 250 kHz in some discharges. The LTCC-3D δBPAR measurements (not previously available in TCV or elsewhere) provide evidence that certain instabilities have a finite parallel δB at the wall, hence at the LCFS, consistent with the recent theoretical results for pressure-driven modes. The LTCC-3D δBNOR measurements improve significantly on the corresponding measurements with the saddle loops, which are mounted onto the wall and have a bandwidth of ∼3 kHz (due to the wall penetration time). A detailed end-to-end system modeling tool has been developed and applied to test on the simulated data the actual measurement capabilities of this new diagnostic system and obtain the ensuing estimates of the intrinsic measurement uncertainties. A detailed error analysis is then performed so that, finally, fully calibrated, absolute measurements of the frequency-dependent amplitude and spectral breaks of coherent eigenmodes and in-coherent broadband magnetic fluctuations are provided for the first time in physical units with quantitative uncertainties.