Recent results from MAST Upgrade are presented, emphasising understanding the capabilities of this new device and deepening understanding of key physics issues for the operation of ITER and the design of future fusion power plants. The impact of MHD instabilities on fast ion confinement have been studied, including the first observation of fast ion losses correlated with Compressional and Global Alfvén Eigenmodes. High-performance plasma scenarios have been developed by tailoring the early plasma current ramp phase to avoid internal reconnection events, resulting in a more monotonic q profile with low central shear. The impact of m / n = 3/2, 2/1 and 1/1 modes on thermal plasma confinement and rotation profiles has been quantified, and scenarios optimised to avoid them have transiently reached values of normalised beta approaching 4.2. In pedestal and ELM physics, a maximum pedestal top temperature of ∼350 eV has been achieved, exceeding the value achieved on MAST at similar heating power. Mitigation of type-I ELMs with n = 1 RMPs has been observed. Studies of plasma exhaust have concentrated on comparing conventional and Super-X divertor configurations, while X-point target, X-divertor and snowflake configurations have been developed and studied in parallel. In L-mode discharges, the separatrix density required to detach the outer divertors is approximately a factor 2 lower in the Super-X than the conventional configuration, in agreement with simulations. Detailed analysis of spectroscopy data from studies of the Super-X configuration reveal the importance of including plasma-molecule interactions and D _2 Fulcher band emission to properly quantify the rates of ionisation, plasma-molecule interactions and volumetric recombination processes governing divertor detachment. In H-mode with conventional and Super-X configurations, the outer divertors are attached in the former and detached in the latter with no impact on core or pedestal confinement.

Real-time magnetic control has been developed to deliver precise control of multiple plasma shape parameters for advanced divertor configurations, including double-null, Super-X, X-point target and X-divertor for the first time on the MAST Upgrade (MAST-U) spherical tokamak. Successful real-time magnetic equilibrium control of different plasma shape variables has been accomplished in the 2022–2023 MAST-U experimental campaign for the advanced MAST-U divertor configurations. Application of the MAST-U boundary reconstruction algorithm, LEMUR, is described and compared with off-line equilibrium reconstruction and diagnostic measurements. The process of design and verification of the axisymmetric magnetic control schemes using a suite of control analysis tools (known collectively as TokSys) is also described.

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.

Reliable vertical position control will be an essential element of any future tokamak-based fusion power plant in order to reduce disruptions and maximize performance. We investigate methods to improve vertical controllability boundary determination in plasma operational space and demonstrate a data-driven approach based on direct pseudoinversion of operational space data that is rigorously quantitative, applicable in real-time plasma control systems, and physically intuitive to interpret. Applied to historical shot data from entire run campaigns on the MAST-U, KSTAR, and NSTX tokamaks, this approach, implemented in DECAF, improves vertical displacement event identification accuracy to 98.9%-100%. Further, we explore the application of a physics-based vertical stability metric as an early warning forecaster for vertical displacement events. The development of a linear surrogate model for the plasma current density profile, with a coefficient of determination of 0.992 on the training dataset, enables potential employment of this forecaster in real-time. The application of this approach on historical data from the MAST-U MU02 campaign yields a forecaster with 62.6% accuracy, indicating promise for this method when further refined and potentially coupled with other stability metrics.

The power exhaust proves to be one of the most challenging and concept–defining aspects in the design of a commercial fusion power plant, while the magnetic coil system, capable of supporting advanced exhaust solutions, emerges as one of the main design and cost drivers. Consequently, much effort should be dedicated to the optimisation of a robust global magnetic configuration, which integrates both the plasma and edge scenarios, while ensuring engineering feasibility and compatibility with the available technology. Here we present a multidisciplinary framework employed to analyse, evaluate, and optimise the Spherical Tokamak for Energy Production (STEP) equilibrium configuration, coupled with a viable divertor solution, and a compatible poloidal field coil system. The complexity of this task leads to a multitude of potentially conflicting requirements and competing constraints. We identify interfaces and conflicts between the aspects of the design that were previously considered independently, and highlight the key benefits, trends, and trade–offs between alternative configurations. We demonstrate that advanced exhaust solutions, simultaneously applied to both inboard and outboard divertors, are accessible with feasible coil sets under conditions relevant for STEP. We show that the most promising inner–X geometry, paired with the outer super–X configuration, can significantly enhance divertor’s power handling capability, allowing access to stable detached regimes. The coil set feasibility is further assessed considering its compatibility with the assumed plasma initiation scenario, and with the most demanding plasma current density profiles utilising alternative heating and current drive schemes.

A key aspect in the design of a spherical tokamak reactor is the optimisation of the plasma equilibrium, together with a compatible divertor configuration, and the corresponding poloidal field system. This is a complex multi-disciplinary problem, integrating plasma physics and engineering in order to satisfy a multitude of often conflicting requirements and constraints. The equilibrium design process employed in this work takes into account the reference plasma operating scenario, the power exhaust solution, and the engineering limits. Managing the heat exhaust proves to be one of the most challenging issues in a compact device such as the UKAEA STEP reactor. With a smaller major radius, the available target area over which the scrape-off layer heat load must be deposited is relatively small as compared to conventional aspect ratio devices. Consequently, alternative and advanced divertor concepts need to be considered, having significant implications for the whole reactor design. Here we address the inner divertor power handling challenges. With very limited inboard space, and the small radius of the inner strike point, the associated heat loads are likely to exceed the power handling capacity of standard divertors (SD). An alternative divertor configuration approaching an X-divertor (XD), created by inducing a secondary X-point near the inner strike point, is compared with an SD configuration optimised for the maximal connection length and maximal poloidal flux expansion. The inner X-divertor, simultaneously achieving strong poloidal flux expansion, increased connection length and higher divertor volume, proved to be advantageous in reducing target heat loads and favouring detachment. Amongst a number of viable exhaust solutions considered, the inner-X divertor is indeed emerging as a promising candidate.

We present the first analysis of the atomic and molecular processes at play during detachment in the MAST-U Super-X divertor using divertor spectroscopy data. Our analysis indicates detachment in the MAST-U Super-X divertor can be separated into four sequential phases: first, the ionisation region detaches from the target at detachment onset leaving a region of increased molecular densities downstream. The plasma interacts with these molecules, resulting in molecular ions ( D 2 + and/or D 2 − → D + D − ) that further react with the plasma leading to molecular activated recombination and dissociation (MAR and MAD), which results in excited atoms and significant Balmer line emission. Second, the MAR region detaches from the target leaving a sub-eV temperature region downstream. Third, an onset of strong emission from electron–ion recombination (EIR) ensues. Finally, the electron density decays near the target, resulting in the bulk of the electron density moving upstream. The analysis in this paper indicates that plasma–molecule interactions have a larger impact than previously reported and play a critical role in the intensity and interpretation of hydrogen atomic line emission characteristics on MAST-U. Furthermore, we find that the Fulcher band emission profile in the divertor can be used as a proxy for the ionisation region and may also be employed as a plasma temperature diagnostic for improving the separation of hydrogenic emission arising from electron-impact excitation and that from plasma–molecular interactions. We provide evidences for the presence of low electron temperatures (≪0.5 eV) during detachment phases III–IV based on quantitative spectroscopy analysis, a Boltzmann relation of the high- n Balmer line transitions together with an analysis of the brightness of high- n Balmer lines.

Managing the heat exhaust presents one of the main challenges in the design of a spherical tokamak reactor. The development of a robust exhaust solution is intrinsically linked to the optimisation of the plasma equilibrium, together with a compatible poloidal field (PF) system. A multidisciplinary design framework should balance a host of often conflicting physics and engineering requirements and identify a set of self -consistent constraints. This work employs an iterative approach which integrates the core plasma scenario and divertor magnetic topology, ensuring compatibility with the available technology. The aim of this paper is to develop and assess a number of optimal magnetic configurations consistent with the UKAEA STEP Prototype Plant designs. Firstly, we use the Fiesta free-boundary equilibrium code to optimise the global configuration for a chosen double-null scenario, employing both standard and alternative exhaust-handling schemes. We then analyse the local scrape-off layer (SOL) magnetic topology in order to preliminarily assess the anticipated divertor performance. We discuss the advantages and disadvantages of each configuration in terms of the selected divertor metrics, the effects on the core plasma scenario, engineering feasibility, and the implications for whole reactor design.The inner divertor proves to be the most challenging issue in a compact device such as STEP. The small target area and reduced SOL width at the small strike-point radius lead to excessive heat loads beyond the power-handling capacity of standard concepts. We propose an alternative inner X-divertor configuration and compare it with the standard divertor optimised for maximal connection length and poloidal flux expansion. The objective is to achieve the optimal global configuration with a realistic coil set, taking into account space constraints and material limitations. We discuss the feasibility of each considered exhaust solution in STEP, where the PF shaping coils can be placed either in unfavourable locations outside of the toroidal field (TF) coil or mounted inside the TF and operating near their engineering limits.

First snowflake (SF) divertor experiments in the MAST-U tokamak demonstrated steady-state snowflake-plus divertor configurations in 450 kA ohmic L-mode plasmas. The SF divertor configuration features a second poloidal field (PF) null in the divertor region close by or overlapping with the main X-point. The resulting low PF region and two additional divertor legs (strike points) may lead to additional power and particle flux sharing via a hypothesized convective cell, and increased plasma-wetted area and radiation. The free-boundary Grad–Shafranov equilibrium code FIESTA was used to design SF configurations with several inter-null distances and orientations. In the experiment, the SF configurations with inter-null distances 0.13–0.20 m and lasting 0.2–0.3 s were obtained. Parallel connection lengths between the midplane and the outer strike point in the SF configurations (evaluated at field lines 1–2 mm from the separatrix in the midplane) were 25–30 m, higher than in the standard divertor (20–25 m) or the Super-X divertor (25 m). Diagnostic measurements highlighted salient SF features. The infra-red video bolometer diagnostic showed that the radiated power peaking in the PF null region was not as pronounces as in the standard divertor. Divertor ion fluxes measured by target Langmuir probes showed increased ion flux in the plate region where a secondary SF strike point landed, concomitantly with the SF formation. These measurements may suggest that some particle and heat redistribution was taking place in the convective SF zone. The first SF experiments provide a basis for future SF studies in MAST-U tokamak with higher input power, improved plasma control and diagnostic measurements, to be compared with the modeling predictions of plasma convective SF mixing and lower density strike point detachment threshold.

The SMall Aspect Ratio Tokamak (SMART) device is a new compact (plasma major radius R-geo >= 0.40 m, minor radius a >= 0.20 m, aspect ratio A >= 1.7) spherical tokamak, currently in development at the University of Seville. The SMART device has been designed to achieve a magnetic field at the plasma center of up to B-phi = 1.0 with plasma currents up to I-p = 500 kA and a pulse length up to tau(ft) = 500 ms. A wide range of plasma shaping configurations are envisaged, including triangularities between -0.50 <= delta <= 0.50 and elongations of kappa <= 2.25. Control of plasma shaping is achieved through four axially variable poloidal field coils (PF), and four fixed divertor (Div) coils, nominally allowing operation in lower-single null, upper-single null and double-null configurations. This work examines phase 2 of the SMART device, presenting a baseline reference equilibrium and two highly-shaped triangular equilibria. The relevant PF and Div coil current waveforms are also presented. Equilibria are obtained via an axisymmetric Grad-Shafranov force balance solver (Fiesta), in combination with a circuit equation rigid current displacement model (RZIp) to obtain time-resolved vessel and plasma currents.

COMPASS Upgrade is a new medium size, high magnetic field tokamak (R = 0.9 m, Bt = 5 T, Ip = 2 MA) currently under design in the Czech Republic. It will provide unique capabilities for addressing some of the key challenges in plasma exhaust physics, advanced confinement modes and advanced plasma configurations as well as testing new plasma facing materials and liquid metal divertor concepts. This paper contains an overview of the preliminary engineering design of the main systems of the COMPASS Upgrade tokamak (vacuum vessel, central solenoid and poloidal field coils, toroidal field coils, support structure, cryostat, cryogenic system, power supply system and machine monitoring and protection system). The description of foreseen auxiliary plasma heating systems and plasma diagnostics is also provided as well as a summary of expected plasma performance and available plasma configurations.

In magnetic fusion devices, unwanted non-axisymmetric magnetic field perturbations, known as error fields (EF), can have detrimental effects on plasma stability and confinement. Such EFs may originate from several sources, such as axi-symmetric coil misalignments, coil feeds, 3D structures in the wall surrounding the plasma, presence of ferromagnetic materials near the plasma surface, blanket materials in future devices. To minimize their impact on plasma performance and on the available operational space, it is important to identify the EF sources and develop EF control strategies. MAST Upgrade (MAST-U) is a spherical tokamak which will operate in the near future after a series of enhancements from the previous MAST experiment (Morris A.W. et al., MAST accomplishments and upgrade for fusion next-steps. IEEE Transactions on Plasma Science, April 2014). To deliver a machine with EF amplitude low enough to allow a high quality experimental programme, systematic analysis of the intrinsic EF sources has been carried out for poloidal field (P) and divertor (D) coils. To deliver a machine with EF amplitude low enough to allow a high quality experimental programme, systematic analysis of the intrinsic EF sources has been carried out for poloidal field (P) and divertor (D) coils, whose magnetic field measurements were available when writing this paper. Such measurements reveal that P and D coils are 3D deformed and thus are responsible for intrinsic EFs, with mainly n = 1 and n = 2 toroidal mode numbers. In preparation to MAST-U operation, both passive and active EF control strategies have been adopted for n = 1 and n = 2 EFs compensation. Passive EF control consisted of installing finely each D and P coil within MAST-U device so as to minimize the intrinsic n = 1 EF amplitude. The optimal coil alignment has been determined based on magnetic field measurements and the corresponding 3D electro-magnetic modelling, and envisaged coil shift and tilt of 3.2 mm and 0.7 mrad, respectively, in the case of P coil named P4. Conversely, active control will be use during MAST-U operation to reduce the n = 2 EF which is associated mainly with P4 and P5, as well (Kirk A. et al. 2014 Plasma Phys. Control. Fusion 56, 104003). Since these coils have been re-used from the MAST device, studies attempting n = 2 EF control, based on MAST plasmas, have been modelled utilizing the MARS-F code (Liu Y.Q. et al. 2000 Phys. Plasmas 7, 3681) to interpret experimental results in MAST and to give hints for future n = 2 EF control studies in MAST-U. A model-based control set for n = 2 EF control has been identified which would allow for the minimization of rotation braking, of the resonant magnetic field at th = agand of the plasma displacement in MAST-U, simultaneously.

The mega amp spherical tokamak (MAST) was a low aspect ratio device (R/a = 0.85/0.65 similar to 1.3) with similar poloidal cross-section to other medium-size tokamaks. The physics programme concentrates on addressing key physics issues for the operation of ITER, design of DEMO and future spherical tokamaks by utilising high resolution diagnostic measurements closely coupled with theory and modelling to significantly advance our understanding. An empirical scaling of the energy confinement time that favours higher power, lower collisionality devices is consistent with gyrokinetic modelling of electron scale turbulence. Measurements of ion scale turbulence with beam emission spectroscopy and gyrokinetic modelling in up-down symmetric plasmas find that the symmetry of the turbulence is broken by flow shear. Near the non-linear stability threshold, flow shear tilts the density fluctuation correlation function and skews the fluctuation amplitude distribution. Results from fast particle physics studies include the observation that sawteeth are found to redistribute passing and trapped fast particles injected from neutral beam injectors in equal measure, suggesting that resonances between the m = 1 perturbation and the fast ion orbits may be playing a dominant role in the fast ion transport. Measured D-D fusion products from a neutron camera and a charged fusion product detector are 40% lower than predictions from TRANSP/NUBEAM, highlighting possible deficiencies in the guiding centre approximation. Modelling of fast ion losses in the presence of resonant magnetic perturbations (RMPs) can reproduce trends observed in experiments when the plasma response and charge-exchange losses are accounted for. Measurements with a neutral particle analyser during merging-compression start-up indicate the acceleration of ions and electrons. Transport at the plasma edge has been improved through reciprocating probe measurements that have characterised a geodesic acoustic mode at the edge of an ohmic L-mode plasma and particle-in-cell modelling has improved the interpretation of plasma potential estimates from ball-pen probes. The application of RMPs leads to a reduction in particle confinement in L-mode and H-mode and an increase in the core ionization source. The ejection of secondary filaments following type-I ELMs correlates with interactions with surfaces near the X-point. Simulations of the interaction between pairs of filaments in the scrape-off layer suggest this results in modest changes to their velocity, and in most cases can be treated as moving independently. A stochastic model of scrape-off layer profile formation based on the superposition of non-interacting filaments is in good agreement with measured time-average profiles. Transport in the divertor has been improved through fast camera imaging, indicating the presence of a quiescent region devoid of filament near the X-point, extending from the separatrix to psi(n) similar to 1.02. Simulations of turbulent transport in the divertor show that the angle between the divertor leg on the curvature vector strongly influences transport into the private flux region via the interchange mechanism. Coherence imaging measurements show counter-streaming flows of impurities due to gas puffing increasing the pressure on field lines where the gas is ionised. MAST Upgrade is based on the original MAST device, with substantially improved capabilities to operate with a Super-X divertor to test extended divertor leg concepts. SOLPS-ITER modelling predicts the detachment threshold will be reduced by more than a factor of 2, in terms of upstream density, in the Super-X compared with a conventional configuration and that the radiation front movement is passively stabilised before it reaches the X-point. 1D fluid modelling reveals the key role of momentum and power loss mechanisms in governing detachment onset and evolution. Analytic modelling indicates that long legs placed at large major radius, or equivalently low B at the target compared with the X-point arc more amenable to external control. With MAST Upgrade experiments expected in 2019, a thorough characterisation of the sources of the intrinsic error field has been carried out and a mitigation strategy developed.

In magnetic fusion devices, unwanted nonaxisymmetric magnetic field perturbations, known as error fields (EF), can have detrimental effects on plasma stability and confinement. To minimize their impact on plasma performance and on the available operational space, it is important to identify the EF sources and develop EF control strategies. MAST Upgrade is a spherical tokamak which will operate in the near future after a series of enhancements from the previous MAST experiment and will contribute to the programme of the exhaust and fast particle physics [Morris A. W. et al, ”MAST Accomplishments and Upgrade for Fusion Next-Steps” IEEE Transactions on Plasma Science, April 2014]. To deliver a machine with EF amplitude low enough to allow a high quality experimental programme, a careful analysis of the intrinsic EF sources in the poloidal field and in the divertor coils has been carried out. The 3D coil deformation has been characterized through high accuracy magnetic field measurements which reveal that the main EF harmonics have n=1 and n=2 toroidal mode numbers. Passive and active control strategies have been adopted to compensate them in preparation for MAST Upgrade operations. The passive EF control consisted in finding the optimal fine-scale coil alignment so as to minimize the intrinsic n=1 EF amplitude. The optimal coil alignment has been determined based on magnetic field measurements and the corresponding 3D electro-magnetic modelling. Conversely, active control will be adopted during MAST Upgrade operation to reduce the n=2 EF which is associated mainly to the manufacturing of the poloidal field coils named P4 and P5 [Kirk A. et al 2014 Plasma Phys. Control. Fusion 56 104003]. Since these coils have been re-used from the MAST device, the experience gained on n=2 EF control, from experimental studies and the corresponding ERGOS [Nardon E. et al 2007 J. Nucl. Mater. 363-365 1071] and MARS-F [Liu Y. Q. et al 2000 Phys. Plasmas 7 3681] modelling, has allowed the identification of the best n=2 EF control technique and to perform a modelbased optimization of its control parameters in preparation to MAST Upgrade operation.

New results from MAST are presented that focus on validating models in order to extrapolate to future devices. Measurements during start-up experiments have shown how the bulk ion temperature rise scales with the square of the reconnecting field. During the current ramp up models are not able to correctly predict the current diffusion. Experiments have been performed looking at edge and core turbulence. At the edge detailed studies have revealed how filament characteristic are responsible for determining the near and far SOL density profiles. In the core the intrinsic rotation and electron scale turbulence have been measured. The role that the fast ion gradient has on redistributing fast ions through fishbone modes has led to a redesign of the neutral beam injector on MAST Upgrade. In H-mode the turbulence at the pedestal top has been shown to be consistent with being due to electron temperature gradient modes. A reconnection process appears to occur during ELMs and the number of filaments released determines the power profile at the divertor. Resonant magnetic perturbations can mitigate ELMs provided the edge peeling response is maximised and the core kink response minimised. The mitigation of intrinsic error fields with toroidal mode number n>1 has been shown to be important for plasma performance.

This paper reports the progress made at JET-ILW on integrating the requirements of the reference ITER baseline scenario with normalized confinement factor of 1, at a normalized pressure of 1.8 together with partially detached divertor whilst maintaining these conditions over many energy confinement times. The 2.5 MA high triangularity ELMy H-modes are studied with two different divertor configurations with D-gas injection and nitrogen seeding. The power load reduction with N seeding is reported. The relationship between an increase in energy confinement and pedestal pressure with triangularity is investigated. The operational space of both plasma configurations is studied together with the ELM energy losses and stability of the pedestal of unseeded and seeded plasmas. The achievement of stationary plasma conditions over many energy confinement times is also reported.

Alfvén cyclotron instabilities excited by velocity gradients of energetic beam ions were investigated in MAST experiments with super-Alfvénic neutral beam injection over a wide range of toroidal magnetic fields from ∼0.34 T to ∼0.585 T. In MAST discharges with high magnetic field, a discrete spectrum of modes in the sub-cyclotron frequency range is excited toroidally propagating counter to the beam and plasma current (toroidal mode numbers n < 0). At lower magnetic field ≤0.45 T, a discrete spectrum of Compressional Alfvén Eigenmodes (CAEs) with n > 0 arises, in addition to the modes with n < 0. At lowest magnetic fields, the CAEs with n > 0 become dominant, they are observed in frequency range from ∼250 kHz for n=1 to ∼3.5 MHz for n=15, well above the on-axis ion cyclotron frequency (∼2.5 MHz). The data is interpreted in terms of normal and anomalous Doppler resonances modified by magnetic drift terms due to inhomogeneity and curvature of the magnetic field. A Hall MHD model is applied for computing the eigenfrequencies and the spatial mode structure of CAEs and a good agreement with the experimental frequencies is found.