The Divertor Tokamak Test facility (DTT) aims at investigating integrated power exhaust solutions that can be relevant for DEMO and future power plants. Such an ambitious goal imposes several constraints on the engineering design of the actively cooled plasma-facing components (PFCs) of DTT. For instance, the First Wall (FW) must withstand thermal and electromagnetic loads that arise during both normal and off-normal operations of various plasma scenarios. In particular, the Limiter Inboard FW (LIFW), covering 50 % of the IFW, has been designed to cope with plasma limited configurations, i.e. when the plasma interacts with the solid wall. Each module consists of seven long (2.3 m) coaxial pipes made of CuCrZr alloy. Owing to the high heat loads expected, the LIFW PFCs are based on the ITER-like W-monoblock design and the plasma-facing surface, protruding radially towards the plasma with respect to the standard IFW, has a toroidal shaping that helps distribute evenly the heat load. In the present work, the technological limits of the proposed LIFW design are assessed. Based on the hydraulic conditions of the cooling water, the maximum power that can be handled by the LIFW system is evaluated under the assumption of a safety margin from the critical heat flux (CHF). Moreover, the thermostructural behavior of a LIFW unit is simulated in ANSYS under realistic boundary conditions. In this context, a parametric distribution of the thermal load is modelled as a function of the input power and the expected spatial-temporal evolution of the plasma "footprint". Moreover, realistic kinematic boundary conditions, representative of the pinned supports, have been included in the structural integrity assessment of the pipe, carried out according to the ITER SDC-IC design criteria (design-by-analysis approach). Preliminary results suggest that the maximum peak heat flux that can be handled by the LIFW design falls in the range 5-8 MW/m2. This range is compatible with the DTT "Day0" scenario, when, due to the lesser knowledge of machine control, the most critical limiter operations may occur. Nonetheless, studies on the full power scenarios confirmed that in the ramp-up phase the maximum conductive heat load shall be lower than 1 MW/m2 therefore the calculated performances can be considered adequately safe. After the fabrication of small-scale mock-ups, the lifetime of such components will be assessed experimentally, by means of cyclic thermal fatigue high heat flux tests.

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.

Superconducting (SC) tokamak JT-60SA plays an essential role in fusion research and development by supporting and complementing the ITER project, providing directions to the DEMO design activity and fostering next generation scientists and engineers. Since the short circuit incident at the terminal joints of equilibrium field coil #1 during the integrated commissioning (IC) in March 2021, both EU and JA implementing agencies (IAs) have examined how to ensure safe operation of JT-60SA by mitigating the risk of possible discharge occurrence inside the cryostat. Based on the experience of the global Paschen tests, the IAs have established a strategy of risk mitigation measures, which is a combination of (i) reinforcement of insulation, (ii) avoiding unnecessary voltage application to the coil systems and (iii) immediate de-energization of the coils when deteriorated vacuum conditions are detected. Thanks to the considerable efforts of the Integrated Project Team members, the IC restarted in May 2023. After confirmation of the SC state of the coil systems (TF, EF and CS), the coil energization test and the plasma operation phase 1 (OP-1) started. The first plasma was successfully achieved on 23 October 2023 with a limited value of voltage and current applied to the coils. The plasma configuration control was also confirmed with low plasma current and low auxiliary heating power conditions. Based on the IO-F4E-QST collaboration, activities of JT-60SA have been shared with the IO and provided an important lesson for ITER assembly and commissioning, and will provide an outstanding contribution to fusion research at large. After OP-1, maintenance & enhancement phase 1 (M/E-1) starts from January 2024, in which in-vessel components are installed, and heating and diagnostic systems are extensively upgraded to allow a high power heating experiment planned in OP-2. In order to make the best use of JT-60SA, a newly organized JT-60SA experiment team will refine the research plan for the future high heating power operation phase.

The PROTO-SPHERA experiment, built at the CR-ENEA laboratory in Frascati, was in part inspired by the jet + torus astrophysical plasmas, a rather common morphology in Astrophysics. This paper illustrates how the said plasma morphology can be reproduced in a laboratory with the setup of the PROTO-SPHERA experiment. The experiment as such displayed the appearance and sustainment of a plasma torus around an internal magnetized plasma centerpost (jet) by self-organisation; an entirely unexplored phenomenon to date. The remarkable ideal MHD stability of the PROTO-SPHERA plasma is extremely significant, as it is obtained in a simply connected geometry, inside a perfectly insulating vacuum vessel, and without the need of a nearby stabilizing conducting shell. The concluding sections of this paper deal with application of force-free fields to the Pulsar Wind Nebulae morphology and present an extension of the well-known split-dipole model. Such an extension provides a natural description of the presence of tori around the Pulsar plasma jets. In addition, similarities and differences between the laboratory and the astrophysical jet + torus plasmas are detailed.

The mission of WEST (tungsten-W Environment in Steady-state Tokamak) is to explore long pulse operation in a full tungsten (W) environment for preparing next-step fusion devices (ITER and DEMO) with a focus on testing the ITER actively cooled W divertor in tokamak conditions. Following the successful completion of phase 1 (2016-2021), phase 2 started in December 2022 with the lower divertor made entirely of actively cooled ITER-grade tungsten mono-blocks. A boronization prior the first plasma attempt allowed for a smooth startup with the new divertor. Despite the reduced operating window due to tungsten, rapid progress has been made in long pulse operation, resulting in discharges with a pulse length of 100 s and an injected energy of around 300 MJ per discharge. Plasma startup studies were carried out with equatorial boron nitride limiters to compare them with tungsten limiters, while Ion Cyclotron Resonance Heating assisted startup was attempted. High fluence operation in attached regime, which was the main thrust of the first campaigns, already showed the progressive build up of deposits and appearance of dust, impacting the plasma operation as the plasma fluence increased. In total, the cumulated injected energy during the first campaigns reached 43 GJ and the cumulated plasma time exceeded 5 h. Demonstration of controlled X-Point Radiator regime is also reported, opening a promising route for investigating plasma exhaust and plasma-wall interaction issues in more detached regime. This paper summarises the lessons learned from the manufacturing and the first operation of the ITER-grade divertor, describing the progress achieved in optimising operation in a full W environment with a focus on long pulse operation and plasma wall interaction.

Pure-tungsten and tungsten-aluminium films deposited by high power impulse magnetron sputtering (HiPIMS) on copper-chromium-zirconium substrates were investigated as protective coatings against liquid tin corrosion, a critical issue for nuclear fusion applications. The growth of pure-tungsten coatings was controlled by using a negative substrate bias synchronized to the HiPIMS pulse onset, resulting in columnar films with various degree of compactness and crystallinity according to the set bias amplitude (0, 400 and 800 V). Differently, the cosputtering of W and Al favored the formation of an amorphous layer with a compact morphology. During liquid tin corrosion experiments at 400 degrees C for up to 600 min, all produced coatings were not dissolved, but different protective performances were observed after localized liquid tin interaction. Pure-W coated samples suffered from tin penetration after brittle failure of the protective layer. On the contrary, under the same experimental condition, W-Al coatings proved to be effective in limiting liquid tin attack.

In astrophysics, the boundary conditions for plasma phenomena are provided by nature and the astronomer faces the problem of understanding them from a variety of observations [Hester J J et al 1996 Astrophys. J. 456 225], on the other hand, in laboratory plasma experiments the electromagnetic boundary conditions become a major problem in the set-up of the machine that produces the plasma, an issue that has to be investigated step by step and to be modified and adapted with great patience, in particular in the case of an innovative plasma confinement experiment. The PROTO-SPHERA machine [Alladio F et al 2006 Nucl. Fusion 46 S613] is a magnetic confinement experiment, that emulates in the laboratory the jet + torus plasma configurations often observed in astrophysics: an inner magnetized jet of plasma centered on the (approximate) axis of symmetry and surrounded by a magnetized plasma torus orthogonal to this jet. The PROTO-SPHERA plasma is simply connected, i.e., no metal current conducting rod is linked to the plasma torus, while instead it is the inner magnetized plasma jet (in the following always called the plasma centerpost) that is linked to the torus. It is mandatory that no spurious plasma current path modifies the optimal shape of the plasma centerpost. Moreover, as the plasma torus is produced and sustained, in absence of any applied inductive electric field, by the inner plasma centerpost through magnetic reconnections [Taylor J B and Turner M F 1989 Nucl. Fusion 29 219], it is required as well that spurious current paths do not surround the torus on its outboard, in order not to lower the efficiency of the magnetic reconnections that maintain the plasma torus at the expense of the plasma centerpost. Boundary conditions have been corrected, up to the point that the first sustainment in steady state has been achieved for the combined plasma.

Wall conditioning plays a key role to assist the discharge start-up and to achieve high-performance plasma scenarios, especially after machine vents and impurity events. This article describes the conditioning systems of the Divertor Test Tokamak (DTT) by focusing on the design of the baking and glow discharge conditioning systems. A general overview of the current design is presented by discussing the main working parameters and the features of each technology under investigation. The integration into the machine is also discussed, by considering the main limitations and interface with other subsystems, particularly with the diagnostic and the pumping system.

Four different tin-wetted, tungsten CPS (Capillary Porous System) targets where exposed to NBI pulses in the OLMAT High Heat flux (HHF) facility. They include two flexible ones placed on a TZM support (W meshes and W felt) and two compact ones (sintered W disk and 3D printed W). A comparative study was performed using a fastframe imaging camera and an infrared pyrometer. Surface temperature increase and homogeneity, particle ejection, CPS damage and overall behaviour were studied for each case. Sn drop/accumulation at the lower part of the targets was observed for all cases when they are heated up to around 400 C except for the 3D printed W target that has its own Sn deposit. The 3D printed W target presented the best results in all aspects, withstanding heat pulses up to 58 MW/m2 in 100 ms without any damage or particle ejection. On the other hand, the W mesh targets displayed damage at 20 MW/m2 due to a bad thermal contact with the deposit, while the sintered W disk developed a crack during a series of 15 MW/m2 NBI pulses. As might be expected, a reduced increase of temperature during pulses is observed for the two compact W targets. The results and their relevance for the design of a Sn wetted W CPS for application as a DEMO-divertor material are discussed in the present work.

Plasma free boundary simulation codes including light transport and current diffusion are an important mean to develop plasma scenarios and control algorithms in view of the operation of large tokamaks with exploratory scientific and technological purposes like ITER and JT-60SA. In particular, the intrinsic flexibility of the latter will offer the possibility to study a variety of scenarios including advanced scenarios which could be extrapolated to a steady-state demonstration of fusion power plants. The development of a numerical simulator is the object of this paper. The ultimate objective of producing a reasonably fast "flight simulator"is twofold: provide to scientists a tool for integrated scenario design including control laws and supply to tokamak operators an instrument for pulse discharge training. The option explored is obtained by coupling METIS with CREATE-NL+.

In this work we present the results of the LIBS surface mapping and depth profiling of the Frascati Tokamak Upgrade (FTU) tiles of the poloidal limiter, made of a Molybdenum alloy (TZM) and removed from the reactor after its last experimental campaign with fusion plasma (late 2019). The measurements were performed by using a spectrometer array that covers the range 200-780 nm and a high-resolution spectrometer (0.1 & ANGS; at 500 nm) covering a spectral window of 10 nm. By applying a depth profiling procedure, the contamination of the tiles by B, Ca, Cr, Fe, K, Li, Mn, Mo, Ni, Ti was observed in their surface layers. Contamination with H, O, N was instead observed in deeper layers after some laser shots. B was attributed by the preliminary boronization procedure of the first wall (FW), carried out with deuterated diborane (D2B6), although D was not detected on the surface. Li was detected as deposited layer coming from previous experiments with liquid metals limiters in FTU. Ca and K are believed to come from the manipulation of the tiles after FTU decommissioning or as residual impurities of the deposited together with Li. H, O and N, from exposure of the tiles to the environmental gases, Mo and Ti were detected as main constituents of the tiles but also as eroded material during exposure to fusion plasma. Cr, Fe, Mn and Ni as eroded materials from the vacuum vessel (VV). In all cases, the surface concentration and thickness of the contaminated layer strongly varies across the TZM tile, showing a tendency of major impurity accumulations in areas more distant from the reactor's equatorial section.

The investigation of the energy frontier in physics requires novel concepts for future colliders. The idea of a muon collider is very appealing since it would allow to study particle collisions at up to tens of TeV energy, while offering a cleaner experimental environment with respect to hadronic colliders. One key element in the muon collider design is the low-emittance muon production. Recently,the Low EMittance Muon Accelerator (LEMMA) collaboration has explored the muon pair production close to its kinematic threshold by annihilating 45 GeV positrons with electrons in a low Z material target. In this configuration, muons are emerging from the target with a naturally low-emittance. In this paper we describe the performance of a system, to study this production mechanism, that consists in several segmented absorbers with alternating active layers composed of fast Cherenkov detectors together with a muon identification technique based on this detector. Passive layers were made of tungsten. We collected data corresponding to muon and electron beams produced at the H2 line in the North Area of the European Organization for Nuclear Research (CERN) in September 2018.

Liquid metal based divertors could be a more robust alternative to a solid tungsten design for DEMO. The liquid is confined in a sponge-like tungsten layer, called a capillary porous structure (CPS). It has been found previously that under certain conditions, many tin droplets eject from a CPS when it is brought into contact with a hydrogen plasma. These would present a contamination issue for the plasma core. Stability analysis suggests that droplet ejection can be suppressed by reduction of the pore size. To test this, stainless-steel CPS targets with pore size ranging from 0.5–100µm filled with tin were exposed to identical loading conditions. This was done in the linear plasma device Magnum-PSI, capable of reaching divertor relevant plasma conditions. Furthermore, the influence of the CPS manufacturing techniques is considered by comparing the performance of a 3D printed, a mesh felts and a sintered CPS, all made from tungsten. Each target was surrounded by four witness plates, which were analysed post-mortem for Sn content by Rutherford backscattering. During plasma exposure, tin droplets were observed using a fast visible camera and plasma light emission via survey optical emission spectroscopy. The results imply that Sn erosion can be reduced by a factor of 50 when reducing the pore size. Moreover, it highlights the importance of avoiding overfilling of CPS targets with Sn.

COMPASS addressed several physical processes that may explain the behaviour of important phenomena. This paper presents results related to the main fields of COMPASS research obtained in the recent two years, including studies of turbulence, L-H transition, plasma material interaction, runaway electron, and disruption physics: Tomographic reconstruction of the edge/SOL turbulence observed by a fast visible camera allowed to visualize turbulent structures without perturbing the plasma. Dependence of the power threshold on the X-point height was studied and related role of radial electric field in the edge/SOL plasma was identified. The effect of high-field-side error fields on the L-H transition was investigated in order to assess the influence of the central solenoid misalignment and the possibility to compensate these error fields by low-field-side coils. Results of fast measurements of electron temperature during ELMs show the ELM peak values at the divertor are around 80% of the initial temperature at the pedestal. Liquid metals were used for the first time as plasma facing material in ELMy H-mode in the tokamak divertor. Good power handling capability was observed for heat fluxes up to 12 MW m(-2) and no direct droplet ejection was observed. Partial detachment regime was achieved by impurity seeding in the divertor. The evolution of the heat flux footprint at the outer target was studied. Runaway electrons were studied using new unique systems-impact calorimetry, carbon pellet injection technique, wide variety of magnetic perturbations. Radial feedback control was imposed on the beam. Forces during plasma disruptions were monitored by a number of new diagnostics for vacuum vessel (VV) motion in order to contribute to the scaling laws of sideways disruption forces for ITER. Current flows towards the divertor tiles, incl. possible short-circuiting through PFCs, were investigated during the VDE experiments. The results support ATEC model and improve understanding of disruption loads.

Since the 2018 IAEA FEC Conference, FTU operations have been devoted to several experiments covering a large range of topics, from the investigation of the behaviour of a liquid tin limiter to the runaway electrons mitigation and control and to the stabilization of tearing modes by electron cyclotron heating and by pellet injection. Other experiments have involved the spectroscopy of heavy metal ions, the electron density peaking in helium doped plasmas, the electron cyclotron assisted start-up and the electron temperature measurements in high temperature plasmas. The effectiveness of the laser induced breakdown spectroscopy system has been demonstrated and the new capabilities of the runaway electron imaging spectrometry system for in-flight runaways studies have been explored. Finally, a high resolution saddle coil array for MHD analysis and UV and SXR diamond detectors have been successfully tested on different plasma scenarios.

A large superconducting machine, JT-60SA has been constructed to provide major contributions to the ITER program and DEMO design. For the success of the ITER project and fusion reactor, understanding and development of plasma controllability in ITER and DEMO relevant higher beta regimes are essential. JT-60SA has focused the program on the plasma controllability for scenario development and risk mitigation in ITER as well as on investigating DEMO relevant regimes. This paper summarizes the high research priorities and strategy for the JT-60SA project. Recent works on simulation studies to prepare the plasma physics and control experiments are presented, such as plasma breakdown and equilibrium controls, hybrid and steady-state scenario development, and risk mitigation techniques. Contributions of JT-60SA to ITER and DEMO have been clarified through those studies.

The problem of identifying and therefore modelling a complex system makes use of various techniques and strategies whose computational efforts change drastically. It is not straightforward to analyse the complexity of a system as a whole because of myriads of factors, such as the way of arranging its constituent items and how they interact mutually. Intuitively, the bigger the set of sub-parts is, the more numerous the degrees of freedom are. Additionally there is not a specific and global criterion for optimally determining an always-working method that makes the identification procedure easier, especially in those contexts where the number of unknown variables can make the difference. In this sense, plasma physics is not an exception, being a field where complex phenomena, such as plasma instabilities, easily arise. From a systemic, high-level perspective, the possibility of employing a model that can describe these behaviours is particularly appealing, since it can be exploited for control applications that have not to neglect the underlying physical nature. So far, most of the work published in literature has focused on more physically-grounded models, which could describe how plasma physics works in detail, but very little has been done as mentioned before, with the aim of providing a computational, yet system-oriented, insight of these physical systems. Starting from real flux measurements recorded thanks to suitable sensors installed inside Tokamak machines, the paper attempts to provide a solution based on already known tools available in literature to solve the aforementioned problem, by combining both machine learning-based strategies for dimensionality reduction and control theory. More in detail, the whole architecture presented in this work is founded on the use of auto-encoders, which are intrinsically capable of compressing input features thanks to their structure, and Hammerstein–Wiener models, which are structurally endowed with both linear and non-linear sub-modellers for better capturing the whole dynamics to identify. By merging these functional blocks, it is possible to address both the issue of establishing the most relevant sub-set of variables for identification and the identification problem itself, resulting in a fully customisable approach to data-driven modelling.

Following ELMy H-mode experiments with liquid metal divertor target on the COMPASS tokamak, we predict the behavior of a similar target on COMPASS Upgrade, where it will be exposed to surface heat fluxes even higher than those expected in the future EU DEMO attached divertor. We simulate the heat conduction, sputtering, evaporation, excitation and radiation of lithium and tin in the divertor area. Measured high-resolution data from COMPASS tokamak were rescaled towards the Upgrade based on many established scalings. Our simulation then yields the amount of released metal which ranges from 4 mg s(-1) upto 12 g s(-1) depending mainly on the geometry and Li/Sn choice, quite independently from active cooling or strike point sweeping.

PROTO-SPHERA is a magnetic confinement device for controlled fusion experiments which forms a plasma spherical torus not around a metal centerpost (as in spherical tokamaks), but around a plasma centerpost. Thanks to a vacuum vessel in polymethylmethacrylate (PMMA) transparent to visible light, a system of six fast visible light cameras has been implemented around the PROTO-SPHERA experiment so as to perform an optical tomography. To do so, synchronous triggered frames acquisition and exposure time control have been performed using the MARTe2 software framework. Also, a proper calibration/reconstruction of each line of sight for all the cameras has been implemented, so as to compute a 3D representation of the vessel. A 2D tomography inversion of the plasma column equatorial plane has been performed under the consideration that the value of each pixel represents the integrated brightness along its line of sight. The camera acquisition system, the experimental setup and the obtained results will be presented.