In magnetic confinement fusion facilities, the reliable operation of COntrol and Data Acquisition Systems (CODAS) requires embedded solutions capable of ensuring deterministic performance within hard real-time constraints. This manuscript aims to introduce and compare different embedded implementations, considered as potential candidates for integration into control and diagnostic systems of magnetic confinement fusion plants. Comparisons are based on quantifying the computational and temporal footprint that each hardware/software configuration introduces in a representative real-time acquisition and transmission loop. The three hardware targets are considered, namely the STMicroelectronics microntroller board STM32H723ZG, in Baremetal and Free Real Time Operating System (FreeRTOS) environments, and PicoZed board running Linux. A standard implementation on Field Programmable Gate Array (FPGA) will serve as a time performance reference. Performance is evaluated using key metrics for fast CODAS applications: sampling jitter, cycle time, round-trip time. The results expose trade-offs between determinism, modularity, and resource efficiency, guiding embedded system selection for fusion-grade CODAS subsystems.

Time synchronization is a fundamental requirement for the reliable operation of Control and Data Acquisition Systems (CODASs) in large-scale fusion experiments such as the Divertor Tokamak Test (DTT). Distributed diagnostics, sensors, and control subsystems must share a unified time reference to guarantee deterministic data acquisition and stable plasma control. This paper presents the FPGA-based implementation and evaluation of a synchronization system that combines the IEEE 1588 Precision Time Protocol (PTP) with Pulse Per Second (PPS) generation. The proposed platform is built on Zynq UltraScale+ Kria KR260 System-on-Modules (SOMs) running a customized PetaLinux distribution with LinuxPTP utilities. Hardware timestamping is enabled through the integrated Timestamping Unit (TSU) in the Gigabit Ethernet MAC, while a hardware logic module generates PPS signals from the synchronized PTP clock. Experimental validation demonstrates nanosecond-level synchronization with an RMS timing accuracy of approximately 8.5 ns. A detailed analysis of PPS offset, network path delay, and servo adjustments confirms stability of the timing system. The proposed design offers a low-cost, flexible, fully customizable and controllable solution for distributed diagnostic and control systems in fusion facilities.

We address a key challenge in plasma diagnostics: reconstructing the three-dimensional (3D) light emission density from projection data. To this end, we employ a method for reconstructing 3D density functions based on a corollary of the 3D Fourier slice theorem. This corollary establishes a connection between 2D plane projections and plane sections of the 3D Fourier transform, enabling a direct and non-iterative solution to the inverse problem. The method is applied to the PROTO-SPHERA experiment, an innovative magnetic confinement plasma setup for controlled nuclear fusion research. Experimental data were acquired using six high-speed cameras arranged with cylindrical symmetry around the plasma chamber, capturing time-resolved pictures that are assumed to be parallel projections of the visible light emissions from Helium and Hydrogen discharges. After a transformation of coordinates, by computing their inverse 3D Fourier transform, we reconstruct the spatial distribution of light emission density. This technique enabled to reveal dynamic features of plasma self-organization — such as torus formation around a centerpost — and provides internal cross-sectional views of the plasma, displaying previously obscured spectral components.

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.

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.

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.

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.

We propose an approach to image reconstruction based on a formulation of the Fourier slice theorem in 3D, which relates a set of projections onto planes -or 3D Radon transform- to a 3D Fourier transform. After a suitable transformation of coordinates, by calculating the 3D inverse Fourier transform of projections, the 3D image of the scanned object is reconstructed. The formulation is applied to the reconstruction of the distribution of light emission density of magnetically confined plasma, employing data provided by ENEA (Italian National Agency for Technology, Energy and Sustainable Economic Development). Data include light emission by Hydrogen and Helium plasma, recorded within the framework of the PROTO-SPHERA experiment, where plasma is produced within a transparent vessel and monitored by six cameras. This experiment displayed the appearance and sustainment of a plasma torus around an internal plasma centerpost by self-organisation; an entirely unexplored phenomenon to date. The 3D image reconstruction enabled observation of the formation of the plasma torus at different time instants and of the emission spectrum of its cross-sections.

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.

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.

Results from the last FTU campaigns on the deuterium large (wrt FTU volume) pellet REs suppression capability, mainly due to the induced burst MHD activity expelling REs seed are presented for discharges with 0.5 MA and 5.3T. Clear indications of avalanche multiplication of REs following single pellet injection on 0.36 MA flat-top discharges is shown together with quantitative indications of dissipative effects in terms of critical electrical field increase due to fan-like instabilities. Analysis of large fan-like instabilities on post-disruption RE beams, that seem to be correlated with low electrical field and background density drops, reveal their strong RE energy suppression capability suggesting a new strategy for RE energy suppression controlling large fan instabilities. We demonstrate how such density drops can be induced using modulated ECRH power on post-disruption beams.

Results from the last FTU campaigns on RE mitigation strategies for quiescent and post-disruption RE beams are presented. We provide experimental evidence that for some RE quiescent scenarios D-2 solid pellets achieve complete RE suppression capability, mainly due to the induced burst MHD activity expelling RE seed, whereas in other cases we report clear indications of avalanche multiplication of RE. Results on the assimilation of solid deuterium pellets on RE quiescent scenarios are provided. Quantitative indications of dissipative effects of anomalous Doppler instabilities (ADI) and MHD activity, in terms of critical electric field increase, is introduced and supported by experimental evidence. Multiple analysis are provided to show the significant energy conversion/dissipation of large ADI on post-disruption RE beams suggesting new strategies for RE energy suppression. We also demonstrate experimentally that modulated ECRH could be used for ADI pacing.

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.

Transport modelling of Joint European Torus (JET) dimensionless collisionality scaling experiments in various operational scenarios is presented. Interpretative simulations at a fixed radial position are combined with predictive JETTO simulations of temperatures and densities, using the TGLF transport model. The model includes electromagnetic effects and collisions as well as E ⃗ × B ⃗ shear in Miller geometry. Focus is on particle transport and the role of the neutral beam injection (NBI) particle source for the density peaking. The experimental 3-point collisionality scans include L-mode, and H-mode (D and H and higher beta D plasma) plasmas in a total of 12 discharges. Experimental results presented in (Tala et al 2017 44th EPS Conf. ) indicate that for the H-mode scans, the NBI particle source plays an important role for the density peaking, whereas for the L-mode scan, the influence of the particle source is small. In general, both the interpretative and predictive transport simulations support the experimental conclusions on the role of the NBI particle source for the 12 JET discharges.

The aim of this paper is to present a signal processing algorithm that, applied to the raw Locked Mode signal, allows us to obtain a disruption indicator in principle exploitable on different tokamaks. A common definition of such an indicator for different machines would facilitate the development of portable systems for disruption prediction, which is becoming of increasingly importance for the next tokamak generations. Moreover, the indicator allows us to overcome some intrinsic problems in the diagnostic system such as drift and offset. The behavior of the proposed indicator as disruption predictor, based on crossing optimized thresholds of the signal amplitude, has been analyzed using data of both JET and ASDEX Upgrade experiments. A thorough analysis of the disruption prediction performance shows how the indicator is able to recover some missed and tardy detections of the raw signal. Moreover, it intervenes and corrects premature or even wrong alarms due to, e.g., drifts and/or offsets.

Beryllium limiter marker tiles were exposed to plasma in the Joint European Torus to diagnose the erosion of main chamber wall materials. A limiter marker tile consists of a beryllium coating layer (7-9 mu m) on the top of bulk beryllium, with a nickel interlayer (2-3 mu m) between them. The thickness variation of the beryllium coating layer, after exposure to plasma, could indicate the erosion measured by ion beam analysis with backscattering spectrometry. However, interpretations from broad beam backscattering spectra were limited by the non-uniform surface structures. Therefore, micro-ion beam analysis (mu-IBA) with 3 MeV proton beam for Elastic back scattering spectrometry (EBS) and PIXE was used to scan samples. The spot size was in the range of 3-10 mu m. Scanned areas were analysed with scanning electron microscopy (SEM) as well. Combining results from mu-IBA and SEM, we obtained local spectra from carefully chosen areas on which the surface structures were relatively uniform. Local spectra suggested that the scanned area (approximate to 600 mu m x 1200 mu m) contained regions with serious erosion with only 2-3 mu m coating beryllium left, regions with intact marker tile, and droplets with 90% beryllium. The nonuniform erosion, droplets mainly formed by beryllium, and the possible mixture of beryllium and nickel were the major reasons that confused interpretation from broad beam EBS.

In ITER and DEMO, various control objectives related to plasma control must be simultaneously achieved by the plasma control system (PCS), in both normal operation as well as off-normal conditions. The PCS must act on off-normal events and deviations from the target scenario, since certain sequences (chains) of events can precede disruptions. It is important that these decisions are made while maintaining a coherent prioritization between the real-time control tasks to ensure high-performance operation. In this paper, a generic architecture for task-based integrated plasma control is proposed. The architecture is characterized by the separation of state estimation, event detection, decisions and task execution among different algorithms, with standardized signal interfaces. Central to the architecture are a plasma state monitor and supervisory controller. In the plasma state monitor, discrete events in the continuous-valued plasma state are modeled using finite state machines. This provides a high-level representation of the plasma state. The supervisory controller coordinates the execution of multiple plasma control tasks by assigning task priorities, based on the finite states of the plasma and the pulse schedule. These algorithms were implemented on the TCV digital control system and integrated with actuator resource management and existing state estimation algorithms and controllers. The plasma state monitor on TCV can track a multitude of plasma events, related to plasma current, rotating and locked neoclassical tearing modes, and position displacements. In TCV experiments on simultaneous control of plasma pressure, safety factor profile and NTMs using electron cyclotron heating (ECH) and current drive (ECCD), the supervisory controller assigns priorities to the relevant control tasks. The tasks are then executed by feedback controllers and actuator allocation management. This work forms a significant step forward in the ongoing integration of control capabilities in experiments on TCV, in support of tokamak reactor operation.

In this paper we validate the finite element code RAPLICASOL, which models radiofrequency wave propagation in edge plasmas near ICRF antennas, against calculations with the TOPICA code. We compare the output of both codes for the ASDEX Upgrade 2-strap antenna, and for a 4-strap WEST-like antenna. Although RAPLICASOL requires considerably fewer computational resources than TOPICA, we find that the predicted quantities of experimental interest (including reflection coefficients, coupling resistances, - and -matrix entries, optimal matching settings, and even radiofrequency electric fields) are in good agreement provided we are careful to use the same geometry in both codes.

A novel multiphysics methodology for the computation of realistic plasma neutron sources has been developed. The method is based on state-of-the-art plasma transport and neutron spectrum calculations, coupled with a Monte Carlo neutron transport code, bridging the gap between plasma physics and neutronics. In the paper two JET neutronics tokamak models are used to demonstrate the application of the developed plasma neutron sources and validate them. Diagnostic data for the record JET D discharge 92436 are used as input for the TRANSP code, modelling neutron emission in two external plasma heating scenarios, namely using only neutral beam injection and a combination of the latter and ion cyclotron resonance heating. Neutron spectra, based on plasma transport results, are computed using the DRESS code. The developed PLANET code package is employed to generate plasma neutron source descriptions and couple them with the MCNP code. The effects of using the developed sources in neutron transport calculations on the response of JET neutron diagnostic systems is studied and compared to the results obtained with a generic plasma neutron source. It is shown that, although there are significant differences in the emissivity profiles, spectra shape and anisotropy between the neutron sources, the integral response of the time-resolved ex-vessel neutron detectors is largely insensitive to source changes, with major relative deviations of up to several percent. However it is calculated that, due to the broadening of neutron spectra as a consequence of external plasma heating, larger differences may occur in activation of materials which have threshold reactions located at DD neutron peak energies. The PLANET plasma neutron source computational methodology is demonstrated to be suitable for detailed neutron source effect studies on JET during DT experiments and can be applied to ITER analyses.