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• The MITICA (Megavolt ITER injector and concept Advancement) experiment, currently in the installation and commissioning phase at the neutral beam test facility in Padua-Italy, will be equipped with a beam source (BS) expected to generate a 40A negative ion beam to the energy of 1MeV (values for D). It is composed of an ion source to generate a h or d plasma and of a multiple-Aperture multiple-grid (MAMuG) accelerator [1] , where ions extracted are accelerated in five consecutive stages of 200 kV each. • The engineering and fabrication of the beam source for MITICA BS started in october 2018, when a contract between fusion for energy (F4E) and alsymex (France) was signed. Presently, all the components and sub-assemblies have been manufactured, except for a few non-critical items. Assembly of the accelerator is well advanced and ion source is on-going. • A review of the fabrication of the main components and aspects of their assembly is reported in this paper, with a focus on the main achievements. The MITICA (Megavolt ITER Injector and Concept Advancement) experiment, currently in the installation and commissioning phase at the Neutral Beam Test Facility in Padua-Italy, will be equipped with a Beam Source (BS) expected to generate a 40A negative ion beam to the energy of 1MeV (values for D). It is composed of an ion source to generate a H or D plasma and of a multiple-Aperture multiple-grid (MAMuG) Accelerator [1] , where ions extracted are accelerated in five consecutive stages of 200 kV each. The engineering and fabrication of the beam source for MITICA BS started in October 2018, when a contract between Fusion for Energy (F4E) and Alsymex (France) was signed. Presently, all the components and sub-assemblies have been manufactured, except for a few non-critical items. Assembly of the accelerator is well advanced and ion source is on-going. A review of the fabrication of the main components and aspects of their assembly is reported in this paper, with a focus on the main achievements. [ABSTRACT FROM AUTHOR]

One of the main challenges for the construction of DEMO, the first fusion demonstration reactor, is the power exhaust issue. To deal with it, EUROfusion has undertaken the construction of DTT (Divertor Tokamak Test) facility which will be held in Frascati, Italy. It aims to test various divertor models which can be integrated and used under different plasma confinement configurations. The current design of DTT embed 54 divertor cassettes which must be anchored to the vacuum vessel using suitable fixation systems. This paper deals with the initial monotonic load assessment of the fixation system of the Fixation System which must ensure compliance with multiple design requirements. The Remote Handling compatibility is one of the most demanding, since it includes a preloading phase of the divertor necessary to mitigate the shaking of the cassette due to dynamic forces. Moreover, the system must withstand the conspicuous loads acting on the Divertor due to the full or partial loss of plasma confinement (Disruptive Events). Structural verifications through Finite Element Method (FEM) have been performed on the Divertor-Fixation System, considering a set of load combinations, with particular focus on the most demanding case related to Fast-transient Vertical Disruption Event-downward. [ABSTRACT FROM AUTHOR]

• Remote control of a pellet facility has been proved repeatable and reliable. • E3S platform has been successfully tested in remote control of a pellet injector. • Performance tests indicate that, for massive I/O data flow, E3S should be scaled-up. The four-barrel, two-stage gun Ignitor Pellet Injector (IPI) was developed in collaboration between ENEA and ORNL. The prototype injector is presently located at Oak Ridge (TN, USA), and is normally operated locally through a control and data acquisition system developed in LabVIEW. More recently, a remote-control system has been set up, based on RealVNC®, which allows to operate the IPI from a control room in Italy. Tools for data transfer and storage into ENEA ICT area have also been provided. A Staging, Storage and Sharing system, named E3S, developed using OwnCloud as architectural component, is used for file synchronization and sharing of the data acquired by the diagnostic systems. It provides a homogeneous platform able to store and share heterogeneous data produced by many data acquisition systems in large nuclear fusion experiments. This paper reports about the implementation of the IPI remote control, and presents the application of E3S to this specific case, allowing easy storage and sharing of experimental data onto a wide-area distributed file-system, as well as remote data access via web-services based on MDS+ tool, integrated with MySQL metadata. A performance analysis of the architectural components is also introduced. [ABSTRACT FROM AUTHOR]

The Divertor Tokamak Test (DTT) facility is in advanced design and construction phase at the ENEA laboratories in Frascati, Italy, to contribute to the study, the design and assessment of systems for the treatment of heat exhaust. The DTT Toroidal Field (TF) magnet system consists of 18 superconducting coils, fed in series by a single power supply (TFPS), which is a 24-pulses thyristor-based converter with an output DC current up to of 44 kA. In case of a quench in the superconductors, the magnetic energy stored in the coils is extracted by 3 Fast Discharge Units (FDUs) in series with 3 sectors of 6 TF coils. After the respective contracts were awarded, the TFPS and the FDUs are currently in the final design phase which is addressed in this paper. In particular, the analysis is focused on the most innovative aspects of the design, as the use of silicon carbide (SiC) varistors instead of discharge resistors and of a fully electronic and completely redundant high-current switch without any mechanical device. [ABSTRACT FROM AUTHOR]

In the framework of the Broader Approach Agreement for the construction of the JT-60SA tokamak, ENEA provides 9 of the 18 toroidal field (TF) coils of the JT-60SA magnet system. The 9 coils are being manufactured by ASG superconductors in Genoa under the supervision of ENEA in collaboration with the JT-60SA European home team. The manufacturing is composed of two main steps: one concerning winding pack assembly and impregnation, and the other devoted to the integration into the casing structure and associated final coil preparation. This paper describes the results of the validation activities set-up for the integration phase. Specifically, welding of casing components has been retained particularly critical for at least three reasons: (i) during welding the WP may be damaged by the intense heating; (ii) distortion caused by heating may determine incorrect coil geometry and then field errors; and (iii) flaws may reduce structural strength and then the overall lifetime of the machine. Similarly, final embedding has been demonstrated on a 1 m long mock-up of the coil. Main results and lessons learned are here described. [ABSTRACT FROM AUTHOR]

To reach fusion conditions and control the plasma configuration in ITER, the next step in tokamak fusion research, two neutral beam injectors (NBIs) will supply 16.5 MW each, by neutralizing accelerated negative hydrogen or deuterium ions. The requirements of ITER NBIs (40A/1 MeV D- ions for ≤1 h, 46A/870 keV H- ions for ≤1000 s) have never been simultaneously attained. So in the Neutral Beam Test Facility (NBTF, Consorzio RFX, Italy) the operation of the full-scale ITER NBI prototype (MITICA) will be tested and optimised up to full performances, focussing on accelerator (including voltage holding), beam optics, neutralisation, residual ion removal. The NBTF includes also the full-scale prototype of the ITER NBI source with 100 keV particle energy (SPIDER), for early investigation of: negative ion production and extraction, source uniformity, negative ion current density and beam optics. This paper will describe the main results of the first two years of SPIDER operation, devoted to characterizing plasma and beam parameters, including investigation of RF-plasma coupling efficiency and magnetic filter field effectiveness in reducing co-extracted electrons. SPIDER is progressing towards the first caesium injection, which aims at increasing the negative ion density. A major shutdown, planned for 2021, to solve the issues identified during the operation and to carry out programmed modifications, will be outlined. The installation of each MITICA power supply and auxiliary system is completed; in-vessel mechanical components are under procurement by Fusion for Energy (F4E). Integration, commissioning and test of the power supplies, procured by F4E and QST, as the Japanese Domestic Agency (JADA), will be presented. In particular, 1.0MV insulating tests were carried out step-by-step and successfully completed. In 2020 integrated tests of the power supplies on the accelerator dummy load started, including the assessment of their resilience to accelerator grid breakdowns using a short-circuit device located in vacuum. The aggressive programme, to validate the NBI design at NBTF and to meet ITER schedule (requiring NBIs in operation in 2032), will be outlined. Unfortunately, in 2020 the coronavirus disease infection affected the NBTF activities. A solution to proceed with integrated power tests despite the coronavirus is presented. [ABSTRACT FROM AUTHOR]

• A new multiscale model for TRIEX-II facility has been presented. • The tool permits to fast compute both inventories and permeation fluxes. • The modelling results show well agreement with the experimental results. • The tool can be applied to more complex systems such as a breeding blanket and its ancillary systems. At ENEA Brasimone Research Centre, Italy, a new experimental facility named TRIEX-II (Tri tium Ex traction) was designed and installed. Its aim is to characterize, in the range of operating conditions foreseen for the European Test Blanket System WCLL-TBS, several extraction technologies for hydrogen isotopes (Q 2) solubilized in the flowing metallic LiPb alloy (15.7 at. % Li). One of these technologies is the packed column, an example of Gas/Liquid Contactors (GLCs). This paper proposes a multiscale modelling tool, combining different scales through two computational tools. The extraction column mock-up is described by a component-detail level model, developed in COMSOL Multiphysics, and integrated into a system level code of the whole TRIEX-II circuit, developed using MATLAB/Simulink. The integration is carried out by implementing the COMSOL component into an S-function of MATLAB/Simulink, preserving the process flow diagram of the loop. In this way, it was possible to quantify the Q 2 concentrations and the permeation fluxes, and to evaluate the theoretical extraction efficiency. The LiPb flow field inside the extractor was also derived. Finally, a comparison with the experimental results was performed. The results suggest that this tool could be adapted for analyses of complex systems, at a multiscale level, in view of design improvements and safety studies for the tritium cycle of ITER. [ABSTRACT FROM AUTHOR]