DTT, Divertor Tokamak Test facility, is currently under construction at the Frascati ENEA Research Center. Its main aim is to explore alternative solutions for the extraction of the heat generated by the fusion process. Its Control and Data Acquisition System (CODAS) will (1) orchestrate and synchronize all the DTT systems during Plasma operation and maintenance; (2) acquire data from the experiment diagnostics and plant systems and store it in an experimental database to be used for on-line and off-line analysis; (3) provide real-time Plasma control. The expected duration of the plasma discharge in DTT is in the order of some tens of seconds and therefore DTT can be considered a long-lasting experiment, involving data streaming technologies for data communication and storage. The main DTT CODAS design is based on three principles: (1) Taking inspiration from other similar experiments currently under development, namely ITER CODAC, (2) relying on proven solutions already adopted in running experiments with similar constraints and (3) taking advantage from practices widely adopted in fusion and, more in general, in industry. Despite using components already adopted in other fusion experiments, DTT CODAS is the first system that seamlessly integrates all of them.

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.

ITER envisages the use of two heating neutral beam injectors plus an optional one as part of the auxiliary heating and current drive system. The 16.5 MW expected neutral beam power per injector is several notches higher than worldwide existing facilities. A Neutral Beam Test Facility (NBTF) was established at Consorzio RFX, exploiting the synergy of two test beds, SPIDER and MITICA. SPIDER is dedicated to developing and characterizing large efficient negative ion sources at relevant parameters in ITER-like conditions: source and accelerator located in the same vacuum where the beam propagates, immunity to electromagnetic interferences of multiple radio-frequency (RF) antennas, avoidance of RF-induced discharges on the outside of the source. Three years of experiments on SPIDER have addressed to the necessary design modifications to enable full performances. The source is presently under a long shut-down phase to incorporate learnings from the experimental campaign. Parallelly, developments on MITICA, the full-scale prototype of the ITER NBI featuring a 1 MV accelerator and ion neutralization, are underway including manufacturing of in-vessel components, while power supplies and auxiliary plants are already under final testing and commissioning. Integration, commissioning and tests of the 1MV power supplies are essential for this first-of-kind system, unparalleled both in research and industry field. The integrated test to confirm 1MV output by combining invertor systems, DC generators and transmission lines extracted errors/accidents in some components. To realize a concrete system for ITER, solutions for the repair and the improvement of the system were developed. Hence, NBTF is emerging as a necessary facility, due to the large gap with existing injectors, effectively dedicated to identify issues and find solutions to enable successful ITER NBI operations in a time bound fashion.

SPIDER, ITER's full-size beam source at the Neutral Beam Test Facility (NBTF), has been operational since 2018. SPIDER's mission is to optimize the operation of the beam source in order to reuse SPIDER's experience on MITICA, the full-size prototype of the ITER Neutral Beam Injector, and in the ITER injectors. The exploitation of SPIDER started with short, low-performance pulses lasting up to a few seconds and progressed to obtain long pulses up to 3000 seconds. Furthermore, the integration of plant and diagnostic systems has grown over time. The amount of data collected per pulse provide a measure of evolution: it has gone from a few tens of MB in the first campaign pulses to the current average value of over 130 GB, most of which produced by infrared and visible cameras. From the first operation onwards, the control systems have also evolved, including components and functions initially not foreseen or developed in a preliminary form. This includes the progressive integration of plant and diagnostic systems and of protection and safety functions. The paper initially focuses on the architecture of the SPIDER control systems, i.e. CODAS, the system delivering conventional control and data management, the central interlock system, delivering plant protection, and the central safety system delivering people and environment safety. Since we developed all systems according to the ITER guidelines for implementing control systems, the integrated SPIDER control, interlock and safety systems may provide an interesting example for developers of ITER plant systems. The paper then describes how the top-down definition and implementation of operating states and operational scenarios provides the framework for the integration of control, interlock and safety systems and the basic elements for successful operation. Finally, the paper reports the lessons learned during these years of operation with particular attention to the progressive, continuous evolution and recommissioning of systems.

RFX-mod2 is an upgrade of RFX-mod that will use a modified shell and mechanical structure in order to enhance plasma-shell proximity and therefore to improve plasma control. The Control and Data Acquisition System for most of the plant systems and diagnostics previously used in RFX-mod will be refurbished, while others will be completely re-built. The most important component that will be completely renewed is the ElectroMagnetic probe (EM) data acquisition system, where a new architecture based on XILINX Zynq FPGA will be used to carry out at the same time both high-speed data acquisition and resampled data streaming for active plasma control. The use of MDSplus will be retained in RFX-mod2, while the MARTe framework used for real-time plasma control will be replaced by MARTe2, a new framework developed under strict software quality standards. Plant control in RFX-mod2 will be supervised by WinCC-OA, replacing the previous FactoryLink SCADA systems. Older plant systems such as vacuum control based on outdated S5 PLCs will be updated and will use OPC-UA for communication with the supervisory control system.

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.