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The present contribution is devoted to the neutral beam injectors (NBIs) for ITER heating and current-drive. First, updated information is provided about the development status of the entire NBI prototype (MITICA); starting in 2021, the first experiments will be dedicated to high-voltage holding tests in vacuum. Then the contribution describes the full-scale prototype of the NBI ion source (SPIDER) and the activities performed in the first two years of operation, devoted to investigating the operational regimes of the ion source and verifying the performances of the various plants in view of extensive beam operation. Some improvements have already been implemented; others are being prepared for the next shutdown. The characterisation of the SPIDER plasma is presented and the results of the first beam operations are reported.

In the ITER neutral beam (NB) system, deuterium negative ion beam with 1 MeV, 40 A for 3600 s is required for plasma heating and current drive. The NB Test Facility (NBTF) is now being constructed in Padova, Italy, to achieve such high requirement prior to the ITER operation. Japan procures DC 1 MV power supply (PS) components for the NBTF and the ITER NB. Since this PS consists of various kinds of component with 100 m long, manufacturing and factory test have been performed over five years, and shipping and installation have been proceeded in parallel step by step as presented in the last FEC [A]. The on-site assembling technique for such complicated 1MV PS components is one of challenges not only for the ITER NB but also typical electrical components. As recent progress in the last two years, completion of assembly and also high-voltage insulation test of DC 1 MV power supply components are reported.

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 17MW each, by neutralizing accelerated negative hydrogen or deuterium ions. The requirements of ITER NBIs (40A/1MeV D- ions for

The negative-ion beam source SPIDER, which is the full-scale prototype source for the ITER neutral beam injector, recently started the operation with caesium. This experimental phase follows three years of volume operation, devoted to the commissioning of the plants and to the integrated test of the ion source and accelerator. The ion source, composed of eight RF drivers connected to a large plasma chamber from which the negative ions are extracted, was operated up to a RF power of 400kW, and beam energy up to 50kV, and plasma discharges limited to less than one minute This contribution will describe the main results of the first campaign with caesium in SPIDER. The repetition of short plasma and beam extraction blips with different injection rates was applied, to study the effect on plasma and beam parameters. The caesiation procedure adopted in SPIDER will be described (caesium injection rate, duty cycle of plasma-on, RF power, source gas pressure) together with the effects of the source parameters on the extracted beam and its uniformity. Even though the use of a much reduced number of beamlets was a strong limitation in terms of total accelerated current (a mask at the plasma grid covered 1252 apertures over 1280 to limit the gas load to the vacuum pumps), it provided advantages in the study of the caesium effect on the beam, such as the introduction of dedicated diagnostics for the single beamlets, and the identification of the beamlet current and optics.

energy. To achieve ITERs objectives to operate with high fusion power, Pfus ~ 500 MW with fusion gain, Q=10 for 400 s in a baseline scenario, and Pfus > 250 MW, Q=5 operation for 3600 s in advanced scenarios ITER relies on the realisation of the auxiliary heating systems, of which one of the most important is the Neutral Beam injection system. A total power of 33 MW from the two heating neutral beam (HNB) injectors is envisaged in the present scenario. To achieve the required heating and current drive necessary from the neutral beams, the injectors are required to provide 1 MeV D0 neutral beams created by accelerating deuterium negative ions to -1 MV across 5 electrostatic grids. These stringent requirement have never been demonstrated on any existing machine to date, making them first of a kind with significant physical and technological challenges. In addition, a beam line at lower energies, 100kV is also envisaged as a diagnostic neutral beam (DNB), to realise charge exchange measurements. One of the advantages of neutral beams as a technique for heating, current drive, and momentum input is that their coupling to the plasma is not strongly dependent upon the edge conditions of the tokamak plasma, and the interaction with the plasma takes place primarily through a relatively wellunderstood mechanism, two-body collisions. As a consequence, a neutral beam test facility (NBTF) can be utilized effectively to develop, test, and improve the neutral beam capabilities at a location remote from the ITER machine, with a high degree of confidence that it will function in a similar fashion when installed on the tokamak, and that it will drive effects within the confined plasma which are largely predictable. The ITER NBTF, is intended to capture these benefits for ITER, and for both the primary heating and current drive beams (HNB) and the DNB. The NBTF will consist of two complementary test facilities at Consorzio RFX in Padua, a 1 MV test bed (MITICA), which differs from an actual injector only in the diagnostic capability, and an ion source test bed (SPIDER) which is capable of testing a full size ITER ion source at full parameters, but with an higher degree of flexibility and at reduced voltage of 100 kV. The NBTF is the centrepiece of a coordinated development effort, including subsidiary test facilities in Japan, Europe and India, that will result in the experience and information necessary to proceed with the finalization of the ITER neutral beams. Having such a facility for the development of neutral beams allows the inherent technological and physics risks of Neutral beam injectors to be mitigated. These risks can be summarised in a few main categories, HV Holding, Optics and Ion source efficiencies. This paper will outline the importance of the NB system to ITER and highlight the importance of the NBTF in ensuring that these systems are successfully realised on the ITER machine.

In the ITER neutral beam (NB) system, deuterium negative ion beam with 1 MeV, 40 A for 3600 s is required for plasma heating and current drive. The NB Test Facility (NBTF) is now being constructed in Padova, Italy, to achieve such high requirement prior to the ITER operation. Japan procures DC 1 MV power supply (PS) components for the NBTF and the ITER NB. Since this PS consists of various kinds of component with 100 m long, manufacturing and factory test have been performed over five years, and shipping and installation have been proceeded in parallel step by step as presented in the last FEC [A]. The on-site assembling technique for such complicated 1MV PS components is one of challenges not only for the ITER NB but also typical electrical components. As recent progress in the last two years, completion of assembly and also high-voltage insulation test of DC 1 MV power supply components are reported.

In the ITER Neutral Beam Test Facility (NBTF) in Padova (Italy), the Ion Source and Extraction Power Supplies (ISEPS) are currently in operation for the SPIDER experiment and in power supply integrated commissioning phase for MITICA. The procurement of ISEPS by OCEM Energy Technology foresees four stages, corresponding to four units: one ISEPS for SPIDER, one ISEPS for MITICA and two ISEPS systems for the Neutral Beam Injectors of ITER. The first two stages have been completed, while the others are currently in progress. The ISEPS systems include 4 RF generators based on self-excited free-running tetrode oscillators. The reasons of the selection of this technology laid on the experience on RF ion sources for NBI at IPP Garching, the belief that self-excited oscillators could assure stability in power and frequency in a wide range of load parameters and on the positive feedbacks from the experimental results until the time of the definition of the procurement scope.. Moreover, at that time, no sufficient evidence was available that alternative technologies could have given better performance. The choice of the technology was anyway left open in the technical specifications, but all potential suppliers proposed solutions based on tetrode oscillators for the above mentioned reasons. However, subsequent feedbacks from ion sources operation at IPP reported on the observation of sudden frequency variations of the oscillator, so-called "frequency-flip", which limited the controllability of the RF generators. Afterwards, RF amplifiers based on solid state components were tested with positive results in BATMAN and ELISE facilities at IPP. Observations of frequency flips came also from the first experimental campaigns of SPIDER; the accompanying analyses explained the intrinsic and unavoidable limits of the application of this technology to the resonant loads of Neutral Beam Injectors based on RF ion sources. The comprehension of the phenomenon allowed properly controlling the oscillators so as to avoid the frequency flip occurrence and to achieve long pulse operation, but the impossibility to operate at the desired load resonance condition was also fully clear, preventing the achievement of the nominal power. All these evidences led to the decision to replace the RF generators in SPIDER and MITICA and change the current ITER baseline from self-excited freerunning tetrode oscillator technology to solid state amplifier technology. The paper will present and discuss the issues found in SPIDER with the operation of the RF generators based on self-excited tetrode oscillators and the advantages expected from their substitution with RF solid state amplifiers. The assessment of the procurement for SPIDER and MITICA will be also described, including the necessary adaptation of the site interfaces and the strategy to test and install the amplifiers.

The Acceleration Grid Power Supply (AGPS) has been procured by ITER-India for the SPIDER experiment hosted in the Neutral Beam Test Facility (NBTF) in Padua, Italy. SPIDER is the prototype of the negative ion source of the ITER Heating Neutral Beam Injectors. The AGPS is devoted to feed the electrostatic accelerator of SPIDER with a dc voltage down to -96 kV, a maximum current of 75 A and a maximum pulse duration of 1 hour. Realized with Pulse Step Modulation technology, its main components are: three oil-insulated multi-secondary transformers fed by the 22 kV ac distribution network, 150 series-connected water-cooled switched power modules based on IGBTs, an output inductor, and a Control and Protection System. The AGPS has been commissioned at NBTF site on dummy load in 2018, reaching the full voltage. Then, ITER Organization (IO) accepted the system and transferred the Responsibility of Use to Consorzio RFX in 2019, in order to start the SPIDER operation with beam acceleration. However, it was not allowed to exceed the acceleration voltage of 30 kV, corresponding to the voltage ratings of the output cable and load disconnectors. In 2020, these components were replaced by new ones, with full voltage ratings. For space constraints, the number of disconnectors were reduced, accepting some compromise on the flexibility in operation. Then, in 2021 the asseveration for an output voltage down to -96 kV has been achieved, and the first SPIDER experimental campaigns with high acceleration voltage started. While at the beginning the beam current was very limited, it increased during the year with the use of Cesium in the ion source, which caused a substantial improvement of negative ion's production. During the operations, some relevant aspects of the AGPS for which an upgrade could be beneficial have been identified, for example the voltage regulation and ripple in no-load conditions, the voltage regulation range, and some control functions. This paper describes the process and the technical solutions which allowed achieving the asseveration to operate the AGPS at full voltage, reports the first experimental results and discusses the advisable improvements which have been identified.

First operation of SPIDER and integrated power tests in MITICA.

As indicated in the European Fusion Roadmap, the main objective of the Divertor Tokamak Test facility (DTT) is to explore alternative power exhaust solutions for DEMO so as to mitigate the risk that the conventional divertor based on detached conditions to be tested on the ITER device cannot be extrapolated to a fusion reactor. The issues to be investigated by DTT include: • demonstrate a heat exhaust system capable of withstanding the large load of DEMO in case of inadequate radiated power fraction; • close the gaps in the exhaust area that cannot be addressed by present devices; • demonstrate that the possible (alternative or complementary) solutions (e.g., advanced divertor configurations or liquid metals) can be integrated in a DEMO device. In this paper, we describe a proposal for such a DTT, presented by ENEA in collaboration with a European team of scientists. The selection of the DTT parameters (a major radius of 2.15 m, an aspect ratio of about 3, an elongation of 1.6-1.8, a toroidal field of 6 T, and a flat top of about 100 s) has been made according to the following specifications: • edge conditions as close as possible to DEMO in terms of dimensionless parameters; • flexibility to test a wide set of divertor concepts and techniques; • compatibility with bulk plasma performance. • an upper bound of 500 M€ for the investment costs. This paper illustrates this DTT proposal showing how the basic machine parameters and concept have been selected so as to make a significant step toward the accomplishment of the power exhaust mission.

Heating neutral beam (HNB) injectors, necessary to achieve burning conditions and to control plasma instabilities in ITER, are characterized by such demanding parameters that a neutral beam test facility (NBTF) dedicated to their development and optimization is being realized in Padua (Italy) with direct contributions from the Italian government (through Consorzio RFX as the host entity) and the ITER international organization (with kind contributions from the ITER domestic agencies of Europe, Japan and India) and technical and scientific support from various European laboratories and universities. The NBTF hosts two experiments: SPIDER, devoted to ion source optimization for the required source performance, and MITICA, the full-size prototype of the ITER HNB, with an ion source identical to SPIDER. This paper gives an overview of the progress towards NBTF realization, with particular emphasis on issues discovered during this phase of activity and on solutions adopted to minimize the impact on the schedule and maintain the goals of the facilities. The realization of MITICA is well advanced; operation is expected to start in 2023 due to the long procurement time of the in-vessel mechanical components. The beam source power supplies, operating at 1 MV, are in an advanced phase of realization; all high-voltage components have been installed and the complex insulation test phase began in 2018. At the same time, construction and installation of SPIDER plant systems was successfully completed with their integration into the facility. The mechanical components of the SPIDER ion source were installed inside the vessel and connected to the plants. Integrated commissioning with the control, protection and safety systems ended positively and the first experimental phase has begun. The first results of the SPIDER experiment, with data from operational diagnostics, and the plans for the 1 MV insulation tests on the MITICA high-voltage components are presented.

To reach fusion conditions and to control the plasma configuration in ITER, the next step in thermonuclear fusion research, two heating and current-drive neutral beam injectors (NBIs) will supply 17MW each, by accelerating negative hydrogen or deuterium ions to 1MeV. The requirements of ITER NBIs (40A negative H or D ions for 1 hour) have never been simultaneously attained. So in the dedicated Neutral Beam Test Facility (NBTF) at Consorzio RFX (Italy) the performances of the ITER NBI (divergence

This paper gives an overview of the progress of the Neutral Beam Test Facility realization, including the experimental status of SPIDER.