Neutral Beam Injectors (NBIs) based on negative ions are the workhorses of future fusion reactors, such as ITER, which they are expected to provide with up to 33 MW of power to heat the fusion plasma. The negative hydrogen ions are extracted from a RF plasma, in which a magnetic filter field cools down the electrons reaching the so-called expansion region and allows the formation and survival of negative ions near the apertures in the plasma grid. To further improve the production of negative ions, cesium is usually evaporated inside the source and deposited onto the plasma walls, reducing the work function of the surfaces. This dramatically increases the density of negative hydrogen ions near the surfaces, causing the transition to an electronegative plasma in the vicinity of the plasma grid. This condition can be observed with Langmuir probes, which can then be used to provide a local meaurement of negative ion density in the ion source. In this paper we use the measurements provided by the Langmuir probe sensors embedded in the plasma grid of SPIDER, the prototype ion source of ITER NBIs, to determine the density of negative ions. A fitting method based on the determination of the collection area of the different plasma species is proposed and adapted to SPIDER experimental condition, taking into account the shape of the probes and the local topology of the magnetic field. The method is then applied to the experimental data, determining the densities of the positive and negative ions and of the electrons during a plasma pulse. Finally, a vertical array of four probes in the plasma grid is used to assess the vertical profile of plasma parameters.

The RFX-mod2 device, the upgraded version of the previous RFX-mod with a modified magnetic boundary, is presently under realization and will start to be operated in 2024. Significant upgrades of the diagnostic capabilities have been proposed and are under development. These include a largely increased number of in-vessel magnetic and electrostatic sensors, a new fast reciprocating manipulator for the exploration of the edge plasma in a wide range of experimental conditions, the improved Thomson scattering and soft X-ray diagnostics system for a detailed determination of the behavior of the electron temperature profile, new dedicated systems for the space and time resolved determination of x-ray spectra and neutron rate, a diagnostics made on reflectometric analysis for real-time determination of plasma position, two diagnostics devoted to the imaging of light impurities and influxes behavior along with arrays of Halo current sensors. These diagnostic upgrades will be accompanied by a significant effort on the control the electron density by means of proper treatment of plasma facing components with in-vessel fixed electrodes distributed along the machine. The described advancements will allow a deeper understanding of physics phenomena in the wide variety of magnetic configurations, including the Tokamak, the Reversed-Field Pinch and the Ultra-low q, which can be produced in RFX-mod2 thanks to its flexibility and unique MHD control capabilities.

Magnetic fusion experiments rely heavily on coil loops as the primary type of magnetic sensors, offering the desired precision, reliability, and robustness. However, in order to obtain the magnetic field evolution, the signals from these sensors need to be time-integrated. Traditionally, analog integrators were employed due to their wide dynamic range, but they present challenges in terms of complexity and the need for a separate channel for the derivative (dB/dt) signals required to measure fast events, plasma instabilities, and magnetic turbulence accurately. In this work, we propose a novel system design based on high-resolution analog-to-digital converters (ADCs) that eliminates the need for analog integrators and the second acquisition channel, simplifying and compacting the overall system. The system utilizes a 1 MSamples/sec, 20-bit resolution ADC, providing a resolution comparable to that of good analog integrators. To ensure accurate measurements, each acquisition channel is electrically isolated individually, effectively eliminating the ground loops generated by the experiment's magnetic fields, which often hamper the measurements. The system architecture is implemented on 6U boards, with each board acting as an autonomous system housing 12 input channels and its own SOC-FPGA, with a total of 144 channels on a 6U sub-rack. Each board simultaneously provides three essential functionalities: a timing synchronization decoder, transient recording of full-speed ADC data, and continuous Ethernet UDP transmission of subsampled signals to the real-time control system. This comprehensive approach allows for efficient data acquisition, analysis, and integration into the experiment's control framework.

Giant negative ion sources are used for neutral beam injectors in fusion devices. A high density of cold negative hydrogen ions is required over the large extraction area of the caesium-seeded plasma source, to provide the required negative ion current, distributed uniformly over thousands of extraction apertures. In this regard, it is expected that the expansion of plasma and neutrals from the driver region provides as uniform as possible plasma properties at the extraction region, for adequate compensation of the space charge of such large negative ion density, and relatively slow precursors for the negative ion conversion at caesiated surfaces. These conditions are difficult to achieve in the presence of the transverse magnetic field necessary to filter the diffusion of electrons to the extraction region. The driver region can be either a large volume multi-cusp filament-arc plasma, or an inductively-coupled plasma discharge realised in multiple drivers with external radiofrequency antennas: neutral beams based on filament sources for negative ions reached impressive performances in the recent decades, and an intense development program is in progress for the rf-driven source plasma to bridge the gap in view of the ITER neutral beam injector. The optimization of the ITER beam source plasma, aiming at extracting 350/290 A/m2 of H- /D- with low-divergence at the low filling pressure of 0.3 Pa, is challenging. A review of the ITER beam source physics is provided, based on experimental measurements obtained until now also on the one-to-one prototype SPIDER, and on results of numerical models. This is in the line of the massive work done until now towards the development of negative ion sources, based on both filament arc and rf sources. An overview of the ongoing R&D physics program for SPIDER is also proposed and results of experiments performed at other test facilities are presented.

Three years of experiments on SPIDER allowed characterization of the main features of the source plasma and of the negative ion beam, in the original design configuration. For the large dimensions of the source chamber, and of the extraction area, the investigation of the single-beamlet currents and of the source plasma uniformity had to be carried out to extend the knowledge gained in smaller prototype sources. The configuration of the multiple RF drivers and filter field topologies were found to cause a peculiar behavior in the plasma confinement in the drivers, creating left-right asymmetries which were also visible in the extracted negative ion currents, even after the early implementation of a new scheme of plasma-grid current send and return busbars that greatly improved performance at high filter fields. The plasma properties in the driver and expansion region as well as the positive ion energy at the extraction region were studied in different experimental conditions, and interpreted also with the support of numerical models, suggesting that an improved plasma confinement could contribute to the increase of the plasma density, and to a certain extent to a lowering of the plasma potential profile; both effects shall contribute to increase the presence of cold negative ions for the formation of low-divergence beamlets. Early results related to unwanted RF discharges on the back of the plasma source and the gas conductance of the beam source suggested the reduction of the vessel pressure as mitigation, leading to the definition of a new pumping system. The difficulties related to the simultaneous operation, stable control and high-power operation of multiple RF self-oscillating vacuum tube based RF generators were an unambiguous obstruction to the experimentation, calling for the implementation of RF solid-state amplifiers. The initial tests related to caesium management, the non-uniform plasma properties at different locations across the plasma grid, and the challenges in the measurement of the current and divergence of the accelerated beamlet, unambiguously resulted in the need of new diagnostic systems to investigate with better resolution the spatial uniformities. This contribution summarises how the main experimental findings in the previous experimental campaigns are driving modifications to the SPIDER experiment, during the present shut down, in view of future operations.

Each ITER Neutral Beam Injector (NBI) will provide 16.5 MW hydrogen/deuterium particles, electrostatically accelerating negative ions to 1 MeV. The challenging ITER NBI requirements have never been simultaneously attained and are expected to be demonstrated at the ITER Neutral Beam Test Facility (NBTF), at Consorzio RFX (Italy). NBTF includes the MITICA experiment, full-scale NBI prototype with 1 MeV particle energy, and SPIDER, with 100 keV particle energy, devoted to test and optimise the full-scale ion source. The four years of SPIDER operation demonstrated beam acceleration with high current density, but also highlighted that the non-uniformity and divergence of the beam are too large to comply with the ITER requirements, and that the current density is still lower; moreover, these findings can be traced back to non-homogeneity of the source plasma, where negative ions are produced. SPIDER then entered a shutdown dedicated to the improvement of the various plants and of the diagnostic system. The proposed additional diagnostic systems are integrated with the existing ones and mainly aim at increasing the spatial resolution of the plasma measurements, at investigating profiles in other directions and at measuring other parameters. In parallel, most of the already existing diagnostics will be refurbished or improved.

Stable and uniform beams with low divergence are required in particle accelerators; therefore, beyond the accelerated current, measuring the beam current spatial uniformity and stability over time is necessary to assess the beam performance, since these parameters affect the perveance and thus the beam optics. For high-power beams operating with long pulses, it is convenient to directly measure these current parameters with a non-intercepting system due to the heat management requirement. Such a system needs to be capable of operating in a vacuum in the presence of strong electromagnetic fields and overvoltages, due to electrical breakdowns in the accelerator. Finally, the measure of the beam current needs to be efficiently integrated into a pulse file with the other relevant plant parameters to allow the data analyses required for beam optimization. This paper describes the development, design and commissioning of such a non-intercepting system, the so-called beamlet current monitor (BCM), aimed to directly measure the electric current of a particle beam. In particular, the layout of the system was adapted to the SPIDER experiment, the ion source (IS) prototype of the heating neutral beam injectors (HNB) for the ITER fusion reactor. The diagnostic is suitable to provide the electric current of five beamlets from DC up to 10 MHz.

The RFX-mod experiment (formerly RFX [1]) is the largest Reversed Field Pinch [2] device in operation, that proved the feasibility of active stabilization of MHD instabilities (Resistive Wall Modes) [3], by enclosing the plasma in a combination of a passive stabilizing shell and a real-time controlled network of saddle coils [4], as originally conceived by J.D. Lawson [5] and later proposed by C.M. Bishop [6]. The core of the experiment was the toroidal vacuum vessel (Inconel 625, Rmajor = 2.0 m, rminor = 0.5 m, thickness = 30 mm), surrounded by a Copper shell (3 mm thick) for the passive stabilization of the MHD instabilities, both enclosed in a toroidal support structure (AISI 304 L, 47 mm thick) embedding a set of 4 × 48 saddle coils for the active MHD control (Fig. 1). The flexibility of the RFX-mod device allowed exploring magnetic configurations at different levels of the safety factor [7]. In RFP regimes, especially at high plasma currents, transitions to improved confinement helical states [8], similar to theoretical and numerical predictions [9], have been observed and characterized. Thanks to active control, stable very-low q (edge q<2) ohmic tokamak discharges have been routinely obtained [10]; moreover, ultra-low q regimes have been studied [8]. H-mode in tokamak plasmas have been obtained by means of a polarized insertable electrode [11]. The properties of RFP plasmas in RFX-mod have been found to be influenced in several ways by the residual MHD instabilities (Tearing Modes), whose amplitude and phase non-linear dynamics are strongly influenced by the characteristics of the toroidal complex containing the plasma. In particular, the very high resistivity of the Inconel vacuum vessel (actually the highest among all RFP devices) was such that Tearing Modes were locked to the wall in all plasma current regimes explored by RFX. RFX-mod active control allowed mitigating the localized interaction due the bulging induced by wall locking of tearing modes and very low plasma current campaigns (Ip<150kA) revealed spontaneous fast rotating tearing modes regimes. On the other hand, the high proximity of the vessel plays an important role in the very-low q ohmic tokamak operations [8]. Having identified the limitations posed by its toroidal complex [12], a substantial modification, of the RFX experiment has been proposed, named RFX-mod2 being the second major modification since its original design. The implementation of the proposed machine modification, involving the components of the whole vessel complex (Fig. 1), has been developed in virtue of an industrial innovation project co-funded by an Italian local authority (Regione Veneto) in the framework of the 2014-2020 European Regional Development Fund. The project, aimed at the development of technologies and innovation of industrial processes for the manufacturing of equipment for energy and environment, has been carried out in partnership between Consorzio RFX (research institution in charge of the conceptual design) and three manufacturing industries with specific competences necessary for the development of the detailed design: o Vacuum vessel and UHV components manufacturing processes (Zanon Pressure Equipment srl, now Brembana & Rolle spa). o Material surface treatments (Alca Technology srl). o Metal additive manufacturing (Sisma spa).

MITICA, the full-scale prototype of the NBI for ITER, is under procurement and is going to be installed and operated in Padova at the Neutral Beam Test Facility at Consorzio RFX. MITICA will be equipped with several diagnostics to assess and optimize the beam production in the beam source and the beam transport in the beamline components of the ITER HNB injectors, which will have a reduced set of these diagnostics. Among this set, the thermal diagnostic is the main in vessel measurement for the protection and calorimetry of the Beam Source (BS) and the beam line components (BLCs). About 700 thermal sensors will be mounted in MITICA with different technologies: thermocouples for distributed measurements on the in vacuum components, fiber Brag gratings (FGSs) as thermal sensors on high voltage biased panels of the BLCs. The full set of the thermocouples can be split into two groups: one foreseen only in MITICA and another one, called ITER-like and subject of this paper, common to both HNB and MITICA. In fact, the requirements about the vacuum compatibility and the radiation hardness to be satisfied in ITER NBIs are more constraining than in MITICA and hence the technical solutions adopted in the design of such sensors have been severely investigated. The ITER-like thermocouples with isolated junction are Type N (Nicrosil-Nisil) without ferromagnetic materials to avoid errors due the strong magnetic fields expected in ITER. Accuracy tolerances are in compliance with ASTM E230 special class. Mineral Insulated Cables (MICs) with MgO as insulating material and Inconel 600 for the metal sheath are mandatory to survive to the high neutron and gamma fluxes. These thermocouples are realized with an Ultra High Vacuum (UHV) compatible termination with ceramic/metallic bushing leak tight tested through helium bombing technique accordingly to 10-9mbar l/s, consistently with the large number of sensors and in accordance with the ITER requirements. The diameter of the MIC is 0.5mm with a sheath thickness of 0.08mm to get a reliable mechanical robustness, compact cabling and compliance with the requirements for remote handling operations. The paper reports the technical specifications for the procurement of the full set of ITER-like sensors. All the manufacturing details are presented and all the tests carried out by Metrologie Srl, an accredited testing and calibration laboratory, are widely described and the results reported demonstrating that the majority of the ITER requirements are satisfied.

The RFX-mod2 device [1], the upgraded version of RFX-mod, will start its operation in 2024 with improved magnetic boundary and diagnostic capabilities. The main device modification is the enhancement of the passive stabilizing shell to plasma proximity. This, coupled to the advanced active feedback control system, is predicted to significantly improve plasma performances in a variety of magnetic configurations, including the reversed-field pinch (RFP), the tokamak and the ultra-low q. For a better characterization of plasma dynamics in all the accessible experimental conditions, several significant diagnostics improvements have been proposed and are presently under implementation. These include the installation of >1000 in-vessel high frequency coils for the characterization of long and very-small scale magnetic fluctuations, of about 500 edge electrostatic probes, distributed throughout the toroidal and poloidal directions, for the analysis of the electron density, temperature, plasma potential and flow in the edge and of plasma-wall interaction and turbulence. A higher repetition rate Thomson scattering system and a strengthened Soft X-ray diagnostics based on the double filter technique, will provide better reconstruction of the topology and the dynamics of the core thermal barriers, which form in helical RFP equilibria. Multiple lines of sight neutron diagnostics based on fast inorganic and organic scintillators and a new Compact Neutral Particle diagnostic system are dedicated to the analysis of anomalous ion heating phenomena in RFP plasmas. A diagnostic neutral Beam (50keV) will investigate the evolution of the ion temperature profile. The electron distribution function will be energy, time and space resolved by means of a soft X-ray imaging system based on a GEM detector in a pinhole configuration. A fast-reciprocating manipulator, housing systems of magnetic and electrostatic probes, will allow the exploration of the edge radial plasma profiles and turbulence even in high current RFP regimes and to characterize SOL and pedestal regions in H-mode tokamak shaped and circular plasmas. An innovative reflectometric technique for plasma position in the tokamak will be tested. The 3D pattern of the plasma wall interaction will be studied with a set of 7 cameras (500 fps) measuring the emission and the Carbon influx. The poloidal distribution of low Z impurities will be obtained with the Light Impurity Tomography (LIT). A cavity-based imaging polychromator designed to resolve 2D absolute intensity images of different emission lines with < 5mm resolution, named MANTIS [2], will gain information on the 2D pattern of electron density and temperature [3]. The edge radial profiles of ne and Te, will be studied with the Thermal Helium Beam [4,5]. The edge characterization is completed by measuring the edge fluctuations due to turbulence [6] thanks to the Gas Puff Imaging diagnostic, already present in RFX-mod. The complex arrays of magnetic coils, along with a system of distributed halo sensors, will allow to validate the electromagnetic modelling of the sideway forces during rapid transients in the tokamak. 1 L. Marrelli et al., Nucl. Fusion 59 (2019) 076027 2 A. Perek et al., Nuclear Materials and Energy 26 (2021) 100858 3 B. Schweer et al., Journal of Nuclear Materials 196-198 (1992) 174-178 4 S. Kajita and N. Ohno, Rev. Sci. Instrum. 82 (2011) 023501 5 M. Agostini et al., Rev. Sci. Instrum. 91 (2020) 113503 6 M. Agostini et al., Rev. Sci. Instrum. 81 (2010) 10D715.

NBTF Progress Report: SPIDER, MITICA and Other

On 08/02/2023 a formal inspection by INAIL+Health minister occurred, with relation to radioprotection "nulla osta" (category A) for NBTF, with positive result. Meeting foreseen on 01/03/2023 for presentation of ICMATE to IO team.

Bilateral agreement with ICMATE signed. Ongoing discussion on MITICA plan wrt preliminary installation+functional test for cryopumps, also involving the cryoplant.

ITER envisages the use of two heating neutral beam injectors plus an optional one as part of the auxiliary heating and current drive system, to reach the desired performances during its various phases of operation. The 16.5 MW expected neutral beam power per injector is several notches higher than worldwide existing facilities. In order to enable such development, a Neutral Beam Test Facility (NBTF) was established at Consorzio RFX, exploiting the synergy of two test beds, called SPIDER and MITICA. SPIDER is dedicated 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, in particular events/issues occurred during operation, which led to the identification of improvement opportunities/necessities (e.g. RF discharges, local burns, water leaks, other damages, configuration/design upgrades to maximize chances/margin to quest target parameters). Parallelly, developments on MITICA, the full-scale prototype of the ITER Neutral Beam Injector (NBI) featuring a 1 MV accelerator and ion neutralization, are underway including manufacturing of the beam source, accelerator and the beam line components, while power supplies and auxiliary plants, already installed, are under final testing and commissioning. Integration, commissioning and tests of the 1 MV power supplies are essential for this first-of-kind system, unparalleled both in research and industry field. 1.2 MV dc insulating tests of high voltage components were successfully completed. The integrated test to confirm 1 MV output by combining invertor systems, DC generators and transmission lines extracted errors/accidents in some components. To realize a concrete system for ITER, said events have been addressed and 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. The lessons learned during the implementation on NBTF and future perspectives are here discussed.

SPIDER is the full-scale prototype of the ion source of the ITER Heating Neutral Beam Injector, where negative ions of Hydrogen or Deuterium are produced by a RF generated plasma and accelerated with a set of grids up to ~100 keV. SPIDER Beam Source design complies with the ITER specific requirements, being fully installed within a vacuum vessel, and with the electrical circuits operating at the residual background pressure. The Power Supply System is composed of high voltage dc power supplies capable of handling frequent grid breakdowns, high current dc generators for the magnetic filter field and RF generators for the plasma generation. During the first 3 years of SPIDER operation different electrical issues were discovered, understood and addressed thanks to deep analyses of the experimental results supported by modeling activities. One of the main issues encountered was the presence of RF discharges on the RF circuit on the backside of the beam source, limiting the operating pressure within the source. The self-excited RF oscillators showed frequency instabilities that were found to be intrinsic limits of the application of this technology to the resonant loads of Neutral Beam Injectors; their understanding allowed assessing the technical basis for the final decision to replace the RF generators in SPIDER and MITICA and change the current ITER baseline to solid-state amplifier technology. Other issues faced regard the effect of the mutual coupling between the RF circuits on board the source, various electromagnetic compatibility problems, the limitation in the operational space to be mitigated improving the SPIDER operating strategies, the revision of the configuration of the magnetic filter field layout to avoid plasma quench. The paper gives an overview on the observed phenomena and relevant analyses to understand them, on the effectiveness of the short-term modifications provided to SPIDER to face the encountered issues and on the design principle of long-term solutions to be introduced during the currently ongoing long shutdown.

An overview of the recent operations and the main results of cesium injection in the Source for the Production of Ions of Deuterium Extracted from Rf plasma (SPIDER) negative ion source are described in this contribution. In experiments without cesium injection, all SPIDER plants were tested to verify the basic expectations on the operational parameters (e.g., electron cooling effectiveness of magnetic filter field) and to determine its operational region. For beam properties, it was shown that the current density varies across the beam in the vertical direction. In preliminary cesium experiments, the expected increase of negative ion current and simultaneous decrease of co-extracted electrons were found, along with the influence of the control parameters (polarization of the plasma electrodes, magnetic filter field) on the SPIDER beam uniformity in the horizontal and vertical directions. It was shown that non-Gaussian tails can be identified in the angular distribution on the plane perpendicular to the beam propagation direction. Stray particles, nonhomogeneous beam and large divergence might result in unexpected heat and particle loads over ITER neutral beam injector (NBI) accelerator grids; it is the goal of SPIDER to assess and possibly to identify suitable methods for controlling these beam features. A major shutdown, planned for late 2021, to solve the issues identified during the operation and to carry out scheduled modifications, is outlined. Such improvements are expected to allow SPIDER to pursue the ITER requirements in terms of negative ion current, electron-to-ion ratio, and beam duration.

Durante lo shutdown lungo di SPIDER verranno installate alcune nuove diagnostiche. Una di queste è la misura della corrente su ciascuno dei quattro segmenti della griglia di estrazione (EG). L'alimentazione della griglia è unica e i quattro segmenti saranno isolati e alimentati in parallelo. La misura di corrente verrà fatta prima che i conduttori di uscita da ciascun segmento si riuniscano nel conduttore di ritorno dell'alimentatore della griglia. La misura deve quindi essere fatta in vuoto e al potenziale della griglia. Le soluzioni ad oggi più soddisfacenti sono uno shunt (100muOhm) con trasmissione in fibra oppure un trasduttore di corrente basato su un sensore di Hall in anello chiuso. Entrambe le soluzioni necessitano di elettronica attiva che potrebbe essere alimentata usando i cavi di estensione delle termocoppie della griglia di estrazione che non saranno più di tipo elettrico. Trattandosi di cavi resistivi, c'è una complicazione dovuta alla caduta di tensione che nel caso di sensore di Hall è fino a 5V perché dipende dalla corrente che il sensore assorbe per compensare il campo magnetico che lo interessa. Oggetto di questa breve nota è una prova volta a valutare le prestazioni di due trasduttori di corrente di LEM e VAC.

Schedules: MITICA plan is being updated after recent developments/decisions on 1 MV insulating transformer reconstruction. Quality: further revision of Quality Plan ongoing, together with some procedures. Internal quality audits have started with primary goal to gather feedback and improvement opportunities. Third parties: preparations for establishment of bilateral agreement with INDA progressing. In parallel, indications have been provided to IO/INDA regarding profiles of specialized resources that could be integrated in the NBTF team.

SPIDER and MITICA plan being reviewed, in view also of the preparation of AWP2023. Updated versions of the SPIDER and MITICA schedules will be made available in view of the next NAC meeting scheduled for October. The RoD of the SC has not yet been finalized. The approval of the RoD is urgent in order to apply the amendment presented to the SC in June.

A progress meeting with DAs was held on 21st of June. The aim is to keep the DAs updated on progress and in particular on lessons learned.