The diagnostic neutral beam injector (DNBI) of the RFX-mod2 experiment (Consorzio RFX, Padua) is expected to provide novel and significant information about the reversed field pinch (RFP) confinement of fusion plasmas. The present DNBI, built by the Budker Institute of Plasma Physics, features an arc discharge H+ source, coupled to a 4 grid 50 keV acceleration system, to produce a 50 ms, 5 A ion beam. This contribution presents the latest upgrades of the DNBI. The high-voltage deck (HVD) was completely restructured, and the power transfer was simplified to a single-phase insulation transformer. The 50 kV distribution circuit was modernized and made safer against breakdowns. Several custom power supplies in the HVD were designed and procured; their electronic boards were developed to be multipurpose in the DNBI, simplifying the system and improving its maintainability. The features of the magnetic insulation power supply (MIPS) and gas valve power supplies (GVPS) are presented in detail. Finally, the new PLC control system was improved for better protection of the CPU against overvoltages and for better scalability and maintainability of the system.

Diagnostic Neutral Beam Injectors (DNBI), through the combined use of Charge Exchange Recombination Spectroscopy (CHERS) and Motional Stark effect diagnostics (MSE), are a well-known tool to access important information about magnetically confined plasmas, such as radial profiles of ion temperature, ion flow, impurity content and intensity and direction of the magnetic field. For this purpose, a DNBI was installed and operated in the RFX-mod experiment, which was designed to confine plasma mainly through the Reversed Field Pinch configuration. The DNBI, designed and built by the Budker Institute of Nuclear Physics (BINP), was based on a source of positive hydrogen ions, accelerated to 50 keV and for a maximum ion current of 5 A. The beam could be modulated and the maximum overall duration was 50 ms. With the upgrade of RFX-mod to the present RFX-mod2 machine, the DNBI is being renovated to solve several power units faults and improve the overall reliability of the system. The 50 kV power supply is being improved, as well as the power supplies in the high voltage deck and its insulation transformer. Magnetic field survival tests were performed on the toroidal-core-based DC-DC converters that should power the electronic boards in a reliable way. The control system, originally based on CAMAC technology, was redesigned to be fully replaced. This contribution reviews the technical criticalities emerged in the DNBI check-up and the new solutions adopted to make the DNBI operative and more reliable.

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 integration of the new plasma position reflectometer in the RFX-mod2 experiment (the upgraded version of the previous RFX-mod that operated until 2015) is presented in this contribution. Particular attention has been devoted to the high field side subsystem where an antenna pair will be installed in the inner midplane. Waveguides, insulated through the application of a ZrO painting, will be routed in between the vacuum vessel and the conductive shell to a vertical port. The severe constraints in terms of physical space available guided the antennas design: a hoghorn antenna model was first numerically modeled and, due to the complex geometry, produced through metal additive manufacturing; then, a post-production surface treatment allowed achieving a surface with characteristic roughness and conductivity comparable to traditional manufactured antennas. Different bench tests are presented to assess the system performance.

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 2025. 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 analysis of x-ray spectra and neutron rate, a reflectometric diagnostic 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 to improve the control of the electron density and of the impurity influxes by means of proper treatment of plasma facing components with in-vessel fixed electrodes distributed over the first wall. 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.

This paper presents the results of experimental tests on samples made of copper coated with alumina layer, performed to assess the reliability of its dielectric properties for applications in the low temperature plasma at the edge of a fusion device. The cue of the study was related to the plasma facing components of the RFX-mod2 fusion device (Marrelli et al., 2019, Peruzzo et al., 2023, Peruzzo et al., 2019), devoted to the experimental study of the magnetic confinement of fusion plasmas in a variety of configurations, including the reversed-field pinch and the tokamak. In RFX-mod2 an in-vacuum copper shell for the passive stabilization of MHD modes will surround the plasma. To avoid potentially harmful electrical discharges, which could be induced by rapid transients of the plasma current, this structure must be covered with an electrically insulating layer. For RFX-mod2 an alumina coating was chosen, whose dielectric properties have been tested both in air and in the presence of weakly ionized plasma. Electrical tests, conducted on copper samples with alumina deposits of about 100 mu m thickness, revealed that the ceramic layer has a high electrical resistance value in air (> > 1 G Omega), but electrical discharges can occur in presence of a weakly ionized plasma, depending on compactness and porosity of the alumina layer, causing local melting of the alumina and expulsion of copper droplets from the substrate. Scanning Electron Microscope (SEM) analyses revealed that in the failed samples the ceramic layer was irregular and rough, with interconnected cavities and cracks, which could reduce its effective thickness and explain the dielectric breakdown at relatively low voltages (< 400 V). The analyses also showed that samples with a more compact layer present a higher dielectric strength in the presence of the plasma, highlighting that compactness and porosity play crucial roles in ensuring good insulation for materials in a plasma. This study led the definition of the requirements for the insulating coating of the plasma facing components of the RFXmod2 fusion machine, however the results can be useful for other fusion and non-fusion plasma applications requiring electrical insulation, which can span from industrial devices to spacecrafts.

The RFX-mod2 installation is planned to be completed by 2024 and the start of operations is expected in 2025. The high flexibility of the machine (already tested in the previous RFX-mod experiment) allows operation in Reversed Field Pinch and tokamak configuration as well as ultra-low q pulses. In this work we present predictive analysis on transport, performances and plasma control in RFX-mod2 in view of the first experimental campaigns.

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.

• Upgrade of RFX-mod experiment proposed for studies of RFP and Tokamak physics. • High-performance polymers and surface coatings for in-vessel components insulation. • In-vessel components made from Metal Additive Manufacturing . • Effective collaboration between research institutions and industries. • Support of local authorities to scientific research and technological innovation. A substantial modification of the toroidal complex of the RFX experiment, named RFX-mod2, is currently under completion, involving the whole core system of the machine and in particular the vacuum vessel, the entire plasma facing components and a wide set of in-vessel diagnostic systems . The combination of challenging requirements, in terms of electrical insulation and vacuum compatibility of in-vessel components, and stringent geometrical constraints to comply with interfaces to existing machine components, in particular external coils and diagnostic systems, called for the adoption of novel technological solutions developed in collaboration with local industries in the framework of an industrial innovation project co-funded by Regione Veneto. The paper presents an overview of the technological solutions implemented and tested during the manufacturing phase of the new components of the RFX-mod2 machine complex.

This paper discusses the electrical insulation of a copper structure, which will be installed in the RFX-mod2 machine, in the presence of a weakly ionized plasma. The effectiveness of aluminum oxide (alumina) coating in preventing electrical discharges was evaluated through tests conducted both in air and in a weakly ionized plasma environment. It was shown that copper samples with alumina surface deposition sustained a voltage of about 2 kV in air but showed dielectric breakdown in the presence of plasma. SEM analysis revealed an irregular microstructure with a high degree of porosity and large internal cavities, which tend to reduce the effective coating thickness. New copper prototypes with more compact and less porous alumina deposition showed effective electrical insulation in the presence of plasma, highlighting how porosity and compactness are important factors in ensuring electrical insulation in the presence of plasma. The use of other materials, such as paints, silicones and resins, to insulate the inner part of holes in the copper structure, due to limitations in applying alumina coatings with plasma spray techniques, is also discussed.

RFX-mod2 (R = 2.0 m, a = 0.49 m), the upgraded version of the previous RFXmod fusion device, will be equipped with a new reflectometry system specifically designed for plasma position control purposes. Featuring a high temporal and spatial resolution and being suitable for harsh fusion reactor experimental conditions (long pulses, high neutron fluxes), reflectometry has been proposed as a good candidate for this task. On RFX-mod2 the diagnostic system will consist of four bistatic ultrafast independent reflectometric units working in the frequency range (16–26.5 GHz) and installed in four different poloidal locations at the same toroidal angle: two on the equatorial plane (high field side/low field side) and two at the vertical top/bottom ports. Standard pyramidal horns will be installed in the external midplane and in the vertical ports, while parabolic hoghorn reflectors have been designed for the internal midplane. Different technical solutions for the integration in the machine, as the additive manufacturing for the antenna production, are presented. Despite the differences with respect to the application on large Tokamaks like DEMO, the RFX-mod2 plasma position reflectometer can contribute to test on a simple machine some of the issues related to the development of a reflectometry based plasma position and shape control.

This article describes the present upgrade of the RGM system, the magnetic fault detection system designed for RFX (named RGM, from the Italian acronym rilevamento guasti sistemi magnetici), to be fully compliant with the new device RFX-mod2, considering both machine and operational modifications. RGM was originally designed for RFX, and it was based on the symmetries of: 1) the magnetic structures of RFX windings and 2) the topology of the electrical circuits. Then, in the early 2000s, a refurbishment of RGM took place in view of the machine upgrade to RFX-mod. Nowadays, the present upgrade to RFX-mod2 requires further modifications to the RGM system, making it compatible with new machine and operational specifications. RGM will be equipped with several additional magnetic sensors to overcome some constraints due to its original working principle and refurbished with redesigned electronic boards based on state-of-the-art technologies and components. The new system is designed to be fully compatible with the existing hardware, in order to keep advantage of the strengths of the original system.

This paper summarizes the results of preliminary insulation tests performed on copper samples covered with a Al2O3 (aluminum oxide, commonly called alumina) film, deposited using the Atmospheric Plasma Spray (APS) technique.The electrical insulation tests were conducted in a vacuum chamber in the presence of weakly ionized plasma, in order to simulate the operating conditions that are expected in the plasma edge region in the RFX-mod2 experiment, in particular in the region included between the first wall and the copper passive stabilization shell.A minimum thickness of alumina deposit is required in order to ensure the dielectric rigidity required by the conditions of use in RFX-mod2. From electromagnetic simulations, it is estimated that the dielectric thickness must be such as to withstand a voltage greater than 1.5 kV.The experiments were carried out by applying a potential difference, generated by a bank of capacitors, between the sample and a copper pin placed in front of the alumina covered surface. The results of the tests conducted show that an alumina thickness of about 0.1 mm is sufficient to guarantee electrical insulation, up to a voltage of about 2.5kV.

In the RFX-mod2 experiment presently under construction, devoted to the study of magnetic confinement of fusion relevant plasmas, significant electric fields, in the kV/mm range, are expected to form in between in-vessel conductive plasma facing components during transient plasma current phases (start-up and fast termination). While such electric fields are of no concern for components in vacuum, the presence of a scrape-off plasma at the edge (electron density ne 1016 ÷1018 m-3, electron temperature Te of few eV) can create the conditions for potentially dangerous arc formation. For this reason part of the plasma facing components (in particular the graphite ‘first wall’ tiles covering the copper ‘stabilizing shell’ placed within the plasma chamber) require a proper conditioning technique capable of maintaining the insulation between conductive components even in presence of the scrape-off plasma.An experimental apparatus has been developed in order to test the conditions for the arc formation and prevention between two electrodes immersed in a plasma generated by a hot emitting filament. The results of an extensive experimental campaign will be presented, aimed at demonstrating the possibility of gaining a sufficient electrical conditioning by applying the standard conditioning technique usually employed for higher voltage ranges. It consists of a sequence of high voltage pulses applied to the pair of electrodes with current limitation, in the presence of a background cold plasma with low ionization degree. The experimental procedure is such that the voltage of the pulses is slightly increased when arcing ceases, until the final desired voltage level is achieved (2.5 kV). Different electrode materials have been tested in a variety of plasma conditions in terms of electron density and working gas pressure.

An upgrade of RFX--mod toroidal machine assembly is under development, aimed at widening the explored operational scenarios both in RFP and Tokamak configuration. The main design driver for this machine upgrade, named RFX-mod2, is the enhancement of the 'shell-plasma proximity', which is expected to provide a significant improvement in the plasma magnetic confinement. In order to achieve this goal, the existing Passive Stabilizing Shell will be adapted in order to be enclosed in the new vacuum vessel. This reconfiguration of the internal components of the experiment requires the design of a new structure aimed at supporting both the shell and the plasma facing components. The paper describes the design aspects of the shell assembly, with particular focus on the thermo-mechanical FEM analyses performed to verify the design of the components at the expected operating conditions.

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.

The requirements of ITER neutral beam injectors (1 MeV, 40 A negative deuterium ion current for 1 h) have never been simultaneously attained; therefore, a dedicated Neutral Beam Test Facility (NBTF) was set up at Consorzio RFX (Padova, Italy). The NBTF includes two experiments: SPIDER (Source for the Production of Ions of Deuterium Extracted from Rf plasma), the full-scale prototype of the source of ITER injectors, with a 100 keV accelerator, to investigate and optimize the properties of the ion source; and MITICA, the full-scale prototype of the entire injector, devoted to the issues related to the accelerator, including voltage holding at low gas pressure. The present paper gives an account of the status of the procurements, of the timeline, and of the voltage holding tests and experiments for MITICA. As for SPIDER, the first year of operation is described, regarding the solution of some issues connected with the radiofrequency power, the source operation, and the characterization of the first negative ion beam.

High heat fluxes are exchanged in fusion machines (up to 50 MW m−2), thus producing elevated temperature and requiring thermal monitoring and control. The design of a temperature measurement system for the RFX experiment is developed through three-dimensional nonlinear transient finite element simulations of the torus assembly under upgrade from “mod” to “mod2” with enhanced magnetic front-end, vacuum confinement barrier, and first wall. Analyses show how heat fluxes applied at the plasma-facing materials are transmitted, attenuated and delayed, through the machine parts. Results identify the passive stabilising shell as the instrumentable component closest to the plasma boundary able to follow the thermal behaviour by the detection of temperature variations at least of 10 °C during plasma pulses with a response time of about 200 s. Allowable temperature limits of materials are verified simulating a full experimental day with 24 plasma pulses, in particular at the shell supporting rings made of polyamide-imide and at the vessel spacers made of polyether-ether ketone-coated stainless steel. Simulations of the pulse discharge cleaning demonstrated the capability of the system to provide the required power for first wall conditioning (25 kW) and the need to realise a duty cycle (1-h on/3-h off) limiting the average heat flux and the maximum temperature (55 °C) at the vacuum vessel sealing elements in order to minimise differential thermal deformations. Proposed layouts of temperature sensors are able to detect the maximum temperatures expected during operation.

RFX-mod2 is the latest upgrade of the reversed field pinch machine operated at Consorzio RFX. A significant modification consists of replacing the first wall files, proposed as a key factor to improve the gas density with reduction of hydrogen retention. Polycrystalline graphite has been identified as file material given its high thermal conductivity (up to about 165 Wm(-1) K-1), uniformity and small grain size (about 20 mu m), and high mechanical strength (100 MPa compressive strength, 30 MPa tensile strength). With respect to RFX-mod, the mechanical resistant section of files has been increased coherently with the magnetic front-end modification that foresees the files supported by the existing MHD passive stabilising shell, so decreasing the maximum stress at 3.5 MPa calculated from finite element analysis that simulates the operating condition. This low stress level together with a measurement of the experimental loads during next RFX-mod2 operations could qualify the use of extruded graphite for a possible further first wall change in the future. Indeed, extruded graphite is considered attractive given its high directional thermal diffusivity (about 50 % better then polycrystalline graphite) to enhance the heat transmission and so improving the gas density control, and the low stress induced may allow use of this mechanically less performing graphite.

The poloidal distribution of electromagnetic loads during fast transients in the vacuum vessel of COMPASS-U tokamak is calculated analytically. The estimates are then compared with CarMaONL numerical simulations. The results show that the poloidal eddy currents in the vacuum vessel must be taken into account for the proper evaluation of disruption forces in the COMPASS-U tokamak.