CuCrZr is one of the most interesting copper-based alloys due to the combination of high conductivity and mechanical properties. In this work, the material was additively manufactured through Laser Powder Bed Fusion (LPBF) technology. Two machines were used to investigate the influence of the building process on the final properties, an EOSint M280 and AMCM M290-1 kW. Several heat treatments were carried out to evaluate their impact on the mechanical properties and the thermal conductivity: solution annealing + aging, direct age hardening, and an innovative low-temperature annealing, as a stand-alone treatment or followed by an aging treatment. A maximum thermal conductivity of 340 W/mK and an electrical conductivity above 93 %IACS were measured on the samples treated with the low-temperature annealing treatment followed by aging at 550 degrees C for 3 h. An ultimate tensile strength (UTS) of almost 600 MPa was achieved by treating the material with a direct age hardening treatment at 500 degrees C for 3 h. In this study, we demonstrate that, for the same powder, different building processes did not affect the material performance. At the same time, heat treatments play a major role in determining the properties. TEM analyses were also carried out on the as-manufactured, solution annealed, and directly aged material, to investigate the precipitation of Cr- and Zr-rich particles and correlate the microstructure and the final properties.

In this study, the processability of CuCrZr alloy with additive manufacturing (AM) technology and the performance achievable with Direct Age Hardening treatments for nuclear fusion applications were investigated. This copper alloy is one of the most interesting for the field: it is easier to manufacture through Laser-based additive manufacturing technology and mechanically superior compared to pure copper, and it ensures values of thermal conductivity high enough to be considered a valid substitute for pure copper in many applications. The investigation on CuCrZr alloy was carried out in order to examine the influence of Direct Age Hardening (DAH) treatments on physical and mechanical properties. Laser Powder Bed Fusion technology was used to produce samples with CuCrZr alloy. The additive manufacturing process involved a machine provided with a 370 W IR laser and a preliminary process optimization was carried out to find the printing parameters that assured the highest density (99.15 %), which confirmed the processability of CuCrZr alloy also with low IR laser power. Then, three different DAH treatments were tested and the performance of DAHed material was compared to that of the alloy in as-built conditions. Precipitation phenomena were investigated with DSC analyses, revealing the effectiveness of the treatment already after 1 h. A deep microstructural investigation revealed a fine cellular structure formed during solidification and the presence of nanometric precipitates starting from the as-built condition. The presence of microstructural defects was also investigated. Mechanical performance and thermal conductivity were tested, too: the as-built samples showed limited properties, while very promising results for the use of additively manufactured CuCrZr components have been obtained after the DAHs. The ultimate tensile strength (UTS) and yield strength (YS) doubled the as-built values after 1 h treatment at 550 degrees C. The thermal conductivity reached three times the initial condition (from 100 W/mK to 300 W/mK).

The ITER Neutral Beam Test Facility (NBTF) is hosted in Padua and includes two experiments: MITICA, the 1 MeV full-scale prototype of the ITER HNB injector, and SPIDER, the 100 keV full-size ITER Radio Frequency (RF) negative ion source. SPIDER and MITICA experiments are actively cooled by Ultrapure Water (UPW) to electrically insulate in-vessel components that are biased to high voltage levels. Water conductivity is an important monitored parameter to ensure components' insulation by limiting the leakage current of active cooled equipment. A very low conductivity water is especially important in MITICA, where components need to operate up to 1 MVdc, an insulation level beyond the actual industrial standard. Careful selection of suitable materials for any in-vessel (vacuum insulation) and out-vessel component (air insulation) is of utmost importance, as their interaction with water and different environment may affect their chemical and mechanical properties. Additionally, materials selection can severely influence water chemical characteristics: cooling circuits are made of metals (copper, aluminium, steel) and insulating materials (plastic, rubber) when connecting parts at different electric potentials.During the first years of cooling plant exploitation, it was shown that water degrades more quickly than estimated by design. The Primary Circuits (PCs) that showed the most severe water degradation during operation are SPIDER and MITICA power supply ones, respectively called PC01 and PC08. This paper describes the results of specific experimental tests performed on MITICA PC08 to evaluate possible causes of water degradation and detect sources of contaminants that might compromise future experimental campaigns. The cooling circuit was subdivided in different sections and water circulation tests were performed at constant temperature and flowrate. Water samples were collected and Inductively Coupled Plasma Mass Spectrometry (ICP-MS) analyses were performed to quantify the type and amount of metals released. Moreover, components samples collected along their cooling circuits were characterized by Scanning Electron Microscope (SEM) technique to detect undesired contaminants and immersed in UPW to study their corrosion behaviour using metal release tests.

Borated water is used as a shield in nuclear fusion cooling circuits. General corrosion, activated corrosion products (ACPs) formation and stress corrosion cracking initiation of 316 L steels exposed to ultrapure water (UPW) or 8000 ppm B water at 80 C were tested. A Ni enriched sub-oxide layer, a transition layer and oxide layer were observed using advanced characterisation (STEM-EELS, APT). The oxide formed in UPW was protective (Cr:O 40:45), the oxide formed in 8000 ppm B was non-passivating. 8000 ppm B led to higher release of Fe, Cr and Mo, 316 L was more prone to SCC initiation and enhanced ACPs formation.

The experimental fusion reactor ITER will feature two (or three) heating neutral beam injectors (NBI) capable of delivering 33(or 50) MW of power into the plasma. A NBI consists of a plasma source for production of negative ions (extracted negative ion current up to 329 A/m 2 in H and 285 A/m 2 in D) then accelerated up to 1 MeV for one hour. The negative ion beam is neutralized, and the residual ions are electrostatically removed before injection. The beamline was designed for a beam divergence between 3 and 7 mrad. The ion source in ITER NBIs relies on RF-driven, Inductively-Coupled Plasmas (ICP), based on the prototypes developed at IPP Garching; RF-driven negative-ion beam sources have never been employed in fusion devices up to now. The recent results of SPIDER, the full size ITER NBI ion source operating at NBTF in Consorzio RFX, Padova, measure a beamlet divergence minimum of 12mrad and highlighted beam spatial non-uniformity. SPIDER results confirmed the experimental divergence found in smaller prototype sources, which is larger compared to filament-arc ion sources. Although prototype experiments have shown that the extracted current requirement can be achieved with minor design improvements, the beamlet divergence is expected to marginally achieve the design value of 7 mrad, which in multi-grid long accelerators results in unexpected heat loads over the accelerator grids. A contributor to the beam divergence is the energy/temperature of the extracted negative ions, so it is believed that plasma differences between the two source types play a role. Research is focused on the plasma parameters in the ion source. One RF driver, identical to the ones used in SPIDER, installed in a relatively small-scale experimental set-up, inherently more flexible than large devices, is starting operations devoted to the investigation of the properties of RF-generated plasmas, so as to contribute to the assessment of negative ion precursors, and of their relationship with the plasma parameters, particularly when enhancing plasma confinement. The scientific questions, that have arisen from the preliminary results of SPIDER, guided the design of the test stand, which are described in this contribution, together with the diagnostic systems and related simulation tools. The test stand, which shares with the larger experiment all the geometrical features and constraints, will allow technological developments and optimized engineering solutions related to the ICP design for the ITER NBIs.

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.

A future nuclear fusion plant DEMO needs neutral beam injection (NBI) systems requiring a vacuum pumping system with a very high pumping speed, in the order of several 1000 m3/s. Large customized cryopumps are actually used to meet the requirements. A promising concept for future NBI systems is based on high capacity getter materials. The ZAO® alloy, developed by SAES Getters, Italy, provides a drastically improved performance for the pumping of hydrogen compared to conventional getter materials. This paper describes the experimental characterization of a large pump of scalable size with 15 kg of ZAO® and the achieved results, in particular the systematic investigation of sorption characteristics and regeneration behaviors. Major findings include a very good repeatability of the sorption performance, a reduced pumping speed at higher pressures only above the NBI relevant level, an improved performance (+20%) with elevated getter temperature and an isotope independent sticking coefficient for hydrogen. Furthermore, improved operation experience and regeneration prediction tools have been developed. Employing the experimental results, a simulation task was performed and the sticking factor of the getter cartridge was determined with 7%.

• 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.

In vacuum technology, capture pumps based on Non Evaporable Getters are commonly applied to ultra-high vacuum systems. Recent improvements in the absorption of hydrogenic species, with the introduction of Zr-V-Ti-Al alloys (ZAO & REG;), make them an appealing and viable solution for the application in fusion research, and in particular for the vacuum system of neutral beam injectors (hydrogen pumping speed of thousands of m3/ s, pressure of tens of mPa). This paper describes the characterization of the new NEG material in pumps of increasing dimensions, including the development, construction and test of a large mockup pump of modular design, to demonstrate the scalability of the technology. Effective pumping speeds of the order of 14 m3/s or higher at a concentration of 130 Pa m3/kg were achieved by the mockup pump, for an installed getter mass of about 16 kg, and a stability within 10% up to 1300 Pa m3/kg The measured effective pumping speed per unit area of sintered disks is of the order of 3.5 m/s, corresponding to 4.9 m/s at the disk surfaces as derived from numerical simulations. General guidelines for the design of large NEG pumps for hydrogen are discussed, including thermal aspects and duty cycle of the pump.

An improvement for the vacuum system of the multidriver radio frequency (RF) prototype negative ion source SPIDER has been developed, to allow operating at high RF power, while minimizing the breakdown probability on the outside of the plasma source. A first-of-its-kind large nonevaporable getter (NEG) pump, based on a modular design of 384 cartridges totaling about 350 kg of ZAO® alloy (composed by Zr-Ti-V-Al) with an installed pumping speed at a room temperature of 330 m3/s for hydrogen, will complement the existing pumping system, based on eight cryogenic pumps and four turbomolecular pumps totaling about 90 m3/s in hydrogen. The vessel pressure during absorption is expected to be between 20 and 40 mPa, while during the getter regeneration, the peak operating pressure will be around 2 Pa. The NEG pump will use an additional vessel module, with integrated thermal shields to protect the in-vessel components during activation and regeneration of the pump, to be carried out at temperatures in the range of 550–600 °C. Integrated thermal analyses were carried out to verify the low heat load on pre-existing in-vessel components with a low limit of acceptable temperature, and to provide boundary conditions for the mechanical verifications of the pump structure. The scenario of cyclic hydrogen load was calculated considering the SPIDER operation modes, the expected gas throughput, and the cumulated load to the pump, to determine the regeneration temperature and auxiliary pumping necessary to make the regeneration duration compatible with the high availability of the system. The upgrade of the auxiliary pumping system is also described, as well as the mitigation of air or water exposure risk during regeneration of the NEG pump.

SPIDER is the negative ion source testbed for ITER neutral beam injector. It is currently the largest negative ion source ever built, equipped with a 100 keV accelerator, aiming at producing a negative ion beam with an extracted current density in hydrogen of 355 A/m2, beam-on time of one hour. During the first 3.5 years of operation several improvements on the SPIDER source and its operating conditions were implemented. This paper focuses on the description of materials degradation phenomena in SPIDER and on the lessons learnt that were derived from components inspection during the already planned shutdown in 2022. The aim of these inspections was to improve the design of some components and to repair or refurbish damaged components. In particular, this paper describes the characterisation of damaged molybdenum coating, testing of surface cleaning techniques and experimental evidence of caesium vapour uneven distribution inside the plasma chamber. Sputtering due to beam particles was also investigated experimentally to observe its effect on materials used in SPIDER.

Additive manufacturing (AM) is revolutionizing the industrial scenario. Four copper samples have been printed via Laser Powder Bed Fusion (LPBF) at DIAM Laboratory (INFN—Sezione di Padova, Padova, Italy). Samples had different geometrical characteristics, to test the feasibility of the AM as a productive technique for the creation of unsupported copper structures that are characterized by surfaces with a very small inclination angle, where supports cannot be placed. Parts have been printed successfully even in case of 18° of inclination of unsupported walls with respect to the horizontal plane, and on the same samples, surface finishing treatments (performed by Rösler Italiana S.r.l. and INFN-LNL) have been performed to reduce the roughness of the down-facing surfaces. Indeed, the down-skin regions are the most critical areas of AM parts. Several surface treatments are under investigation: mass-finishing treatments (mechanical and chemically assisted mechanical processes), chemical polishing, and electropolishing, and for some of them, the results are extremely positive: from an initial roughness (Ra) of 30–35 µm, the treatments allowed us to achieve a Ra value lower than 1 µm. The study here exposed presents a good way to rapidly reduce the roughness of 3D-printed parts, reaching a mirror-like aspect.

The Divertor Tokamak Test (DTT) vacuum vessel (VV) is a toroidal chamber that ensures an enclosed vacuum environment for the plasma, a confinement barrier, and temperature control. Water flows in the double-shell D-shaped VV cross-sectional structure made from stainless steel to operate at a set temperature. Beyond temperature control, the water cooling circuit needs to act as a neutron shielding media to protect the structures installed outside of the VV, namely, the superconducting coils. The shielding function is achieved due to the addition of boric acid in the water. The requirement on the borated water for the DTT VV is to have 8000-ppm B solution highly enriched in (SUPERSCRIPT)-S-1 ZEROB (95% (SUPERSCRIPT)-S-1 ZEROB). Given the lack of water chemistry guidelines for fusion power plants, the water chemistry requirements from fission power plants were investigated. In this work, general corrosion of stainless steel, 316L type family, in concentrated borated water solutions was investigated experimentally using metal release test. Samples were exposed to ultrapure water (UPW) and 8000-ppm B borated water at 80 degrees C for one week to quantify the amount of ions released in solution. 316L general corrosion was studied considering the different water chemistries, UPW versus borated water, and steels microstructures. DTT VV presents many welded joints, so general corrosion of welding-induced microstructure was here investigated compared to 316L base microstructure. The release of ions from general corrosion was found to be more influenced by water chemistry than microstructure.

SPIDER is the 100 keV full-size Negative Ion Source prototype of the ITER Neutral Beam Injector, operating at Consorzio RFX in Padova, Italy. The largest Negative Ion Source in the world, SPIDER generates an RF driven plasma from which Deuterium or Hydrogen negative ions are produced and extracted. At the end of 2021, a scheduled long-term shutdown started to introduce major modifications and improvements aiming to solve issues and drawbacks identified during the first three years of SPIDER operations. The first action of the shutdown period was the disassembly and characterization of the SPIDER beam source after removal from the vacuum vessel and its placement inside the clean room. Each component was carefully assessed and catalogued, following a documented procedure. Some source components, i.e., the Plasma Grid, Extraction Grid and Bias Plate, revealed the presence of different and non-uniform red, white and green coatings that might be correlated to back-streaming positive ions impinging on grid surfaces, electrical discharges and caesium evaporation. Thus, several analyses have been carried out to understand the nature of such coatings, with the study still ongoing. The evidence of caesium evaporation and deposition on molybdenum-coated SPIDER components, such as the formation of oxides and hydroxides, is demonstrated through surface characterization analyses with the use of the Scanning Electron Microscope (SEM), X-ray Diffraction (XRD) and X-ray Photoelectron Spectroscopy (XPS).

Non Evaporable Getter (NEG) technology has been studied as an alternative pumping solution in facilities for fusion research since the 80 s. The recent development of a new generation of Zr-V-Ti-Al (ZAO®) getter alloys with improved features makes NEG technology an attractive option in several fusion applications, including in particular, Neutral Beam Injectors (NBI). The main advantages of ZAO® NEG pumps are related to the high pumping speed for H2, the high capacity that results into a very promising availability, the ease of integration and the low power consumption. From the vacuum point of view, research fusion facilities are in the high vacuum pressure range (e.g. 1∙10−6 ÷1∙10-5 Pa background pressure) and may present leaks and significant outgassing from internal components. Specific activities have been devoted to test the NEG pump robustness, i.e. the ability to maintain adequate performances in conditions very far from UHV. In particular, the effects of carbon oxides and hydrocarbon adsorption have been investigated, as well as the outgassing from materials like polyimide or even fluorinated polymers as PTFE. The results show that the ZAO® getter material is intrinsically robust, able to perform repeated H2 adsorption/desorption cycles with stable performances, and recover most of the original pumping speed even after significant exposure to hydrocarbons and fluorine species. These outcomes give promising indications for the future use of NEG solutions in real fusion facilities.

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.

The main purpose of the Divertor Tokamak Test (DTT) is to study solutions to mitigate the issue of power exhaust in conditions relevant for ITER and DEMO. The key feature of such a study is to equip the machine with a significant amount of auxiliary heating power (45 MW) in order to test different divertor solutions. According to the Italian project, the experiment is foreseen to operate with the following main parameters: B-T = 6 T, I-P= 5.5 MA, R-O= 2.1 m, a = 0.65 m and a pulse duration of 95 s. It shall be able to study different divertor magnetic configurations and reach a reactor relevant power flow to the divertor. The proposed mix of heating power foreseen to achieve the target value of 45 MW delivered to the plasma will be provided by Electron Cyclotron Resonant Heating (ECRH), Ion Cyclotron Resonant Heating (ICRH) and Negative-ion-based Neutral Beam Heating (NNBH). In this framework, the conceptual design of a NNBH system for DTT is here presented, with a particular focus on the technical solutions adopted to fulfil the requirements and maximize the performances. The proposed system features two beamlines providing deuterium negative ions (D-) with an energy not smaller than 300 keV and an injected power of 5-8 MW each. The design of the main components of the injectors is described in detail, explaining the motivations behind the main design choices and the related evaluations by means of physics and engineering simulations.