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The Divertor Tokamak Test (DTT) is a new experimental facility whose construction is starting in Frascati, Rome, Italy; its main goals are improving the understanding of plasma-wall interactions and supporting the development of ITER and DEMO. DTT will be equipped with a Neutral Beam Injector (NBI) based on negative deuterium ions, designed to inject 10 MW of power to the tokamak. A fundamental system for the good operations of the DTT NBI will be its Gas injection and Vacuum System (GVS). Indeed, the efficiency of the entire NBI strongly depends on the good performance of its GVS. The GVS for DTT NBI will be composed of two systems working in parallel: a grounded section connected to the main vacuum vessel, and a high voltage part connected to the ion source vessel and working at −510 kV voltage. The grounded part will feature a fore vacuum system (given by screw and roots pumps) plus a high vacuum system based on turbo-molecular pumps located on the side walls of the vessel and Non-Evaporable Getter (NEG) pumps located inside the vessel on the upper and lower surfaces. On the other hand, the high voltage part will feature a fore vacuum system (given by two compact screw pumps mounted on the external surface for the ion source vessel) plus a high vacuum system based on turbo-molecular pumps also located on the sidewalls of the ion source vessel. A dedicated deuterium gas injection will feed the process gas to the ion source and the neutralizer. This paper gives a description of the conceptual design of the GVS for DTT NBI, and of the procedure followed to optimize this system considering the operational requirements and the other constraints of the DTT NBI. • A conceptual design for a Gas injection and Vacuum System (GVS) has been developed for DTT NBI. • The current solution features two parts, one at ground potential and one at high voltage. • The design has been optimized in terms of the most relevant aspects. • With the design solution here presented, the main requirements of the vacuum pumping system are satisfied. [ABSTRACT FROM AUTHOR]

• Heat loads calculations for the accelerator of the ITER HNB have been performed with fully 3D codes. • A different focal point was found for the beam halo, and thus a different heat load repartition among the grids. • The total heat load on each grid was confirmed to be within the acceptable range. • New scenarios were considered: out of perveance conditions and larger temperature for the negative ions. Among the heating and current drive system which are being developed for the ITER experimental reactor, 2 neutral beam injectors with a power of 16.5 MW each will be based on the acceleration of H −/ D − ions up to the energy of 870 keV / 1 MeV. Dealing with an accelerated beam power up to 40 MW, beam divergence, aiming and homogeneity have to be optimized to avoid damaging the beamline components or the grids of the accelerator. The design of the ITER HNB prototype, MITICA, was based on the coupling of several codes starting from experimental inputs available at that time. Heat loads, in particular, were estimated by the 2D version of the code Monte Carlo particle tracing code EAMCC. In the last years, a 3D version of the same code was developed and the operation of negative ion sources for fusion increased the knowledge concerning those parameters which are fundamental to estimate the expected heat loads. In this work, calculations are performed by EAMCC3D and the role of different source and accelerator parameters, such as the negative ion temperature and the beam halo, is highlighted. [ABSTRACT FROM AUTHOR]

The Divertor Tokamak Test (DTT) is a new, super-conducting device, being constructed in Frascati, Italy. DTT will be capable of plasma operations at high density and high heating power, in conditions relevant to address the power exhaust issue in support of ITER operation and DEMO design. DTT foresees the installation of a mix of auxiliary heating systems to couple up to 45 MW to the plasma, including Neutral Beam Injection (NBI). The neutral beam injector is currently being designed, aiming at delivering tangentially to the plasma neutral particles at energy of 510 keV, with a total power of ∼10 MW. In the present work, we apply for the first time the orbit-following Monte Carlo code ASCOT to DTT, in order to analyse with more details the interaction of the high-energy beam, described in real geometry beamlet by beamlet, and the plasma. The results of the simulation give an insight of the behaviour of beam energetic particles in DTT. Thanks to the flexibility of DTT, different plasmas can be generated, e.g. in terms of plasma shape due to different divertor concepts. We present the comparison of two cases with different plasma vertical positions and we analyse the effect on beam absorption in the plasma. We then present a sensitivity scan on plasma density, to verify the coupling of beam power at densities lower than the reference target scenario. These investigations are crucial to provide feedback and suggestions to DTT design and to assess the beam fast ion physics for plasma scenario developments. [ABSTRACT FROM AUTHOR]

• Work completed by the Heating and Current Drive Workpackage for DEMO during the pre-conceptual design phase (2014-2020) • The Electron Cyclotron System is the baseline for DEMO • Description of the Electron Cyclotron System • Description of the Ion Cyclotron System • Description of the Neutral Beam Heating System The European DEMO is a pulsed device with pulse length of 2 hours. The functions devoted to the heating and current drive system are: plasma breakdown, plasma ramp-up to the flat-top where fusion reactions occur, the control of the plasma during the flat-top phase, and finally the plasma ramp-down. The EU-DEMO project was in a Pre-Concept Design Phase during 2014-2020, meaning that in some cases, the design values of the device and the precise requirements from the physics point of view were not yet frozen. A total of 130 MW was considered for the all phases of the plasma: in the flat top, 30 MW is required for neoclassical tearing modes (NTM) control, 30 MW for burn control, and 70 MW for the control of thermal instability (TI), without any specific functions requested from each system, Electron Cyclotron (EC), Ion Cyclotron (IC), or Neutral Beam (NB) Injection. At the beginning of 2020, a strategic decision was taken, to consider EC as the baseline for the next phase (in 2021 and beyond). R&D on IC and NB will be risk mitigation measures. In parallel with progresses in Physics modelling, a decision point on the heating strategy will be taken by 2024. This paper describes the status of the R&D development during the period 2014-2020. It assumes that the 3 systems EC, IC and NB will be needed. For integration studies, they are assumed to be implemented at a power level of at least 50 MW. This paper describes in detail the status reached by the EC, IC and NB at the end of 2020. It will be used in the future for further development of the baseline heating method EC, and serves as starting point to further develop IC and NB in areas needed for these systems to be considered for DEMO. [ABSTRACT FROM AUTHOR]

• Experimental campaign on MTF carried out in QST lab in Naka, Ibaraki, Japan. • Opera, SLACCAD, COMSOL, EAMCC3D correctly predict beam divergence. • Semi-empirical Opera model with current density asymmetry correctly predicts the beam deflection. • COMSOL and EAMCC3D slightly underestimate heat loads on the grids. It is a widely recognized phenomenon that, in negative ion sources for NBI, the magnetic fields in the source cause an asymmetry of the current density at the meniscus, which affects the beam optics. In order to achieve a more realistic estimation of the negative ion beam optics, the ion current density profile at the beamlet meniscus is sometimes assumed to be non-symmetric with respect to the beamlet axis. The new hypothesis tested in this paper is an empirical law obtained by recent investigations on a similar Kamaboko ion source, which relates this asymmetry to the magnetic field strength at the meniscus and to the power injected in the ion source. For the first time this hypothesis is used to directly and correctly predict the value of the beamlet deflection. The MeV ion source Test Facility (MTF) is a high voltage negative ion beam accelerator dedicated to fusion experiments in QST, Naka, Ibaraki, Japan. MTF can be operated as a three- or five-stage multi-aperture accelerator with a configuration and beam energy comparable to the future ITER NBI. In 2019, experiments were conducted on MTF aiming at accelerating a H- beam with high energy and good optics for a long duration of the beam (around 100 s). These experiments allowed to measure the beamlet divergence and deflection as well as the heat loads due to particle impinging on the accelerator grids, over a wide range of operating parameters. The predictions of several beam simulation codes (Opera, COMSOL, SLACCAD, EAMCC3D) have been compared to the experimental measurements, which allows to obtain a good agreement in terms of beam optics and a fair agreement in terms of heat loads. [ABSTRACT FROM AUTHOR]

In a nuclear fusion reactor, some components of the system are subjected to huge thermal loads. These components, such as grids, neutralizers, calorimeters, ducts, divertors, blankets etc. have to be actively cooled in order to avoid damage, functionality problems and specially to ensure their reliability. The design process of such components requires taking into account several physical phenomena, with the goal of not only satisfying the functionality conditions, but also with a focus on the structural verifications and component lifetime. For assessment of the structural design for components in ITER, the ITER Structural Design Criteria for In-vessel Components (SDC-IC) represent the reference verification criteria. The aim of this work is to investigate the relationship between the cooling system performance and the fatigue life. In this work a generalized framework for the evaluation of cooling system components is presented, taking into account the necessary input parameters, performing numerical analyses, division of component into several interest regions and assessing the fatigue life. The performed numerical analyses are a set of coupled CFD and thermo-structural analyses. A case study is presented for a critical component, the Grounded Grid (GG), inside the SPIDER beam source experiment at Consorzio RFX, Padova, Italy. The presented test case includes cooling system effectiveness analysis, assessment of the temperature distribution on housing during operations, mapping the pressure distribution inside the cooling system and finally assessment of the maximum number of allowed cycles. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]