Search

The Reversed Field Pinch (RFP) configuration looks to be an attractive option for fusion-fission hybrid reactors: the toroidal magnetic systems would be made of copper coils instead of more expensive superconductive magnets; fusion conditions could be reached by ohmic heating only, therefore additional heating systems would not be required; the fission blanket could be located in the most external part of the torus thus facilitating maintenance operations. The paper aims at assessing the potentialities, such as fuel fertilization and/or nuclear waste transmutation and electricity production of a hybrid reactor with a RFP fusion core (R = 6 m, a = 1) whose conceptual design and plasma performances are based on RFX-mod, the largest RFP experiment currently in operation. Fusion conditions can be reached by heating a D-T plasma up to 9.6 keV by ohmic heating, generated by a 20 MA plasma current induced and sustained by flux swing only. The neutron flux (2.1 × 1013 fast neutron/cm2/s) is used to breed tritium in both the inner and outer blanket sections and induce fission reactions in dedicated areas in the external blanket section where Pu + MA (60%)-Zr (40%) rods are located. Both neutronic and safety analyses corroborate the viability of a FFH reactor with a RFP core. [ABSTRACT FROM AUTHOR]

• The main purpose of DTT is to study solutions to mitigate the issue of power exhaust in conditions relevant for ITER and DEMO. • The design of the NBI for DTT was driven by integrated physics and engineering evaluations, aiming at optimizing all the main critical aspects at the same time. • Flexibility has been considered as the main design guideline, following the needs of the DTT experiment. • Efficiency and RAMI have also been considered as guidelines. 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 0 = 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. [ABSTRACT FROM AUTHOR]

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]

In the EU DEMO design (Romanelli, 2012; Federici et al., 2014), due to the large number of complex systems inside the tokamak vessel it is of vital importance to address the in-vessel integration at an early stage in the design process. In the EU DEMO design, after a first phase in which the different systems have been developed independently based on the defined baseline DEMO configuration, an effort has been made to define the interface requirements and to propose the strategies for the mechanical integration of the auxiliary heating and fuelling systems into the Vacuum Vessel and the Breeding Blanket. This work presents the options studied, the engineering solutions proposed, and the issues highlighted for the mechanical in-vessel integration of the DEMO fuelling lines, auxiliaries heating systems, and diagnostics. [ABSTRACT FROM AUTHOR]

A new technique, called Vacuum Tight Threaded Junction (VTTJ), has been developed and patented by Consorzio RFX, permitting to obtain low-cost and reliable vacuum-compatible non-welded junctions. The technique has been tested up to an internal pressure of 500 bar and up to a temperature of 700 °C [1] . The main advantages with respect to existing technologies are an easy construction, a low cost and a high repeatability of the process. Due to these advantages, the new technique has been adopted for several in-vacuum components of the SPIDER experiment and has been also recently accepted by the ITER vacuum group for the usage in the MITICA experiment, the full prototype of the ITER Neutral Beam Injectors. This paper reports the test and qualification, according to the ITER criteria, of the VTTJ technique with geometry and materials compatible with the divertor and other components of future fusion reactors. Namely, a set of junction samples have been manufactured, joining CuCrZr to AISI 316L stainless steel and using tube-to-tube geometry. [ABSTRACT FROM AUTHOR]

The heating & current drive (H&CD) systems in a DEMOnstration fusion power plant are one of the major energy consumers. Due to its high demand in electrical energy the H&CD efficiency optimization is an important goal in the DEMO development. The H&CD power for DEMO, based on physics scenarios for the different plasma phases, is needed for plasma initiation phases (incl. breakdown), current ramp-up, heating to H-mode, burn control, controlled current ramp-down, MHD control and other functions. Plasma control will need significant installed H&CD power, though not continuously used. Previously, in the DEMO1 2015 baseline definitions, optimistic forecasted H&CD efficiencies had been assumed in the corresponding system code (i.e. PROCESS) module. Realizing that there is a high uncertainty in the assumptions the efficiencies have been modified and the impact on the DEMO power plant and basic tokamak configuration are discussed in this article. A comparison of the various H&CD systems NBI (Neutral Beam Injection), Electron Cyclotron (EC), Ion Cyclotron (IC) in terms of impact on Tritium Breeding Ratio (TBR) due to various openings for the H&CD front end components in the breeding blanket (BB) is presented. For increasing the reliability as major features the power per system unit and the redundancy are identified leading to a new proposal for clusters for EC and modular ion-sources for NB. [ABSTRACT FROM AUTHOR]

A common problem of the multi-beamlet, negative-ion based Heating Neutral Beam (HNB) Injectors presently used in fusion research, is the undesired deflection of negative ions due the permanent magnets embedded in the Extraction Grid (EG) for the suppression of the unwanted co-extracted electrons. A new solution has been recently developed at Consorzio RFX, which makes use of an additional set of permanent magnets, also embedded in the EG, with a series of advantages in terms of performances and versatility with respect to the traditional electrostatic compensation. This solution is now the new standard for ITER. After a thorough validation of the design by different numerical models, it was decided to experimentally test this solution for the first time on the Negative Ion Test Stand (NITS) at National Institute for Quantum and Radiological science and technology (QST), in Japan, within the framework of a scientific cooperation agreement between QST and Consorzio RFX. To this purpose, an EG having an ITER-like profile and compatible with NITS accelerator was designed and constructed at Consorzio RFX. This paper describes the design solutions adopted for this grid and its related permanent magnets, the construction phase, and a brief report on the first experimental results. [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]

The MITICA experiment (Megavolt ITER Injector & Concept Advancement) is the prototype and the test bed of the Heating and Current Drive Neutral Beam Injectors, which will be necessary for the full-performance exploitation of ITER. MITICA injector experiments shall demonstrate the reliable and accurate emission of a 17 MW beam of neutral particles for duration up to 1 hour, fulfilling ITER specific requirements. The accelerator grids are among the most critical parts of this experiment, because they must fulfill several operational requirements and at the same time satisfy the fatigue verifications according to the ITER Structural Design Criteria for In-vessel Components (SDC-IC). After about two years of continuous development, two design improvements were found to effectively increase the fatigue life of the grids up to the requested values. The first method was to adopt a novel shape of the cooling channels inside the grids, called Nozzle Island Cooling Enhancement (NICE) and able to provide a high performance cooling without exceeding the limits on the pressure drop through the grids. The second, called Stress Relieving Slits (SRS), was to introduce suitable slits in the grids, whose design was iteratively optimized until they were able to significantly reduce the stress/strain peaks due to thermal gradients. The NICE and SRS design solutions, here described in detail, were found to be key improvements in order to obtain a design of the grids able to satisfy all the operating requirements and all the structural verifications according to the ITER criteria. [ABSTRACT FROM AUTHOR]

A new technique, called Vacuum Tight Threaded Junction (VTTJ), has been developed and patented by Consorzio RFX, permitting to obtain low-cost and reliable non-welded junctions, able to maintain vacuum tightness also in heavy loading conditions (high temperature and high mechanical loads). The technique can be applied also if the materials to be joint are not weldable and for heterogeneous junctions (for example, between steel and copper) and has been tested up to 500 bar internal pressure and up to 700 °C, showing excellent leak tightness in vacuum conditions and high mechanical resistance. The main advantages with respect to existing technologies (for example, friction welding and electron beam welding) are an easy construction, a low cost, a precise positioning of the junction and a high repeatability of the process. Due to these advantages, the new technique has been adopted for several components of the SPIDER experiment and it is proposed for ITER, in particular for the ITER Heat and Current Drive Neutral Beam Injector and for its prototype, the MITICA experiment, to be tested at Consorzio RFX. This paper gives a detailed description of the VTTJ technique, of the samples manufactured and of the qualification tests that have been carried out so far. [ABSTRACT FROM AUTHOR]

MITICA is a prototype of the heating neutral beam (HNB) Injectors for ITER, built with the purpose of validating the injector design and optimizing its operation. Its goal is to produce a focused beam of neutral particles (H or D) with energy up to 1 MeV and power of 16 MW for 1 h. MITICA includes a Radio Frequency (RF) Plasma Source for the production of negative ions, a multi-stage electrostatic accelerator (up to 1 MV and 40 A), a neutralizer, a residual ion dump and a calorimeter. A transverse magnetic field in the Ion source and accelerator, including both a long-range component and a local component is crucial for obtaining the required Ion current and accelerator efficiency. The long-range component is produced by the current flowing through the plasma grid (PG) and related bus-bars. The PG current distribution and the uniformity of the resulting magnetic field have been optimized by detailed finite element (FEM) models. Hollow volumes in the thick copper part of the PG among beamlet groups allow a more uniform PG current distribution and a consequently uniform magnetic field in front of the grid. The paper describes in detail the PG geometry optimization procedure and the related magnetic and thermo-structural FEM analyses. [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]

Abstract: MITICA (Megavolt ITER Injector Concept Advancement) is a test facility for the development of a full-size heating and current drive neutral beam injectors for the ITER Tokamak reactor. The optimized electrostatic and magnetic configuration has been defined by means of an iterative optimization involving all the physics and the engineering aspects. The acceleration grids have been designed considering optical performances and mechanical constraints related to embedded magnets, to cooling channels, to the grid stiffness and manufacturability. A combination of “local” vertical field and horizontal “long range” field has been found to be the most effective set-up for ion extraction, beam focusing and minimization and equalization of thermo-mechanical loads and minimal number of electrons exiting the accelerator. [Copyright &y& Elsevier]

Abstract: The ITER [1] fusion device is expected to demonstrate the feasibility of magnetically confined deuterium–tritium plasma as an energy source which might one day lead to practical power plants. Injection of energetic beams of neutral atoms (up to 1MeV D0 or up to 870keV H0) will be one of the primary methods used for heating the plasma, and for driving toroidal electrical current within it, the latter being essential in producing the required magnetic confinement field configuration. The design calls for each beamline to inject up to 16.5MW of power through the duct into the tokamak, with an initial complement of two beamlines injecting parallel to the direction of the current arising from the tokamak transformer effect, and with the possibility of eventually adding a third beamline, also in the co-current direction. The general design of the beamlines has taken shape over the past 17 years [2], and is now predicated upon an RF-driven negative ion source based upon the line of sources developed by the Institute for Plasma Physics (IPP) at Garching during recent decades [3–5], and a multiple-aperture multiple-grid electrostatic accelerator derived from negative ion accelerators developed by the Japan Atomic Energy Agency (JAEA) across a similar span of time [6–8]. During the past years, the basic concept of the beam system has been further refined and developed, and assessment of suitable fabrication techniques has begun. While many design details which will be important to the installation and implementation of the ITER beams have been worked out during this time, this paper focuses upon those changes to the overall design concept which might be of general interest within the technical community. [Copyright &y& Elsevier]

Abstract: The optimal behaviour of copper as heat sink material makes it suitable for many components of nuclear fusion devices subjected to cyclic heat loads with high power densities. In particular, this material is considered for several water cooled components for the ITER Neutral Beam Injector, like the ion source and the acceleration grids. The electro-deposition technique permits to obtain complex geometric shapes (with small cooling channels and embedded magnets) and to have good mechanical properties at the same time, due to the high purity and to the very small grain size. Consequently, the thermo-mechanical properties of electro-deposited pure copper are considered as a crucial aspect for the design of the new generation fusion experimental devices. This paper presents an exhaustive test campaign aimed at investigating the properties of this material. [Copyright &y& Elsevier]

Abstract: The analytical interpretative models for calorimetric measurements currently available in the literature can consider close systems in steady-state and transient conditions, or open systems but only in steady-state conditions. The PCCE code (Predictive Code for Calorimetric Estimations), here presented, introduces some novelties. In fact, it can simulate with an analytical approach both the heated component and the cooling circuit, evaluating the heat fluxes due to conductive and convective processes both in steady-state and transient conditions. The main goal of this code is to model heating and cooling processes in actively cooled components of fusion experiments affected by high pulsed power loads, that are not easily analyzed with purely numerical approaches (like Finite Element Method or Computational Fluid Dynamics). A dedicated mathematical formulation, based on concentrated parameters, has been developed and is here described in detail. After a comparison and benchmark with the ANSYS® commercial code, the PCCE code is applied to predict the calorimetric parameters in simple scenarios of the SPIDER experiment. [Copyright &y& Elsevier]

Abstract: The SPIDER and MITICA experiments, planned to be built at Consorzio RFX for the development and optimization of the ITER Heating and Current Drive Neutral Beam Injectors, feature a number of components intercepted by high heat fluxes, that must be cooled during operations by means of suitable high performance cooling systems. As the design for such cooling systems presents some technological and heat transfer issues, a specific R&D program has been carried out, particularly referred to the accelerator grids that are among the most critical components. These components are foreseen to be manufactured by electrodeposition of pure copper onto a copper base plate to realize cooling channels and magnets slots, while the connection with the feeding manifolds is foreseen to be realized by friction welding or electron beam welding. Suitable manufacturing parameters and production methodologies have been identified by constructing and testing a first series of prototypes. A set of thermocouples have been embedded in some prototypes, to allow the evaluation of local heat transfer coefficients and the identification of local spots where dry-out phenomena might occur. To this purpose, a dedicated electrodeposition process has been developed. [Copyright &y& Elsevier]

Abstract: The ITER Neutral Beam Test Facility (PRIMA—Padova Research on Injector Megavolt Accelerated) is planned to be built at Consorzio RFX (Padova, Italy). PRIMA includes two experimental devices: a full size plasma source with low voltage extraction called SPIDER (Source for Production of Ion of Deuterium Extracted from RF plasma) and a full size neutral beam injector at full beam power called MITICA (Megavolt ITER Injector Concept Advancement). SPIDER is the first experimental device to be built and operated, aiming at testing the extraction of a negative ion beam (made of H− and in a later stage D− ions) from an ITER size ion source. The main requirements of this experiment are a H−/D− current of approximately 70A/50A and an energy of 100keV. This paper presents an overview of the SPIDER beam source design, with a particular focus on the main design choices, aiming at reaching the best compromise between physics, optics, thermo-mechanical, cooling, assembly and electrical requirements. [ABSTRACT FROM AUTHOR]

Abstract: The total power loads applied to inertially cooled components are commonly calculated starting from calorimetric measurements as cooling fluid temperature increase and flow rate during and after the power load application. In this paper, some generalizations to the standard models of measure interpretation are introduced. A method is presented to take into account the effect of thermal conduction and radiation between the component and the surrounding material. It consists in evaluating the conduction and convection thermal resistances and the radiation heat flux by Computational Fluid Dynamics (CFD) numerical models. An analytical model permits to use these evaluations to obtain, from the measured convection flux, the corresponding fluxes by conduction and radiation. Sometimes the durations of pulses and duty cycles are such that the calorimetric measurements after one pulse are actually influenced by the previously applied pulse. A correction factor is introduced to cancel this systematic error. The cooling down phase can be long compared to the duty cycles between the pulses, and consequently the signals cannot be completely recorded till the thermal equilibrium. In order to evaluate the total heat flux on a component also in this case, a model is presented which considers a proper extrapolation curve to estimate the integral of the energy exhausted by the cooling systems. As an example, the models introduced in the paper have been used to interpret experimental measurements carried out on the RADI experiment at IPP Garching and some results are presented in the paper. [Copyright &y& Elsevier]

Abstract: In the framework of the strategy for the development and the procurement of the NB systems for ITER, it has been decided to build in Padova a test facility, including two experimental devices: a full size plasma source with low voltage extraction and a full size NB injector at full beam power (1MV). These two different devices will separately address the main scientific and technological issues of the 17MW NB injector for ITER. In particular the full size plasma source of negative ions will address the ITER performance requirements in terms of current density and uniformity, limitation of the electron/ion ratio and stationary operation at full current with high reliability and constant performances for the whole operating time up to 1h. The required negative ion current density to be extracted from the plasma source ranges from 290A/m2 in D2 (D−) and 350A/m2 in H2 (H−) and these values should be obtained at the lowest admissible neutral pressure in the plasma source volume, nominally at 0.3Pa. The electron to ion ratio should be limited to less than 1 and the admissible ion inhomogeneity extracted from the grids should be better than 10% on the whole plasma cross-section having a surface exposed to the extraction grid of the order of 1m2. The main design choices will be presented in the paper as well as an overview of the design of the main components and systems. [Copyright &y& Elsevier]

Abstract: The SINGle Aperture–SINgle GAP (SINGAP) accelerator for ITER neutral beam injector foresees four grids for the extraction and acceleration of negative ions, instead of the seven grids of the Multi-Aperture Multi-Grid (MAMuG) reference configuration. The grids have to fulfil specific requirements coming from ion extraction, beam optics and thermo-mechanical issues. This paper focuses on the thermo-hydraulic and thermo-mechanical design of the grids carried out by Consorzio RFX for the design of the first ITER NB injector and the ITER NB Test Facility. The cooling circuit design (position and shape of the channels) and the cooling parameters (water coolant temperatures, pressure and velocity) were optimized with sensitivity analyses in order to satisfy the grid functional requirements (temperatures, stresses, in plane and out of plane deformations). The design required a complete modelling of the grids and their support frames by means of 3D FE and CAD models. [Copyright &y& Elsevier]

Abstract: A radio frequency (RF) driven negative ion source has been designed for the ITER neutral beam injectors, as an alternative to the traditional arc driven solution. The main advantage of this technology is to avoid the presence of the filaments, that require periodic maintenance and consequently frequent shutdowns. The requirements for the ion source of the ITER NBI are to provide a uniform flux of D−/H− to the plasma grid of the accelerator that will result in a beam current of 40A at 1MeV. The present specification is for a filling pressure of 0.3Pa. The ion source needs to provide 20/28mA/cm2 D−/H− current density across the 0.58m×1.54m aperture array for 3600s. The source, consisting of a main chamber facing the plasma grid, of eight RF drivers and the auxiliary systems for power transfer, cooling and diagnostic purposes, is housed in the same quasi-cylindrical structure that supports the arc driven solution. Specific electric and hydraulic circuits have been designed and verified. In the paper the analyses performed for the design of the components are presented in detail. [Copyright &y& Elsevier]

• RAMI has to be taken into consideration for the conceptual design of the DEMO NBI. • The RAMI scores of two design options for the DEMO NBI beam source were compared. • This has allowed for further design developments. DEMO is a first-of-a-kind DEMOnstration fusion power plant [1,2] and is intended to follow the ITER experimental reactor. The main goal of DEMO will be to demonstrate the possibility to produce electric energy for the grid from the fusion reaction early in the second half of the century. The injection of high energy neutral (1 MeV) particle beams is one of the main tools to heat the plasma up to fusion conditions, control the plasma burn phase and ramp the plasma down. Within the EUROfusion Framework a conceptual design of the Neutral Beam Injector (NBI) for the DEMO fusion reactor is currently being developed. Thereby, Reliability, Availability, Maintainability and Inspectability (RAMI) have to be taken into consideration for the conceptual design of the DEMO NBI, together with the exploitation of the currently available return of experience from the ITER NBIs. Comparing the failure risk of two different source concepts due to the considered failure modes has allowed for further design developments aiming at exploiting the advantages of the modular approach while minimizing its drawbacks. [ABSTRACT FROM AUTHOR]

Abstract: SPIDER is a full-size Negative Ion Source and a multi-beamlet 100kV accelerator, presently under construction in Padova and is part of PRIMA, the test bed for the full development of the Heating Neutral Beam system for ITER. During the final design, the SPIDER extractor and accelerator system has been improved with respect to the original configuration. In particular, the magnetic configuration has been optimized in order to improve the performances in terms of ion beam optics and aiming, and to obtain an efficient filter for the extracted electrons. [Copyright &y& Elsevier]

• 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]

• Status of MITICA facility. • SPIDER experimentation started. • First scan of plasma light and spectroscopic signals vs RF power and filter field. • SPIDER short-term planning. To reach fusion conditions and control plasma configuration in ITER, a suitable combination of additional heating and current drive systems is necessary. Among them, two Neutral Beam Injectors (NBI) will provide 33 MW hydrogen/deuterium particles electrostatically accelerated to 1 MeV; efficient gas-cell neutralisation at such beam energy requires negative ions, obtained by caesium-catalysed surface conversion of atoms inside the ion source. As ITER NBI requirements have never been simultaneously attained, a Neutral Beam Test Facility (NBTF) was set up at Consorzio RFX (Italy), including two experiments. MITICA is the full-scale NBI prototype with 1 MeV particle energy. SPIDER, with 100 keV particle energy, aims at testing and optimising the full-scale ion source: extracted beam uniformity, negative ion current density (for one hour) and beam optics (beam divergence <7 mrad; beam aiming direction within 2 mrad). This paper outlines the worldwide effort towards the ITER NBI realisation: the main results of the ELISE facility (IPP-Garching, Germany), equipped with a half-size source, are described along with the status of MITICA; specific issues are investigated by small specific facilities and by joint experiments at QST and NIFS (Japan). The SPIDER experiment, just come into operation, will profit from strong modelling activities, to simulate and interpret experimental scenarios, and from advanced diagnostic instruments, providing thorough plasma and beam characterisation. Finally, the results of the first experiments in SPIDER are presented, aimed at a preliminary source plasma characterisation by plasma light detectors and plasma spectroscopy. [ABSTRACT FROM AUTHOR]

The limits for the heat loads on the DEMO first wall are significantly stricter compared to those of ITER due to cooling and breeding blanket requirements. Fast particles in the form of NBI ions can escape the confinement due to magnetic perturbations and produce a significant power load on the first wall. In this contribution, we use the fast ion orbit-following code ASCOT to study NBI losses under the effects of 3D magnetic perturbations, including toroidal field ripple and ferritic inserts. The losses were found to not endanger the integrity of the wall, with loads remaining below 40 kW/m2 even with unmitigated ripple, and losses did not increase above 2.5 % even with 16 toroidal field coils. Ionization in the scrape-off layer was not found to increase the losses significantly for realistic density profiles, and the design of the wall allows for some flexibility in the distance between the plasma and the first wall tiles. [ABSTRACT FROM AUTHOR]

• SPIDER Beam Source manufacturing and assembly completed in October 2017. • Visual inspection, pressure/leak tests, electric and magnetic tests performed on site. • Some issues were discovered and fixed, on site. • The installation of the Beam Source inside the vessel was completed in February'18. • Final tests after installation were passed, commissioning started in May'18. The SPIDER Beam Source (BS), the first prototype of a full scale ion source for the ITER Heating Neutral Beam injector, was delivered to the Neutral Beam Test Facility (NBTF) site in Padova (Italy) after about five years procurement phase. A huge effort was devoted during the procurement for quality controls and testing at the supplier's workshops. Several activities were also carried out on NBTF site for verification/adjustment of interfaces, solution of still open issues, as well as final tests before and after installation inside the vacuum vessel. The NBTF Team undertook the BS site acceptance tests including: pressure and leak tests of the hydraulic circuits; electrical tests; measurement of magnetic field profiles; functionality tests of diagnostics installed on the BS; checks of grids alignment by means of laser tracker. Accurate positioning of the BS inside the vacuum vessel was performed and various service lines were connected in a tight space. Several improvements were undertaken in order to guarantee reliability and reduce the risks during the commissioning and experimental campaign in 2018. After installation, the integrated commissioning phase was initiated, powering the RF and high voltage circuits, followed by the first operation in vacuum. [ABSTRACT FROM AUTHOR]

• Joint experimental campaigns carried out in QST lab in Naka, Ibaraki, Japan. • Full compensation of magnetic criss-cross deflection of beamlets obtained. • Non-uniform extraction assumed for interpreting the experimental results. • Numerical simulations can reproduce the experimental results with good accuracy. In the framework of the collaboration between Consorzio RFX (Padova, Italy) and QST (Naka, Japan) two experimental campaigns have been organized on the Negative Ion Test Stand (NITS) in Naka, employing an ITER-like multi-beamlet configuration. These campaigns were aimed at compensating for the undesired magnetic deflection of the ion beam. The experiments have clarified the influence of operating parameters on the optical properties of the ion beam. The physical phenomena acting on the optics have been investigated, in particular the effect of an asymmetry of the ion current density at meniscus. Comparisons with simulations have been made and the modeling has been improved consequently. During the second experimental campaign, a perfect compensation of the ion deflection has been achieved. [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]

Substantial progresses have been achieved in the realization of the ITER Neutral Beam Test Facility (NBTF) hosted in Padova, Italy; the buildings, completed by the end of 2015, are being progressively filled with new systems and components. The realization of SPIDER, the ITER full-size negative ion source, is well advanced and important progress is also recorded for MITICA, the full-scale prototype of the ITER HNB injector. The paper gives an overview of the achieved results, highlighting the main challenges faced. [ABSTRACT FROM AUTHOR]

Abstract: The TPE-RX Neutral Beam Injector, which provides a 25keV positive ion beam energy with a maximum current of 50A for a pulse duration of 30ms, will be installed on RFX-mod thanks to the agreement with the AIST Institute of Tsukuba (Japan). The main scientific objective is the study of the behavior of the fast ions, which in the RFP helical equilibrium have exhibited very long confinement times. The integration of TPE-RX NBI on RFX-mod requires the development of several new components: a mechanical interface between the RFX-mod vacuum vessel and neutralizer; a Magnetic Residual Ion Dump; a new vacuum pumping system designed to maximize pumping and minimize beam stopping due to reionization. As regards the power supplies the compliance of the Japanese equipment to the Italian safety rules has been considered and layout studies have been carried out; the integration of the NBI control system in the RFX timing sequence has been studied as well. [Copyright &y& Elsevier]

Abstract: In 2007 the radio frequency driven negative hydrogen ion source developed at IPP in Garching was chosen by the ITER board as the new reference source for the ITER neutral beam system. In order to support the design and the commissioning and operating phases of the ITER test facilities ISTF and NBTF in Padua, IPP is presently constructing a new test facility ELISE (Extraction from a Large Ion Source Experiment). ELISE will be operated with the so-called “half-size ITER source” which is an intermediate step between the present small IPP RF sources (1/8 ITER size) and the full size ITER source. The source will have approximately the width but only half the height of the ITER source. The modular concept with 4 drivers will allow an easy extrapolation to the full ITER size with 8 drivers. Pulsed beam extraction and acceleration up to 60kV (corresponding to pre-acceleration voltage of SINGAP) is foreseen. The aim of the design of the ELISE source and extraction system was to be as close as possible to the ITER design; it has however some modifications allowing a better diagnostic access as well as more flexibility for exploring open questions. Therefore one major difference compared to the source of ITER, NBTF or ISTF is the possible operation in air. Specific requirements for RF sources as found on IPP test facilities BATMAN and MANITU are implemented [A. Stäbler, et al., Development of a RF-driven ion source for the ITER NBI system, SOFT Conference 2008, Fusion Engineering and Design, 84 (2009) 265–268]. [Copyright &y& Elsevier]

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]

• First complete performance analysis of neutron diagnostics in DTT for half and full power scenarios. • Neutron flux monitors assessment for non-magnetic plasma position measurements. • Core ion temperature and thermal/non thermal fuel ion population assessment based on a time of flight neutron spectrometer. • Role of triton burn-up estimated to be higher than present day devices. A conceptual design and performance studies for a collimated neutron flux monitor and neutron spectrometer for the Divertor Tokamak Test (DTT) facility are presented. This study is based on the single-null divertor configuration and for "Half Power" and "Full power" scenarios with 15 MW of negative-ion NBI, 29 MW of ECH and 3 MW of ICRF heating with a maximum neutron yield of 1.5 × 10 17 s − 1. Fast ion distributions (both from auxiliary heating systems and fusion born) have been simulated in TRANSP/NUBEAM and the corresponding neutron energy spectra have been calculated using DRESS. Synthetic diagnostics have been implemented to determine the neutron fluxes and spectra at the detector location. Neutron emissivity profiles, plasma position, core ion temperature and the ratio of thermal and non-thermal D ion populations can be obtained with good accuracy and time resolution. [ABSTRACT FROM AUTHOR]