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Neutral beam injection (NBI) is one of the auxiliary power systems considered for the EU DEMO pulsed plasma ('DEMO1'). In this paper, we discuss the characteristics of optimized NBI in terms of the DEMO1 requirements, relating physics and engineering contexts in a novel parameter range compared to current NBI systems and in a larger plasma volume than ITER. Different injection options are investigated to account for various concepts discussed in the literature. The investigation is carried out using a wide range of sensitivity studies by means of the METIS 0.5D transport code and ASCOT Monte Carlo simulations of injected neutral-beam particles. This investigation has generated a series of recommendations for the NBI design, contributing to the system optimization process. We show how tangential injection aimed at the plasma core, with an energy 800 keV, is recommended to maintain high fusion power performance and fulfill the requirements for bulk heating. Compliance with engineering constraints on the NBI design, e.g., the least interference with breeding blanket modules or compatibility with solutions for the beamline components, imposes some restrictions when discussing the beam geometry. The minimum density at which an NBI can be safely operated without harmful shine-through losses is investigated for different injection energies and compared to the ITER case. The NBI operational window for DEMO is shown to be significantly extended to transient, low-density phases, also highlighting the importance of NBI systems designed for modular energy and power output. NBI can therefore sustain the plasma during a considerable portion of the transient phases, i.e., the current ramp-up and ramp-down phases. The final decision on the DEMO heating mix will be made on the basis of system operability and performance: the optimized NBI is shown to be a suitable and effective option for the EU DEMO inductive scenario. [ABSTRACT FROM AUTHOR]

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]

Heating and current drive (H&CD) systems are being investigated for a demonstration fusion power plant DEMO to deliver net electricity for the grid around 2050. Compared to ITER, which has to show the generation of 500-MW thermal power, the target of DEMO is the successful production of 300 to 500 MW electrical power to the grid and to aim for a self-sufficient tritium fuel cycle. Three H&CD systems are under development for DEMO in Europe, the electron cyclotron (EC) system, the neutral beam injection (NBI) system, and the ion cyclotron system. Based on present studies for plasma ramp-up, ramp-down, and flat top phases, to be further validated in more detailed simulations, the assumed total launched power needed from the H&CD system in DEMO is in the range of 50–100 MW, to be provided for plasma heating and control. This paper describes the design and Research and Development status of selected H&CD systems, considered for their deployment in the EU DEMO. It was always considered that different H&CD configurations and design variants will have an impact on the performances for the whole fusion plant. It shall be noted that the basis for the H&CD integrated design and system development is the actual version of the European fusion electricity roadmap. The project also elaborates on H&CD efficiency improvements which will reduce the recirculating power fraction in the future fusion power plants. Different studies under investigation will be discussed such as for NBI the photoneutralization and for EC novel concepts for gyrotron multistage-depressed collector. [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]

A new magnetic configuration is proposed for the suppression of co-extracted electrons in a negative-ion accelerator. This configuration is produced by an arrangement of permanent magnets embedded in one accelerator grid and creates an asymmetric local magnetic field on the upstream and downstream sides of this grid. Thanks to the "concentration" of the magnetic field on the upstream side of the grid, the resulting deflection of the ions due to magnetic field can be "intrinsically" cancelled by calibrating the configuration of permanent magnets. At the same time, the suppression of co-extracted electrons can be improved. [ABSTRACT FROM AUTHOR]

Two Neutral Beam Injectors (NBI) are foreseen to provide a substantial fraction of the heating power necessary to ignite thermonuclear fusion reactions in ITER. The development of the NBI system at unprecedented parameters (40 A of negative ion current accelerated up to 1 MV) requires the real-ization of a full scale prototype, to be tested and optimized at the Test Facility under construction in Padova (Italy). The beam source is the key component of the system and the design of the multi-grid accelerator is the goal of a multi-national collaborative effort. In particular, beam steering is a challenging aspect, being a tradeoff between requirements of the optics and real grids with finite thickness and thermo-mechanical constraints due to the cooling needs and the presence of permanent magnets. In the paper, a review of the accelerator physics and an overview of the whole R&D physics program aimed to the development of the injector source are presented. [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]

The ion sources play an important role in the framework of the nuclear fusion, as a fundamental part of the neutral beam injection systems needed for the auxiliary heating of the plasma magnetically confined in fusion devices. In the case of large experiments, such as ITER and beyond, the use of negative ions was found necessary to increase the neutralization efficiency of high-energy beams, calling for an additional effort in the research on the physical mechanisms regulating their formation and extraction as a beam from the source. This paper summarizes the analyses investigating the physical phenomena occurring in the source and their effect on the optics of the extracted beam. The results, carried out by means of advanced numerical tools, are particularly useful to yield a better evaluation of the real performances of the system, and particularly of the accelerator, whose modeling usually neglect some characteristics of the source. These analyses were performed for the specific case of the SPIDER experiment (Source for Production of Ion of Deuterium Extracted from an RF plasma), representing the major test bed for negative ion source worldwide. The choice of a specific device for the source modeling was found useful to support its operational phase and to help the experimental data interpretation and the identification of dangerous operating conditions. Nonetheless, the validity of the results here presented still holds in more general cases. [ABSTRACT FROM PUBLISHER]

The ion sources play an important role in the framework of the nuclear fusion, as a fundamental part of the neutral beam injection systems needed for the auxiliary heating of the plasma magnetically confined in fusion devices. In the case of large experiments, such as ITER and beyond, the use of negative ions was found necessary to increase the neutralization efficiency of high-energy beams, calling for an additional effort in the research on the physical mechanisms regulating their formation and extraction as a beam from the source. This paper summarizes the analyses investigating the physical phenomena occurring in the source and their effect on the optics of the extracted beam. The results, carried out by means of advanced numerical tools, are particularly useful to yield a better evaluation of the real performances of the system, and particularly of the accelerator, whose modeling usually neglect some characteristics of the source. These analyses were performed for the specific case of the SPIDER experiment (Source for Production of Ion of Deuterium Extracted from an RF plasma), representing the major test bed for negative ion source worldwide. The choice of a specific device for the source modeling was found useful to support its operational phase and to help the experimental data interpretation and the identification of dangerous operating conditions. Nonetheless, the validity of the results here presented still holds in more general cases. [ABSTRACT FROM PUBLISHER]

The megawatt-level TPE-RX neutral beam injector is planned to be installed on the RFX-mod reversed field pinch (RFP) experiment. To properly integrate the TPE-RX injector on RFX-mod, a 3-D Monte Carlo simulation of the injector has been carried out. The code simulates the charged and neutral particle trajectories from the acceleration grids of the beam source to the RFP plasma. The simulation includes the neutralization stage, the interaction between the accelerated particles and the magnetic field produced by RFX and the residual ion dump (RID), and the interactions between the accelerated beam and the background gas. The simulation has allowed to estimate the neutralizer and RID efficiency and specifications, the geometry of the beam duct between the injector and RFX-mod, and the necessary degree of magnetic screening. [ABSTRACT FROM AUTHOR]

An electron dump (ED) has been developed and added to the SPIDER negative ion beam source in order to dispose of the electrons exiting from the accelerator, which are mainly generated by stripping reactions. This component prevents from having unwanted and possibly concentrated heat load from electrons on the downstream components and permits to obtain on the diagnostic calorimeter a detailed measurement of the negative ion beam. The detail design of the ED is here described, together with the main design choices that have driven the design process. The physics and thermomechanical analyses, carried out to compare the different design proposals, are also presented. [ABSTRACT FROM PUBLISHER]

The megavolt ITER injector concept advancement neutral injector test facility will be constituted by a RF-driven negative ion source and by an electrostatic Accelerator, designed to produce a negative Ion with a specific energy up to 1 MeV. The beam is then neutralized in order to obtain a focused 17 MW neutral beam. The magnetic configuration inside the accelerator is of crucial importance for the achievement of a good beam efficiency, with the early deflection of the co-extracted and stripped electrons, and also of the required beam optic quality, with the correction of undesired ion beamlet deflections. Several alternative magnetic design concepts have been considered, comparing in detail the magnetic and beam optics simulation results, evidencing the advantages and drawbacks of each solution both from the physics and engineering point of view. [ABSTRACT FROM AUTHOR]

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]

The SPIDER experiment's main goal is to test the extraction of negative ions from an ITER-size ion source. It is designed to extract 1 280 negative ion beamlets and accelerate them up to a 100 kV potential. The negative ion beam at exit and the operating parameters will be carefully measured and optimized in order to match the ITER requirements for the Neutral Beam Injector (NBI) ion sources. Inside a negative ion accelerator, there are generally two main factors that can cause the deflection of the ion beamlets: the repulsion among beamlets and the electron suppression magnetic field. These two effects are both to be considered highly detrimental for the ITER NBI since they are expected to cause higher heat loads on the ITER NBI neutralizer and decrease the overall beam quality (in terms of aiming and divergence). Hence, they should also be considered and minimized for the SPIDER device, where it will be possible to precisely investigate the beamlet footprint using an instrumented calorimeter relatively close to the accelerator exit. This paper presents a design optimization process aiming at compensating the two described effects. To make this, a mechanical offset of the grounded grid apertures is considered. The OPERA-3-D code (Vector Fields Co. Ltd.) is used as the main tool for this optimization process because it can take into account beamlet repulsion and the interaction among the beamlets and grids. This is made by solving the electrostatic Poisson's equation with a finite element approach to calculate the particle trajectories of the negative ions under the influence of electrostatic fields, magnetic fields, and space charge. [ABSTRACT FROM AUTHOR]

In view of the realization of the negative ion beam injectors for ITER, a test facility, named SPIDER, is under construction in Padova (Italy) to study and optimize production and extraction of negative ions. The present paper is devoted to the analysis of the expected first operations of SPIDER in terms of single-beamlet and multiple-beamlet simulations of the hydrogen beam optics in various operational conditions. The effectiveness of the methods adopted to compensate for the magnetic deflection of the particles is also assessed. Indications for a sequence of the experimental activities are obtained. [ABSTRACT FROM AUTHOR]

The Radio Frequency (RF) negative hydrogen ion source prototype has been chosen for the ITER neutral beam injectors due to its optimal performances and easier maintenance demonstrated at Max- Planck-Institut fur Plasmaphysik, Garching in hydrogen and deuterium. One of the key information to better understand the operating behavior of the RF ion sources is the extracted negative ion current density distribution. This distribution-influenced by several factors like source geometry, particle drifts inside the source, cesium distribution, and layout of cesium ovens-is not straightforward to be evaluated. The main outcome of the present contribution is the development of a minimization method to estimate the extracted current distribution using the footprint of the beam recorded with mini-STRIKE (Short-Time Retractable Instrumented Kalorimeter). To accomplish this, a series of four computational models have been set up, where the output of a model is the input of the following one. These models compute the optics of the ion beam, evaluate the distribution of the heat deposited on the mini-STRIKE diagnostic calorimeter, and finally give an estimate of the temperature distribution on the back of mini-STRIKE. Several iterations with different extracted current profiles are necessary to give an estimate of the profile most compatible with the experimental data. A first test of the application of the method to the BAvarian Test Machine for Negative ions beam is given. [ABSTRACT FROM AUTHOR]

The megavolt ITER injector and concept advancement experiment is the prototype and the test bed of the ITER heating and current drive neutral beam injectors, currently in the final design phase, in view of the installation in Padova Research on Injector Megavolt Accelerated facility in Padova, Italy. The beam source is the key component of the system, as its goal is the generation of the 1 MeV accelerated beam of deuterium or hydrogen negative ions. This paper presents the highlights of the latest developments for the finalization of the MITICA beam source design, together with a description of the most recent analyses and R&D activities carried out in support of the design. [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]

In negative ion electrostatic accelerators a considerable fraction of extracted ions is lost by collision processes causing efficiency loss and heat deposition over the components. Stripping is proportional to the local density of gas, which is steadily injected in the plasma source; its pumping from the extraction and acceleration stages is a key functionality for the prototype of the ITER Neutral Beam Injector, and it can be simulated with the 3D code AVOCADO. Different geometric solutions were tested aiming at the reduction of the gas density. The parameter space considered is limited by constraints given by optics, aiming, voltage holding, beam uniformity, and mechanical feasibility. The guidelines of the optimization process are presented together with the proposed solutions and the results of numerical simulations. [ABSTRACT FROM AUTHOR]

• 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]