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The development of High Voltage Direct Current Gas-Insulated components that will be employed for the power supply of Neutral Beam Injectors, requires specific numerical tools to analyze the electric field distribution, with the purpose of supporting the engineering and operational phases.The tools must comply with the nonlinearity of the material properties of solid insulators and the transport mechanisms in the dielectric gas. Despite its high Global Warming Potential, SF6 will be considered as the baseline dielectric gas for the development of HVDC gas-insulated components to be used for the Acceleration Grid Power Supply of Neutral Beam Injectors. Attention especially to the mechanisms driving to an increase of leakage current flowing in the gas, thus giving rise to possible discharge phenomena is here placed, mainly focusing on radiation and injection regimes. The described methodologies are adopted for the numerical modeling of the 500 kV DC gas-insulated components that will be used for the power supply of the Neutral Beam Injector of the Divertor Tokamak Test facility. In particular, a conceptual design of the SF6 Transmission Line and the SF6-air bushing is conceived.

Stable and uniform beams with low divergence are required in particle accelerators; therefore, beyond the accelerated current, measuring the beam current spatial uniformity and stability over time is necessary to assess the beam performance, since these parameters affect the perveance and thus the beam optics. For high-power beams operating with long pulses, it is convenient to directly measure these current parameters with a non-intercepting system due to the heat management requirement. Such a system needs to be capable of operating in a vacuum in the presence of strong electromagnetic fields and overvoltages, due to electrical breakdowns in the accelerator. Finally, the measure of the beam current needs to be efficiently integrated into a pulse file with the other relevant plant parameters to allow the data analyses required for beam optimization. This paper describes the development, design and commissioning of such a non-intercepting system, the so-called beamlet current monitor (BCM), aimed to directly measure the electric current of a particle beam. In particular, the layout of the system was adapted to the SPIDER experiment, the ion source (IS) prototype of the heating neutral beam injectors (HNB) for the ITER fusion reactor. The diagnostic is suitable to provide the electric current of five beamlets from DC up to 10 MHz. [ABSTRACT FROM AUTHOR]

This paper outlines the main features of a new high efficiency ( η > 62%) high power (∼18 MW D ^0 ) neutral beam (NB) concept based on pragmatic solutions suitable with the reactor requirements. The injector is modular (several beamlines in parallel) with independent ion sources referenced to the ground potential and gas neutralizers held at +1 MV. This topology leads to numerous simplifications; it overcomes the main issues of conventional NB systems, such as the complex 1 MV electrical setup, the difficult ion source remote maintenance, the high caesium consumption. The other key parameter is the gas neutralization concept which minimizes the amount of gas by operating at a low gas target and low neutralizer duct conductance. The implementation of an energy recovery system for the residual 1 MeV D ^− is essential to attain a high wall-plug efficiency. These specific features require thin laminar D ^− beams provided by a pre-acceleration up to 100 keV in slotted grid apertures to form thin blade-like beamlets, followed by the post-acceleration to 1 MeV by merging the beamlets in a single beam in five gaps (+200 kV per gap). All these specific aspects minimize the beams losses and thermal loads along the beamline and enhance the injector reliability and availability.

In the ITER neutral beam injectors (NBI), the presence of an external variable magnetic field generated by the ITER tokamak itself, could deflect the ion beam during acceleration and cause a loss of beam focusing. For this reason, the ion source, the accelerator and the neutralizer will be shielded from external magnetic field by means of a passive magnetic shield and a system of active correction and compensation coils (ACCC). The ACCC will operate in a feedback control loop and thus require the measurement of magnetic field inside the NBI vessel. Magnetic sensors for this application must be capable of measuring DC and slow variable magnetic fields, and be vacuum-compatible, radiation-hard and robust, since they will be subjected to neutron flux produced by fusion reactions in the tokamak and inaccessible for maintenance. This paper describes the realization and tests of fluxgate magnetic sensors prototypes specifically designed for this purpose before the installation in MITICA and ITER. [ABSTRACT FROM AUTHOR]

BackgroundNegative ion source based on neutral beam injector is an important auxiliary heating and current driving system for the China Fusion Experimental Test Reactor (CFETR). Due to the high beam energy and large beam spot of negative ion based neutral beam injection system (NNBI), electrical deflection of residual ions has become the preferred scheme for residual ion dumping.PurposeThis study aims at the conceptual design of NNBI electrostatic residual ion dump system to deflect residual ions for neutralization of the beam.MethodsAccording to the development objectives of NNBI, three polar plates and two channels were adopted for the conceptual design of electrostatic residual ion dump (ERID). A negative voltage was applied to the intermediate plate to form a deflection electric field between the channels. When the mixed beam passed through the deflection channel, the residual ions were deflected to the panel due to the action of electric field force. Through theoretical calculation and beam deposition simulation, the structure of the ERID and more uniform heat load distribution on its surface were preliminarily determined.Results & ConclusionsThe simulation results show that the ERID can deflect the residual ions under the parameter design, providing a reference for future engineering design.

A set of novel design solutions for high performance cooling systems have been developed and tested by Consorzio RFX, achieving, with experimental tests, the challenging heat transfer conditions foreseen for Heating and Current Drive Systems of present and future nuclear fusion devices. The project, called Multi-design Innovative Cooling Research & Optimization (MICRO), has the triple objective of: • Verifying the full qualification of the manufacturing process and assess the cooling performance of the present solutions applied on the acceleration grids designed for ITER Heating Neutral Beam Injectors and MITICA experiment, which have similar heat loads as for DEMO1. • Identifying alternative designs realizable with presently qualified manufacturing processes for improving heat transfer mechanisms with limited pressure drops. • Developing further optimized solutions by means of the investigation of the interrelated effects of geometric parameters and thermofluid dynamic properties of the coolant fluid. The main aim of present work is to extend the fatigue life-cycle of high thermal stress components. In the first part of the paper the design description and constraints are presented together with the results obtained from numerical analyses carried out on the full-scaled model of the electrostatic grids; the outcomes of the experimental campaign, together with the characterization of the properties and performance of a set of alternative dielectric fluids, are shown in the second one. [ABSTRACT FROM AUTHOR]

The Acceleration Grid Power Supply Conversion System (AGPS-CS) of the MITICA Neutral Beam Injector has been installed between late 2017 - early 2018 and commissioned in the first half of 2018. This paper is focused on the installation and commissioning activities, with emphasis on the integration issues. The results of the site and acceptance tests are presented, proving the capability of the system to deliver the specified performance. [ABSTRACT FROM AUTHOR]

The use of non-evaporable getter (NEG) pumps is common in many UHV applications, and the properties of the new sintered getters based on the ZAO®alloy make them appealing for applications dealing with large fluxes of hydrogen and its isotopes, as typically occurs in fusion research. In particular, the use of NEG pumps is interesting for the vacuum system of negative ion-based neutral beam injectors, which require huge deuterium gas throughputs for long durations. The key advantages of a NEG pumping solution are the high specific pumping speed and capacity, the ability to withstand a large number of absorption/desorption cycles, the ease of integration. In addition, NEG do not release hydrogen unless power is supplied to heat them, addressing an important safety issue related to uncontrolled release in case of power outage or subsystem failures. This paper reports the experimental characterization of ZAO®sintered getters, in pressure regimes and sorption amounts relevant for the use in the next generation NBI. Single discs and complete pumps were characterized in different conditions of H 2 /D 2 pressure between 10−4 and 10-1Pa and operating temperatures up to 150°C. In particular, the feasibility of getter regeneration within time windows compatible with current NBI operation scenarios was demonstrated. [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]

• Spectroscopy is main diagnostic of ITER prototype negative ion beam source SPIDER. • Photodiodes measure visible radiation from source faster than spectrometers. • Custom electronics has low-noise variable gain amplifier, FPGA-based remote control. • Photodiode has14 bit dynamic range, 1.2–135 kHz bandwidth, 100 kΩ to 5 GΩ gain. • First measurements show dependencies with RF power and filter field current. The source prototype SPIDER in the Neutral Beam Test Facility at Consorzio RFX has been developed to demonstrate the target performances of the RF plasma source of ITER Heating Neutral Beam injectors. SPIDER diagnostics characterize the plasma in the source, optimize performances and provide signals for machine protection. Photodiodes measure the time evolution of lines of sight integrated visible radiation emitted by the plasma in the source. Some are equipped with interference filters to select specific spectral lines, e.g. those of cesium to control its evaporation, which assists the H¯ production, and of metal impurities, which might indicate wall damages. Other lines of sight are fed to grating spectrometers to measure the spectrum at a slower rate, either at low resolution for spectral survey of hydrogen Balmer lines and impurities or at high resolution for molecular Fulcher bands. The overall layout of the photodiode diagnostic and preliminary results during the first SPIDER operation are presented. The custom photodiode electronics is described in detail, including a low-noise variable gain amplifier, FPGA-based remote control and suitably shaped digital output for feedback control. Main characteristics are: linear response up to 10 V, 14 bit dynamic range and 1.2–135 kHz bandwidth depending on gain in range 100 kΩ to 5 GΩ. Measured light through the drivers has allowed to recognise plasma features and dependencies with RF power and filter field current, as well as dominant frequencies in signal fluctuations. [ABSTRACT FROM AUTHOR]

The PRIMA project aims at the construction of two ITER-NBI facilities in Padova (Italy). The first one is called SPIDER which is negative H/D 100 keV RF source, while the second one (MITICA) will be a full scale 1 MeV deuterium beam injector as the one that will be used in ITER. In order to resolve the horizontal beam intensity profile in MITICA and one of the eight beamlets groups in SPIDER, the Close-contact Neutron Emission Surface Mapping (CNESM) system is being developed. The goal of this device is to reconstruct the D− beam evaluating the map of the neutron emission due to interaction of the deuterium beam with the deuterons implanted in the beam dump surface. For this reason, the CNESM diagnostic, which is based on nGEM detectors for fast neutrons, will be placed right behind the SPIDER and MITICA beam dump, i.e. in an UHV (Ultra High Vacuum) environment. Since the nGEM detectors need to operate at atmospheric pressure a vacuum sealed detector box has been designed to be installed inside the vacuum vessel and able to sustain atmospheric pressure inside. This paper describes the status of the CNESM diagnostic and underlines the different phases followed during the realization and installation of the diagnostic on the SPIDER beam dump as well as its imaging performances. [ABSTRACT FROM AUTHOR]

• Neutral Beam Injectors make use of large vacuum pumps to pump hydrogenic species. • Non-Evaporable Getters materials remove hydrogen in passive way. • Strategy to choose the design of the mock-up of a large getter pump. Non-evaporable getter (NEG) technology is a potential candidate as pumping system for future neutral beam injector applications. The strategy to validate the use of getter pump technology is based on the realization of a relatively large pump mock-up that has to be tested in fusion-relevant conditions. The objectives of this mock-up are to demonstrate that a pump of large dimensions and capacity is usable. This paper deals with the design of the mock-up, based on conceptual studies which involve at first 3D gas flow simulations considering different modular mock-up pumps based on NEG sintered disks. In addition transient thermal simulations with FE method have been performed with the aim to analyze the thermal response of the mock-up. The conceptual design has been carried out in order to define the best configuration to obtain high pumping speed with low spatial gradient of gas concentration inside the getter material. The suggested solution will exhibit a modular structure of getter disks, which on one hand simplifies the mechanical assembling, and on the other hand allows interpretative modelling at different scales. It is foreseen to test the pump mockup in the TIMO facility at KIT Karlsruhe. [ABSTRACT FROM AUTHOR]

• Influence of magnetic field in RF plasma source of CRAFT was investigated. • Checkboard and symmetric arrangement of magnets were simulated and tested. • Symmetric arrangement with three rows of magnets gave the best performance in term of plasma parameters. • The results are helpful to optimize the performance of high power ion sources. Radio Frequency (RF) driven negative ion source is the preferred ion source for negative ion based Neutral Beam Injection system (NNBI). The plasma generator, which consists of RF driver(s) and an expansion chamber, is one of the key parts for NNBI. The distribution of cusp magnetic field which is generated by permanent magnets installed at the lateral wall of the expansion chamber is a common issue, due to the great influence on electron density, electron temperature and spatial uniformity of plasma density in the extraction area. Two kinds of configuration of cusp magnetic field (checkerboard type and symmetric type) and three types of arrangement of permanent magnets (with three, four and five rows of magnets) have been carried out with the approach of finite element and the experimental performance. On the RF negative ion source testbed, a moveable Langmuir probe has been adopted to investigate the characteristic of electron density and temperature near the extraction area. Since the symmetric type with three rows of permanent magnets generates the distribution of cusp magnetic field that contribute to extract the plasma from RF driver and improve the probability of particle collision easily, the higher electron density and the lower electron temperature was obtained via simulation. The experimental conclusion is that the symmetric type with three rows of permanent magnets leads to a higher electron density and lower electron temperature, showing a great agreement with simulation results. The results are helpful to optimize the structure and performance of high power ion source, and provide support for the development of large area RF negative ion sources with multi-driver. [ABSTRACT FROM AUTHOR]

• A -800 kV transmission line for neutral beam injector was design in this study, mainly about the insulating gap. • The ITER-like transmission line structure and coaxial structure both has been considered. • The simulation results shows that the coaxial has more advantages in terms of volume and electric field uniformity. • The design basis in accordance with specifications has been presented in the manuscript. The high-voltage transmission line carries electricity, water, and gas for the high-potential components in the NNBI system. High voltage transmission lines used in -800 kV acceleration grade NNBI systems require the transmission of multiple high voltages of different voltage levels. Its insulation design not only guarantees the reliability of insulation but also requires a small occupation volume. The FEA method is used for insulation design of multi-conductor transmission lines. In the design of -800 kV transmission line, coaxial and non-coaxial structures are mainly considered. In the case of SF6 filling space, conductor radius and conductor position are the main factors affecting insulation strength. Both structures can make its insulation strength meet the requirements. However, it is found that the coaxial transmission line is more compact and concise when it meets the insulation requirements, but it is difficult to install and maintain. [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]

Toward the 1 MV power supply system for the ITER neutral beam injector consisting of various components with oil insulation, gas insulation and vacuum insulation, design and manufacturing of dc ultra-high-voltage component have been progressed. Thermal design of long transmission line up to 100 m was established. In order to design the high-temperature water channel for the plasma grid at 1 MV potential, the resistivity of high-purity water at 180 °C was measured for the first time. With the obtained resistivity, realistic high-temperature water circulating system was designed. A high-voltage bushing to transmit multiple voltages simultaneously has been also developed through a careful experimental study on vacuum insulation characteristics with multi-gaps coaxial electrodes. With the empirical scaling the electrostatic screen was designed taking into account effect of multi-gap and expected deformation such as thermal elongation during operation. After the assembly, stable voltage holding of 240 kV at each stage was confirmed in the manufactured HV bushing. [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]

ITER Neutral Beam Injectors (NBIs) need to be shielded from the relatively strong stray magnetic field generated by the Poloidal Field Coils of the Tokamak. For this reason both the Heating Neutral Beams (HNB) and the Diagnostic Neutral Beam (DNB) will be provided with a Passive Magnetic Shield and with a system of Active Correction and Compensation Coils (ACCC). The ACCC will operate in feedback control and thus require the measurement of magnetic field inside the NBI vessel, i.e. in an environment subjected to the neutron flux coming from the Tokamak. To this purpose, magnetic sensors which are robust, radiation hard, drift-immune and remote-handling compatible are required. Flux-gate magnetic sensors are a good candidate for this task, but commercial sensors of this kind have typically a limited measured range (below 0.1 mT). The feasibility of a flux-gate sensor for the ITER NBI has been studied by developing a numerical model which includes magnetic core hysteresis, and which demonstrated that, by suitable choice of the core magnetic properties and geometry, it is possible to increase the measurement range by at least 2 orders of magnitude. On this basis, a flux-gate sensor prototype has been realized at Consorzio RFX. Experimental tests carried out so far have demonstrated that the measurement range can be increased to ∼10 mT with acceptable accuracy and frequency bandwidth. [ABSTRACT FROM AUTHOR]

A long pulse acceleration of high-power-density hydrogen negative ion beams of 184 MW/m 2 (0.97 MeV, 190 A/m 2 ) has been successfully achieved up to a facility limit of 60 s by using a 5 stage multi-aperture multi-grid accelerator, which is the first demonstration of the ITER relevant beam with a pulse length over the time constant of cooling capability of the water-cooled acceleration grids. In order to extend the pulse length of such high power density beams, the extraction and acceleration geometries were tuned to match the beam optics to high power density by increasing the transmission at the extractor and the acceleration electric field at the first acceleration gap. After the modifications, the available current density at the same acceleration voltage was increased by 36%, and grid heat loads at each acceleration grid were reduced to less than 3% of the electric input power, which satisfied an allowable level for long pulse acceleration. Furthermore, since the conditioning procedure has been established for long pulse acceleration, a rapid conditioning of total 16 operational days up to 1MeV, 60 s class has been achieved, which was about half of previous results. Because JT-60SA and ITER NBIs are designed by the same concept, these results are applicable to the modifications and operations toward these NBIs directly. [ABSTRACT FROM AUTHOR]

The Acceleration Grid Power Supply (AGPS) is a system devoted to supply the acceleration grids of the MITICA experiment, the full scale prototype of the ITER Neutral Beam Injector (NBI), being built in Padova (Italy) to test the NBI in advance of the operation in ITER. The procurement of the AGPS is split in two parts: the low voltage conversion system, namely the AGPS-Conversion System (AGPS-CS,) being procured by the European domestic agency, and the high voltage DC Generators (AGPS-DCG), being procured by the Japanese domestic agency. After the contract award, in late 2015, the executive design phase of the AGPS-CS has been finalized. The paper will present the final design of the AGPS-CS and will discuss the main choices in particular concerning the protection from internal faults of the inverters. [ABSTRACT FROM AUTHOR]

Several novel design solutions for high performance cooling systems have been developed by Consorzio RFX, permitting to simulate the challenging heat transfer conditions foreseen in the future fusion devices. The project, called Multi-design Innovative Cooling Research & Optimization (MICRO), aims, on one hand, to verify the present solution applied inside the NBI accelerator grids (i.e. MITICA experiment) and, on the other, to perform further improvements in the heat transfer process with an acceptable pressure drop and reliable manufacturing process. The main advantages rely on the possibility to extend the fatigue life-cycle of different high thermal stress components and to investigate the possibility to employ alternative dielectric fluids instead of water. Such design solutions would in fact allow the exploitation of less performing fluids in terms of cooling capability. This is particularly relevant in view of DEMO and future power plants characterized by higher efficiency and reliability. The paper gives a description of the Computation Fluid Dynamics (CFD) analysis and the samples manufacturing process. [ABSTRACT FROM AUTHOR]

The Heating Neutral Beam Injectors (HNBs) for ITER will have to deliver 16.7 MW beams of H/D particles at 1 MeV energy. The beams will be produced from H −/ D − ions, generated by a radiofrequency plasma source coupled to an ion acceleration system. A prototype of the ITER HNB ion source is being tested in the SPIDER experiment, part of the ITER Neutral Beam Test Facility at Consorzio RFX. Reaching the design targets for beam current density and fraction of coextracted electrons is only possible by evaporating cesium in the source, in particular on the plasma facing grid (PG) of the acceleration system. In this way the work function of the surfaces decreases, significantly increasing the amount of surface reactions that convert neutrals and positive ions into H −/ D −. It is then of paramount importance to monitor the density of negative ions and the density of Cs in the proximity of the PG. Monitoring the Cs spatial distribution along the PG is also essential to guarantee the uniformity of the beam current. In SPIDER, this is possible thanks to the Cavity Ringdown Spectroscopy (CRDS) and the Laser absorption Spectroscopy diagnostics (LAS), which provide line-integrated measurements of negative ion density and neutral, ground state Cs density, respectively. The paper discusses the CRDS and LAS measurements as a function of input power and of the magnetic and electric fields used to reduce the coextraction of electrons. Negative ion density data are in qualitative agreement with the analogous measurements in Cs-free conditions. In agreement with simulations, Cs density is peaked in the center of the source; a top/bottom non uniformity is also present. Several effects of plasma on Cs deposition and negative ion production are presented. [ABSTRACT FROM AUTHOR]

• Feasibility analysis of SEE diagnostic system is given based on FEM. • A SEE diagnostic system for CRAFT NBI is designed. • Preliminary experiments of SEE diagnostic system is obtained on test facility. Neutral beam injection is one of the main auxiliary heating methods in Tokamak nuclear fusion device. The calorimeter which will be applied in Neutral beam injection system at Comprehensive Research Facility for Fusion Technology (CRAFT) is design for beam cutoff measurement, but limited by the design of double-layer water-cooling copper tubes paralleled to each other, the diagnostic system based on thermocouples cannot give the horizontal beam profile. To satisfy the requirement of diagnostic system, a beam diagnostic system based on Secondary Electron Emission (SEE) is designed. In order to achieve this goal, the rules of secondary electron emission are analyzed based on finite element methods (FEM) and the primary experiment has been carried out. Firstly, the laws of secondary electron emission during the beam bombarding on the calorimeter are analyzed; secondly, a set of signal acquisition system is set up, including sampling and filtering; finally, a preliminary experiment is carried out and the relationships between the signal of secondary electrons and the parameters, such as beam energy, beam current and bias voltage, are obtained. These results lay a foundation for the subsequent construction of the beam diagnostic system for CRAFT neutral beam injector (NBI). [ABSTRACT FROM AUTHOR]

A radio frequency (RF) ion source was designed and developed for neutral beam injector in Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP). Recently, the RF ion source was upgraded with new RF power generator (50 kW with frequency of 1 MHz), a new matching unit and the faraday shield with water cooling. Two new methods were used for the ion source tests on the test bed. The high pressure was used for the ignition of plasma and change to low pressure of 0.3 Pa to maintain the plasma. The RF plasma can be generated without the start filament successfully. In order to avoid the plasma oscillation with high power, the RF power was set to increase with two stages without change the matching unit. High power of 46 kW with pulse length of 22 s was achieved on the test bed and the plasma duration can be extended too. [ABSTRACT FROM AUTHOR]

High Voltage Bushing (HVB) is the key component of Diagnostic Neutral Beam (DNB) system of ITER as it provides access to high voltage electrical, hydraulic, gas and diagnostic feedlines to the beam source with isolation from grounded vessel. HVB also provides primary vacuum confinement for the DNB system. Being Safety Important Class (SIC) component of ITER, it involves several configurational, technological and operational challenges. To ensure its operational performance & reliability, particularly electrostatic behavior, half scale down Prototype High Voltage Bushing (PHVB) is designed considering same design criteria of DNB HVB. Design optimization has been carried out followed by finite element (FE) analysis to obtain DNB HVB equivalent electric stress on different parts of PHVB, taking into account all design, manufacturing & space constraints. PHVB was tested up to 60 kV without breakdown, which validates its design for the envisaged operation of 50 kV DC. This paper presents the design of PHVB, FEA validation, manufacturing constraints, experimental layout with interfacing auxiliary systems and operational results related to functional performance. [ABSTRACT FROM AUTHOR]

The ITER Neutral Beam (NBI) Test Facility, under construction in Padova, Italy consists in the ITER full scale ion source for the heating neutral beam injector, referred to as SPIDER, and the full size prototype injector, referred to as MITICA. The Control and Data Acquisition System (CODAS) for SPIDER has been developed and is going to be in operation in 2016. The system is composed of four main components: Supervision, Slow Control, Fast Control and Data Acquisition. These components interact with each other to carry out the system operation and, since they represent a common pattern in fusion experiments, software frameworks have been used for each (set of) component. In order to reuse as far as possible the architecture developed for SPIDER, it is important to clearly define the boundaries and the interfaces among the system components so that the implementation of any component can be replaced without affecting the overall architecture. This work reports the experience gained in the development of SPIDER components, highlighting the importance in the definition of generic interfaces among component, showing how the specific solutions have been adapted to such interfaces and suggesting possible approaches for the development of other ITER subsystems. [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]

Highlights • The details of design of a large area RF plasma generator for the EAST neutral beam injector was presented. • The performance of large area RF plasma generator was reported for the first time. • The parameters of large area plasma generator were measured and presented. Abstract A large area radio frequency (RF) ion source was developed for neutral beam injector in Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP). An experiment of RF source was installed in the test facility. The dimension of large area RF source is 862 mm (L) × 472 mm (W) × 420 mm (D). Six turns of coil is used to transfer RF power to the plasma. The plasma density was measured by the movable Langmuir probe in front of the plasma grid. The electron density of 6 × 1017 m−3 was achieved with RF power of 30 kW and source pressure of 0.5 Pa. The density uniformity was estimated as 95%. The heat loading on the Faraday shield and coil were also measured and acceptable for the long pulse operation with high power in the future. The details of the first tests of large area RF plasma source are presented in this paper. [ABSTRACT FROM AUTHOR]

The Acceleration Grid Power Supply of the MITICA test facility in Padova (Italy) is currently under commissioning. The power conversion system, the DC generator, and the High Voltage equipment have been individually commissioned, whereas the integration tests are ongoing. It is a challenging process due to the unconventional application, to the variety of different electrical technologies involved and to the complexity of the interfaces. During the integrated tests of the power supplies the achievement of 700 kV stable operation has been demonstrated for the first time in a Neutral Beam Injector, but an unexpected event occurred, most likely a breakdown in the HV part, which resulted in a fault of the DC generator. A subsequent test using an auxiliary power supply was performed to check the voltage withstanding capability of the HV plant, but another breakdown occurred at around 1MV. This paper describes the activity performed to identify the location of the breakdowns affecting the integrated tests. A test campaign has been devised with increased diagnostic capabilities and specific strategy conceived to trigger intentional breakdowns in specific locations and collect measurement patterns for different cases. The results of the campaign will be presented and the current understanding of the issue will be described, with a view on future tests and further improvements of diagnostics. [ABSTRACT FROM AUTHOR]

The Divertor Tokamak Test facility (DTT) is the new tokamak under construction in Frascati (Italy), with the main aim of testing different divertor magnetic configurations and technologies in view of DEMO. The Heating and Current Drive function will be provided also by one Neutral Beam Injector (NBI), accelerating negative Deuterium ions with an energy of 510 keV, giving an output power of 10 MW maximum. The Acceleration Grid Power Supply (AGPS), with 3 stages rated 167 kV each, will feed the NBI accelerator. For optical reasons, the insulating gap among the acceleration grids is at the limit of the voltage holding in vacuum, therefore frequent arc breakdowns (BD) among the grids will occur, requiring the quick switch-off of the AGPS to limit the arc energy. For the AGPS, beside the ITER-like scheme including gas-insulated HV components, an innovative approach based on the Modular Multilevel Converter (MMC) technology is being considered. This would include air-insulated HV converters, requiring a huge building volume, which at present is not available in close proximity to the tokamak building. The alternative of installing the MMC converters in a new building to be erected in an available area located far from the tokamak, and connecting the AGPS to the load with HV coaxial cables, is under study. This paper presents the studies carried out to verify the impact of the cables length on the arc energy, in case of breakdown among the grids, and to estimate the magnitude of the voltage oscillations due to resonance effects. These oscillations can occur not only between each hot pole and the return conductor, endangering the cable insulation, but also between the return conductor and local ground, with potential consequences on the laying method, the human safety and the electromagnetic compatibility with sensitive apparatus nearby, which have all been evaluated. [ABSTRACT FROM AUTHOR]

The long-pulse acceleration of the high-power positive ion beam has been demonstrated with the JT-60 positive ion source in the joint experiment among Japan Atomic Energy Agency (JAEA), Korea Atomic Energy Research Institute (KAERI) and National Fusion Research Institute (NFRI) under the collaboration program for the development of plasma heating and current drive systems. In this joint experiment, the increase of the heat load and the breakdowns induced by the degradation of the beam optics due to the gas accumulation was one of the critical issues for the long-pulse acceleration. As a result of development of the long-pulse operation techniques of the ion source and facilities of the neutral beam test stand in KAERI, 2 MW 100 s beam has been achieved for the first time. The achieved beam performance satisfies the JT-60SA requirement which is designed to be a 1.94 MW ion beam power from an ion source corresponding to total neutral beam power of 20 MW with 24 ion sources. Therefore, it was found that the JT-60 positive ion sources were applicable in the JT-60SA neutral beam injectors. Moreover, because this ion source is planned to be a backup ion source for KSTAR, the operational region and characteristic has been clarified to apply to the KSTAR neutral beam injector. [ABSTRACT FROM AUTHOR]

In order to optimize the design and operation of the Ion Source for the ITER Neutral Beam Injector (NBI), a dedicated 100 keV Ion Source, identified as Source for the Production of Ions of Deuterium Extracted from RF plasma (SPIDER), is under construction in the Neutral Beam Test Facility, at the Consorzio RFX premises, in Padua, Italy. The Ion Source, polarized at −112 kVdc voltage, will produce negative ions (Deuterium D − or Hydrogen H − ) which, extracted from the Plasma Grid with an extraction voltage up to 12 keV, are accelerated up to 100 keV by the 100 kVdc Power Supply (100 kV PS). The required Ion Source and the Extraction Power Supplies (ISEPS) system and the associated diagnostics are hosted inside a High Voltage Deck (HVD), a −100 kVdc air-insulated Faraday cage. These power supplies and various diagnostics and control equipment are connected to the Ion Source by means of a High Voltage Transmission Line (TL). The HVD (procurement started mid 2013; delivery on site in the second half of 2014) will consist of a wide structure (13 m ( L ) × 11 m ( W ) × 5 m ( H )), designed to support the weight of the ISEPS components, mounted on supporting insulators and clad with a conductive metal sheet in order to reduce the electromagnetic interference (EMI). The paper reports on the design solutions of the HVD focusing on insulation, mechanical, thermal and EMI issues. The details of the main HVD interfaces with the TL and with insulating transformer are also described. Finite Element (FE) analyses have been performed, on the one hand, to verify the configuration from the electrostatic point of view and, on the other, to evaluate electrical screen effectiveness from the EMI point of view. [ABSTRACT FROM AUTHOR]

Toward ITER, the long-pulse acceleration of MeV-class negative ion beams has been successfully demonstrated and R&Ds and the procurement of high-voltage components of 1 MV power supply system have been progressed. In the accelerator development after the production of 1 MeV beams, an extraction grid, where a water cooling is reinforced and the beam is steered by aperture offset to suppress grid heat loads, has been newly developed to extend the pulse duration of the MeV-class beams. As the result, a total grid heat load has been reduced from 20% to 10% of the total beam power. The beam pulse was successfully increased from 0.4 s to 60 s at 0.68 MeV, 100 A/m 2 and 12 s at 0.89 MeV, 156 A/m 2 beams. As for the procurement of the 1 MV power supply system, 1 MV insulating transformer as the critical component has been developed. In the mock-up, a dc 1.2 MV for 1 h with the margin of 20% was stably sustained. Through the R&Ds of the critical components, design of five power supplies procured by JAEA has been finalized, and out of which a testing power supply has been completed in June 2014. [ABSTRACT FROM AUTHOR]

The SPIDER experiment, currently under construction at the Neutral Beam Test Facility in Padua, Italy, is a full-size prototype of the ion source for the ITER Neutral Beam Injectors. The Ion Source and Extraction Power Supplies (ISEPS) for SPIDER are supplied by OCEM Energy Technology s.r.l. (OCEM) under a procurement contract with Fusion for Energy (F4E) covering also the units required for MITICA and ITER injectors. The detailed design of SPIDER ISEPS was finalized in 2011 and manufacture of most components completed by end 2013. The Factory Acceptance Tests took place early 2014. ISEPS, with an overall power rating of 5 MVA, form a heterogeneous set of items including solid state power converters and 1 MHz radiofrequency generators of 200 kW output power. The paper presents the main features of the detailed design developed by OCEM, focusing in particular on the high output voltage pulse step modulators, the high output current resonant converters, the radiofrequency generators by HIMMELWERK GmbH and the architecture and implementation of the complex control system. Details are given on non-standard factory tests verifying the insulation requirements specific to this application. Performance of ISEPS during the factory acceptance tests is described, with emphasis on demonstration of the load protection requirements, a crucial point for all neutral beam power supplies. Finally, key dates of SPIDER ISEPS installation and site testing schedule are provided. [ABSTRACT FROM AUTHOR]

The SPIDER experiment (Source for the Production of Ions of Deuterium Extracted from an RF plasma) aims to qualify and optimize the full size prototype of the negative ion source foreseen for MITICA (full size ITER injector prototype) and the ITER Heating and Current Drive Injectors. Both SPIDER and MITICA experiments are presently under construction at Consorzio RFX in Padova (I), with the financial support from IO (ITER Organization), Fusion for Energy, Italian research institutions and contributions from Japan and India Domestic Agencies. The vacuum vessel hosting the SPIDER in-vessel components (Beam Source and calorimeters) has been manufactured, assembled and tested during the last two years 2013–2014. The cylindrical vessel, about 6 m long and 4 m in diameter, is composed of two cylindrical modules and two torispherical lids at the ends. All the parts are made by AISI 304 L stainless steel. The possibility of opening/closing the vessel for monitoring, maintenance or modifications of internal components is guaranteed by bolted junctions and suitable movable support structures running on rails fixed to the building floor. A large number of ports, about one hundred, are present on the vessel walls for diagnostic and service purposes. The main working steps for construction and specific technological issues encountered and solved for production are presented in the paper. Assembly sequences and tests on site are furthermore described in detail, highlighting all the criteria and requirements for correct positioning and testing of performances. [ABSTRACT FROM AUTHOR]

In ITER, each heating neutral beam injector (HNB) will deliver about 16.5 MW heating power by accelerating a 40 A deuterium negative ion beam up to the energy of 1 MeV. The ions are generated inside a caesiated negative ion source, where the injected H 2 /D 2 is ionized by a radio frequency electromagnetic field. The SPIDER test bed, currently being manufactured, is going to be the ion source test facility for the full size ion source of the HNBs and of the diagnostic neutral beam injector of ITER. The SPIDER beam source comprises an ion source with 8 radio-frequency drivers and a three-grid system, providing an overall acceleration up to energies of about 100 keV [1] . SPIDER represents a substantial step forward between the half ITER size ion source, which is currently being tested at the ELISE test bed in IPP-Garching, and the negative ion sources to be used on ITER, in terms of layout, dimensions and operating parameters. The SPIDER beam source will be housed inside a vacuum vessel which will be equipped with a beam dump and a graphite diagnostic calorimeter. The manufacturing design of the main parts of the SPIDER beam source has been completed and many of the tests on the prototypes have been successfully passed. The most complex parts, from the manufacturing point of view, of the ion source and the accelerator, developed by galvanic deposition of copper are being manufactured. The manufacturing phase will be completed within 2015, when the assembly of the device will start at the PRIMA site, in Padova (I). The paper describes the status of the procurement, the adaptations operated on the design of the beam source for the fabrication, with particular emphasis to the engineering development to enable the fulfillment of the tight requirements set in the technical specifications. Moreover the tests performed on the prototypes are reported. [ABSTRACT FROM AUTHOR]

Magnetic materials play an important role in the magnetic fields distribution inside Tokamaks, and their contributions must be carefully computed. In some applications, when 3D geometries are involved and high accuracy must be achieved, finite elements may reveal too computationally demanding and different approaches must be considered. In this paper, two possible solutions are presented. The first one is based on the discretization of magnetic parts into rectangular prisms with constant magnetization density; analytical formulas are then used to compute the field contribution due to each brick. The second one is based on the representation of the effect of magnetic materials as a set of dipoles, by using again analytical formulas to evaluate the contribution to the magnetic field. Iterative procedures are applied to evaluate the magnetization density in prisms. The efficiency and accuracy of both models are validated against commercial codes on a test case, and then applied to the computation of magnetic field in presence of Neutral Beam Injector in an ITER-like geometry. [ABSTRACT FROM AUTHOR]

The neutral beam system for ITER consists of two heating and current drive neutral ion beam injectors (HNB) and a diagnostic neutral beam (DNB) injector. The proposed physical plant layout allows a possible third HNB injector to be installed later. The HNB Passive Magnetic Shield (PMS) works in conjunction with the active compensation/correction coils to limit the magnetic field inside the Beam Line Vessel (BLV), Beam Source Vessel (BSV), High Voltage Bushing (HVB) and Transmission Line (TL) elbow to acceptable levels that do not interfere with the operation of the HNB components. This paper describes the current design of the PMS, having had only minor modifications since the preliminary design review (PDR) held in IO in April 2013, and the assembly strategy for the vessel PMS. [ABSTRACT FROM AUTHOR]

The neutral beam system for ITER consists of two heating and current drive injectors plus a diagnostic neutral beam injector. The proposed physical plant layout allows for a possible third heating injector to be installed later. For correct operation of the beam source, and to avoid deflections of the charged fraction of the beam, the magnetic field along the beam path must be very low. To minimize the stray ITER field in critical areas (ion source, acceleration grids, neutralizer, residual ion dump), a Magnetic Field Reduction System will envelop the beam vessels and the high voltage transmission lines to ion source. This whole system comprises the Passive Magnetic Shield, a set of thick steel plates, and the Active Correction and Compensation Coils, a set of coils carrying currents which depend on the tokamak stray field. This paper describes the status of the coil design, terminals and support structures, as well as a description of the calculations carried out. Most coils are suitable for removal from their final position to be replaced in case of a fault. Conclusions of the chosen design highlight the strategy for the system feasibility. [ABSTRACT FROM AUTHOR]