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SPIDER is the full-scale prototype of the ion source of the ITER Heating Neutral Beam Injector, where negative ions of Hydrogen or Deuterium are produced by a RF generated plasma and accelerated with a set of grids up to ~100 keV. The Power Supply System is composed of high voltage dc power supplies capable of handling frequent grid breakdowns, high current dc generators for the magnetic filter field and RF generators for the plasma generation. During the first 3 years of SPIDER operation different electrical issues were discovered, understood and addressed thanks to deep analyses of the experimental results supported by modelling activities. The paper gives an overview on the observed phenomena and relevant analyses to understand them, on the effectiveness of the short-term modifications provided to SPIDER to face the encountered issues and on the design principle of long-term solutions to be introduced during the currently ongoing long shutdown., 8 pages, 12 figures. Presented at SOFT 2022

In SPIDER, the full-scale prototype of the negative ion source of the ITER Heating Neutral Beam Injectors, negative ions of hydrogen or deuterium are extracted and accelerated by a set of grids. This article describes the process and the technical solutions that allowed achieving the asseveration to operate the Acceleration Grid Power Supply of SPIDER at full voltage (-96 kV), reports the first experimental results, and discusses some possible enhancements that have been identified.

Supervisory control system for superconducting magnet power supplies in JT-60SA is designed and developed. This system is in charge of many kinds of control; discharge sequence control, real-time magnet current control, data acquisition, machine protection and human safety. In the design, communication interface has to be optimized according to all requirements. The most important feature is real-time control with 4 kHz which is required for equilibrium control of plasma. In order to satisfy this requirement, reflective memory (RFM) network is applied for high speed data sharing with multiple controllers. This paper describes the architecture of the control system and the operation of the power supply system under the supervision.

The main purpose of the Divertor Tokamak Test facility (DTT) [1], whose construction is starting in Frascati, Italy, is to study solutions to mitigate the issue of power exhaust in conditions relevant for ITER and DEMO. DTT will be equipped with a significant amount of auxiliary heating power (45 MW) to reach PSEP/R = 15 MW m-1 required to be DEMO-relevant [2]. DDT is characterized by high flexibility for the assembling and testing of divertor components and for the different magnetic configurations to address the integrated physics and technology problems. The current conceptual design of the beamline for the DTT Neutral Beam Heating system is here presented, with a particular focus on the effect on the DTT plasma and on the technical solutions adopted to maximize the RAMI indexes (Reliability, Availability, Maintainability and Inspectability) and minimize complexity and costs. Various design options were considered, and a comprehensive set of simulations was carried out using several physics and engineering codes to drive the choice of the most suitable design options and optimize them, aiming at finding a good compromise among different requirements. This paper describes the design of the main components of the DTT NBI beamline, explaining the motivations for the main design choices.

The European DEMO will make use of a significant additional heating power, which could be partly provided by Neutral Beam Injectors (NBI), which accelerate negative ions by means of grids placed at increasing potentials. These grids will be fed by the Acceleration Grid Power Supply (AGPS), divided in a number of stages, which has to provide an overall dc voltage down to -1 MV and currents in the order of tens of Amperes. The AGPS will have to satisfy a set of static and dynamic requirements, in terms of ripple, accuracy, rise time and so on. In addition, during the NBI operation, frequent losses of insulation between the grids are expected, with consequent arc discharges, called breakdowns. The AGPS will be able to handle such events by nullifying the output currents as fast as possible in order to limit the energy discharged onto the grids. Adopting the Modular Multilevel Converter (MMC) technology for the AGPS of DEMO and future tokamaks seems promising, due to its intrinsic properties of modularity, high efficiency, fast dynamic response and small energy transferred to the arc in case of breakdown. One of the main drawbacks of this technology is the large volume occupied by the converter, which is air-insulated. This can be partly reduced by adopting alternative MMC schemes, to minimize the number of components and optimize the counter-voltage applied at breakdown. In this paper, alternative topologies for the MMC submodules or combinations of different schemes (full-bridge, half-bridge) are investigated. After a preliminary design of the converter, the results of numerical simulations carried out with circuit models are shown, with control schemes customized for the NBI operation. The performance of the different solutions in steady-state, dynamic and anomalous conditions are discussed, with particular focus on breakdown events. Finally, thermal analyses on the converter are carried out, to verify whether natural convection of air can be a suitable cooling method for the power components.

The Divertor Tokamak Test facility (DTT) is the new tokamak whose construction is starting in Frascati (Italy), with the main aim of testing different divertor magnetic configurations and technologies in view of DEMO. In the present design, the DTT heating mix includes one Negative Neutral Beam Injector (NBI), with a power injected in the plasma of 10 MW and a beam energy of 500 keV. The NBI conceptual design is under development and includes solutions already adopted for ITER (RF-driven ion source, gas neutralizer, electrostatic Residual Ion Dump), but with some simplifications to reduce costs and complexity deemed feasible for DTT. In particular, the beam source is air-insulated and fed by the gas-insulated Transmission Line through gas-air bushings. This paper presents the status of the conceptual design of the main power supplies of the DTT NBI. The main ratings are collected from physics requirements and NBI design. On this basis, the reference technical solutions are selected, considering the experience gained at the ITER Neutral Beam Test Facility in Padova (Italy), the specific design of the DTT NBI and the layout constraints of the DTT Site. For the Acceleration Grid Power Supply, besides the ITER-like design, an innovative approach based on the Modular Multilevel Converter technology, never adopted in this application, is considered; pros and cons of the two solutions are evaluated.

Heating neutral beam (HNB) injectors, necessary to achieve burning conditions and to control plasma instabilities in ITER, are characterized by such demanding parameters that a neutral beam test facility (NBTF) dedicated to their development and optimization is being realized in Padua (Italy) with direct contributions from the Italian government (through Consorzio RFX as the host entity) and the ITER international organization (with kind contributions from the ITER domestic agencies of Europe, Japan and India) and technical and scientific support from various European laboratories and universities. The NBTF hosts two experiments: SPIDER, devoted to ion source optimization for the required source performance, and MITICA, the full-size prototype of the ITER HNB, with an ion source identical to SPIDER. This paper gives an overview of the progress towards NBTF realization, with particular emphasis on issues discovered during this phase of activity and on solutions adopted to minimize the impact on the schedule and maintain the goals of the facilities. The realization of MITICA is well advanced; operation is expected to start in 2023 due to the long procurement time of the in-vessel mechanical components. The beam source power supplies, operating at 1 MV, are in an advanced phase of realization; all high-voltage components have been installed and the complex insulation test phase began in 2018. At the same time, construction and installation of SPIDER plant systems was successfully completed with their integration into the facility. The mechanical components of the SPIDER ion source were installed inside the vessel and connected to the plants. Integrated commissioning with the control, protection and safety systems ended positively and the first experimental phase has begun. The first results of the SPIDER experiment, with data from operational diagnostics, and the plans for the 1 MV insulation tests on the MITICA high-voltage components are presented.

In the present pre-conceptual design, the base converters to supply the main DEMO poloidal magnets are rated for 45 kA and about 10 kV. If the traditional design approach, based on thyristor bridges, was adopted, this would result in very large reactive power exchanged when low voltage values are required by the load. To satisfy the limitations imposed by National Grid Operators, the reactive power should be compensated inside the DEMO Plant Electrical System by means of a Reactive Power Compensation and Harmonic Filtering system. This could represent an additional cost and takes up a large area of the plant. This paper shows that the reactive power demand, estimated with an analytical model starting from the current/voltage waveforms foreseen in DEMO magnets, is huge. Therefore, an innovative approach, based on Active Front End converters, is proposed and compared with the traditional solution for coil power supplies.

The Residual Ion Dump Power Supply (RIDPS) is devoted to feed the Electrostatic Residual Ion Dump, a device which shall collect the residual ions at the output of the gas Neutralizer in the ITER Heating Neutral Beam Injectors (HNBI) and their full-scale prototype (MITICA), presently under construction in the Neutral Beam Test Facility (NBTF) in Padua, Italy. The procurement of the three RIDPS, part of the Procurement Arrangement signed between Fusion for Energy (F4E) and ITER Organization (IO) in 2009, has been committed by F4E to OCEM Energy Technology s.r.l. (OCEM). This power supply has to provide an average voltage of up to 25 kV, plus an ac low-frequency voltage component with maximum amplitude of 5 kV. The nominal output current is 60 A, and the maximum pulse duration is 1 hr. The selected topology is the Pulse-Step-Modulator (PSM), including a multi-winding dry type transformer and 42 water-cooled switching power modules, connected in series at the output. Among the various advantages of this technology, the most important for this application are the modularity, allowing redundancy at module level, and the light output filter and short switch-off time in case of fault, both limiting the energy delivered to possible arches among RID plates. The manufacturing of the RIDPS for MITICA has been concluded in summer 2018. The Factory Tests on the single components and on the full system were finished in October 2018 and in the same month started the Site Installation activities at NBTF. At the end of November the installations were completed and the Site Tests started, planned to be completed by beginning of 2019. A close follow-up of the whole procurement process has been carried out by both F4E and Consorzio RFX. Particular effort has been devoted to the verification of the design, the definition of the test procedures and the integration of the system with the existing installations, including the management of layout interferences and the detailed scheduling of site works. In particular, during the factory tests, some problems appeared, requiring the definition of proper solutions. For example, in the original specification the frequency of the ac component of the output voltage was set to 50 Hz, while it has been found that this caused unwanted input current peaks due to synchronization with the grid voltage, with possible saturation of the transformer. Electromagnetic immunity issues have output filter caused a voltage ripple slightly exceeding the limit with reversed polarity, solved with the improvement of the filtering action. Finally, a suitable test procedure has been devised for the Electromagnetic Emission Tests at Site. In this paper, the most significant issues faced during the design, integration and tests and the solutions identified will be described. The layout adopted in the final installation will be commented and the most interesting results of the Factory and Site Tests will be presented, with particular emphasis to those proving the compliance of the RIDPS with the most critical requirements.

JT-60SA is the satellite tokamak under construction in Naka, Japan, in the framework of the EU-JA -Broader Approach? Agreement. In JT-60SA, to attain steady-state high-beta plasmas, suppression of Resistive Wall Modes (RWM) is necessary. At this purpose, a passive stabilizing plate (SP) and an active control system based on 18 in-vessel sector coils (SC) are foreseen. In the past years, physics studies have been carried out to produce the inputs for the design of the RWM control system and the SC design evolved accordingly. In the present arrangement, the coils are placed on the plasma side of SP, to minimize the magnetic shielding effects of passive structures. Electromagnetic analyses on SC have been carried out in support to the evaluation of the different coil design solutions and to derive the requirements for the related power supply system (RWM-PS). To achieve the highest flexibility in RWM control, it has been devised to feed each SC independently, with currents following 18 real-time references. In 2012, the SC design has been further improved by increasing the number of turns from 2 to 8. This allowed reducing the rated current, thus the voltage drops on the feeders and the power required from RWM-PS. To keep the RWM under control with relatively low Ampere-turns (2.2 kAt), the RWM-PS has to guarantee very fast dynamics (current bandwidth: 3 kHz; maximum latency: 50 us), largely beyond that of standard industrial applications. This paper firstly reports the main requirements for the RWM control system. Then, the reference design of the RWM-PS is described. It includes an ac/dc conversion system, dc-link capacitor banks and a set of 18 fast inverters. The advantages of the proposed scheme are discussed and the main electrical parameters are shown in detail. The main requirements of the control section are given, with details on possible implementation and interfaces with JT-60SA central control.

In June 2018 the SPIDER device, which is the full-size prototype of the negative ion source and extractor for the ITER Heating Neutral Beam has entered the first operation phase at the Neutral Beam Test Facility (NBTF) in Padova, Italy. This paper describes the present status of the device, the experimental plans and the results obtained during the first experimental campaign of the radio-frequency driven plasma source.

An advanced technology has been developed and employed for the main circuit breakers (CB) of the quench protection circuits (QPC) of the superconducting coils of JT-60SA: it consists in a Hybrid mechanical-static CB (HCB) composed of a mechanical Bypass switch (BPS) for conducting the continuous current, in parallel to a static circuit breaker (SCB) based on integrated gate commutated thyristor (IGCT) for current interruption. It was the result of a R&D program carried out since 2006 to identify innovative solutions for the interruption of high dc current, able to improve the maintainability and availability of the CB. The HCB developed for the JT-60SA QPC is the first realization of a dc circuit breaker based on this design approach for interrupting current of some tens of kA with reapplied voltage of some kV. It also represents the first application of hybrid technology with IGCT for protection of superconducting magnets in fusion experiments. The paper aims at giving a comprehensive overview of the main R&D activities devoted to the development of this new technological approach; then, the key aspects of the design, manufacturing and testing of the QPCs for JT-60SA, successfully completed in Naka Site in summer 2015 are presented. Finally, the significance of this research is discussed and the possible future developments, in particular in view of DEMO fusion reactor, are outlined.

One of the main objectives of JT-60SA, the satellite tokamak under construction in Naka (Japan), is to confine steady-state high-beta plasmas. To reach the desired plasma performance, the control of the instabilities called Resistive Wall Modes (RWMs) is crucial. At this purpose, besides the combination of an in-vessel passive structure (stabilizing plate) and plasma rotation, a dedicated active control system based on 18 in-vessel sector coils has been devised. These coils are placed just behind the tiles of the first wall and around the ports; therefore, due to the plasma proximity and the low shielding effect of the surrounding passive structures, they can efficiently generate fast magnetic fields if properly driven. The strategy is to succeed in controlling RWMs (which grow exponentially) when their amplitude is still low, so that low magnetic field components and relevant current to produce them are sufficient. As a consequence, the power rating of the power supply system is not so demanding, but at the price of high dynamic performance. To achieve an effective control, each coil will be independently fed by a dedicated inverter, rated for 300 A, 240 V, with a closed-loop bandwidth of 3 kHz and a latency between reference and output voltage lower than 50 μs. These performances exceed the standard industrial requirements and call for the adoption of innovative solutions. To exploit the simple and compact H-bridge topology, a switching frequency higher than that sustainable by standard silicon IGBTs is necessary; the adoption of new power semiconductors based on Silicon-Carbide, capable to switch at 30 kHz (60 kHz equivalent at the output), and a very fast control system have been the key design choices to satisfy the requirements. To prove the feasibility at reasonable cost of this design, an inverter prototype has been developed in 2014, whose performance was very satisfactory. Therefore, the same inverter design has been confirmed also for the final power supply system (called RWM-PS), which is now under manufacturing. This will be composed of an ac disconnector, a step-down transformer, two ac/dc rectifiers, a distributed dc-link capacitor bank, 18 water-cooled fast inverters and a Local Control Cubicle. Both differential-mode and common-mode filters are foreseen at the inverter outputs; their main aims are to reduce the current ripple and minimize the electro-magnetic interferences due to the very long coaxial cables connecting the fast inverters to the coils. The control system supervises all the plant. Each inverter can be individually controlled in current control loop or in open-loop, the reference being arbitrary and generated by the JT-60SA MHD Controller. The RWM-PS will be the first power supply system for fast control of plasma instabilities in fusion experiments adopting SiC semiconductors. The paper will give an overview of the final design of the RWM-PS, with particular emphasis on its special features and the solutions adopted to satisfy the critical requirements. The interface issues for an effective integration with the JT-60SA power and control systems will be also treated., 27th IEEE Symposium on Fusion Engineering (SOFE-2017)

This paper deals with the qualification process of the full scale prototypes of the Hybrid mechanical-static dc Circuit Breaker (HCB) for the Quench Protection Circuits (QPC) of the Toroidal Field (TF) and Poloidal Field (PF) superconducting coils of the Satellite Tokamak JT-60SA. The HCB developed for JT-60SA QPC is the first dc circuit breaker based on hybrid mechanical-static design at this level of power (25.7 kA-1.93 kV, +/- 20 kA-+/- 3.8 kV). Moreover, the JT-60SA QPC represents the first application of protection for superconducting magnets based on this hybrid technology. Special type tests have been designed to verify the performance of the device up to the nominal ratings and beyond, thus proving the suitability or the technology, the design margins, and the reliability; the results of the most significant tests are presented and discussed. The qualification program also includes the validation of the electrical models developed during the design phase, which are described in the paper, too.

The realization of the ITER Neutral Beam Test Facility (NBTF) and the start the experimental phase are important tasks of the fusion roadmap, since the target requirements of injecting to the plasma a beam of Deuterium atoms with a power up to 16.5 MW, at 1MeV of energy and with a pulse length up to 3600s have never been reached together before. The ITER NBTF, called PRIMA (Padova Research on ITER Megavolt Accelerator), is hosted in Padova, Italy; it includes two experiments: MITICA, the full-scale prototype of the ITER injector and SPIDER, the full-size negative ion source. The realization promoted by the ITER organization is carried out with the contribution of the European Union, channeled through the Joint Undertaking for ITER (F4E), of the Consorzio RFX which hosts the Test Facility, the Japanese and the Indian ITER Domestic Agencies (JADA and INDA) and several European laboratories, such as IPP-Garching, KIT-Karlsruhe, CCFECulham, CEA-Cadarache. The early start of operation of PRIMA experiments is urgent because sufficient experimental time is necessary to face and solve the issues related to the achievement of the desired performance in time for the ITER operation, requiring NBI since the beginning. Substantial progresses have been recently achieved: the buildings construction, begun in October 2012, has been completed by the end of 2015 and the installation of some components has been started since the end of 2014. The SPIDER realization is well advanced: the installation phase is proceeding in good agreement with the general plan; it is expected to be almost completed by the end of 2016. In parallel, the commissioning of the SPIDER power supply (PS) and auxiliary plant systems is being proceeding. Tests at full power and remote control are planned, including also those addressed to reproduce the grid breakdowns and to test the relevant protections. The design of the MITICA injector components was completed in 2015their procurement is being made through a number of tenders, some of them already launched. The HV Power Supply system of the MITICA 1MV accelerator, provided by JADA, was delivered on site in December 2015. The challenging installation of these components, including the step-up transformers and the SF6 gas insulated transmission line, started soon after and will go on throughout 2016. The present phase, with the PRIMA buildings continuously filled with new components, with the installation activities progressing and with also the commissioning and testing phase starting represents a substantial step forward toward the main target. The paper will describe the main challenges the Project Team has dealt during this phase and the important feedback derived for the ITER HNB systems both from the technical and the organizational standpoints.

The Residual Ion Dump Power Supply (RIDPS) is part of the Ground Related Power Supplies (GRPS) of the MITICA experiment of the ITER Neutral Beam Test Facility (NBTF) and the two ITER Heating Neutral Beam Injectors (HNBI). The GRPS will be manufactured by OCEM Energy Technology s.r.l. (OCEM) via a procurement contract with F4E. The RIDPS is devoted to feed the electrostatic Residual Ion Dump (RID), which deflects and collects the beam residual ions after the neutralization process. The maximum average voltage of the RIDPS is 25 kV, to which can be superimposed a sinusoidal or trapezoidal alternate voltage at 50 Hz, 5 kV maximum. The nominal current is 60 A, with a maximum pulse length of 1 h. This paper describes the detailed design of the RIDPS, highlighting its peculiar aspects, and the expected performance resulting from simulations.

SPIDER is one of two projects of the ITER Neutral Beam Test Facility under construction in Padova, Italy, at the Consorzio RFX premises. It will have a 100 keV beam source with a full-size prototype of the radiofrequency ion source for the ITER neutral beam injector (NBI) and also, similar to the ITER diagnostic neutral beam, it is designed to operate with a pulse length of up to 3600 s, featuring an ITER-like magnetic filter field configuration (for high extraction of negative ions) and caesium oven (for high production of negative ions) layout as well as a wide set of diagnostics. These features will allow a reproduction of the ion source operation in ITER, which cannot be done in any other existing test facility. SPIDER realization is well advanced and the first operation is expected at the beginning of 2018, with the mission of achieving the ITER heating and diagnostic NBI ion source requirements and of improving its performance in terms of reliability and availability. This paper mainly focuses on the preparation of the first SPIDER operations - integration and testing of SPIDER components, completion and implementation of diagnostics and control and formulation of operation and research plan, based on a staged strategy

This paper presents a technique to model electric devices and systems, which cannot be represented adequately with rational transfer functions, to be integrated in electric transient simulations. Such technique has been successfully validated and exploited in the modeling and optimization phase of a fast switching converter with its load, which is a coil subject to skin, proximity and shielding effects. Furthermore, the possibility to use this technique in order to control such converter has been investigated, obtaining promising results.

The ITER Neutral Beam Test Facility (NBTF), called PRIMA (Padova Research on ITER Megavolt Accelerator), is hosted in Padova, Italy and includes two experiments: MITICA, the full-scale prototype of the ITER heating neutral beam injector, and SPIDER, the full-size radio frequency negative-ions source. The NBTF realization and the exploitation of SPIDER and MITICA have been recognized as necessary to make the future operation of the ITER heating neutral beam injectors efficient and reliable, fundamental to the achievement of thermonuclear-relevant plasma parameters in ITER. This paper reports on design and R&D carried out to construct PRIMA, SPIDER and MITICA, and highlights the huge progress made in just a few years, from the signature of the agreement for the NBTF realization in 2011, up to now - when the buildings and relevant infrastructures have been completed, SPIDER is entering the integrated commissioning phase and the procurements of several MITICA components are at a well advanced stage

The RFX-mod experiment is a fusion device designed to operate as a reversed field pinch (RFP), with a major radius R = 2 m and a minor radius a = 0.459 m. Its high versatility recently allowed operating it also as an ohmic tokamak, allowing comparative studies between the two configurations in the same device. The device is equipped with a state of the art MHD mode feedback control system providing a magnetic boundary effective control, by applying resonant or non-resonant magnetic perturbations (MP), both in RFP and in tokamak configurations. In the fusion community the application of MPs is widely studied as a promising tool to limit the impact of plasma filaments and ELMs (edge localized modes) on plasma facing components. An important new research line is the exploitation of the RFX-mod active control system for ELM mitigation studies. As a first step in this direction, this paper presents the most recent achievements in term of RFX-mod tokamak explored scenarios, which allowed the first investigation of the ohmic and edge biasing induced H-mode. The production of D-shaped tokamak discharges and the design and deployment of an insertable polarized electrode were accomplished. Reproducible H-mode phases were obtained with insertable electrode negative biasing in single null discharges, representing an unexplored scenario with this technique. Important modifications of the edge plasma density and flow properties are observed. During the achieved H-mode ELM-like electromagnetic composite filamentary structures are observed. They are characterized by clear vorticity and parallel current density patterns.

This paper presents the development and testing of the prototype of a hybrid mechanical-static dc circuit breaker (CB) for quench protection of superconducting magnets based on a mechanical switch paralleled to a static CB. In normal operation, the mechanical switch is closed and it handles the continuous flow of dc current, minimizing the on-state losses. In case of intervention, the mechanical switch opens, commutating the current into the static CB that quickly interrupts and transfers the current into a discharge resistor, withstanding the reapplied voltage. This paper was carried out to face and solve some issues of this design solution being not available in the literature examples of similar applications rated for significant power levels. The operation of the hybrid CB has been successfully tested up to a current of 10 kA with a recovery voltage of about 1 kV. Special tests have been dedicated to characterize the reliability and the repeatability of the current commutation from the mechanical switch to the static CB and to characterize the arc voltage under different conditions.

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

The paper describes the main features of the Superconducting Magnets Power Supply to generate the toroidal and poloidal magnetic fields in JT-60SA tokamak, with special regard to coil current regulation mode and magnets protection.

The paper presents an overview on the power supply (PS) systems for plasma instabilities control in fusion experiments, based on active control coils. First, the MHD instabilities and the approach to their control in Tokamaks and Reversed Field Pinches (RFPs) are described. Then, the features of MHD modes controls presently used in fusion experiments are reviewed. For the control systems based on active coils fed by fast power supplies, the typical requirements in terms of power, dynamics, accuracy and delay are summarized and discussed. Then, a survey on the technology available to design these types of PSs is given, together with the most suitable circuit topologies and guidelines for the design, on the basis of solutions adopted in existing experiments.

In the framework of EU – JA “Broader Approach” Agreement, the JT-60SA satellite tokamak will be built in Naka, Japan. In JT-60SA, to achieve the steady-state high beta plasma, suppression of Resistive Wall Modes (RWM) is necessary. Therefore, a passive stabilizing plate and a RWM active control system based on in-vessel coils are foreseen for JT-60SA. In the present design, this system consists in 18 sector coils, 6 in the toroidal, 3 in the poloidal direction, rated for a maximum current of 2.5 kA and fed by a dedicated fast power supply system. The coils shall produce sufficient radial magnetic field at the plasma edge, and the active control system shall be fast enough to react in real time to the plasma instabilities. In addition, during fast transients of the plasma current, significant loop voltage could be induced along the sector coils, which could result in over-currents in the coils and in the power supplies. This paper deals with the electromagnetic analyses on the sector coils of JT-60SA, with the main aim to determine the voltage/current characteristic of the coils and, at the same time, to have a first estimation of amplitude and shape of the produced magnetic field; different sector coil design solutions were considered and evaluated. The studies on the coil over-currents during plasma disruption or vertical displacement events are also presented. The results give useful indications in support to the development of the sector coils and power supply design and the identification of their electrical and functional requirements.

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 (1 MV). These two different devices will separately address the main scientific and technological issues of the 17 MW 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 1 h. The required negative ion current density to be extracted from the plasma source ranges from 290 A/m2 in D2 (D−) and 350 A/m2 in H2 (H−) and these values should be obtained at the lowest admissible neutral pressure in the plasma source volume, nominally at 0.3 Pa. 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 1 m2. The main design choices will be presented in the paper as well as an overview of the design of the main components and systems.

The effective control of magnetohydrodynamic instabilities in fusion experiments by means of actively controlled coils calls for challenging dynamic requirements to the relevant power supply (PS) systems. The first part of this paper gives an overview on typical requirements and technological solutions adopted in some present fusion experiments. Then, the advantages of SiC devices with respect to the Si ones are described in this paper, including lower switching losses, higher switching speed, higher operating junction temperature, and higher voltage capability. As an application example, this paper reports on the case of the PS system for resistive wall mode control in the JT-60SA satellite tokamak, where analyses showed the difficulty of satisfying the demanding requirements in terms of high current bandwidth (3 kHz) and short latency ( $ ) with a simple H-bridge topology adopting standard IGBT. On the contrary, the use of Si–SiC IGBT with the same topology can allow meeting the specifications, as demonstrated by the development and test of such a power amplifier, rated for $300~A_{{\mathrm{pk}}}$ and 240 V.

In this paper, the design and experimental results of a prototype of a 72 kVA very fast converter for physics research are presented. Hybrid Si-SiC IGBT power modules and an advanced fast control scheme were used to achieve very small response latency (

The paper, after an overview of the typical requirements of Power Supply (PS) systems to control MagnetoHydroDynamic (MHD) instabilities and of technological solutions adopted in some present fusion experiments, reports on the last progress in the production of SiC power devices, which features can be interesting for this type of applications. The advantages of the SiC devices with respect to the Si ones include lower switching losses, higher switching speed, higher operating junction temperature and a higher voltage capability. The paper reports on the case of the PS system for RW M control in the JT - 60SA satellite tokamak, where analyses showed the difficulty of satisfying the demanding requirements in terms of high current bandwidth (3 kHz) and short latency (

The ITER project requires additional heating by two neutral beam injectors, each accelerating to 1 MV a 40 A beam of negative deuterium ions, to deliver to the plasma a power of about 17 MW for one hour. As these requirements have never been experimentally met, it was recognized as necessary to setup a test facility, PRIMA (Padova Research on ITER Megavolt Accelerator), in Italy, including a full-size negative ion source, SPIDER, and a prototype of the whole ITER injector, MITICA, aiming to develop the heating injectors to be installed in ITER. This realization is made with the main contribution of the European Union, through the Joint Undertaking for ITER (F4E), the ITER Organization and Consorzio RFX which hosts the Test Facility. The Japanese and the Indian ITER Domestic Agencies (JADA and INDA) participate in the PRIMA enterprise; European laboratories, such as IPP-Garching, KIT-Karlsruhe, CCFE-Culham, CEA-Cadarache and others are also cooperating. Presently, the assembly of SPIDER is on-going and the MITICA design is being completed. The paper gives a general overview of the test facility and of the status of development of the MITICA and SPIDER main components at this important stage of the overall development; then it focuses on the latest and most critical issues, regarding both physics and technology, describing the identified solutions.

JT-60SA, the superconducting tokamak under construction in Japan, will be equipped with a mix of new and reused Power Supply Systems. Most of the new Power Supply Systems are procured by European Voluntary Contributors under the framework of Broader Approach agreement between Europe and Japan. For the toroidal circuit, the 25.7 kA steady state ac/dc converter is procured by CEA. For the poloidal circuits, the procurement of 10 ac/dc converters, almost all rated ±20 kA and about ±1 kV is shared between CEA and ENEA. The two in-vessel coils for fast control of plasma position are supplied with two independent thyristor converters procured by ENEA, each one rated ±5 kA and ±1 kV. The operation of four Switching Network Units procured by ENEA inserts settable resistors in series to each Central Solenoid module, producing up to 5 kV at the nominal current of 20 kA and allowing stable plasma initiation. The protection of toroidal and poloidal superconducting magnets is assured by thirteen Quench Protection Circuits procured by Consorzio RFX, rated ±20 kA and ±3.8 kV for poloidal coils and 25.7 kA and 2.8 kV for toroidal ones, inserting a discharge resistor in the related circuit for fast dissipation of the magnetic energy stored in the coil. Finally, the suppression of Resistive Wall Modes is actively performed by 18 in-vessel coils, independently supplied by a corresponding number of fast inverters procured by Consorzio RFX, each one rated for 300 A and 240 V. The key aspects of the design of the aforementioned Power Supply Systems are described in the paper, informing also about the present status of the procurement: their design was completed and some systems were already manufactured, tested and delivered to Japan. © 2015 Elsevier B.V. All rights reserved.

JT-60SA will be provided with a set of power supply systems procured by Europe and Japan under the framework of Broader Approach Agreement. The toroidal circuit is supplied by an ac/dc thyristors converter rated for 25.7 kA in steady state, and the toroidal superconducting coils are protected by three Quench Protection Circuits (QPC) assuring fast dissipation of the stored magnetic energy of about 1 GJ in case of fault. The poloidal circuits are supplied by ten ac/dc thyristor converters, almost all rated for ±20 kA and ±1 kV; ten QPC rated for the same nominal current and ±3.8 kV assure the protection of the poloidal superconducting coils. The high voltage required for the plasma breakdown is generated in the poloidal circuits by four Booster ac/dc converters and by six Switching Network Units (SNU). The Booster converters are rated +4/-14.5 kA and ±5 kV for short time, thus are inserted in the circuits only when needed and then bypassed. The SNU are operated in order to insert in the circuits settable resistors, producing up to 5 kV at the nominal current of 20 kA, and then to by-pass them after plasma initiation phase. Two in-vessel coils for fast control of plasma position are supplied by two independent ac/dc thyristor converters, rated ±5 kA and ±1 kV, and the in-vessel coils for Resistive Wall Mode control are supplied by 18 fast inverters rated for 300 A and 240 V. After a preliminary definition of the reference schemes and main requirements of the components, the procurement of the Power Supply systems has been started by means of contracts awarded to industrial suppliers including detailed design, manufacturing and test. Besides highlighting the main characteristics of Power Supply systems of JT-60SA as resulting after the detailed design, the paper describes the present status of their procurement: The QPC units have been already installed in Japan for the final acceptance test; the manufacture and factory tests of some SNU have been successfully completed; the detailed design of ac/dc converters for toroidal and poloidal circuits has been finalized, and the first units have been manufactured and are ready for factory testing. © 2015 IEEE.

This paper reports the main recent results of the RFX-mod fusion science activity. The RFX-mod device is characterized by a unique flexibility in terms of accessible magnetic configurations. Axisymmetric and helically shaped reversed-field pinch equilibria have been studied, along with tokamak plasmas in a wide range of q(a) regimes (spanning from 4 down to 1.2 values). The full range of magnetic configurations in between the two, the so-called ultra-low q ones, has been explored, with the aim of studying specific physical issues common to all equilibria, such as, for example, the density limit phenomenon. The powerful RFX-mod feedback control system has been exploited for MHD control, which allowed us to extend the range of experimental parameters, as well as to induce specific magnetic perturbations for the study of 3D effects. In particular, transport, edge and isotope effects in 3D equilibria have been investigated, along with runaway mitigations through induced magnetic perturbations. The first transitions to an improved confinement scenario in circular and D-shaped tokamak plasmas have been obtained thanks to an active modification of the edge electric field through a polarized electrode. The experiments are supported by intense modeling with 3D MHD, gyrokinetic, guiding center and transport codes. Proposed modifications to the RFX-mod device, which will enable further contributions to the solution of key issues in the roadmap to ITER and DEMO, are also briefly presented. © 2017 IAEA, Vienna.

This paper first recalls the main characteristics, design criteria, and strong points of the ac/dc conversion system of RFX-mod. Then, after having described some innovative experimental scenarios and performance improvements which have been devised, the corresponding impact on the ac/dc conversion system and the compatibility with the present design are discussed. Some possible upgrades are identified and the pros and cons are evaluated, also through dedicated simulations.

JT60SA, the superconducting tokamak under construction in Japan, will be equipped with a mix of new and reused Power Supplies (PS). Most of the new PS are procured by European Voluntary Contributors under the framework of Broader Approach agreement between F4E and JAEA. For the toroidal circuit, the 6 pulses ac/dc converter will be procured by CEA. It is rated 25.7kA and 80V dc, and will work in steady state condition. For the poloidal circuits the procurement of ten new ac/dc converters, rated ±20kA and about ±1kV is shared between CEA and ENEA. They are 24 pulses four quadrant converters, with back to back thyristor bridges. In order to allow a smooth crossing of zero current, a particular control mode with circulating current operation will be implemented. The fast control of plasma position is obtained by two in-vessel coils supplied with independent thyristor converters procured by ENEA, each one rated ± 5kA and ±1kV. Plasma initiation requires a fast variation of current in the Central Solenoids, obtained with the insertion of a settable resistor in series to the coils. This is achieved with the operation of four Switching Network Units procured by ENEA, producing up to 5kV at the nominal current of 20 kA. The protection of superconducting magnets, both toroidal and poloidal, is assured by thirteen Quench Protection Circuits procured by Consorzio RFX, rated ±20kA and ±3.8 kV for poloidal QPCs and 25.7kA and 2.8kV for toroidal ones: in case of coil quench or other fault requiring a fast de-magnetization of the coils, each QPC inserts in the circuit a dump resistor discharging the energy stored in the coils. Finally the suppression of RWM is actively performed by 18 in-vessel coils, independently supplied by a corresponding number of IGBT inverters procured by Consorzio RFX, characterized by high dynamic performances and low latency of 50 ?s, and rated for 300A and 240V. The present status of the aforementioned PS is described in the paper: their detailed design has been completed and some systems have been already manufactured and tested.

This paper firstly recalls the main characteristics, design criteria and strong points of the ac/dc conversion system of RFX-mod. Then, after having described some innovative experimental scenarios and performance improvements which have been devised, the corresponding impact on the ac/dc conversion system and the compatibility with the present design are discussed. Some possible upgrades are identified and the pros and cons are evaluated, also through dedicated simulations.

This paper reports the progress in the reference design of the Acceleration Grid Power Supply (AGPS) of the ITER Neutral Beam Injector (NBI). The design of the AGPS is very challenging, as it shall be rated to provide about 55 MW at 1 MV dc in quasi steady-state conditions; moreover, the procurement of the system is shared between the European Domestic Agency (F4E) and the Japanese Domestic Agency (JADA), resulting in additional design complication due to the need of a common definition of the interface parameters. A critical revision of the main design choices is presented also in light of the definition of some key interface parameters between the two AGPS subsystems. Moreover, the verification of the fulfillment of the requirements in any operational conditions taking into account the tolerance of the different parameters is also reported and discussed.

The JT-60SA satellite tokamak will be built in Naka, Japan. One of the main aims of this machine is to achieve steady-state high-beta plasmas. To reach this result, passive stabilizing plate (SP) and resistive wall modes (RWM) active control system based on 18 in-vessel coils will be installed. In the present design, these coils are placed on the plasma side of the SP, behind the first wall. This solution maximizes the efficiency in producing fast magnetic fields into the plasma by minimizing the shielding effect of the passive structures. Then, if the power supply (PS) and the control system have sufficient dynamic performance, it is possible to control the RWM with very low magnetic fields. This allows minimizing the Ampere-turns and the power requested to control the RWM. Conversely, the very fast dynamics required represents one of the main issues for the design of the RWM control system. This paper, after having recalled the main specification data for the RWM control system deriving from the physics studies, describes the analyses performed to complete the set of requirements necessary for the PS design. The characterization of coils and feeders is shown and the voltage necessary to produce the required current and bandwidth is quantified. Possible connections among PS and coils are analyzed in order to achieve the highest possible flexibility in controlling the RWM with a reduced set of independent PS. Finally, considerations on reasonable voltage margins to cope with load uncertainties are given.

To explore the influence of sociodemographic factors on access to appointments with physicians in primary, secondary, and tertiary health care in a publicly funded health care system.A population-based registry study.Different health care settings in Västernorrland county, Sweden.All residents in the county at the end of 2006.The number of people per 1000 residents who had at least one appointment with a physician in an average month in different health care settings.A total of 87 people had appointments with a physician in primary health care, 44 in outpatient clinics at a regional hospital, 20 in an emergency department, 14 in home care, and two in a university hospital outpatient clinic. Twelve were hospitalized at a regional hospital and1 at the university hospital. Being young or elderly, female, divorced, widowed, and having a contractor as usual source of care were all independently associated with higher odds of receiving primary care.The physician's office in primary care is the setting that has the potential to affect the largest number of people. The extent of the use of health care was independently influenced by all sociodemographic characteristics studied, which highlights the importance of individual factors in future resource allocation. Regarding availability the ecology model provides superior information as compared with the absolute number of physicians' appointments. The prerequisites in Sweden of high-quality registries and unique personal identification numbers encourage future research on the ecology model to optimize accessibility of health care.

The Acceleration Grid Power Supply of the ITER Neutral Beam Injector is designed to deliver up to 200 kV to each of the five stages of the electrostatic accelerator, producing a total voltage up to −1 MV with respect to ground. In the present design, the output voltages of the dc generators are independently controlled by five closed control loops. The output voltage measurements are obtained from the difference between the grid potential measurements, referred to local ground. Therefore, the control precision for each stage is affected by the amplification of the measurement errors. Thus, an alternative control scheme, based on a single closed control loop and regulator, has been considered. In this paper, the performance of the two control methods in terms of systematic errors, dynamic performance and sensitivity to load variations are evaluated and compared.

This paper summarizes the main achievements of the RFX fusion science program in the period between the 2008 and 2010 IAEA Fusion Energy Conferences. RFX-mod is the largest reversed field pinch in the world, equipped with a system of 192 coils for active control of MHD stability. The discovery and understanding of helical states with electron internal transport barriers and core electron temperature >1.5 keV significantly advances the perspectives of the configuration. Optimized experiments with plasma current up to 1.8 MA have been realized, confirming positive scaling. The first evidence of edge transport barriers is presented. Progress has been made also in the control of first-wall properties and of density profiles, with initial first-wall lithization experiments. Micro-turbulence mechanisms such as ion temperature gradient and micro-tearing are discussed in the framework of understanding gradient-driven transport in low magnetic chaos helical regimes. Both tearing mode and resistive wall mode active control have been optimized and experimental data have been used to benchmark numerical codes. The RFX programme also provides important results for the fusion community and in particular for tokamaks and stellarators on feedback control of MHD stability and on three-dimensional physics. On the latter topic, the result of the application of stellarator codes to describe three-dimensional reversed field pinch physics will be presented. © 2011 IAEA, Vienna.

The recent start-up improvements brought to RFX-mod have allowed increasing the plasma current in the range 1.7–1.9 MA, but still they proved to be not sufficient to routinely and reliably reach higher values of plasma current that would be very useful for the present studies on enhanced confinement. The need for higher poloidal flux variation together with the overabundant available toroidal flux has led to the study of a rearrangement of the power supply system with the aim of increasing the first at the expenses of the latter. Thanks to the flexibility of the power supply system, composed of modular thyristor converters, the rearrangement has been easily obtained and the resulting enhanced poloidal flux capability scenario described in the paper allows a two step plasma current ramp-up. In order to increase as much as possible the plasma current up to the nominal value of 2 MA it is necessary to fully exploit the thyristor converter power capability, in particular in the first ramp-up phase. In the paper a detailed verification of their maximum performance is presented. The first results obtained after power supply reconfiguration are very good, with plasma current kept constant at about 2 MA for 100 ms. The enabled operation at very high plasma current requires also an optimization of the magneto hydro dynamics (MHD) mode coil power supplies that are required to provide higher current values. Therefore a dedicated study of their control system has been worked out, which allowed understanding how to increase as much as possible the output current without losing the dynamic performance, so keeping the efficiency in the control of dominant and secondary modes, which is essential to obtain good and reproducible discharges.

The acceleration grids of the ITER Neutral Beam Injector (NBI) are likely to be subjected to frequent breakdowns. As a consequence, large voltage oscillations arise on the power supply terminals, which could stress the system and may propagate to the power supply. This paper deals with the analyses carried out to assess the impact of such voltage oscillations on the Acceleration Grid Power Supply (AGPS) of the NBI. A detailed high-frequency model of the power supply is setup and simulations are run, with the aim of estimating the voltage and current stresses at the primary side of the AGPS. The results are analyzed and discussed, giving inputs for the design and testing of the equipments.

With the exploration of the MA plasma current regime in up to 0.5 s long discharges, RFX-mod has opened new and very promising perspectives for the reversed field pinch (RFP) magnetic configuration, and has made significant progress in understanding and improving confinement and in controlling plasma stability. A big leap with respect to previous knowledge and expectations on RFP physics and performance has been made by RFX-mod since the last 2006 IAEA Fusion Energy Conference. A new self-organized helical equilibrium has been experimentally achieved (the Single Helical Axis - SHAx - state), which is the preferred state at high current. Strong core electron transport barriers characterize this regime, with electron temperature gradients comparable to those achieved in tokamaks, and by a factor of 4 improvement in confinement time with respect to the standard RFP. RFX-mod is also providing leading edge results on real-time feedback control of MHD instabilities, of general interest for the fusion community. © 2009 IAEA, Vienna.

This paper deals with the analyses carried out to assess the performance of the acceleration grid power supply of the ITER neutral beam injector; a model of the power supply is setup and simulations are run in both normal and anomalous operating conditions, with the aim of demonstrating the fulfilment of the requirements and to give inputs for the design of the equipments. The dynamic performance and the output voltage ripple are studied in normal operation on the basis of a conceptual design of the system; anomalous operating conditions, such as breakdown of the grids and sudden interruption of the beam current, are also analyzed and discussed.

This paper reports the highlights of the RFX-mod fusion science programme since the last 2010 IAEA Fusion Energy Conference. The RFX-mod fusion science programme focused on two main goals: exploring the fusion potential of the reversed field pinch (RFP) magnetic configuration and contributing to the solution of key science and technology problems in the roadmap to ITER. Active control of several plasma parameters has been a key tool in this endeavour. New upgrades on the system for active control of magnetohydrodynamic (MHD) stability are underway and will be presented in this paper. Unique among the existing fusion devices, RFX-mod has been operated both as an RFP and as a tokamak. The latter operation has allowed the exploration of edge safety factor q edge < 2 with active control of MHD stability and studies concerning basic energy and flow transport mechanisms. Strong interaction has continued with the stellarator community in particular on the physics of helical states and on three-dimensional codes. © 2013 IAEA, Vienna.

The displacements of double bonds by basic catalysts (N-lithioethylene diamine with an excess of ethylene diamine, potassium t-butylate in dimethylsulfoxide, benzyl-sodium in boiling xylene) are studied on p-menthenes and p-menthadienes. The isomerizations of p-menthadienes are accompanied by dehydrogenation. A representative mechanism of isomerization and dehydrogenation according to the catalyst is proposed for limonene.

The displacements of double bonds and dehydrogenation by basic catalysts are examined by analogy with the preceding studies on p-menthenes and p-menthadienes [1] with bicyclic terpenes: α- and β-pinenes, sabinene and 3-carene.