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To reach fusion conditions and control the plasma configuration in ITER, the next step in tokamak fusion research, two neutral beam injectors (NBIs) will supply 16.5 MW each, by neutralizing accelerated negative hydrogen or deuterium ions. The requirements of ITER NBIs (40A/1 MeV D- ions for ≤1 h, 46A/870 keV H- ions for ≤1000 s) have never been simultaneously attained. So in the Neutral Beam Test Facility (NBTF, Consorzio RFX, Italy) the operation of the full-scale ITER NBI prototype (MITICA) will be tested and optimised up to full performances, focussing on accelerator (including voltage holding), beam optics, neutralisation, residual ion removal. The NBTF includes also the full-scale prototype of the ITER NBI source with 100 keV particle energy (SPIDER), for early investigation of: negative ion production and extraction, source uniformity, negative ion current density and beam optics. This paper will describe the main results of the first two years of SPIDER operation, devoted to characterizing plasma and beam parameters, including investigation of RF-plasma coupling efficiency and magnetic filter field effectiveness in reducing co-extracted electrons. SPIDER is progressing towards the first caesium injection, which aims at increasing the negative ion density. A major shutdown, planned for 2021, to solve the issues identified during the operation and to carry out programmed modifications, will be outlined. The installation of each MITICA power supply and auxiliary system is completed; in-vessel mechanical components are under procurement by Fusion for Energy (F4E). Integration, commissioning and test of the power supplies, procured by F4E and QST, as the Japanese Domestic Agency (JADA), will be presented. In particular, 1.0MV insulating tests were carried out step-by-step and successfully completed. In 2020 integrated tests of the power supplies on the accelerator dummy load started, including the assessment of their resilience to accelerator grid breakdowns using a short-circuit device located in vacuum. The aggressive programme, to validate the NBI design at NBTF and to meet ITER schedule (requiring NBIs in operation in 2032), will be outlined. Unfortunately, in 2020 the coronavirus disease infection affected the NBTF activities. A solution to proceed with integrated power tests despite the coronavirus is presented.

An upgrade of the RFX-mod experiment is presently in the final design phase, aimed at widening the explored operational scenarios both in RFP and Tokamak configuration. The main design driver for this machine upgrade is the enhancement of the ‘shell-plasma proximity’, which is expected to provide a significant improvement in the plasma magnetic confinement. The achievement of this aim implies a major change of the internal components of the machine such as the removal of the present vacuum vessel, transferring the function of vacuum barrier to the duly modified toroidal support structure, and the integration of a new in-vessel support system to sustain the conductive stabilizing shell and the whole first wall. The paper presents an overview of the design choices and the proposed implementations, assessed on the base of engineering analyses and results of experimental tests performed on mock-ups of the new components. The solutions conceived to fulfill vacuum and electrical requirements of the in-vessel components, to guarantee their reliability during normal and abnormal operating conditions events, and interface compatibility with existing components and torus assembly sequence are finally highlighted.

An upgrade of RFX-­mod toroidal machine assembly is under development, aimed at widening the explored operational scenarios both in RFP and Tokamak configuration. The main design driver for this machine upgrade, named RFX-mod2, is the enhancement of the ‘shell-plasma proximity’, which is expected to provide a significant improvement in the plasma magnetic confinement. In order to achieve this goal, the existing Passive Stabilizing Shell will be adapted in order to be enclosed in the new vacuum vessel. This reconfiguration of the internal components of the experiment requires the design of a new structure aimed at supporting both the shell and the plasma facing components. The paper describes the design aspects of the shell assembly, with particular focus on the thermo-mechanical FEM analyses performed to verify the design of the components at the expected operating conditions.

The results produced by experimental campaigns of RFX-mod shows the need to improve passive MHD control and minimise braking torque on plasma. These improvements require major mechanical changes on the present components of the machine including first wall, vacuum vessel, and toroidal support structure (TSS). The vacuum vessel will be removed and the first wall tiles will be directly supported by the passive stabilising shell, so increasing the poloidal cross section by 28 mm radially and bringing the passive stabilising shell as close as possible to the plasma boundary. This paper presents the mechanical design modifications of the torus assembly. Composite rings are designed to support the passive stabilising shell on the TSS. The vacuum boundary is shifted at the TSS by developing different joint solutions compatible with the stringent requirements of the present components: ceramic-metal brazed rings at the two poloidal joints, fully welded solution at the external equatorial joint, and resistive weld plate at the internal equatorial joint. The vacuum vessel portholes for pumping, fuelling, diagnostics, and sensor cable routing are redesigned to be integrated onto the TSS. The design is supported by thermo-mechanical analyses and verifications carried out applying load combinations corresponding to the expected load cases.

RFX-mod2 is the last upgrade of the reversed field pinch machine operated at Consorzio RFX. A significant modification consists of the replacement of the first wall tiles, proposed as a key factor for the improvement of the gas density with the reduction of the hydrogen retention, and designed in coherence with the magnetic front-end modification that foresees the tiles supported by the existing MHD passive stabilising shell. The main choices in the design of the new first wall tiles are the polycrystalline graphite as bulk material, the use of the existing fixing keys with fastening bayonets, and the tile width that shall be less than the diameter of the larger port holes to allow remote handling operations for the maintenance of the first wall; the latter fix the number of tiles to 2016 as in the original configuration. The expected decrease of the plasma-wall interaction determined the first wall design with expected power densities up to 50 MW/m2 considering the deformation of the last magnetic surface in both reversed field pinch and tokamak configurations. At the other side, the need of sensor integration and shielding of the passive stabilising shell from the plasma addressed the tile thickness and surface extension. Local prominences have been modelled on tiles, based on visual inspections of actual surfaces after previous operations, in order to limit the plasma in regions far from openings and supporting structures. The tile resistant section has been increased coherently with all the interfaces and constraints, so decreasing the maximum stress at 3.5 MPa calculated from finite element analysis that simulates the operating condition. This low stress level together with a measurement of the experimental loads during next RFX-mod2 operations could qualify the use of extruded graphite for a further first wall change in the future. Indeed, extruded graphite is considered attractive given its high directional thermal diffusivity (about 50% better then polycrystalline graphite) to enhance the heat transmission and so improving the gas density control, and the low stress induced may allow this mechanically less performing grade of graphite.

The poloidal distribution of electromagnetic loads during fast transients in the vacuum vessel of COMPASS-U tokamak is calculated analytically. The estimates are then compared with CarMa0NL numerical simulations. The results show that the poloidal eddy currents in the vacuum vessel must be taken into account for the proper evaluation of disruption forces in the COMPASS-U tokamak.

The design of RFX-mod2 experiment implies a major change and reconfiguration of the internal components of the present machine assembly. Critical aspects of the design regard the technical solutions needed to fulfil geometrical, vacuum and electrical requirements imposed by interface with existing components and by specific operating conditions. The paper presents an overview of the design choices and the proposed implementations, assessed by means of engineering analyses and experimental tests performed on mock-ups of the new components.

A major modification of the RFX-mod toroidal load assembly has been decided in order to improve passive MHD control and to minimise the braking torque on the plasma, thus extending the operational space in both RFP and Tokamak configurations. With the removal of the vacuum vessel, the support structure will be modified in order to obtain a new vacuum-tight chamber and the first wall tiles will be directly in front of the passive stabilising shell inside of it, so increasing both the poloidal cross section (by 28 mm radially) and the plasma-shell proximity. This implies the design of a new vacuum fit electromagnetic measurement system. The spatial resolution is constrained by the number of graphite tiles, 28 (poloidal) x 72 (toroidal), which cover the inner surface of the copper shell. The new local probes (of size 42 x 37 x 7 mm) will be installed in vacuum onto the copper shell, behind the graphite tiles, and shall operate up to a maximum temperature of 180°C to allow for baking cycles for first wall conditioning. Because of the reduced room available, triaxial pickup probes have been designed, with the additional advantage of allowing the minimization of alignment errors. The paper describes the detailed design of the new probe set, in particular highlighting the technological issues complying with different functional requirements: the thermo-mechanical interface with the copper shell, their protection, the electrical connection for the transmission of the signal with a minimum level of noise and the choice of materials suitable to the vacuum environment at relatively high temperature. The extensive tests carried out on probe prototypes to characterise their electromagnetic, thermal and vacuum behaviour are also reported.

The externally applied loop voltage, required to drive current carrying magnetic configurations, such as tokamaks or reversed field pinches (RFP), can induce currents between the plasma and the surrounding conducting components, such as limiters, divertor sections, stabilizing plates or enclosing stabilizing shells with insulated electrical gaps. An appropriate modeling of this problem can quantify these currents and give constraints for the design of new devices. In this paper a methodology based on lumped element circuit is used to evaluate this problem for the case of in vessel components of the RFX-mod2, which led to a redesign of the electrical connections of plasma facing components. In the new design some parts, having applied potential in the kV range, are still exposed to the weakly ionized scrape off plasma. The process of arc formation between them and their prevention had been experimentally investigated and possible arc suppression strategies identified.

The measurement of the plasma current in ITER plays an outstanding role as it is part of the machine protection and is a safety-relevant measurement: it will be used in relation with regulatory limits to show that the operation remains within the safe envelope defined in the ITER license. The Continuous External Rogowski (CER) is an inductive sensor designed for current measurements and located in the casing of 3 Toroidal Field Coils (TFCs). After the completion of the design of the CER, 4 prototypes of the sensor were manufactured and R&D activities were performed under a Grant with the European Domestic Agency (F4E-GRT-012). The work was carried out between 2010 and 2011 by the ITERMAG consortium comprising 3 laboratories: CRPP (Switzerland) as leader, CEA (France) and RFX (Italy). The R&D campaign on CER prototypes consisted in the measurement of about 100 parameters to characterize the CER in terms of electrical, thermal, mechanical and also of vacuum compatibility. From these results, electromagnetic modeling of the CER response was performed. It is demonstrated that the CER fulfills ITER requirements. However, the vacuum compatibility of the prototype has to be improved and solutions to cope with this issue are proposed. (C) 2013 Elsevier B.V. All rights reserved.

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.

This paper describes the final design of the Short-Time Retractable Instrumented Kalorimeter Experiment (STRIKE) for the SPIDER experiment (Source for Production of Ions of Deuterium Extracted from Radio frequency plasma) under construction at the Consorzio RFX premises. The STRIKE diagnostic will be used to characterise the SPIDER beam during short pulse operation (several seconds) to verify the degree of attainment of ITER requirements about the maximum allowed beam non-uniformity. After a preliminary design developed in the last few years, the complete STRIKE diagnostic system has been recently subjected to a final review and is now ready for construction. The main components of the system are: an array of 16 tiles made of Carbon Fibre Composite, which will be exposed to the high power density beam (of the order of 20 MW/m22) exiting from the SPIDER Beam Source; a supporting and positioning system, based on a set of structures made of stainless steel, with controlled moving systems, to be installed within the SPIDER vacuum vessel; a set of thermal, electric and thermo-graphic sensors to properly detect the operating conditions. The paper will focus in particular on the development of the engineering design of the supporting and positioning system, with a description of the relevant CAD and FEM analyses, and will give an overview of the complete system, with reference to the manufacture of CFC tile prototypes and to the choice of the complete set of sensors.

A major refurbishment of the toroidal load assembly is under study to extend the operational space of RFX-mod in both RFP and Tokamak configurations. It includes the removal of the vacuum vessel to increase the plasma-shell proximity and modifications of the support structure to obtain a new vacuum-tight chamber. This entails the design of a new electromagnetic measurement system, taking into account the requirements of both the equilibrium and MHD control system and physics analyses requirements in both RFP and shaped Tokamak configurations. The spatial resolution is constrained by the number of graphite tiles, 28 (poloidal) x 72 (toroidal), which cover the inner surface of the toroidal vacuum chamber. Alternative layouts of 4 x 72 and 7 x 72 are being considered to assure the calculation of the MHD mode spectrum up to (m = 0,1,2 or 3; n = 0,…,±35). A substantial advantage in terms of harmonic spectrum reconstruction is expected also in case of sensor faults with respect to the present 4x48 arrangement. Moreover, 12 poloidal arrays will include 14 sensors to compute the harmonic content needed by the plasma boundary reconstruction algorithm (m = 0,…,6). Due to the reduced room between the copper shell and the graphite tiles, the design of triaxial, vacuum fit pickup probes have started also for their convenience in compensating alignment errors.

The system of diagnostics, data acquisition and control foreseen on the Divertor Test Tokamak experiment (DTT) is presented. Conceived in an integrated way, the control system meets the specifications of a fusion experiment devoted to the study of the power exhaust problem in view of DEMO. Diagnostics and feedback control are particularly functional to the need of maintaining the plasma close to equilibrium in situations prone to instabilities where the plasma wall interaction is optimized. Strongly oriented to the exploration of control methods suitable for DEMO, DTT will specifically experiment on physics and engineering model based control systems. Control and data flow schemes are inspired by those of ITER.

After 10 years of operation since its major modification, an upgrade of the RFX-mod experiment is presently under design. The scientific objective is the improvement of 3D physics studies through a more robust transition to higher confinement regimes in both Reversed Field Pinch (RFP) and Tokamak configuration obtained thanks to an advanced system for the active control of MHD instabilities. The main design driver requirements for this machine upgrade are the removal of the present resistive vacuum vessel and the enhancement of the ‘shell-plasma proximity’, to reduce the deformation of the last close magnetic surface and to improve the self-organized helical plasma regimes. The fulfillment of these requirements implies a major change of the internal components of the machine such as the replacement of the whole first wall, the change of the support system of the stabilizing shell and the modification of the present toroidal support structure to provide the function of vacuum barrier. In combination, other components of the machine will be upgraded, such as magnets and power supply, diagnostic systems and a NBI will be integrated. The paper presents an overview of the engineering design of the new components and highlights the critical aspects of the new torus assembly.

This paper deals with an R&D activity in progress in support of the design of the ITER NBI prototype (MITICA), aimed at the identification of the most appropriate design solutions and installation guidelines to keep under control and minimize the effects of ionizing radiation on the electronic and passive components related to control, diagnostics, and electrical insulation located inside the MITICA vessel and its concrete biological shield. The method followed to approach the problem is based on three steps: estimation of expected radiation levels map, identification of electronic and passive components database for MITICA diagnostics and finally the assessment of the ionizing radiation compatibility for the identified components. The ultimate aim of this activity is the definition, on the basis of this assessment, of a list of preventive or mitigating actions (such as relocation, shielding or preventive maintenance) for the most high-risk components.

This paper presents the results of a feasibility study aimed at the design of a mechanical support and electrical connector for in-vessel magnetic sensors compatible with remote handling maintenance. The work starts with the review of the reference design proposed in the ITER Design Description Documents and highlights the issues still open. A new design proposal is then developed taking into account the experience of similar applications developed for JET. The design proposal is supported by CAD modelling and thermal and electromagnetic analyses. In conclusion the developments necessary to qualify the proposal before considering it mature for production are outlined.

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 failure mode and effects analysis (FMEA) is a widely used analytical technique that helps in identifying and reducing the risks of failure in a system, component or process. The application of a systematic method like the FMEA was deemed necessary and adequate to support the design process of the ITER NBI (neutral beam injector). The approach adopted was to develop a FMEA at a general “system level”, focusing the study on the main functions of the system and ensuring that all the interfaces and interactions are covered among the various subsystems. The FMEA was extended to the whole NBI system taking into account the present design status. The FMEA procedure will be then applied to the detailed design phase at the component level, in particular to identify (or define) the ITER Class of Risk. Several important failure modes were evidenced, and estimates of subsystems and components reliability are now available. FMEA procedure resulted essential to identify and confirm the diagnostic systems required for protection and control, and the outcome of this analysis will represent the baseline document for the design of the NBI and NBTF integrated protection system. In the paper, rationale and background of the FMEA for ITER NBI are presented, methods employed are described and most interesting results are reported and discussed.

This paper presents the preliminary design of the "In-Vessel Magnetic platform", which is a subsystem of the magnetic diagnostics formed by all the components necessary for guaranteeing the thermomechanical interface of the actual magnetic sensors with the vacuum vessel (VV), their protection and the electrical connection to the in-vessel wiring for the transmission of the detected signal with a minimum level of noise. The design has been developed in order to comply with different functional requirements: the mechanical attachment to the VV; the electrical connection to the in-vessel wiring; efficient heat transfer to the VV; the compatibility with Remote Handling (RH) system for replacement; the integration of metrology features for post-installation control; the Electro Magnetic Interference (EMI) shielding from Electron Cyclotron Heating (ECH) stray radiation without compromising the sensor pass band (15 kHz). Significant effort has been dedicated to develop the CAD model, integrated within the ITER In-Vessel configuration model, taking care of the geometrical compliance with the Blanket modules (modified in order to accommodate the magnetic sensors in suitable grooves) and the RH compatibility. Thorough thermo-mechanical and electro-magnetic Finite Element Method (FEM) analyses have been performed to assess the reliability of the system in standard and off-normal operating conditions for the low frequency magnetic sensors. (C) 2016 Consorzio RFX. Published by Elsevier B.V. All rights reserved.

A new system of magnetic probes was installed during the 2005 shutdown and was commissioned during the 2005/2006 restart phase of JET. The system is made of traditional pick-up coils and Hall sensors, all located outside the vessel. Direct field measurements from Hall sensors are used to correct the drift of the integrators of the pick-up coils signals. This feature will be crucial for future ITER-like devices, where long-lasting flat top phases are expected, in a high neutron yield and a high temperature environment. After a general overview of the system, the paper describes the major manufacturing and installation issues, including the construction of the supports and probes as well as the acceptance tests before and after installation. The functional commissioning of the system includes the integration of the new signals in the JET Data Acquisition System and the analysis of several discharges with and without plasma. The critical aspects of the assessment of the reliability of the signals are shown and commented on.

Disruptions represent one of the main concerns for Tokamak operation, especially in view of fusion reactors, or experimental test reactors, due to the electro-mechanical loads induced by halo and eddy currents. The development of a predictive tool which allows to estimate the magnitude and spatial distribution of the halo current forces is of paramount importance in order to ensure robust vessel and in-vessel component design. With this aim, two numerical codes (CARIDDI, CAFE) have been developed, which allow to calculate the halo current path (resistive distribution) in the passive structures surrounding the plasma. The former is based on an integral formulation for the eddy currents problem particularized to the static case; the latter implements a pair of 3D FEM complementary formulations for the solution of the steady-state current conduction problem. A simplified plasma model is adopted to provide the inputs (halo current injected into the first wall). Two representative test cases (ITER symmetric and asymmetric VDEs) have been selected to cross check the results of the proposed approaches.

Experimental results and theoretical studies call for Reversed Field Experiment (RFX) machine and power supply improvements to allow studies that go beyond those of a conventional Reversed Field Pinch (RFP) with passively stabilized turbulent MHD dynamo. The new paths opened by recent results in RFX and other RFP machines are introduced; then the goals and the design lines of the technical modifications of RFX, mainly addressed to improve the first wall, the plasma magnetic boundaries and to increase the operational flexibility of the toroidal field circuit power supply, are reported.

This paper deals with the electromagnetic analyses carried out for the design of a new system developed to actively control the magnetic configuration of RFX. The system, which consists of a new conductive shell and a new set of saddle coils closely fitting the plasma region, is intended for replacing the present shell, keeping the existing vacuum vessel and principal coils of the machine unchanged. In the first part of the paper, the basic concepts representing the aims and the constraints of the system are described. Following this, the results of the electromagnetic analyses over the operating frequency range, performed by means of both analytical and finite element codes, are reported and discussed. The efficacy of the system is pointed out with the evaluation of the magnetic field generated by the new saddle coils and the reduction of field errors provided by the new shell with an overlapped gap. Finally, the calculation of the electrical parameters of the coils, taking into account the eddy current effect, is reported.

This paper summarizes the progress of R&D activities related to the design of in-vessel magnetic sensors for measurement of the equilibrium field in ITER and the associated mechanical support and connection system. The background to the design activities, developed in the last few years in the framework of several collaborations, is provided in the introduction. The paper focuses on the experimental results obtained from a set of sensor prototypes, recently manufactured and tested, and on numerical simulations performed in order to verify the compliance of the sensors with ITER requirements. The paper finally illustrates the status of the engineering analyses performed to progress the design of the support and connection system, conceived to be replaceable by remote handling. In conclusion, specific issues which require further developments to achieve the final design are outlined.

The ITER magnetic equilibrium sensors will be installed in a hostile environment characterized by severe neutron radiation and heat loads, which can cause undesired spurious signals and compromise the accuracy of the measurements obtained by time-integration. The paper deals with the construction and tests of prototype in-vessel pick-up coils, which have been developed aiming at the reduction of the adverse effects of neutron radiation by using suitable materials and by minimizing the thermal gradients. A first solution based on a winding made of a flexible ceramic-coated conductor (instead of a Mineral Insulated Cables) and “impregnated” with ceramic filler has been implemented and tested. A second solution based on coils constituted by a stack of circuits printed on a ceramic substrate sintered in one piece has also been developed and tested with positive results. The paper describes the design developments, the rationale for the test and discusses the results and the developments necessary for the production of reliable equilibrium sensors for ITER.

The current status of the ITER magnetics diagnostic is summarised. The various risks are discussed on the basis of when they could be present during ITER operation and appropriate mitigation is reviewed. An implementation risk or common mode operation failure risk to the magnetic diagnostic can only be mitigated by diversity. All identified risks are being handled with ongoing research and development to minimise these risks. A reasonable level of design and manufacturing diversity has been therefore introduced in the procurement and installation plans, in order to optimise overall reliability avoiding dispersion of resources. (C) 2009 Elsevier B.V. All rights reserved.

The ITER magnetic diagnostics must provide essential information for plasma diagnostic purposes and for the machine control. The sensors will be installed in a hostile environment characterized by severe neutron irradiation and heat loads, which can reduce the sensor lifetime and also generate undesired DC signals, which might compromise the accuracy of the measurements obtained by time-integration. The paper is focused on the design development and optimisation of the in-vessel tangential pick-up coils and blanket halo sensors. The work is aimed at achieving the required measurement precision in spite of the thermal effect due to the neutron irradiation. Within the limits of several tight engineering constraints, a new concept of magnetic pick-up coil has been developed. A winding made of a ceramic-coated conductor (instead of a mineral insulated cables) and “impregnated” with ceramic filler is proposed. Similar issues are also related to the halo sensor design.

This paper is focused on the assessment of the thermal behaviour in nominal operating conditions of two components of the ITER-NBI ion source: the arc-discharge chamber and the filament cassette assembly. The study is based on hydraulic, thermo-mechanical and thermo-electrical calculations performed by means of 2D and 3D finite element (FE) models, with inputs coming partly from the ITER reference design documentation and partly from the design review activities presently in progress. The paper describes the results of the analyses, comparing the alternative cooling design with the reference one, and highlights some open issues that should be worked out before finalizing the design.

This paper presents the status of the design and manufacture of the main in-vessel subsystems of the magnetic diagnostics enhancement project for JET. The subsystems are conceived as several sets of pick-up coils supported by mechanical structures attached to existing components of the first wall, to be installed by remote handling. The design constraints and fundamental choices at the base of the project are illustrated. The technological solutions proposed for the construction, presently in progress, are described.

Recent studies on RFP and Tokamak devices call for an active control of the MHD and resistive wall modes to induce plasma mode rotation and to prevent mode phase locking. The results obtained on RFX, where slow rotation of phase locked modes has been induced, support the possibility of extending active MHD mode control through a substantial modification of the device. A new first wall with an integrated system of electric and magnetic transducers has been realised. A close fitting 3 mm thick Cu shell replaces the 65 mm Al shell. A toroidal support structure (TSS) made of stainless steel replaces the shell in supporting all the forces acting on the torus. A system of 192 saddle coils is provided to actively control the MHD modes. This system completely surrounds the toroidal surface and allows the generation of harmonic fields with m=0 and m=1 poloidal wave number and with a toroidal spectrum up to n=24.

The paper presents the various aspects of the new set of 192 saddle coils (SC), designed for the active control of MHD plasma modes in the RFX experiment. The choices necessary to fulfil the electromagnetic specifications and the geometrical constraints of the system are described. The methods and results of thermal, electrodynamic and mechanical analyses carried out for the proper design of the coils are also reported. The technological solutions proposed for the construction, presently in progress, are illustrated.

This paper deals with the upgrade and the revision of the RFX fast protection system, which follow the recent modifications to the power supply circuits and the installation of a new copper shell and of 192 saddle coils for MHD mode control. The experimental evidences and the new requirements due to the machine and circuit modifications, which called for the revision of the system, are described, together with its new architecture.

To reach fusion conditions and control plasma configuration in ITER, a suitable combination of additional heating and current drive systems is necessary. Among them, two Neutral Beam Injectors (NBI) will provide 33 MW hydrogen/deuterium particles electrostatically accelerated to 1 MeV; efficient gas-cell neutralisation at such beam energy requires negative ions, obtained by caesium-catalysed surface conversion of atoms inside the ion source. As ITER NBI requirements have never been simultaneously attained, a Neutral Beam Test Facility (NBTF) was set up at Consorzio RFX (Italy), including two experiments. MITICA is the full-scale NBI prototype with 1 MeV particle energy. SPIDER, with 100 keV particle energy, aims at testing and optimising the full-scale ion source: extracted beam uniformity, negative ion current density (for one hour) and beam optics (beam divergence

In ITER, an accurate measurement of plasma current, with high reliability, is mandatory as this parameter is used to demonstrate licensing compliance with regulatory limits. For that purpose, several independent measurements based on magnetic diagnostics have been proposed. Rogowski coils are standard inductive sensors for current measurement in many applications. In ITER, three continuous external Rogowski coils are to be installed in the casing of the toroidal field coils. These sensors are remarkable from several points of view: overall length is about 40 m, high sensitivity needed, located in the toroidal field coil casing at 4.5 K and complex 3D routing with tight bending radius of 50 mm. Since 2005 an extensive work has been carried out to develop and analyze several design options complying with ITER specifications. Prototypes of a selected continuous external Rogowski design were built and tested successfully in terms of electrical, thermal, mechanical and vacuum characteristics. Finally a detailed assembly procedure inside the toroidal field coil casing has been defined according to the coil manufacturing and assembly constraints. (c) 2015 Elsevier B.V. All rights reserved.