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The ITER Heating Neutral Beam injector (HNB) will be equipped with a beam source that will provide a negative beam of 40A (H or D). The R&D activities undertaken in Europe to pursue this challenging goal comprise three experiments: -ELISE – the half size ion source experiment operating in IPP Garching -SPIDER – the full size ion source experiment at Neutral Beam Test Facility (NBTF) site in Padua -MITICA - the full size full energy ITER injector prototype being also established at the NBTF The procurement of the beam source for SPIDER started in October 2012, when a contract between Fusion for Energy and Thales, Microwave & Imaging Sub-Systems (as group leader of a Consortium created with CECOM Srl, Galvano-T GmbH and Zanon SpA) was signed. The scope of the contract has been completed in October 2017 with the delivery of the SPIDER beam source at the NBTF site in Padova Italy. A review of the fabrication of the main parts and of the aspects of their assembly is reported in the paper, with a focus on some of the major issue encountered and the achievement of tight alignment tolerances of the accelerator.

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.

The heating neutral beam (HNB) systems at ITER are designed to inject a total of 33 MW of either 1 MeV D0 or 870 keV H0 beams into the ITER plasma using two injectors with a possible addition of a third injector later to increase the injected power to ∼50 MW. The injectors become radioactive due to the neutron flux from ITER and, in order to avoid the resulting complex remote maintenance, the design, choice of materials and the manufacturing process of each component of the injector is, wherever possible, such that they survive the life time of ITER. To ensure a smooth operational phase of neutral beams at ITER a neutral beam test facility (NBTF) is under construction at Consorzio RFX, Padova, (hereinafter referred to as RFX), which consists of 2 test beds, the 100 kV “SPIDER”, and a 1 MV “MITICA” facilities, which will be used to optimize the source operation for H and D beams. MITICA is essentially a full scale ITER prototype injector for the ITER beam parameters. The manufacturing and operation of the facility will allow validation of the operational space of the injectors and provide valuable information about the manufacturing processes applicable to HNB components. Operation of the two facilities is expected to begin in 2016 and 2019 respectively. Currently experiments on the ELISE facility with a half ITER sized RF beam source are underway. ITER relevant parameters for the H beams have almost been achieved. Efforts are underway to optimise the same with D beams. The experimental database from ELISE will be an important input for establishing the ITER relevant parameter space on the SPIDER source. This paper discusses the present status of the design and development of the injectors for ITER and the progress on the test facilities.

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

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.

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

The design of the beamline components of MITICA, the full prototype of the ITER heating neutral beam injectors, is almost finalised and technical specifications for the procurement are under preparation. These components are the gas neutraliser, the electrostatic residual ion dump, and the calorimeter. Electron dump panels are foreseen each side of the upstream end of the neutraliser to protect the cryo-panels from electrons, created by stripping and other processes, that exit the 1 MeV accelerator. As the design of the components must fulfil requirements on the beam physics, insight on physical processes is required to identify performance trade-offs and constraints. The spatial gas distribution was simulated to verify the pumping requirements with electron dump panels and local conditions for breakdown voltage. Electrostatic analyses were carried out for the insulating elements of the RID to verify the limits of the electric field intensity. Different criteria were approached to investigate the fracture behaviour of ceramics considering the manufacturing implications and extrapolating the conditions for proof testing. Severe heating conditions will be applied steadily, as the maximum pulse duration is 1 h, and cyclically so requiring to fulfil fatigue and ratcheting verifications. High heat fluxes, up to 13 MW/m 2 on the calorimeter, with enhanced heat transfer in subcooled boiling conditions will occur in the actively cooled CuCr1Zr panel elements provided with twisted tapes as turbulence promoters. Special R&D activities were undertaken to support the design: manufacturing of thick twisted tapes leading to an increased cooling performance while maintaining flow rate requirements, bending of swirl tubes, verification for permanent deformations due to stress relaxation after heating of swirl tubes, double side deep drilling of 2 m long CuCr1Zr plates.

The ITER baseline foresees 2 Heating Neutral Beams (HNB’s) operating at 1 MeV 40 A D0, each capable of delivering 16.5 MW of deuterium neutrals to the plasma, with a 3rd HNB injector foreseen as an upgrade option that would bring up the total neutral beam power to 50MW [1]. In addition a dedicated Diagnostic Neutral Beam (DNB) will be injecting 100 keV 60 A of H0 for charge exchange recombination spectroscopy (CXRS) [2]. This paper highlights the impact of the staged approach on the ITER NBI, presents the new installation strategy and reports on the status of the adopted design refinements and the integration of the HNB’s and the DNB into the ITER plant.The ITER baseline foresees 2 Heating Neutral Beams (HNB’s) operating at 1 MeV 40 A D0, each capable of delivering 16.5 MW of deuterium neutrals to the plasma, with a 3rd HNB injector foreseen as an upgrade option that would bring up the total neutral beam power to 50MW [1]. In addition a dedicated Diagnostic Neutral Beam (DNB) will be injecting 100 keV 60 A of H0 for charge exchange recombination spectroscopy (CXRS) [2]. This paper highlights the impact of the staged approach on the ITER NBI, presents the new installation strategy and reports on the status of the adopted design refinements and the integration of the HNB’s and the DNB into the ITER plant.

The neutral beam (NB) system for ITER is composed of two heating neutral beam injectors (HNBs) and a diagnostic neutral beam injector (DNB). A third HNB can be installed as a future up-grade. This paper will present the design development of the components between the injectors and the tokamak; the so-called ‘front end components’: the drift duct consists of the NB bellows and the drift duct liner, the vacuum vessel pressure suppression system box (VVPSS box), the absolute valve, and the fast shutter. These components represent the key links between the ITER tokamak and the vessels of the NB injectors. The design of these components is demanding due to the different loads that these components will have to stand. The paper will describe the different design solutions which have to be implemented regarding the primary vacuum confinement, the power handling capability and the remote maintenance operations. The sizes of the components are determined by the large cross section of the neutral beam. The power handling capability is driven by the anticipated re-ionization of the neutral beam and the electromagnetic fields in this region. The drift duct bellows (with an inner diameter of 2.5 m) shall guarantee a leak tight vacuum enclosure during the vertical and radial displacements of the ITER vacuum vessel. The conductance of the VVPSS box must be maximized in the available space. The absolute valve remains a challenging development. The total leak rate through the valve must be ≤1 × 10−8 Pa m3/s when the valve is closed. Due to the radiation environment, the seals of the gate valve will be metallic. An R&D program has been launched to develop a suitable metallic seal solution with the required dimensions. The maximum allowed closing time for the fast shutter shall be less than 1 s. For all these components the leak tightness will be guaranteed by a welded lip seal and the mechanical stability by bolted structures.

The 100 kV negative hydrogen ion source based Diagnostic Neutral Beam (DNB) injector is part of the Indian procurement for ITER. The DNB is designed to deliver an 18–20 A hydrogen neutral beam to the ITER plasma. The exit scraper (ES) defines the profile of the beam after it leaves the calorimeter, the last Beam Line Components (BLCs) during operation. BTR 1 and PDP 1 codes are used to obtain the optimum entry and exit opening dimensions of the ES, so as to deliver the maximum power to the tokamak plasma, and to limit the power to downstream components due to beam interception, to protect front end components from the interception of re-ionized devious particles. Each horizontal heat transfer (HT) panel of ES receives a total power of 85 kW for each symmetric 10 mrad beam (worst case scenario) with peak power density of 1.17 MW/m 2 . The thermo hydraulic design is carried out to withstand the heat flux due to the beam interception on the ES. The mechanical design of the ES is carried out by considering the spatial constraints in the DNB system and the remote handling (RH) system. In the present design, the copper heat transfer panels are directly bolted to the structural member. Protection shall be incorporated to ensure that there is no exposure of the bolt head. The thermo-mechanical analysis has been carried out for the normal, off-normal (accidental/worst-case) events of the ITER DNB. During normal operation, the maximum possible von Mises stress of the order of 62.4 MPa can be expected on the bottom heat transfer panel, and the corresponding deflection of the structure is of the order of 0.62 mm. All the stress checks were carried out with respect to the Structural Design Criteria for In-vessel Components (SDC-IC) criterion. The paper shall present the ES design highlighting its salient features.

The HV bushing has a function of a bulk head between insulation gas (SF6) of 0.6 MPa and vacuum, which is needed to insulate −1 MV in vacuum in ITER heating neutral beam (NB) injection system. Mechanical analyses were carried out to verify the mechanical integrity of the HV bushing composed of five-stage double-layered insulator columns with large brazed ceramic ring and fiber reinforced plastic (FRP) ring of which non-metallic material is required to satisfy a safety factor of ≥3.5. As for the FRP ring required sustaining the pressure load, seismic load and dead weight, it was confirmed that isotropic fiber cloth FRP rings having sufficient strength against shear stress should be used. The shape and fixation area of the Kovar sleeve were modified to lower the stress at the brazing area having vacuum leak tightness and pressure tightness against the air filled at 0.6 MPa. As a result, a design of the insulator for the HV bushing was established satisfying the requirement.

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.

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

The NB ITER Remote Maintenance System (NB IRMS) provides the means for the remote maintenance within the NB Cell by removal and replacement of the plant equipment. The NB IRMS will be installed and removed with the assistance of human workers during the preparation, and post-operation phase. During the maintenance operation after opening the Passive Magnetic Shield (PMS) and vessels, the maintenance activity and recovery from failure should be conducted remotely. This paper describes the concept design of the NB IRMS operating inside the NB cell for maintenance of the plant equipment such as NB components, and Upper Port Plugs (UPP). The main tasks of the IRMS, the description of the sub-systems and their specification, and deployment/operation principles are presented. The transportation concept of the NB IRMS to the hot cell facility for storage and maintenance is presented, which is to avoid unnecessary exposure on the equipment inside the NB cell during the machine operation.

SPIDER is one of the two projects of the ITER Neutral Beam Test Facility (NBTF) under construction in Padova, Italy, at Consorzio RFX premises; it is the full-size prototype of the radiofrequency (RF) ion source for ITER Neutral Beam Injector (NBI), with beam energy of 100keV, designed to operate with pulse length up to 3600 s, featuring ITER-like filter field configuration, caesium oven layout and a wide set of diagnostics. These features will allow reproducing the ion source operation like in ITER as 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 to prove the possibility of achieving the ITER NBI ion source target 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 workout of operation and research plan, based on a staged strategy.

The MITICA experiment (Megavolt ITER Injector & Concept Advancement) is the prototype and the test bed of the Heating and Current Drive Neutral Beam Injectors (HNB), which will be necessary for the full-performance exploitation of ITER. MITICA injector experiments shall demonstrate the reliable and accurate emission of a 17 MW beam of neutral particles of deuterium or hydrogen for duration up to 1 hour, fulfilling ITER specific requirements. MITICA test bed is in the final design phase and will be procured and installed in PRIMA facility (Padova Research on Injector Megavolt Accelerated) in Padova, Italy. The beam source is the key component of the system, as its goal is the generation of the 1 MeV accelerated beam of deuterium or hydrogen negative ions. The beam source is a complex system, having overall dimensions in the order of 4x4x3 m3 and a weight of around 20 tons. A major effort has been put in place in the recent months for the design closure, in collaboration with ITER Organization, F4E and European and Japanese laboratories (IPP, CEA, JAEA), solving the remaining outstanding issues, namely the optimization of the thermo-mechanical behavior of the accelerator grid segments subjected to a substantial heat power deposition, the finalization of the mechanical structure of the accelerator coherently identifying a reliable assembly and alignment procedure, the completion of the design of all involved interfaces with surrounding components and maintenance tools, and the conclusion of last R&D campaign on large ceramic insulators. Comprehensive analyses have also been carried out to finally verify the structural and electrostatic design of the overall structure with regard to normal operation, main fault conditions and seismic events. This paper presents the highlights of the latest developments for the finalization of the MITICA beam source design, together with a description of the most recent analyses and R&D activities carried out in support of the design.

The heating and current drive beam lines (HNB) at ITER are expected to deliver ∼16.7 MW power per beam line for H beams at 870 keV and D beams at 1 MeV during the H-He and the DD/DT phases of ITER operation respectively. On the other hand the diagnostic neutral beam (DNB) line shall deliver ∼2 MW power for H beams at 100 keV during both the phases. The path lengths over which the beams from the HNB and DNB beam lines need to be transported are 25.6 m and 20.7 m respectively. The transport of the beams over these path lengths results in beam losses, mainly by the direct interception of the beam with the beam line components and reionisation. The lost power is deposited on the surfaces of the various components of the beam line. In order to ensure the survival of these components over the operational life time of ITER, it is important to determine to the best possible extent the operational power loads and power densities on the various surfaces which are impacted by the beam in one way or the other during ...

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.

In ITER, each heating neutral beam injector (HNB) will deliver about 16.5 MW heating power by accelerating a 40 A deuterium negative ion beam up to the energy of 1 MeV. The ions are generated inside a caesiated negative ion source, where the injected H2/D2 is ionized by a radio frequency electromagnetic field. The SPIDER test bed, currently being manufactured, is going to be the ion source test facility for the full size ion source of the HNBs and of the diagnostic neutral beam injector of ITER. The SPIDER beam source comprises an ion source with 8 radio-frequency drivers and a three-grid system, providing an overall acceleration up to energies of about 100 keV [1]. SPIDER represents a substantial step forward between the half ITER size ion source, which is currently being tested at the ELISE test bed in IPP-Garching, and the negative ion sources to be used on ITER, in terms of layout, dimensions and operating parameters. The SPIDER beam source will be housed inside a vacuum vessel which will be equipped with a beam dump and a graphite diagnostic calorimeter. The manufacturing design of the main parts of the SPIDER beam source has been completed and many of the tests on the prototypes have been successfully passed. The most complex parts, from the manufacturing point of view, of the ion source and the accelerator, developed by galvanic deposition of copper are being manufactured. The manufacturing phase will be completed within 2015, when the assembly of the device will start at the PRIMA site, in Padova (I). The paper describes the status of the procurement, the adaptations operated on the design of the beam source for the fabrication, with particular emphasis to the engineering development to enable the fulfillment of the tight requirements set in the technical specifications. Moreover the tests performed on the prototypes are reported.

The first boundary feedthroughs of the ITER heating neutral beam (HNB) injectors are key components of the neutral beam system. They allow the penetration of signals and power from/to the beam line components (BLC) and front end components (FEC), through the confinement boundary of ITER. The design status of the two types of feedthroughs is presented in the first part. This includes the design requirements, cabling strategy, the analysis done and the maintenance strategies. The second part deals with the qualification of the feedthroughs. Being made of brittle material and not covered by mechanical standard a qualification by testing strategy is proposed based on existing rules and norms. The design of the feedthroughs is almost finished and satisfies all the requirements. The next step is the procurement for their integration in the MITICA [1] test bed and the detailed definition of the qualification tests to be carried out for the HNB feedthroughs.

The Vacuum vessel of the ITER Heating Neutral Beam injector is a key component of the neutral beam system. It supports the injector components and is also an extension of the ITER Vacuum Vessel. This paper presents the status of the design of the vessel. In the first part of this paper we present the main design requirements. Besides the pressure and thermal loads, the various interfaces with other components are important constraints for the design. Some have specific requirements concerning displacement, or require specific features for their maintenance. Indeed all maintenance involving opening of the vessel needs to be performed using remote handling systems. The second part aims at presenting the present design of the NB vessel. Various concepts were studied before coming to the present design. Structures were optimised to satisfy the chosen design code (RCC-MR) under the various loading scenarios applied to the vessel, both for normal and accidental cases. In addition, the vessel incorporates two sets of long lip seals which tolerate limited displacements. Limiting the displacement is a real challenge considering the size of the flange and the forces involved. An innovative solution has been developed to improve the mechanical connection between the lids and the vessel walls in order to fulfil the requirements. The vessel displacement is also important for the alignment of the Beam Line Components with respect to the neutral beam and for the cryopumps interfaces. The geometry of the vessel was optimised to reduce the electric field stresses and to increase the voltage holding of the Beam Source, which is at 1MV potential. The present design satisfies all the requirements. The main issues have been solved, and the proposed design is now being optimised to facilitate manufacturing.

Calorimeter for Diagnostic Neutral Beam (DNB) consists of two movable panels, forming V-shaped configuration with an included angle of 45 Deg. w.r.t. neutral beam axis. Panel's closing and opening motion will be achieved through Calorimeter Motion Mechanism (CMM). The DNB calorimeter design has evolved after analyzing several design proposals serving the basic functionality of the component, but these designs had compatibility issues for the ITER Remote Handling (RH) scenarios, operational maintenance, ITER Vacuum Handbook (IVH) compliance for hydraulic integration, design of flexible elements for large displacements, neutron irradiation susceptibility and joint materials for Vacuum Quality Class 1 (VQC1) application. In the design concept the movement mechanism and hydraulic line flexible element have been deliberately kept outside the DNB vessel, hence there is no sliding and complicated links inside the vessel which cannot be maintained through available RH tools. The paper proposes to present state of the art design of the ITER-DNB Calorimeter.

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

Positioning of Cesium (Cs) oven modules in the complex interface dominated space envelope of a negative ion source such as Diagnostic Neutral Beam (DNB) source for ITER is a challenge not only for the designer of the ion source, but also that of remote handling. A more user friendly design of the Cs delivery could emerge from the consideration of a possibility of injecting the Cs from an oven located outside the vacuum envelope of the ion source, thereby ensuring an ease of Cs refilling and oven maintenance. The design of such a delivery system involves long transmission path of lengths ∼4 m, from ambient to vacuum. System design involves incorporation of a low loss transmission tube enveloped by highly reflective inner surface pipe to reduce the heat losses and therefore heating of the nearby systems. A combination of all metallic valves operated at high temperatures has been incorporated in such a way that the Cs refilling or oven maintenance can be done without breaking the ion source vacuum. Removable joints in the oven heating elements are provided at specific locations to remove the Cs oven for ion source maintenance. Experimental data on Cs transmission over such a long length, required for an effective design of a co-axial transmission, is not presently available. However, an experiment has been carried out in ITER-India making measurements of Cs distribution in coaxial transmission of a length of more than 5 m. These experiments incorporate an additional feature of multiple nozzle distributor based Cs delivery into the ion source which might help in reducing the need of multiple Cs ovens in large ion sources like ITER. The Cs flux from the oven is measured by surface ionization detector (SID). The angular distribution of the Cs flux is measured by a movable SID in linear direction and has been found in good agreement with the calculations. The Cs inventory in the oven reservoir was measured by electrical resistivity measurements methods. The paper proposes to present the measurement results and also proposes a possible configuration of the Cs oven for ITER DNB ion source.

The Vacuum vessel of the ITER Heating Neutral Beam injector is a key component of the neutral beam system. It is host to the injector components and is also an extension of the ITER Vacuum Vessel. This article presents the s tatus of the design of the vessel. The first part of this article is dedicated to the presentation of the main design requirements. Besides the pressure and thermal load s, the various interfaces with other components are important constraints for the design . Some have specific requirements concerning displacement, or require specific featur es for their maintenance. Indeed all maintenance involving an opening of the vessel need s to be performed using remote handling systems. The second part aims at presenting the present desi gn of the NB vessel. Various concepts were studied before coming to the present day design. Structures were optimised to satisfy the RCC-MR code under the various loadin g scenarios applied to the vessel, both for normal and accidental cases. In addition, the vessel incorporates two sets of long lip seals which tolerate limited displacements. Lim iting the displacement is a real challenge considering the size of the flange and th e forces involved. An innovative solution has been developed to improve the mechanic al connection between the lids and the vessel walls in order to fulfil the requirement s. The vessel displacement is also important for the alignment of the Beam Line Compon ents with respect to the neutral beam and for the cryopumps interfaces. The geometry of the vessel was optimised to reduce the electric field stresses and to increase the voltage holding of the Beam Source, which is at 1MV. The present design satisfies all the requirements. The main issues have been solved, and the proposed design is now being optimised to f acilitate manufacturing.

The ITER fusion device is intended to demonstrate the viability of magnetically confined deuterium-tritium plasma as an energy source. One of the principal methods of heating and driving current in the plasma will be energetic beams of neutral atoms of D° at up to 1 MeV and of H° at up to 870 KeV, with a total injectable neutral beam power after transit through the neutralizer and downstream beamline elements of 16.5 MW from each ion source for pulse lengths of up to 3600 seconds. These requirements far exceed those of any previous positive or negative ion source, thus spawning a substantial development program to ensure that they will be met with a robustly reliable system. The ion source will consist of a large plasma expansion region fed by 8 RF driver units, and will be cesiated to enhance surface production of negative ions, followed by a multi-aperture multi-grid extractor and electrostatic accelerator. The plasma portion of the source is derived from a line of RF sources developed at IPP Garching,1 and the extractor/accelerator from development work at JAEA2, with the integrated design of the ITER source being done at Consorzio RFX. Unlike the first generation of high power negative hydrogen ion isotope sources, the ITER source will have the major advantage of a succession of progressively more comprehensive test facilities, culminating in a full power and pulse length test bed at Consorzio RFX. This talk will discuss the major beamline components, including the ion source and accelerator, the neutralizer cell that converts a portion of the negative ions to neutrals, the residual ion deflection system, and the tokamak field compensation system. Some remaining physics and engineering issues, along with their expected resolution, will be discussed, as well as the development and testing strategy.

Activities towards the establishment of the NB Test Facility (NBTF) in Padua-Italy and those related to the procurement of the heating neutral beams for ITER have recently reached a good level ofprogress thanks to the finalization of the agreements on theNBTF between F4E (the EU Domestic Agency for ITER), Consorzio RFX (the host of the NB test facility) and the ITER organization. This paper presents the status of the design of the various components within the EU scope of procurement, with a focus on the modificationsimplemented in the last years as a result of intense R&D activity undertaken in EU.

The beam source of neutral beam heating/current drive system for ITER is needed to accelerate the negative ion beam of 40A with D- at 1MeV for 3600 sec. In order to realize the beam source, design and R&D works are being developed in many institutions under the coordination of ITER organization. The development of the key issues of the ion source including source plasma uniformity, suppression of co-extracted electron in D beam operation and also after the long beam duration time of over a few 100 sec, is progressed mainly in IPP with the facilities of BATMAN, MANITU and RADI. In the near future, ELISE, that will be tested the half size of the ITER ion source, will start the operation in 2011, and then SPIDER, which demonstrates negative ion production and extraction with the same size and same structure as the ITER ion source, will start the operation in 2014 as part of the NBTF. The development of the accelerator is progressed mainly in JAEA with the MeV test facility, and also the computer simulation of beam optics also developed in JAEA, CEA and RFX. The full ITER heating and current drive beam performance will be demonstrated in MITICA, which will start operation in 2016 as part of the NBTF.

The ITER neutral beam (NB) injectors are the first injectors that will have to operate under conditions and constraints similar to those that will be encountered in a fusion reactor. These injectors will have to operate in a hostile radiation environment and they will become highly radioactive due to the neutron flux from ITER. The injectors will use a single large ion source and accelerator that will produce 40 A 1 MeV D− beams for pulse lengths of up to 3600 s. Significant design changes have been made to the ITER heating NB (HNB) injector over the past 4 years. The main changes are: Modifications to allow installation and maintenance of the beamline components with an overhead crane. The beam source vessel shape has been changed and the beam source moved to allow more space for the connections between the 1 MV bushing and the beam source. The RF driven negative ion source has replaced the filamented ion source as the reference design. The ion source and extractor power supplies will be located in an air insulated high voltage (−1 MV) deck located outside the tokamak building instead of inside an SF6 insulated HV deck located above the injector. Introduction of an all metal absolute valve to prevent any tritium in the machine to escape into the NB cell during maintenance. This paper describes the status of the design as of December 2008 including the above mentioned changes. The very important power supply system of the neutral beam injectors is not described in any detail as that merits a paper beyond the competence of the present authors. The R&D required to realize the injectors described in this paper must be carried out on a dedicated neutral beam test facility, which is not described here.

The electron cyclotron (EC), ion cyclotron (IC), heating-neutral beam (H-NB) and, although not in the day 1 baseline, lower hybrid (LH) systems intended for ITER have been reviewed in 2007/2008 in light of progress of physics and technology in the field. Although the overall specifications are unchanged, notable changes have been approved. Firstly, it has been emphasized that the HC the possibility to operate IC with 2 antennas with increased robustness (no moving parts); the possible increase to 2 MW of key components of the EC transmission systems in order to provide an easier upgrading of the EC power as may be required by the project; the addition of a building dedicated to the RF power sources and to a testing facility for acceptance of diagnostics and heating port plugs. Thirdly, the need of a plan for developing, in time for the active phase, a CD system such as LH suitable for very long pulse operation of ITER was recognised. The review describes these changes and their rationale.

The ITER baseline foresees 2 Heating Neutral Beams (HNB's) based on 1 MeV 40 A D(-) negative ion accelerators, each capable of delivering 16.7 MW of deuterium atoms to the DT plasma, with an optional 3rd HNB injector foreseen as a possible upgrade. In addition, a dedicated diagnostic neutral beam will be injecting ≈22 A of H(0) at 100 keV as the probe beam for charge exchange recombination spectroscopy. The integration of the injectors into the ITER plant is nearly finished necessitating only refinements. A large number of components have passed the final design stage, manufacturing has started, and the essential test beds-for the prototype route chosen-will soon be ready to start.