Beam Line Components (BLCs) for ITER Diagnostic Neutral Beam (DNB) system have been designed with respect to ITER Structural Design Criteria for In-Vessel Components (SDC-IC) and manufactured to meet EN/ISO standards, ITER Vacuum Handbook and ASME code requirements. The manufacturing of these components was preceded by a dedicated R&D on the material development and manufacturing technology developments. Though, this route of technology development through prototyping provided considerable understanding in terms of defining the manufacturing processes and established the manufacturing feasibility of building components for complex systems like Neutral Beams, the production related challenges still remained unpredicted and they had to be addressed considering the practical aspects along with the specification requirements. The paper is aimed to provide the insight of technologies implemented for the production and some of the significant learnings, based on the experience of manufacturing of BLCs for ITER DNB system.

ITER envisages the use of two heating neutral beam injectors plus an optional one as part of the auxiliary heating and current drive system. The 16.5 MW expected neutral beam power per injector is several notches higher than worldwide existing facilities. A Neutral Beam Test Facility (NBTF) was established at Consorzio RFX, exploiting the synergy of two test beds, SPIDER and MITICA. SPIDER is dedicated to developing and characterizing large efficient negative ion sources at relevant parameters in ITER-like conditions: source and accelerator located in the same vacuum where the beam propagates, immunity to electromagnetic interferences of multiple radio-frequency (RF) antennas, avoidance of RF-induced discharges on the outside of the source. Three years of experiments on SPIDER have addressed to the necessary design modifications to enable full performances. The source is presently under a long shut-down phase to incorporate learnings from the experimental campaign. Parallelly, developments on MITICA, the full-scale prototype of the ITER NBI featuring a 1 MV accelerator and ion neutralization, are underway including manufacturing of in-vessel components, while power supplies and auxiliary plants are already under final testing and commissioning. Integration, commissioning and tests of the 1MV power supplies are essential for this first-of-kind system, unparalleled both in research and industry field. The integrated test to confirm 1MV output by combining invertor systems, DC generators and transmission lines extracted errors/accidents in some components. To realize a concrete system for ITER, solutions for the repair and the improvement of the system were developed. Hence, NBTF is emerging as a necessary facility, due to the large gap with existing injectors, effectively dedicated to identify issues and find solutions to enable successful ITER NBI operations in a time bound fashion.

The article will present and discuss the issues found in Source for Production of Ion of Deuterium Extracted from RF plasma (SPIDER) experiment with the operation of the radio frequency (RF) generators based on self-excited tetrode oscillators and the advantages expected from their replacement with RF solid-state amplifiers. The assessment of the solution based on new solid-state amplifiers for both SPIDER and Megavolt ITER Injector and Concept Advancement (MITICA) experiments operating in Padua, Italy, in the Neutral Beam Test Facility, will be described, outlining the main design challenges and open points.

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 ITER Neutral Beam Test Facility, in an advanced stage of construction in Padova, includes the installation, tests, and optimization of the full prototype of the ITER Heating Neutral Beams injector (HNBs), named MITICA. The MITICA Neutral Beam Injector will host its main components in a SS304L vacuum vessel composed of two modules, connected between them on site: the Beam Source Vessel (cubic shape of 5 m side and 67 tons weight) containing the Beam Source and the Beam Line Vessel (section 4.5 m x 4.5 m, length 11 m and 76 tons weight) containing the Beam Line Components and the Cryopumps. The manufacturing is described, going through the FE analyses performed to assess the structural integrity, the materials selection, the welding qualifications, the implementation of the double barrier sealings and the control of the deformations. The Factory Acceptance Tests of the individual vessels are presented, including their Helium Leak Tests. The two vessels have been assembled on-site inside the MITICA bio-shield. The main outcomes of on-site final assembly and Site Acceptance Tests are described. Both the vessels have been detail designed, manufactured, installed and tested by De Pretto Industried, from Schio, (VI) Italy, supporting fusion technology’s applications from more than 50 years. The design and the technical support were provided by Consorzio RFXa, while the procurement was managed by Fusion For Energye.

The requirements of ITER neutral beam injectors (1 MeV, 40 A negative deuterium ion current for 1 h) have never been simultaneously attained; therefore, a dedicated Neutral Beam Test Facility (NBTF) was set up at Consorzio RFX (Padova, Italy). The NBTF includes two experiments: SPIDER (Source for the Production of Ions of Deuterium Extracted from Rf plasma), the full-scale prototype of the source of ITER injectors, with a 100 keV accelerator, to investigate and optimize the properties of the ion source; and MITICA, the full-scale prototype of the entire injector, devoted to the issues related to the accelerator, including voltage holding at low gas pressure. The present paper gives an account of the status of the procurements, of the timeline, and of the voltage holding tests and experiments for MITICA. As for SPIDER, the first year of operation is described, regarding the solution of some issues connected with the radiofrequency power, the source operation, and the characterization of the first negative ion beam.

The ITER project is a major step in the development of fusion energy aiming to demonstrate the feasibility of exploiting magnetic confinement fusion for the production of energy in an experiment close to the dimensions of a future commercial fusion reactor. To achieve high fusion gain and control several aspects of the plasma behavior, three heating systems capable of injecting a total power up to 73 MW will be installed. This paper will summarize the requirements of the three systems and will highlight the main achievements in their design and construction.

The Indian test facility (INTF) is a R&D facility under development at the Institute for Plasma Research, Gandhinagar, as part of the neutral beam development from negative ion source program. The major advantage of the INTF, besides developing beams from large ion sources, is to characterize and benchmark the ITER diagnostic neutral beam (DNB) to the desired specifications over transport lengths of ~21 m, a unique feature of this test bed. Such a study will establish the expected power to be delivered by DNB into the ITER plasma, an important parameter to estimate the S/N ratio expected from the He ash measurements by charge exchange recombination spectroscopy. It may be noted that DNB is a procurement package for India as an in-kind contribution to ITER. Extensive physics and thermomechanical calculations have been performed to finalize the component design and layout. Adequate choice of materials, and manufacturing and jointing processes compatible to ITER safety standards have been made in order to make the components adhere to the safety and quality classification, thereby ensuring that the components survive the ITER lifetime while operating in harsh nuclear environments. The components are currently in various phases of manufacturing, and the first operations of the INTF are anticipated in Q4 of 2019.

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 <7 mrad; beam aiming direction within 2 mrad). This paper outlines the worldwide effort towards the ITER NBI realisation: the main results of the ELISE facility (IPP-Garching, Germany), equipped with a half-size source, are described along with the status of MITICA; specific issues are investigated by small specific facilities and by joint experiments at QST and NIFS (Japan). The SPIDER experiment, just come into operation, will profit from strong modelling activities, to simulate and interpret experimental scenarios, and from advanced diagnostic instruments, providing thorough plasma and beam characterisation. Finally, the results of the first experiments in SPIDER are presented, aimed at a preliminary source plasma characterisation by plasma light detectors and plasma spectroscopy.

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 (NBTE) 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.

The negative ion source SPIDER produces a beam power that has to be absorbed by a beam dump component made of stacked hypervapotrons. The manufacturing process produced dimension errors that have given rise to a possible shine-through issue. Finite volume models were developed to carry out transient thermal analyses aiming at validating the design of a system made of copper screens to shield the heat loads passing through the beam dump. The evaluation of the best solution was conducted by looking both at the thermal performances and the technical feasibility. The results show the capability of the system to safely withstand the heat power up to one hour.

Substantial progresses have been achieved in the realization of the ITER Neutral Beam Test Facility (NBTF) hosted in Padova, Italy; the buildings, completed by the end of 2015, are being progressively filled with new systems and components. The realization of SPIDER, the ITER full-size negative ion source, is well advanced and important progress is also recorded for MITICA, the full-scale prototype of the ITER HNB injector. The paper gives an overview of the achieved results, highlighting the main challenges faced. (C) 2017 Elsevier B.V. All rights reserved.

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 accelerator for the Diagnostics Neutral Beam (DNB) beam source is composed of a multi-aperture grid system with three water cooled grids made from Oxygen free Copper. To achieve the focusing requirements at the distance of >20 m, the grid segments are designed with two stage angles (0.222 degrees and 0.665 degrees) from the centerline in the horizontal direction. The configuration of this kind of 'angled segment' includes the water cooling channels milled in the angular form, subsequently closed by copper electrodeposition, providing the angles on front and back surface and then drilling of apertures on the angular plane. The long beam path and low energy beam demands the tight tolerances on each of these mechanical features and therefore demands the high degree of manufacturing controls on each of the processes. To unveil the challenges those could appear during the production of such grid, a 1:1 prototype of the most complex type of grid has been manufactured. This paper shall present the technical data generated out of manufacturing of this prototype, summarizing the recommendations for real grid production on: optimization of the sequence of manufacturing, effect of each of the operations, post-manufacturing handling and identifying the measurement techniques. The experience gathered here provides a recipe for the best manufacturing practices for the accelerators of NB system for ITER and upcoming devices. (C) 2017 Elsevier B.V. All rights reserved.

Prototype High voltage bushing (PHVB) is a scaled down configuration of DNB High Voltage Bushing (HVB) of ITER. It is designed for operation at 50 kV DC to ensure operational performance and thereby confirming the design configuration of DNB HVB. Two concentric insulators viz. Ceramic and Fiber reinforced polymer (FRP) rings are used as double layered vacuum boundary for 50 kV isolation between grounded and high voltage flanges. Stress shields are designed for smooth electric field distribution. During ceramic to Kovar brazing, spilling cannot be controlled which may lead to high localized electrostatic stress. To understand spilling phenomenon and precise stress calculation, quantitative analysis was performed using Scanning Electron Microscopy (SEM) of brazed sample and similar configuration modeled while performing the Finite Element (FE) analysis. FE analysis of PHVB is performed to find out electrical stresses on different areas of PHVB and are maintained similar to DNB HV Bushing. With this configuration, the experiment is performed considering ITER like vacuum and electrical parameters. Initial HV test is performed by temporary vacuum sealing arrangements using gaskets/O-rings at both ends in order to achieve desired vacuum and keep the system maintainable. During validation test, 50 kV voltage withstand is performed for one hour. Voltage withstand test for 60 kV DC (20% higher rated voltage) have also been performed without any breakdown. Successful operation of PHVB confirms the design of DNB HV Bushing. In this paper, configuration of PHVB with experimental validation data is presented.

The Indian Test Facility (INTF) for Neutral Beams is designed with the objective to not only test the performance of an ion source but also to characterize the transmission of the Neutral Beam to a length of 20.66 m and thereby generate adequate database on the quality of the beam that is required for the CXRS diagnostics for measurement of Helium ash content in the fusion plasma.

Indian Test Facility (INTF) is a test bed to characterize ITER Diagnostic Neutral Beam (DNB) system, is being constructed at ITER-India, IPR [1]. The operation is aimed to generate 18-20 A of 100 KeV, hydrogen neutral beam with a 3 s ON/20 s OFF duty cycle at 5 Hz modulation. The final focused beam will be characterized on a beam dump or a heat sink system named Second Clorimeter (SC), situated at a distance of 20.66 m from the Grounded Grid (GG) of the Beam Source (BS). It is designed to withstand a total beam power or heat load of 2.84MW, corresponding peak heat flux is similar to 93.15MW/m(2) and the required number of thermal cycle of 6.13 x 10(4). To handle such high heat flux, a V shaped design is adopted with a suitable apex angle. In INTF, First Calorimeter (FC) is placed just after Residual Ion Dump (RID) at a distance 5.95 m from the GG of BS [2]. In the present work a heat absorbing building block called hypervapotron was adopted from [3] and modified to meet the above mentioned heat load requirements. The paper presents the conceptualization of SC and a thermo-mechanical assessment of the heat transfer element by Finite Element method (FE), carried out using ANSYS. FE results are used to validate the HTE design as per Structural Design: Criteria for In vessel Component (SDC-IC) for monotonic as well as cycling loading (Fatigue) damage criteria. This paper also reports the key challenges and usual solutions adopted in the design.

For the first time the nature of response of thermal fatigue damage (TFD) in Cu-Cr-Zr alloys, considered for the High Heat Flux components of Tokamak and its subsystems in International Thermonuclear Experimental Reactor application has been studied. Temperature cycling between 290 degrees C and 30 degrees C, similar to the service condition, has been carried out on two differently aged Cu-Cr-Zr alloys. The TFD has been assessed by damage mechanics approach using damage parameters, and by surface characteristics. The damage parameters increase exponentially during initial fatigue cycles and saturates, whilst surface characteristics shows continuous increase with increase in thermal fatigue cycles. Damages are different in the aged alloys depending upon the aging conditions. (C) 2016 Elsevier B.V. All rights reserved.

INdian Test Facility (INTF) is envisaged to characterize ITER diagnostic neutral beam system and to establish the functionality of its eight inductively coupled RF plasma driver based negative hydrogen ion source and its beamline components. The beam quality mainly depends on the ion source performance and therefore, its diagnostics plays an important role for its safe and optimized operation. A number of diagnostics are planned in INTF to characterize the ion source performance. Negative ions and its cesium contents in the source will be monitored by optical emission spectroscopy (OES) and cavity ring down spectroscopy. Plasma near the extraction region will be studied using standard electrostatic probes. The beam divergence and negative ion stripping losses are planned to be measured using Doppler shift spectroscopy. During initial phase of ion beam characterization, carbon fiber composite based infrared imaging diagnostics will be used. Safe operation of the beam will be ensured by using standard thermocouples and electrical voltage-current measurement sensors. A novel concept, based on plasma density dependent plasma impedance measurement using RF electrical impedance matching parameters to characterize the RF driver plasma, will be tested in INTF and will be validated with OES data. The paper will discuss about the overview of the complete INTF diagnostics including its present status of procurement, experimentation, interface with mechanical systems in INTF, and integration with INTF data acquisition and control systems.