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The first ever built full-scale prototype of the ITER heating neutral beam injector is the MITICA experiment at PRIMA-NBTF (Neutral Beam Test Facility), under realization in Padua, Italy. The MITICA experiment includes many auxiliary plants; this paper is focused on the simulation and verification of the air cooling system of the High Voltage Power Supply (−1 MVdc), hosted in a Faraday cage (HVD1, High Voltage Deck1) inside a dedicated Building (High Voltage Hall, HVH) at PRIMA-NBTF. The HVD1 hosts the MITICA ISEPS (Ion Source Power Supply) in a two-floor cage, air insulated to ground for -1MVdc, with huge dimensions 12.5 m (L) × 8.4 m (W) × 9.6 m (H). Severe cleanliness constraints are required in the HVH hosting the HVD1, due to the high voltage operative conditions (the −1 MVdc Faraday cage is air insulated). The thermal behavior and air distribution in HVH are investigated as described in this paper.

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.

The ITER Neutral Beam Test Facility (NBTF) is currently under construction at Consorzio RFX, Padova, Italy. The NBTF includes two experimental devices: SPIDER for the study of the ion source and MITICA for the development of the full-size HNB prototype. Even if the two experiments share many architectural aspects in their Control and Data Acquisition System (CODAS), there are nevertheless some significant differences because MITICA must adhere to ITER's directives for plant control design. This paper describes the architecture of the MITICA timing system that represents one of the first examples of integration of IEEE1588 timing functions in fusion research medium-scale experiments. Timing functions are required in MITICA CODAS to: 1) Synchronize data acquisition and assign a timestamp to each data sample; 2) Supervise the generation of events, trigger a reaction of the control system and/or provide enhanced data acquisition in a time window centered on the event occurrence time. The IEEE1588 time synchronization protocol ensures sub-microsecond clock alignment among devices connected via a LAN. Therefore, it has the potential advantage of avoiding direct cabling among the involved actors, being all the synchronization messages exchanged over the LAN. A feature that IEEE1588 does not cover is the management of asynchronous events that may trigger some actions. Asynchronous events may trigger: 1) Enhanced data acquisition (e.g. with a higher sampling speed) during a time window centered on the trigger occurrence time; 2) Real-time response from the control system. While the management of events triggering control actions is under the responsibility of the real- time control system, for events triggering enhanced data acquisition, the "lazy triggers" solution is adopted in MITICA. This solution allows events to be communicated via LAN, compensating the delay in network communication will be compensated by means of circular buffers at the ADC side.

Heat generated in the ITER fusion reactor is deposited in the tokamak vacuum vessel, in-vessel components, and in the components of the neutral beam injector during plasma operations and du...

Cryosorption pumps create a vacuum by adsorbing gas at low temperature through porous solid adsorbents. The transmission probability of gas molecules and heat loads of cryosorption pumps are important factors affecting its performance. Herein, Molflow software based on the Monte Carlo principle is used to analyze the effects of the structural design of cryosorption pumps on transmission probability. The influence of structural design on radiation heat transfer is analyzed by ANSYS Steady-State Thermal software. This provides a reference for the design of a cryosorption pump to validate the prototype of a neutral beam injector for the China Engineering Fusion Experimental Reactor (CFETR).

Neutral beam injection is recognized as one of the most effective methods for plasma heating. According to the requirements for physics experiments of the Experimental Advanced Superconducting Tokamak (EAST), two beamlines for neutral beam injection were built in 2014. In order to meet the diverse requirements of magnetic confinement experiments, recent attempts have been carried out for arc power control of a neutral beam injection ion source. A test system was added to the arc power control system. It provides a new method to smoothly change arc power during a discharge. According to the experimental results of controlling arc power, the fesibility of varying the arc power was verified in the duration of discharge. Meanwhile, the test system achieves the remote setting of arc power parameters.

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

The ITER Neutral Beam Test Facility, in an advanced construction in Padova, includes the installation, tests, and optimization of the full prototype of the ITER Heating Neutral Beams injectors (HNBs), named MITICA. The MITICA Neutral Beam Injector will host its main items in a SS304L vacuum vessel composed of two modules, connected between them on site: the Beam Source Vessel (cubic of 5 m side and 67 tons weight) containing the ion 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 project requirements were transferred in the manufacturing design and in the fabrication process of these large vessels produced through welding and machining. The following steps were in particular carried out before and during manufacturing: o FE analyses to assess the structural integrity and limited deformations under vacuum loading conditions (leading to ribs reinforcement) o materials selection suitable for the specific application o qualification of special processes (welding and NDE) to ensure control of welding distortions and leak tight joints o implementation of double barrier sealings for the most critical connections o control of deformations under vacuum loads comparing displacement measurements with results of FE analyses. 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. They have been connected through a double sealing barrier composed of elastomer and a leak tight welded flange connection, previously qualified at the factory. The main outcomes of on-site final assembly and Site Acceptance Tests are described. Both vessels have been manufactured, installed and tested by De Pretto Industrie; the design and the technical support were provided by Consorzio RFX, while the procurement was managed by F4E.

The ITER neutral beam test facility, currently under the construction at Padua, Italy, consists of two experiments: source for production of ions of deuterium extracted (SPIDER) to test the ITER size ion source and megavolt iter injector and concept advancement (MITICA) to test the full ITER neutral beam injector. While the former represents an intermediate step, also finalized to the ITER Diagnostic Neutral Beam, the latter will represent the final design and implementation of the full prototype of the ITER heating neutral beam (HNB) injectors. Consequently, for the control and data acquisition system (CODAS) of SPIDER, there are no requirements for compliance with ITER, whereas the MITICA HNB Plant System will comply with the ITER directives for plant instrumentation and control because final system will be eventually integrated in ITER. SPIDER CODAS is currently under commissioning and the integration of the different frameworks that operate in an integrated way is proven to be successful. The experience and solutions gained in SPIDER CODAS development will be reused as far as possible in MITICA and to provide at the same time the required ITER plant system compatibility, a set of ITER-like layers (networks) will be defined, using exactly the same protocol defined for ITER interfaces.

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.

A DC −1 MV insulating transformer has been developed for the power supply of the neutral beam injector (NBI) at ITER (International Thermonuclear Experimental Reactor) . A DC long-pulse −1 MV insulation structure between the primary and secondary windings, considering potential variation in the insulator for DC long pulse, has been designed. Simultaneous development of a compound bushing, which consists of an AC 500 kV condenser bushing inside a cylindrical fiber-reinforced plastic tube, has been carried out. The DC −1 MV insulation design has been verified by high-voltage insulation tests of AC 50 Hz 1300 kV peak for 1 min, DC −1200 kV for 1 h, and a long-duration test of DC −1060 kV for 5 h. The results showed that the insulating transformer satisfies the power supply requirements for the ITER NBI. Copyright © 2016 Institute of Electrical Engineers of Japan. Published by John Wiley & Sons, Inc.

Neutral beam injection is recognized as one of the most effective means for plasma heating. According to the research plan of the EAST physics experiment, two sets of neutral beam injector (4–8 MW, 10–100 s) were built and operated in 2014. Neutralization efficiency is one of the important parameters for neutral beam. High neutralization efficiency can not only improve injection power at the same beam energy, but also decrease the power deposited on the heat-load components in the neutral beam injector (NBI). This research explores the power deposition distribution at different neutralization efficiencies on the beamline components of the NBI device. This work has great significance for guiding the operation of EAST-NBI, especially in long pulse and high power operation, which can reduce the risk of thermal damage of the beamline components and extend the working life of the NBI device.

A system for automated liquid nitrogen refilling of cryopumps in the neutral beam injector of the TCV tokamak (Switzerland) is described. The system includes industrially produced temperature and level monitors for cryogenic liquids, as well as an intelligent controller developed by the Budker Institute of Nuclear Physics on the basis of a field programmable gate array. The controller is used to control refilling of two cryopumps connected to a common nitrogen line and prevent emergency situations without operator attendance. Additional protection is ensured by the “timeout” mechanism used in different phases of the refilling cycle. The communication interface is available, and the user software is developed.

This overview summarizes the R&D activities aimed to secure for the ITER Neutral Beam Injectors a stable voltage holding, one of the most challenging issue for the NBI systems operation. The huge size of the accelerator and the highest voltage (1 MV) ever used in such systems make prominent some aspects that are, in other application, almost not influent or well under control, like the Area Effect, Total Voltage Effect, the highly complex electrostatic configuration and, last but not least, the cleaning degree of the electrodes surfaces. A clear explanation of the mechanism leading to the voltage breakdown in these conditions is far from being well established. But in the last years a great effort have been deployed by the NBI community to reliably achieve 1 MV voltage holding that is required for the NB injectors for ITER. Theoretical models and new tools have been derived to explain the mechanism of voltage conditioning and for the design of the complex electrodes geometries that supersede the simple electric field optimization. Presently, the milestone of the 1 MV voltage holding has been achieved but only with multi-gap configuration; activities are nevertheless in progress aimed to obtain a reliable single gap 1 MV voltage holding in the accelerator.

The integration of plant systems involving penetrations into the in-vessel components, such as heating and current drive, fuel cycle, and diagnostics, is a complex task constrained by top-level requirements of remote maintainability and high reliability. Within the EUROfusion Power Plant Physics and Technology Program, some activities are ongoing to assess the integration of different systems into the breeding blanket (BB), specifically neutral beam injector (NBI), electron cyclotron resonance heating launchers, diagnostics sightlines, fueling lines, and specific protections for the first wall (like startup limiters). This paper describes the integration of the NBI system into the Dual-Coolant Lithium-Lead (DCLL) BB for the EU DEMO. After identifying the major issues impacting the mechanical, thermal-hydraulic and neutronic behavior of the blanket, the integration efforts have been focused on minimizing the invasiveness of the NBI system and exploring different NBI options for the best compromise between plasma heating and BB performance. This paper describes the adaptation of the DCLL BB design to allocate the neutral beam duct. A particular attention is devoted to the redistribution of breeding and shielding functions, the new path of fluid circuits, and the additional cooling needs. The consequences of design modifications on key neutronic aspects like tritium breeding ratio and shielding capability are addressed. Besides, the thermal loads transferred to the BB walls from the neutral beam and the plasma are discussed, and a preliminary thermal assessment of the proposed solution is presented.

The injection of high energy neutral beams is one of the main tools to heat the DEMO plasma up to fusion conditions. A conceptual design of the neutral beam injector (NBI) for the DEMO fusion reactor is currently being developed by Consorzio RFX in collaboration with other European research institutes. High injector efficiency and RAMI, which are fundamental requirements for the success of DEMO, have been taken into special consideration for the DEMO NBI. A novel design of the beam source for the DEMO NBI is being developed featuring multiple subsources, following a modular design concept, capable of increasing the reliability and availability of the DEMO NBI. A full comprehensive model of the extraction/acceleration system has been developed, able to estimate all the main physics and engineering aspects (beam optics, coextracted and stripped electrons, heating power deposited on the grids, working temperature of the grids, and related stress and deformations) in a self-consistent way. This paper describes the results obtained applying the full comprehensive model to the present design of the extraction/acceleration system.

The accurate determination of the gas density distribution in negative ion accelerators plays a substantial role for the protection of the components and the efficiency of the system. The presence of background gas in between the electrodes has the highest impact on the beam properties and on the heat loads on the electrodes. The full-scale ITER beam source and extractor test facility SPIDER is studied considering the large vacuum vessel (4 m diameter), the pumping system, the plasma source (hydrogen gas filling pressure of ∼0.3 Pa) and the geometry of the in-vacuum components. On a smaller scale, the beam source and the multi-aperture electrodes (provided with apertures having an inner diameter of ∼12 mm) is accurately modeled. In these applications, the gas-surface interaction plays an important role and is therefore studied with dedicated models (on the nano-scale) to improve the predictive capability of molecular gas flow simulations.

In the Experimental Advanced Superconducting Tokamak (EAST) experiment, part of beam energy passes through the plasma and deposits in the interior wall of EAST. In order to protect the interior wall of EAST and obtain the temperature of it. A set of armor tile protection system (ATPS) based on PCI extensions for instrumentation was established. The function of ATPS is: (1) data acquisition; (2) anti false triggering; (3) output Digital Input Output alarm signal; The system is important to neutral beam injector even to whole EAST experiment. Experiments show that: the protection system fully meet the experimental requirements.

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

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

A common problem of the multi-beamlet, negative-ion based Heating Neutral Beam (HNB) Injectors presently used in fusion research, is the undesired deflection of negative ions due the permanent magnets embedded in the Extraction Grid (EG) for the suppression of the unwanted co-extracted electrons. A new solution has been recently developed at Consorzio RFX, which makes use of an additional set of permanent magnets, also embedded in the EG, with a series of advantages in terms of performances and versatility with respect to the traditional electrostatic compensation. This solution is now the new standard for ITER. After a thorough validation of the design by different numerical models, it was decided to experimentally test this solution for the first time on the Negative Ion Test Stand (NITS) at National Institute for Quantum and Radiological science and technology (QST), in Japan, within the framework of a scientific cooperation agreement between QST and Consorzio RFX. To this purpose, an EG having an ITER-like profile and compatible with NITS accelerator was designed and constructed at Consorzio RFX. This paper describes the design solutions adopted for this grid and its related permanent magnets, the construction phase, and a brief report on the first experimental results.

As one of the most effective methods for plasma heating, a neutral beam injector (NBI) achieved plasma heating and current driving for the first time in EAST 2014 experimental campaign. According to the research plan of the EAST physics experiment, the first NBI (EAST-NBI-1) has been built and become operational in 2014. In this article, the latest experiment results of EAST-NBI-1 are reported as follows: (1) EAST achieves H-mode plasma in the case of NBI heating alone, (2) EAST achieves 22 s long pulse stable H-mode plasma in the case of simultaneous NBI and lower hybrid wave (LHW) heating. The measurement data show that the loop voltage decreased and the plasma stored energy increased obviously. The results indicate that EAST-NBI-1 has achieved plasma heating and current driving, and thus lays a foundation for the construction of EAST-NBI-2, which will be built in a few months this year.

In the framework of the program for the construction-in Padova, Italy-of the first prototype of the International Thermonuclear Experimental Reactor 1 MeV-16 MW negative neutral beam injector (NNBI), an R&D activity on voltage holding in vacuum has been initiated since 2009, aimed at supporting the design, construction, and development of the NNBI accelerator. For this purpose, the high-voltage Padova test facility, designed to operate up to 800 kV, has been built, in parallel to the development of a multivoltage multielectrode voltage holding prediction model (VHPM). This paper focuses on the results of the experimental campaign (up to 270 kV) carried out on a number of electrostatic configurations, aimed at validating the VHPM under a well-controlled condition. The prediction capability of the model is partially confirmed. This result makes it worthwhile to review the physical basis of the model-based on the Cranberg-Slivkov clump theory-while also taking into consideration an alternative breakdown initiation mechanism.

The transformer core snubber (CS), as one of the most important components in the EAST (experimental advanced superconducting tokamak) NBI (neutral beam injector) system, is designed to limit grid damage and protect the ion source during periods of electrical breakdowns. A transformer core snubber is analyzed in detail in this paper. Several kinds of soft magnetic cores are presented and compared. With analysis and experiment on the basic characteristics of the cores, the most suitable materials are suggested. The circuit simulation code is established which could simulate faulty conditions with concentrated and distributed CS concepts. Based on the above work, an ion source CS is developed with series type of distributed topology. The CS has been subjected to experimental validation at 80 kV with a peak short-current of approximately 400 A in a real NBI system, which proves the accuracy of the adopted assumptions and the analysis method.

MITICA is the prototype of the Heating Neutral Beam Injectors (HNB) for ITER and is constituted by a negativeion source, a multi-grid electrostatic accelerator, a neutralizer and a residual ion dump. The injector is designed to produce a 17 MW beam of neutral particles (deuterium or hydrogen) up to 1 MeV. Two HNBs are planned to be installed in ITER at Cadarache, France. MITICA is under construction in the PRIMA facility in Padova, Italy. The design of MITICA/HNB [1] has required extensive optimizations in order to guarantee that the thermomechanical stresses in the materials do not exceed the tolerable limits for the required fatigue life. This result has been achieved by: (a) reducing and uniforming the heat loads on the accelerator grid and beam line components, and (b) improving the heat removal capability of the active cooling system of the grids. This involved an extensive optimization of the electric and magnetic field configuration and of the geometry of the accelerator grids, suppressing the electrons before they are accelerated at higher energy, and improving the optics of the ion beamlets. However, deviation from the expected magnetic and electrostatic configuration can be caused by "off-normal" or "failure" operating conditions of the accelerator, such as undesired mechanical displacement of the grids, nonuniform extracted ion current, demagnetization of magnets, partial failure of the accelerator grid power-supplies and also operation with different ion species. Purpose of the present work is to analyse and identify the "off-normal" operating conditions which could possibly become critical in terms of thermomechanical stresses or of degradation of the optical performances of the beam, and to give indications on protective actions to be taken.

The ITER Heating Neutral Beam (HNB) injectors shall be protected from stray magnetic field (several hundreds of mT) produced by the ITER PF coils and plasma current. Such stray field would hamper the production of negative ions, deflect ion trajectories in the accelerator and cause intolerable heat load on neutralizer and beam line components. In order to keep the residual magnetic field below acceptable levels (1 mT in the ion source and accelerator and 0.3 mT in the neutralizer), each injector will be surrounded by Passive Magnetic Shield and by six Active Correction and Compensation Coils (ACCC). The ACCC will be feedback controlled using magnetic field sensors located inside the HNB vessel, an environment subjected to considerable neutron flux (~ 1099n/cm22/s) during the ITER operation. Therefore, magnetic sensors that are robust, radiation hard, drift-immune and remote-handling compatible are required. Flux-gate magnetic sensors are good candidate, as their active part includes no semiconductor or other radiation-sensitive component, and consists of a ferromagnetic core and two insulated-wire coils. Commercial flux-gate sensors, used for precise measurements of weak magnetic fields, have very good sensitivity, but measurement range below 0.1 mT, and also include on-board electronics. A flux-gate sensor for the ITER HNB has been studied using a numerical model of the magnetic core hysteresis, essential to describe the sensor operation, sensitivity and measurement range. This model indicated that, by suitable choice of the core magnetic properties and gap geometry, the measurement range can be extended by at least 2 orders of magnitude. On this basis, a prototype flux-gate sensor has been realized at Consorzio RFX. Experimental tests carried out so far have confirmed the results of the numerical model and have demonstrated that the measurement range can be increased to ~ 10 mT with acceptable accuracy and frequency bandwidth.

The MITICA experiment (Megavolt ITER Injector & Concept Advancement) is the prototype and the test bed of the Heating and Current Drive Neutral Beam Injectors, which will be necessary for the full-performance exploitation of ITER. MITICA injector experiments shall demonstrate the reliable and accurate emission of a 17 MW beam of neutral particles for duration up to 1 hour, fulfilling ITER specific requirements. The accelerator grids are among the most critical parts of this experiment, because they must fulfill several operational requirements and at the same time satisfy the fatigue verifications according to the ITER Structural Design Criteria for In-vessel Components (SDC-IC). After about two years of continuous development, two design improvements were found to effectively increase the fatigue life of the grids up to the requested values. The first method was to adopt a novel shape of the cooling channels inside the grids, called Nozzle Island Cooling Enhancement (NICE) and able to provide a high performance cooling without exceeding the limits on the pressure drop through the grids. The second, called Stress Relieving Slits (SRS), was to introduce suitable slits in the grids, whose design was iteratively optimized until they were able to significantly reduce the stress/strain peaks due to thermal gradients. The NICE and SRS design solutions, here described in detail, were found to be key improvements in order to obtain a design of the grids able to satisfy all the operating requirements and all the structural verifications according to the ITER criteria.

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

The 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.

The 100 keV Ion Source Test facility – Source for the Production of Ions of Deuterium Extracted from RF plasma (SPIDER) – is aimed to test the full scale prototype of the Ion Source for the ITER 1 MeV Neutral Beam Injector (NBI). The SPIDER facility requires the construction of a High Voltage Deck (HVD) and of a High Voltage Transmission Line (TL) respectively to host the Ion Source Power Supplies system polarized at 100 kV and to carry the power and signal conductors to the beam accelerator. In already existing NBI systems with beam energy above 100 keV, the TL is realized with the SF6 Gas Insulated Line technology. In the SPIDER TL case, the presence of a large inner conductor (half meter diameter), would make the pressurized TL a complex and costly component; therefore a free air insulated solution has been proposed. The paper focuses on the design of this TL, which has to host inside the complex high potential (100 kV) inner electrode a number of power and measuring conductors and has to minimize the Electro Magnetic Interferences (EMI) produced by the frequent grids breakdowns. Finite Element (FE) analyses have been performed to verify the configuration from the electrostatic point of view, to evaluate EMI screening effectiveness and to assess the impact of the relatively high thermal dissipation of power conductors located inside the high potential electrode. Moreover, an experimental test campaign has been carried out on a TL mockup to validate the TL electrostatic configuration under DC voltage. Finally, the paper reports on the status of procurement activities for the Transmission Line.

Excellent vacuum performance ensures a high beam transmission efficiency of the neutral beam injector (NBI). The vacuum performance is mainly determined by the cryoperformance of the cryopanel of the cryocondensation pumps which are the main vacuum pumps of NBI. In order to optimize the cryoperformance, the requirements for the temperature distribution and the heat load of the cryopanel are analysed and the factors that affect the cryopanel's temperature distribution are studied. The results indicate that the temperature difference of the cryopanel can be reduced by fabricating the cryopanel with high thermal conductivity material, increasing its thickness and cutting the distance between the two upward cooling pipes. The results may be applied to a cryopanel cooled by forced flow liquid helium.

This article presents studies on the long term pumping behavior of a cryopump. It is shown that the pumping speed does not deteriorate on a time scale of 4200s for a gas load of 2.4×105torrl, corresponding to ∼106 ML of hydrogen. It has also been observed that the need for regeneration of the pump is dictated by the safety limits of operation rather than its pumping capability. No sudden boil off of the cryogen takes place during the regeneration phase.

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 1 MV bushing of the ITER NB injector is probably one of the most challenging components of the whole accelerator, since such a high voltage (HV) in a contaminated and radiated environment requires the development of non-standard technologies. This paper intends to provide a guideline for the design of 1 MV bushing for the ITER neutral beam injector test bed. Although mainly focused on the SINGAP alternative concept of the injector, the reported general criteria can be useful for the design of other HV components, such as post insulators of the acceleration grids. For every different kind of required insulation, basic physical models describing breakdown phenomena are reported together with the main technological implications and the practical recommendations necessary to improve insulation performance. The results of the finite element analyses of 1 MV bushing, operating at CEA—Cadarache laboratories, are used to validate general criteria in a typical geometry. The effects of radiation, such as neutrons and gamma rays, on solid insulators and gases have also been taken into account, describing the most important phenomena and the related implications in terms of voltage hold-off capabilities. Finally the design of the bushing in the SINGAP option and the solutions adopted are described.

This paper presents the results of the design investigations and the manufacturing of cryosorption pumps for the Neutral Beam Test Facility ‘MANITU’ at Max Planck Institute for Plasma Physics (IPP) and the cryopump design analyses performed for the development of a test facility for the first ITER Neutral Beam Injector. The two identical cryopumps for IPP are foreseen to pump a hydrogen-throughput of 3 Pa m 3 /s (H 2 ) with a pumping speed of 350 m 3 /s per pump. The pressure conditions must be maintained over 4 h pumping without regeneration of the cryopanels. For the ITER Neutral Beam Test Facility, design calculations to assess heat loads and pumping parameters of the cryopumps will be briefly described.

The operational performance of the JET neutral beam injector (NBI) system during 2003 is presented and compared with NBI operation from 1994 to 2002. The paper also addresses different demands imposed on NBI operation during the JET Joint Undertaking (until the end of 1999) and the European Fusion Development Agreement (EFDA) JET operating contract (from 2000). The material presented shows new operational performance records achieved in 2003, derived from data focused on average and maximum pulse lengths, pulse power and injected pulse energy. Over the last 10 years, the issue of JET NBI positive ion neutral injector (PINI) reliability and availability has also been of significant interest. A discussion is presented where terminology is defined, technical systems causing unreliability and non-availability are analysed and operational practices are reviewed. The performance analysis shows that during the period of JET operation under the EFDA contract, the NBI facility has successfully changed from high power-short pulse to high power-long pulse (10 s) operation. It also shows that the sources of unreliability and non-availability have largely remained constant during this change.

To characterize ITER Diagnostic Neutral Beam (DNB) system with full specification and to support IPR's negative ion beam based neutral beam injector (NBI) system development program, a R&D facility, named INTF is under commissioning phase. Implementation of a successful DNB at ITER requires several challenges need to be overcome. These issues are related to the negative ion production, its neutralization and corresponding neutral beam transport over the path lengths of ~ 20.67 m to reach ITER plasma. DNB is a procurement package for INDIA, as an in-kind contribution to ITER. Since ITER is considered as a nuclear facility, minimum diagnostic systems, linked with safe operation of the machine are planned to be incorporated in it and so there is difficulty to characterize DNB after onsite commissioning. Therefore, the delivery of DNB to ITER will be benefited if DNB is operated and characterized prior to onsite commissioning. INTF has been envisaged to be operational with the large size ion source activities in the similar timeline, as with the SPIDER (RFX, Padova) facility. This paper describes some of the development updates of the facility.

Considerable R&D effort has gone into the development of a Torus Exhaust Vacuum Pumping System over the past years. The selected primary pump is of cycling type with discontinuous regeneration, solid cryosorbent for helium pumping and no fuel separation. The pump includes a throttle/regeneration valve to maintain the divertor pressure, minimize the thermal load on the pump and permit regeneration. Modelling of various torus cryopump configurations has been carried out and a research program aimed at developing a primary torus cryopump with rapid regeneration cycles is in progress. Panel tests with tritium will provide additional data required for ITER design. The design, fabrication and demonstration of a model cryopump with integral inlet valve is in progress. Experimental work and modelling to develop a high speed pumping system for the neutral beam injector is under way. Practical experience under ITER relevant conditions and in the presence of tritium has accumulated on the performance of oil-free pumps in various tritium laboratories. Recent world-wide developments in the field of fuel clean-up have been impressive. Several concepts based on palladium/silver diffusors combined with catalytic as well as with electrochemical process steps have the potential to achieve the high overall tritium decontamination factors and low tritium inventory levels required by ITER. Basic research activities aimed at characterizing and optimizing the envisioned essential components for the various ITER fuel clean-up concepts are in progress. Tests with an industrially manufactured facility of technical scale employing realistic concentrations of tritium and using an advanced operation and safety concept have been performed. Optimization of process components with respect to performance, operability and safety criteria is proceeding in parallel with integrated loop tests. Endurance tests to demonstrate complete processes with tritium are in preparation in several laboratories. Basic research aimed at developing a process for the recovery of tritium from the helium glow discharge cleaning exhaust stream is under way. The routine operation with tritium of large civilian research laboratories under conditions specific to the respective sites will contribute significantly to a realistic design of the ITER Tritium Plant. For the operation of the Tritium Plant and tritium accountancy much effort has gone into the improvement of known analytical techniques and the further development of calorimetry.

The power supply system of ion source for the Neutral Beam Injector (NBI) in the HT-7 superconducting tokamak is based on a single injector with one ion source that can deliver 700 kW of neutral beam power. Experiments and a discharges test on the ion source were successfully performed. In this paper, the circuit structures and features of every power supply are described and the results of the discharges test are presented.

The ion source test facility ELISE ("Extraction from a Large Ion Source Experiment") has recently being built at IPP Garching (Germany) and is operational since December 2012. ELISE is an important intermediate step in the R&D roadmap that the European domestic agency Fusion for Energy has defined for the ITER neutral beam injectors. This new test facility for a large radio-frequency ion source builds up on the experience gained during many years of successful operation of smaller testbeds at IPP. Purpose of ELISE is to extract a negative ion beam, at ITER relevant parameters, by means of an ion source with the full width (1 m) and half the height of the ITER source. ELISE is designed to generate plasma for up to 1 h and extract a H- or D- beam at up to 60 kV. Available power supplies allow beam pulses of 10 s every 150 s. Although ELISE is designed to be as close as possible to the ITER source, technical solutions have been implemented that allow considerable experimental flexibility. Focus was put on easier diagnostic access and ease of maintenance. The results gained during the construction and operation of ELISE are delivering very important informations to support the design and operation of the neutral beam test facility PRIMA to be built in Padua (Italy). This paper gives an overview of the project, presents experiences made during commissioning and shows results obtained so far.

MITICA is a negative-Ion-based 1 MeV Neutral Beam Injector. The transverse magnetic field in the plasma source and in the accelerator are crucial for preventing the electrons in the plasma source from entering the accelerator and deflecting the secondary electrons, generated inside the accelerator by stripping reactions with background gas, before they are accelerated at higher energy. The scientific objective of MITICA is essentially the experimental validation of the ITER 1 MeV Neutral Beam injector design and indeed a flexible solution is necessary in order to allow an independent variation of the magnetic field in the plasma source and in the accelerator during the experimental campaign. To this purpose a specific magnetic design has been developed and optimized, essentially based on current-carrying busbars with the possible addition of external coils or magnets. The validated magnetic field configuration will then be replicated in the ITER HNB, using only a current-carrying busbar system. The paper presents the rationale of the various steps of this process and the final design obtained.

The MITICA (Megavolt ITER Injector Concept Advancement) experiment is the full scale prototype of the Heating and Current Drive Neutral Beam Injectors for ITER, planned to be built at Consorzio RFX. The injector is designed to accelerate and neutralize high density beams of H − or D − up to 1 MeV. During the acceleration stage the beam is expected to interact with the residual background gas and with the material surfaces of the electrodes, resulting in the production of secondary particles, mainly electrons, which are accelerated together with the primary beam. These particles, having a large divergence, tend to impact on the MITICA components located after the accelerator, causing significant heat loads. Moreover, as there is a high probability of backscattering some particles can also reach components which are not in the direct line of sight of the accelerator, making the evaluation of the distribution of heat loads on the various components not straightforward. This paper discusses the main sources of power loads to the MITICA beamline components downstream of the accelerator, and the implementation of a 3D code tracking the particles in order to evaluate the associated power deposition.

MITICA is a prototype of the Heating Neutral Beam (HNB) Injectors for ITER [1]. The MITICA scientific objective is to validate the design and optimize the operating performances of the injector before installation on ITER. The magnetic field configuration in the injector is critical for obtaining the required efficiency and beam optics. In order to quickly achieve the optimal operating condition, it is highly advisable to provide a flexible control of the most critical features of the magnetic field, without opening the vacuum vessel. The possibility of introducing an independent variation of the magnetic field in the plasma source and in the accelerator has been analyzed and a feasible solution proposed, based on small lateral coils. The possibility of recreating the expected stray magnetic field conditions of the ITER HNB was already foreseen in the past by using a further set of external coils, whose design has been revised and updated. The paper describes the rationale and the magnetic design optimization of these systems.

MITICA (Megavolt ITER Injector Concept Advancement) is a test facility for the development of a full-size heating and current drive neutral beam injectors for the ITER Tokamak reactor. The optimized electrostatic and magnetic configuration has been defined by means of an iterative optimization involving all the physics and the engineering aspects. The acceleration grids have been designed considering optical performances and mechanical constraints related to embedded magnets, to cooling channels, to the grid stiffness and manufacturability. A combination of “local” vertical field and horizontal “long range” field has been found to be the most effective set-up for ion extraction, beam focusing and minimization and equalization of thermo-mechanical loads and minimal number of electrons exiting the accelerator.

A set of physics and engineering codes is currently used and developed at Consorzio RFX to simulate the negative ion beam accelerators of the next experiments SPIDER and MITICA. In order to verify the accuracy of the simulations, an experimental validation has been undertaken, that consist in benchmarking the codes with data from ITER-relevant experiments. As a first step, the SLACCAD code have been applied to simulate the beam optics in the MANITU testbed at IPP, which features a Radio Frequency ion source and is able to operate with ITER-like conditions in terms of extracted ion current density and pulse length. A description of the modeling activities and a discussion of the main results obtained so far are here reported. In particular, the effect of beam halo, space charge compensation and particle angle at the emitter are investigated. Some results will be used to identify reasonable operating parameters as inputs for the design optimization of the SPIDER and MITICA experiments.

The consice description of neutral beam injection system module is given. The analysis of the most severe accident in module, initiated by high voltage insulator damage is carried out on the base the probabilities of NBI module components failure. The accident scenarious with the most unacceptable probabilities are revealed.

A high speed gas pumping system is required to minimize the reionization loss of neutral beams in a neutral beam injector; until now the beam line design has made use of a condensation cryopump system. The system becomes more expensive as size increases and needs high maintenance and has a high operation cost. Instead of the cryocondensation pump a cryosorption pump is considered for the beam line. This paper describes the development of a high speed cryosorption pump for a neutral beam injector. A module test has been carried out to have a fundamental data base. A cryosorption pump with a pumping speed of about 300 m3 s−1 has been designed on the basis of the result of the module test and has been applied on the neutral beam injector of JIPP T-IIU Tokamak. In this paper, the design and the experimental results are discussed.

A design for a power supply system for the neutral beam injection (NBI) test stand of the large helical device (LHD) is presented. The acceleration power supply of −250 kV, 50 A is composed of three power supplies connected in series to deliver two intermediate outputs. It has a −40 kV power supply, a −90 kV power supply and a −120 kV power supply. A beamline and the major part of the power supply system for the NBI test stand were constructed in 1992. The −120 kV power supply was moved from Higashiyama to Toki in 1994. Another power supply of −40 kV will be added for −250 kV acceleration in the future. A brief outline of the control system of the NBI test stand is also presented.

In the framework of the program for the construction in Padova - Italy of the first prototype of the ITER 1 MeV-16 MW Negative Neutral Beam Injector a R&D program on voltage holding in vacuum has been initiated since 2009, aimed at supporting the design, construction and development of the NNBI accelerator. To this purpose, the testing facility HVPTF, designed to operate up to 800 kV, has been built, in parallel to the development of a Multi Voltage-Multi-Electrode-Voltage Holding Prediction Model-VHPM. The paper focuses the results of the campaign (up to 270 kV) carried out in the Facility on different electrode configurations, with the aim of validating the VHPM in well controlled condition. The prediction capability of the model has been partially confirmed. This result made worthy to make a critical review of the physical basis of the model-based on the Cranberg-Slivkov clump theory - taking into consideration also other breakdown initiation mechanism.