Search

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 Ion Source of the ITER Heating Neutral Beam Injector (HNB) is polarized at −1MVdc to ground. A large-1MVdc air-insulated Faraday cage (12.5 m (L) × 8.4 m (W) × 9.6 m (H)), called High Voltage Deck 1 (HVD1), hosts the Ion Source and Extractor Power Supplies (ISEPS) and the associated HNB diagnostics; the HVD1 is interfaced with a SF6 insulated Transmission Line (connecting the main power supplies to the HNB) through a High Voltage (HV) Bushing Assembly (HVBA) carrying inside ISEPS power conductors and diagnostics. The HVD1 and HVBA are unique devices for size, electrostatic and mechanical features. This paper presents the main design choices adopted to comply with the challenging technical requirements and the supporting electrostatic, seismic and thermal analyses. Finally, next procurement steps and definition of special tests to verify the key design choices are also reported.

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.

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.

MITICA is the full scale prototype of the ITER Heating Neutral Beam (HNB) currently under construction at the Neutral Beam Test Facility in Padova (Italy). In the ITER HNB, negative ions (H - /D - ) are produced in the Ion Source (IS) polarized to ground at - 1012 kV, then extracted and accelerated to ground at 1 MeV energy and fi nally neutralized. The complex power supply system feeding the IS includes two non-standard equipment, beyond the current industrial standard for insulation voltage level (-1 MVdc) and dimensions: 1) the High Voltage Deck1 (HVD1), a large air insulated Faraday cage (12.5 m (L) x 8.4 m (W) x 9.6 m (H)) hosting the IS power supplies and diagnostics; 2) the High Voltage Bushing Assembly (HVBA), a -1 MVdc air to SF6 Bushing which interfaces the HVD1 with the SF 6 insulated Transmission Line (TL) connecting the Acceleration Grid Power Supply system with the IS, carrying inside its High Voltage electrode all IS power and diagnostics con- ductors. The HVD1 and HVBA are installed inside a large HV hall. The main design choices leading to the fi nal hall layout, integrating the HV experimental equipment and the conventional building plants assuring the ne- cessary clearance required by such very high electric insulation level, are presented. Moreover, the paper reports on the manufacturing design developed by the supplier, the factory type tests to validate the design of the HVD1, the factory tests carried out on the HVBA and the on-site installation, commissioning and testing activities.

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

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 EU 1 MW, 170 GHz gyrotron with hollow cylindrical cavity have been designed within the European GYrotron Consortium (EGYC) in collaboration with the industrial partner Thales Electron Devices (TED) and under the coordination of Fusion for Energy (F4E). In the frame of the EU program, the short-pulse (SP) version of this tube was designed and manufactured by KIT in collaboration with TED. The experimental verification of the SP gyrotron prototype was successfully completed in 2015. The achieved experimental results show a very stable gyrotron operation with RF output power above 1 MW at good cavity interaction efficiency around 35%. The gyrotron was operated up to 10 ms pulse length; the nominal cavity mode TE32,9 has been excited at the frequency 170.1 GHz in agreement with the corresponding ITER specification. The Gaussian mode content of the output RF beam was about 98% and the total level of internal stray radiation was in the range of 2-3%. The manufacturing of the first industrial continuous-wave (CW) prototype gyrotron, based on the SP gyrotron design, was completed in November 2015 at TED. The tube was delivered to KIT and the experimental verification started in February 2016. Operation of the gyrotron with ms pulse duration resulted in stable excitation of the nominal mode at 170.22 GHz for a wide range of operating parameters and with a level of stray radiation comparable with the SP version. The maximum RF power achieved up to now in short-pulse operation at a level of 1 MW. The measured Gaussian mode content of the RF beam is 97%. Currently, further preparation for the CW operation of the gyrotron is in progress. (C) 2017 Published by Elsevier B.V.

The first series of experiments with the EU 1 MW, 170 GHz CW industrial prototype gyrotron1 for ITER have been completed at the KIT test facility at the end of 2016. Depressed collector operation with 180s pulses (pulse length limited by the HV power supply at KIT) at the Low Voltage Operating Point (I b ~45 A, V acc ~71 kV) resulted in 0.8 MW RF output power with 37 % total efficiency. Similar results were obtained for the High Voltage Operating Point (I b ~40 A, V acc ~80 kV) also. The experiments on the CW gyrotron with pulse duration of up to 1 hour will be continued in 2017 at SPC, Lausanne. In parallel, further experiments with the almost identical, modular short-pulse (SP) version of the gyrotron2, which delivered more than 1.2 MW, are currently ongoing at KIT in order to identify ways to further improve the performance of the CW prototype.

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

The EU 1 MW, 170 GHz gyrotron for ECH&CD on ITER has been developed by EGYC (European GYrotron Consortium) under the coordination of Fusion for Energy (F4E) and in cooperation with the industrial partner Thales Electron Devices (TED). The manufacturing of the first industrial continuous-wave (CW) prototype gyrotron, has been completed in November 2015 at TED. The tube was delivered to KIT and the experimental verification started in February 2016 in the short-pulse (SP) regime. Since October 2016 the gyrotron is in CW operation. In this work, details regarding the experiments at the KIT test-stand with pulses up to 180 s are presented.

The European 1 MW, 170 GHz CW industrial prototype gyrotron for ECRH&CD on ITER was under test at the KIT test facility during 2016. In order to optimize the gyrotron operation, the tube was thoroughly tested in the short-pulse regime, with pulse lengths below 10 ms, for a wide range of operational parameters. The operation was extended to longer pulses with a duration of up to 180 s. In this work we present in detail the achievements and the challenges that were faced during the long-pulse experimental campaign.

An industrial 1 MW, 170 GHz continuous-wave (CW) gyrotron prototype has been manufactured in 2015 at Thales. The physical design is provided by the European GYrotron Consortium (EGYC) and is supported by the construction and the measurement of a modular short-pulse (SP) prototype. In this presentation, the latest experimental results are discussed, with respect to the previous results acquired for the modular SP prototype gyrotron.

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.

The RF heating and current drive (H&CD) systems to be installed for the ITER fusion machine are the electron cyclotron (EC), ion cyclotron (IC) and, although not in the first phase of the project, lower hybrid (LH). These systems require high voltage, high current power supplies (HVPS) in CW operation. These HVPS should deliver around 50 MW electrical power to each of the RF H&CD systems with stringent requirements in terms of accuracy, voltage ripple, response time, turn off time and fault energy. The PSM (Pulse Step Modulation) technology has demonstrated over the past 20 years its ability to fulfill these requirements in many industrial facilities and other fusion reactors and has therefore been chosen as reference design for the IC and EC HVPS systems. This paper describes the technical specifications, including interfaces, the resulting constraints on the design, the conceptual design proposed for ITER EC and IC HVPS systems and the current status.

The RF heating and current drive (H&CD) systems that are to be installed in ITER during the construction phase, are the electron cyclotron (EC) and ion cyclotron (IC) systems. They are complex assemblies of high voltage power supplies (HVPS), RF generators, transmission lines and antennas. Their design and integration are constrained by many interfaces, both internal, between the subsystems, and external, with the other ITER systems. In addition, some components must be compatible with a nuclear environment and are classified as Safety Important Component. This paper describes the processes implemented in ITER to ensure proper integration.

The Electron Cyclotron system for ITER is an in-kind procurement shared between five parties and the total installed power will be 24 MW, corresponding to a nominal injected power of 20 MW to the plasma, with a possible upgrade up to 48 MW (corresponding to 40 MW injected). Some critical issues have been raised and changes are proposed to simplify these procurements and to facilitate the integration into ITER. The progress in the design and the integration of the EC system into the whole project is presented in this paper, as well as some issues still under studies and some recommendations made by external expert committees.

The electron cyclotron (EC) heating and current drive (H&CD) system developed for the ITER is made of 12 sets of high-voltage power supplies feeding 24 gyrotrons connected through 24 transmission lines (TL), to five launchers, four located in upper ports and one at the equatorial level. Nearly all procurements are in-kind, following general ITER philosophy, and will come from Europe, India, Japan, Russia and the USA. The full system is designed to couple to the plasma 20 MW among the 24 MW generated power, at the frequency of 170 GHz, for various physics applications such as plasma start-up, central H&CD and magnetohydrodynamic (MHD) activity control. The design takes present day technology and extends toward high-power continuous operation, which represents a large step forward as compared to the present state of the art. The ITER EC system will be a stepping stone to future EC systems for DEMO and beyond. The development of the EC system is facing significant challenges, which includes not only an advanced microwave system but also compliance with stringent requirements associated with nuclear safety as ITER became the first fusion device licensed as basic nuclear installations as of 9 November 2012. Since the conceptual design of the EC system was established in 2007, the EC system has progressed to a preliminary design stage in 2012 and is now moving forward toward a final design.

In the frame of a collaborative effort between the European GYrotron Consortium (EGYC), the European domestic agency Fusion for Energy (F4E) and the Industrial partner Thales Electron Devices (TED), the first continuous wave (CW) prototype of the EU gyrotron for the ITER Electron Cyclotron Heating system has been manufactured by TED and delivered to the Karlsruhe Institute of Technology (KIT) by end of 2015, where it will be tested up to pulse lengths of 180 s before being studied up to CW operation at Swiss Plasma Center (SPC, formerly CRPP) in Lausanne. The experimental campaign started early 2016. So far, the tube has been operated in ms pulses, allowing a detailed characterization and an extensive parametric study. It has been shown that the tube operates stably in the design mode up to the MW level and that the output beam profile has a high Gaussian content, compliant with the ITER requirements. This paper will report on the recent experimental results.

A 170 GHz, 2 MW, steady-state gyrotron is being developed in collaboration between European research associations and industries to be used for the electron cyclotron resonance heating (ECRH) system of ITER, the gyrotron is presently in the prototype state. EFDA entrusted ENEA with the design of a new power supply for the gyrotron's body; in which, the traditional power vacuum tubes are replaced with solid-state components (IGBT); furthermore ENEA had the task of developing the new body power supply (BPS), following the initial conceptual design. The BPS is formed by 50 identical bidirectional (in current) modules, connected in series; the BPS's control system regulates the output voltage by changing the modulation index. OCEM Spa, under ENEA's supervision, has manufactured the BPS. Tests, done at the start of this year, showed the substantial agreement between the initial technical specifications and the values achieved. Particularly the ramp-up/down-emergency time (50–10 μs), the accuracy (±0.5%), the sine wave modulation (45–20 kV at 5 kHz) and the maximum phase shift ( In the article, the new stair-way modulation (SWM) scheme will be sketched and the test results will be described in detail.

The activities in Europe towards the development of the EC power sources for ITER are centered on the development of a 170GHz, 2MW, CW coaxial cavity gyrotron of collector potential depressed (CPD) type. A gyrotron with a higher unit power than the ITER reference (I MW) would yield a reduction of the installation costs, a more compact launcher design and, if required, an increase of the power delivered through one port. Tests proving the principle were successfully performed on a short-pulse experimental gyrotron delivering up to 2.2 MW in single mode. Following this success, a coordinated and fully consistent programme of development has been launched. The first industrial 2 MW prototype is now at an advanced stage of construction. The associated superconductive magnet producing 6.86 T on the cavity axis is also being procured. Dummy loads suitable for short and CW operation are also part of the development effort. Finally, a new EC test facility, with the features necessary for the testing of the gyrotron up to full power in CW, has been established and includes a fully solid-state power supply system. (c) 2007 Published by Elsevier B.V.

The progress in the development of the European 170 GHz, 1 MW/CW gyrotron for electron cyclotron heating & current drive (ECH&CD) on ITER is reported. A continuous wave (CW) prototype is being manufactured by Thales Electron Devices (TED), France, while a short-pulse (SP) prototype gyrotron is in parallel under manufacture at Karlsruhe Institute of Technology (KIT), with the purpose of validating the design of the CW industrial prototype components. The fabrication of most of the sub-assemblies of the SP prototype has been completed. In a first step, an existing magnetron injection gun (MIG) available at KIT was used. Despite this non-ideal configuration, the experiments provided a validation of the design, substantiated by an excellent agreement with numerical simulations. The tube, operated without a depressed collector, is able to produce more than 1 MW of output power with efficiency in excess of 30%, as expected, and compatible with the ITER requirements. (C) 2015 Karlsruhe Institute of Technology. Published by Elsevier B.V. All rights reserved.

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.

An electron cyclotron system is one of the four auxiliary plasma heating systems to be installed on the ITER tokamak. The ITER EC system consists of 24 gyrotrons with associated 12 high voltage power supplies, a set of evacuated transmission lines and two types of launchers. The whole system is designed to inject 20 MW of microwave power at 170 GHz into the plasma. The primary functions of the system include plasma start-up, central heating and current drive, and magneto-hydrodynamic instabilities control. The design takes present day technology and extends towards high power CW operation, which represents a large step forward as compared to the present state of the art. The ITER EC system will be a stepping stone to future EC systems for DEMO and beyond. The EC system is faced with significant challenges, which not only includes an advanced microwave system for plasma heating and current drive applications but also has to comply with stringent requirements associated with nuclear safety as ITER became the first fusion device licensed as basic nuclear installations as of 9 November 2012. Since conceptual design of the EC system established in 2007, the EC system has progressed to a preliminary design stage in 2012, and is now moving forward towards a final design. The majority of the subsystems have completed the detailed design and now advancing towards the final design completion. (C) 2014 Elsevier B.V. All rights reserved.

EU is presently developing a 1 MW, 170 GHz cylindrical cavity gyrotron for ITER. This activity was initiated in 2008 as a risk mitigation measure in parallel with the development of a 2 MW coaxial-cavity gyrotron, and in 2012 became the main line. In the previous year, the scientific designs of the 1 MW gyrotron components, such as the electron gun, beam tunnel, cavity, quasi-optical output coupler and single-stage depressed collector have been finalized. In addition, in collaboration with the European industrial partner Thales Electron Devices (TED), after completion of the industrial design of the technological parts of the gyrotron, the procurement and manufacturing of subassemblies for a 1 MW CW prototype will start in 2014. Its specifications correspond to the ITER requirements.

Summary form only given. Europe is spending joint efforts to develop and to manufacture MW-level gyrotrons for electron cyclotron heating and current drive (ECRH&ECCD) of future plasma experiments. The two most important are the stellarator Wendelstein W7-X at Greifswald and the tokamak ITER at Cadarache. The construction of the 140 GHz, 1 MW CW gyrotrons for the 10 MW ECRH system of W7-X is proceeding well. It is expected that the next series tube will have finished final acceptance tests well in spring 2013. According to plan, production of the gyrotrons for W7-X will be finalized by 2014. This report will summarize the actual status. In parallel to the production and testing of the W7-X gyrotrons, the European GYrotron Consortium (EGYC) is presently developing the EU-1 MW, 170 GHz gyrotron for ITER. That design had been initiated in 2007 already, as a risk mitigation measure during the development of the advanced EU ITER-2 MW coaxial-cavity gyrotron. The basic idea of the EU ITER-1 MW design is to benefit from the experiences made during development and series production of the W7-X gyrotrons and of the experiences gained from the 2 MW coaxial-cavity gyrotron design. Preliminary designs of the cavity and the magnetron injection gun have been presented earlier. During 2012, the scientific design of the EU ITER-1 MW gyrotron components has been finalized. In collaboration with the industrial partner Thales Electron Devices, the industrial design of the technological parts of the gyrotron is being completed. A short pulse prototype gyrotron is under development to support the design of the CW prototype tube. The technological path towards the EU ITER-1MW gyrotron and the final design will be presented.

EU is developing a 1 MW cylindrical cavity gyrotron. In the last year the design of the components of the new gyrotron has been finalized while the technological design of the new tube has been defined. In the present paper, the main characteristics of the new EU gyrotron for ITER are presented.

___

The ITER electron cyclotron (EC) system having capability of up to 26 MW generated power at 170 GHz is being procured by 5 domestic agencies via 10 procurement arrangements. This implies diverse types of equipment and complex interface management. It also places a challenge on control system architecture to entertain the constraints of procurement slicing and meeting the overall functional requirement. The envisioned architecture is to use the local control units (supplied with each procurement) and a supervisory plant controller (by ITER). This offers a reliable control configuration for such delicate and complex EC plant system. The control system is envisioned to monitor the whole plant and perform automated tasks that are today performed via direct human intervention. For example, the automated gyrotron conditioning and active control of the EC plant to respond to requests from the plasma control system (PCS). This later aspect requires rapid shut down of the gyrotrons and power supplies, deviation of the actuators to direct the power from an equatorial to upper launcher and then restart of the power generation for rapid stabilization of the magneto hydrodynamic (MHD) instabilities that occur in high performance plasma operation. The plant controller will be designed for optimized performance with the PCS and the feedback control system used to actively control the power (with modulation capability up to 5 kHz) and launching direction for MHD stabilization. (C) 2011 ITER Organization. Published by Elsevier B.V. All rights reserved.

____

____

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 neutral beam (NB) system presents several challenges and a robust program is necessary in order to achieve the requirements within the tight constraints due to the ITER construction plan. The establishment of full scale NB test facilities (NBTF) has therefore become a centre piece of the international NB development strategy. This paper describes the status of the design activities that have been undertaken in Europe to develop the components for the heating NB injectors along with the main plans and results of the R&D activities. A description of the programme towards the establishment of the test facilities and the planned activities is also reported.

A preliminary gun design for the EU 170 GHz, 1 MW conventional cavity gyrotron, for the TE32,09 mode [1], will be presented. The ASG magnet of the EU 170 GHz, 2 MW coaxial cavity gyrotron for ITER is used for the required magnetic field. A parameterization of the diode gun has been defined. Using the Ariadne++ code [2] the geometry has been optimized with respect to several criteria, such as the small velocity spreads, the acceptable width of the beam radius at the cavity, etc. The neutralization effect has been taken into account using a simple model.

A collaborative effort between European research Associations and Thales Electron Devices (TED) has been launched by the European Fusion Development Agreement (EFDA) in 2003, aiming at the development of an industrial 170 GHz/2 MW/CW coaxial cavity gyrotron. The first prototype is expected to reach 2 MW/1s and is presently being tested in Lausanne at a dedicated test facility. The test facility has been designed to be flexible: allowing the possible commissioning of tubes with different characteristics, as well the tests of a version of the ITER upper launcher antenna at full performances. The test facility has been commissioned during the first test phases on the gyrotron which include: high voltage stand-off, coaxial insert alignment and cathode conditioning without depressed collector power supply at a slightly reduced electron beam power of 85 kV/78 A/2 ms. The short pulse (

In the reference ECH design, ITER requires a total of 20 MW/CW power at 170 GHz using gyrotrons with a unit power of 1 MW. A higher power per unit (2 MW/gyrotron) would result in a strong reduction of the cost of the whole ECRH system, and would also relax the room constraints on the launcher antenna design. The high power capability of coaxial cavity gyrotrons has been demonstrated with short pulse experiments at FZK. A collaborative effort between European research associations CRPP-EPFL, FZK and Thales Electron Devices (TED) has been launched by the European Fusion Development Agreement (EFDA) in 2003, aiming at the development of an industrial 170 GHz/2 MW/CW coaxial cavity gyrotron. The first prototype, although designed to be CW compatible is expected to reach 2 MW/ls and has been delivered by end of 2006. It will be tested in Lausanne, where a specially dedicated test facility has been built. The test facility has been designed to be flexible enough, allowing the possible commissioning of tubes with different characteristics, as well the tests of the launcher antenna at full performances. Initial experiments are planned for the end of the third quarter 2007.

The delivery of a first industrial prototype of a 170 GHz coaxial cavity gyrotron is expected in September 2006. RF tests with the gyrotron are foreseen to start in January 2007 when the superconducting (SC) magnet has been delivered. Experimental studies have been performed with a 170 GHz pre-prototype of the coaxial gyrotron in order to prove the design of main gyrotron components.

The European Fusion Programme included since the mid-1990s, the development of a full scale, full rating current commutating unit (CCU) to be used as a centre-piece of the ITER magnet protective circuits. The CCU is mainly composed of a by-pass switch (BPS) in parallel to a vacuum circuit breaker (VCB). Both full scale switches were subjected to extensive testing in conditions simulating the actual ITER operation. The operation in absence of the saturable reactor series connected in the ITER reference design was successfully assessed. Interruption tests at low current and full counter-pulse capacitor voltage, were also satisfactorily performed. The tests included an extensive and novel characterisation of the interrupting capability of the vacuum circuit breaker in the presence of an external magnetic field as in the actual location of installation in ITER.

In ITER, EC heating and current drive (H&CD) is foreseen not only as a principal auxiliary system for plasma heating and as assist for plasma start-up, but is considered essential in meeting the key requirement of neoclassical tearing mode (NTM) stabilisation, by localized current drive. In the reference ECH design, ITER requires a total of 20 MW/CW power at 170 GHz using gyrotrons with a unit power of 1 MW. A higher power per unit (2 MW/gyrotron) would result in a strong reduction of the cost of the whole ECRH system, and would also relax the room constraints on the launcher antenna design. In view of the capability of coaxial cavity gyrotrons demonstrated with short pulse experiments at FZK, the European Fusion Development Agreement (EFDA) has started in 2003 the development of an industrial 170 GHz 2 MW/CW coaxial cavity gyrotron, in a collaborative effort between European research associations CRPP/EPFL, FZK, TEKES and Thals Electron Devices (TED). The development plan includes three steps to reach successively 2 MW/1s, 2 MW/60s and finally 2 MW/CW operation. The procurement of the first prototype is in progress and it scheduled to be delivered during the first quarter of 2006. The experimental tests of the prototypes will be carried out at CRPP/EPFL, where an ITER relevant test facility is presently under construction and will be achieved during the second half of 2005. The test facility is designed to be flexible enough, allowing the possible commissioning of tubes with different characteristics, as well the tests of the launcher antenna at full performances.

A 170 GHz, 2 MW, steady state gyrotron is being developed in a collaboration between European research associations and industries to be used for the electron cyclotron resonance heating (ECRH) system of ITER, the next step in nuclear fusion research. The gyrotron is presently in the prototype state. For the test facility dedicated to the testing of this gyrotron, new power supplies need to be procured. A feasibility study was performed aimed at using new designs based on present day techniques and at performing and constructing a prototype for further testing and evaluation. The prototype BPS (body power supply) which is connected to the body of the gyrotron is now in an advanced stage of construction and the first test results are in close agreement with the design objectives of the feasibility study

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

The European Gyrotron Consortium (EGYC) is developing the European 1 MW, 170 GHz Continuous Wave (CW) industrial prototype gyrotron for ITER in cooperation with Thales Electron Devices (TED) and Fusion for Energy (F4E). This conventional, hollow-cavity gyrotron, is based on the 1 MW, 170 GHz Short-Pulse (SP) modular gyrotron that has been designed and manufactured by the Karlsruhe Institute of Technology (KIT) in collaboration with TED. Both gyrotrons have been tested successfully in multiple experiments. In this work we briefly report on the results with the CW gyrotron at KIT and we focus at the experiments at the Swiss Plasma Center (SPC). In addition, we present preliminary results from various upgrades of the SP tube that are currently tested at KIT.

Europe is devoting significant joint efforts to develop and to manufacture MW-level gyrotrons for electron cyclotron heating and current drive of future plasma experiments. The two most important ones are the stellarator Wendelstein W7-X at Greifswald and the Tokamak ITER at Cadarache. While the series production of the 140 GHz, 1 MW, CW gyrotrons for the 10-MW electron cyclotron resonance heating system of stellarator W7-X is proceeding, the European GYrotron Consortium is presently developing the EU-1 MW, 170 GHz, CW gyrotron for ITER. The initial design had already been initiated in 2007, as a risk mitigation measure during the development of the advanced ITER EU-2-MW coaxial-cavity gyrotron. The target of the ITER EU-1-MW conventional-cavity design is to benefit as much as possible from the experiences made during the development and series production of the W7-X gyrotron and of the experiences gained from the earlier EU-2-MW coaxial-cavity gyrotron design. Hence, the similarity of the construction will be made visible in this paper. During 2012, the scientific design of the ITER EU-1-MW gyrotron components has been finalized. In collaboration with the industrial partner Thales electron devices, Velizy, France, the industrial design of the technological parts of the gyrotron is being completed. A short-pulse prototype is under development to support the design of the CW prototype tube. The technological path toward the EU ITER-1 MW gyrotron and the final design will be presented.

The European 1-MW, 170-GHz continuous wave industrial prototype gyrotron for electron cyclotron resonance heating and current drive on international thermonuclear experimental reactor was during 2016 under test at the Karlsruhe Institute of Technology (KIT) test facility. In order to optimize the gyrotron operation, the tube was at first thoroughly tested in the short-pulse regime, with pulses that did not exceed 10 ms, for a wide range of operational parameters. Then, and after proper conditioning of the tube, the operation was extended to longer pulses with duration up to 180 s, which is the maximum pulselength possible at the KIT test facility. In this paper, we present in detail the achievements of the long-pulse experimental campaign.

Reference CRPP-CONF-2009-009View record in Web of Science Record created on 2009-01-23, modified on 2017-05-12

The EU is working towards providing 2 MW, coaxial-cavity, CW, 170 GHz gyrotrons for ITER. Their design is based on results from an experimental pre-prototype tube in operation at FZK for several years, having a pulse length of several milliseconds. The first industrial prototype tube is designed for CW operation, but, in a first phase, will be tested out to Is at the European Gyrotron Test Facility in Lausanne, Switzerland as part of a phased testing/development program (1s, 60 s, CW). It is known that RF beam profile shaping, stray radiation handling, and collector cooling at these high power levels are three issues for the gyrotron. The gyrotron, magnet and body power supply have been delivered and successfully installed at the test stand, hosted by the CRPP. The main high voltage power supply delivery is delayed, so one of the power supplies dedicated to 3 of 9 gyrotrons in the TCV EC system is being used as a backup power source (all 3 TCV power sources can be interfaced with the test stand). Cathode conditioning began in November 2007 followed by collector conditioning in December. Parasitic low frequency oscillations have not hindered operation, and the tests have progressed to conditioning out to 0.14 s pulses by March 2008. During this period, the performance concerning microwave generation has been characterised and the RF beam profile has been measured at several planes to allow reconstruction of the phase and amplitude profile at the gyrotron window and to provide the necessary information permitting proper alignment of the compact RF loads prior to pulse extension. The power will be measured, according to the pulse length, using either a very-short pulse (