The European contribution to the development of the ITER NB injector (original) (raw)

European programme towards the 1MeV ITER NB injector

Fusion Engineering and Design, 2009

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.

EU development of the ITER neutral beam injector and test facilities

2013

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

The European contribution to the development of the ITER NB injectorDisclaimer

Fusion Engineering and …, 2011

This paper reviews the ongoing design, R&D and procurement activities, mostly conducted within the ITER framework, ongoing in Europe under the coordination of Fusion for Energy (F4E), in cooperation with the European Fusion Associations and aimed at the establishment of the ITER Heating Neutral Beam (HNB) system.

Overview of the design of the ITER heating neutral beam injectors

New Journal of Physics, 2017

The heating neutral beam injectors (HNBs) of ITER are designed to deliver 16.7 MW of 1 MeV D 0 or 0.87 MeV H 0 to the ITER plasma for up to 3600 s. They will be the most powerful neutral beam(NB) injectors ever, delivering higher energy NBs to the plasma in a tokamak for longer than any previous systems have done. The design of the HNBs is based on the acceleration and neutralisation of negative ions as the efficiency of conversion of accelerated positive ions is so low at the required energy that a realistic design is not possible, whereas the neutralisation of H − and D − remains acceptable (≈56%). The design of a long pulse negative ion based injector is inherently more complicated than that of short pulse positive ion based injectors because:

Status of the ITER heating neutral beam system

Nuclear Fusion, 2009

The ITER neutral beam (NB) injectors are the first injectors that will have to operate under conditions and constraints similar to those that will be encountered in a fusion reactor. These injectors will have to operate in a hostile radiation environment and they will become highly radioactive due to the neutron flux from ITER. The injectors will use a single large ion source and accelerator that will produce 40 A 1 MeV D− beams for pulse lengths of up to 3600 s. Significant design changes have been made to the ITER heating NB (HNB) injector over the past 4 years. The main changes are: Modifications to allow installation and maintenance of the beamline components with an overhead crane. The beam source vessel shape has been changed and the beam source moved to allow more space for the connections between the 1 MV bushing and the beam source. The RF driven negative ion source has replaced the filamented ion source as the reference design. The ion source and extractor power supplies w...

European contributions to the beam source design and R&D of the ITER neutral beam injectors

Nuclear Fusion, 2000

The paper reports on the progress made by the European Home Team in strong interaction with the ITER JCT and JAERI regarding several key aspects of the beam source for the ITER injectors:-integration of the SINGAP accelerator into the ITER injector design. This is a substantially simpler concept than the MAMuG accelerator of the ITER NBI 'reference design', which has potential for significant cost savings, and which avoids some of the weaknesses of the reference design such as the need for intermediate high voltage potentials from the HV power supply and pressurised gas insulation.-high energy negative ion acceleration using a SINGAP accelerator-long pulse (i.e. >1000 s) negative ion source operration in deuterium-RF source development, which could reduce the scheduled maintenance of the ITER injectors (as it uses no filaments), and simplify the transmission line and the auxiliary power supplies for the ion source

Status of the ITER neutral beam injection system (invited)

Review of Scientific Instruments, 2008

The ITER neutral beam (NB) injectors are the first injectors that will have to operate in a hostile radiation environment and they will become highly radioactive due to the neutron flux from ITER. The injectors will use a single large ion source and accelerator that will produce 40 A 1 MeV Dbeams for pulse lengths of up to 3600 s. Significant changes have been made to the ITER heating NB injector (HNB) over the past 4 years. The main changes are: o Modifications to allow installation and maintenance of the beamline components with an overhead crane. o The RF driven negative ion source developed by IPP Garching has replaced the filamented ion source from JAEA, Naka as the reference design. o The ion source and extractor power supplies will be located in an air insulated high voltage (-1 MV) deck located outside the tokamak building instead of inside an SF 6 insulated HV deck located above the injector. The development of the ITER accelerators and ion sources has been carried out on relatively low powered test stands, making impossible the full demonstration of the ITER requirements. Padua Research on Injectors with Megavolt Acceleration (PRIMA, ex-NBTF) will be built to allow the R&D necessary to finalise the development of the full power system

The Full-Size Source and Injector Prototypes for ITER Neutral Beams

Plasma and Fusion Research, 2016

The development of the NBI systems for ITER requires unprecedented parameters (40 A of negative ion current accelerated up to 1 MV for one hour) so that a test facility is in the final phase of construction at Consorzio RFX (Padova, Italy), housing two experiments. A full-size negative ion source, SPIDER, aims at demonstrating the creation and extraction of a D-/H-current up to 50/60 A on a wide surface (more than 1 m 2) with uniformity within 10 %. The second experimental device is the prototype of the whole ITER injector, MITICA, aiming to develop the knowledge and the technologies to guarantee the successful operation of the two injectors to be installed in ITER, including the capability of 1 MV voltage holding at low pressure. The key component of the system is the beam source, whose design results from a trade-off between requirements of the optics and real grids with finite thickness and thermo-mechanical constraints due to the cooling needs and the presence of permanent magnets. Numerical simulations are a necessary supplement to the experimental effort to optimise the accelerator optics and to estimate heat loads and currents on the various surfaces. In this paper the main requirements for ITER NBI will be discussed. The design and the status of the main components and systems will be described. Particularly a review of the accelerator physics and a comparison between the designs of the SPIDER and MITICA accelerators are presented. Complex network theory will be applied to the NBI system in order to identify the hidden functional relationships and the most important parameters for the operation.

Heating Neutral Beams for ITER: Present Status

IEEE Transactions on Plasma Science, 2016

An auxiliary heating power of >70 MW is envisaged at ITER in order to obtain the plasma temperatures and plasma profiles required to achieve Q > 10 for 400 s (inductive ELMy H-mode) and Q = 5 for a pulse duration of 3600 s (noninductive discharge), for which a potential upgrade may be necessary. In addition to providing the desired heating, the systems are also expected to drive current, tailor the plasma profile, and control the plasma instabilities. It is difficult to realize such a broad range of functionalities through a single system. As a result, three systems are planned during the first operational phase of ITER, which include the neutral beam (NB), electron cyclotron (EC), and ion cyclotron (IC) systems. While the NB systems are expected to deliver 33 MW of heating power to the plasma and drive current through it by the use of two NB injectors, the IC and the EC systems deliver 20 MW each. IC covers a wide variety of heating and current drive schemes and EC heats the electrons providing local heating and current drive, which can be steered across the plasma cross section. These systems, operating in a hostile radiative environment, become radioactive due to the neutron flux from ITER. Coupled to this are the features of long operating pulse lengths at high powers, reduced maintainability, and increased remote handling requirements. As a result, these devices are challenging to realize compared with their present operational counterparts worldwide. In order to help mitigate the risks involved with the manufacturing, setting up, and operation of these systems at ITER, extensive prototyping and research and development (R&D) activities are underway at various laboratories of the participating domestic agencies (DAs) for each of these systems. This paper provides a brief description of the requirements of each of the three heating systems for the various plasma scenarios foreseen for ITER operation, their functional and technical advantages, the various developments over the period of time, and the present status of the prototype and R&D activities underway to realize these systems and the overall development schedule. agency (USDA), and Russian domestic agency (RFDA). This paper provides a brief overview of the requirements, development efforts, and the present status of the EC, IC, and NB systems planned for ITER.