E. Di Pietro

Design of neutral beam system for ITER-FEAT☆

Abstract The neutral beam (NB) system in ITER-FEAT provides heating and current drive (H&CD) by two NB injectors, each delivering 16.7 MW of D 0 beam to the plasma at 1 MeV. The NB system retains the basic concept of the ITER 1998 design, but there are certain modifications that will be described: the beam transmission is improved by a four beam channel design of the neutralizer and the RID. Also the layout of the NB injector integrated in ITER allows both on- and off-axis current drive. The improved performance of the NB system is discussed from the system efficiency and the current drive capability points of view.

ITER neutral beam system

Since the main features of the design of the neutral beam (NB) system for the International Thermonuclear Experimental Reactor (ITER) were first reported, integration with the tokamak and with the rest of the plant has been the main priority. Moreover, operational requirements and maintainability have been considered in the evolution of the design. Each of the three NB injectors is connected to the tokamak vacuum vessel with the NB duct on an equatorial port. The article describes the integration of the NB port/duct with the blanket, the vacuum vessel, the toroidal field and poloidal field coils, the cryostat and the bioshield. Two main design modifications are reported. The insulation of the source, originally done with compressed gas, is now achieved with vacuum to limit the power losses caused by the radiation induced conductivity. Large cylindrical insulators are still required but their inner diameter has been reduced from 2.7 to 1.8 m. The improvements on the compensation system needed to reduce the magnetic field in the NB volume are also described. Finally, the progress in R&D for the ITER NB system is reported, including an overview of the achievements in the critical areas of negative ion production at high current density (tests of a large size, low pressure, steady state caesiated ion source), acceleration up to 1 MV (tests of two alternative accelerator concepts) and neutralization (tests of an experimental plasma neutralizer to investigate it as an alternative to the gas target neutralizer).

JT-60SA superconducting magnet system

The most distinctive feature of the superconducting magnet system for JT-60SA is the optimized coil structure in terms of the space utilization as well as the highly accurate coil manufacturing, thus meeting the requirements for the steady-state tokamak research: a conceptually new outer inter-coil structure separated from the casing is introduced to the toroidal field coils to realize their slender shape, allowing large-bore diagnostic ports for detailed plasma measurements. A method to minimize the manufacturing error of the equilibrium-field coils has been established, aiming at the precise plasma shape/position control. A compact butt-joint has been successfully developed for the Central Solenoid, which allows an optimized utilization of the limited space for the Central Solenoid to extend the duration of the plasma pulse.

Completion of TF Strand Production and Progress of TF Conductor Manufacture for JT-60SA Project

In the framework of the JT-60SA project, aiming at upgrading the present JT-60U tokamak, Europe, as part of its in-kind contribution within the Broader Approach agreement, will provide the toroidal field (TF) magnet system. For this purpose, Fusion for Energy is committed to procure about 29 km of TF conductor. The TF conductor is cable-in-conduit type and includes 486 strands (2/3 NbTi-1/3 copper) wrapped with a thin stainless steel foil and compacted into a rectangular stainless steel jacket. The procurement is split into two main contracts: one for TF strand manufacturing and the other for TF conductor cabling and jacketing. TF strand manufacture was completed while the TF conductor one is being finished. In the present paper, we draw an overview of both productions emphasizing on the quality control (QC) approaches and on aspects relevant to risk management of the JT-60SA tokamak operation. For the NbTi strand, the complete production overview is provided including extensive statistical considerations on NbTi strand critical performances ( IC, TCS). For the TF conductor, the overview also deals with collected results from acceptance tests and peripheral tests led for consolidating the QC (hydraulic tests, nondestructive examination, full-size sample cold tests in SULTAN facility).

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

The article reports on the progress made by the ITER European Home Team in strong interaction with the ITER Joint Central Team and the Japan Atomic Energy Research Institute regarding several key aspects of the beam source for the ITER injectors: (1) Integration of the SINGAP accelerator into the ITER injector design. This is a substantially simpler concept than the multiaperture, multigap (MAMuG) accelerator of the ITER NBI reference design that has the potential for significant cost savings and that avoids some of the weaknesses of the reference design such as the need for intermediate high voltage potentials from the high voltage power supply and pressurized gas insulation. (2) High energy negative ion acceleration using a SINGAP accelerator. (3) Long pulse (i.e. >1000 s) negative ion source operation in deuterium. (4) 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.

JT-60SA TF Magnets Industrial Production Status at Alstom

The general design of the JT-60SA TF system was defined by all the participants in the project (CEA, ENEA, F4E); the detailed design was issued by the voluntary contributors. For the French part, including the procurement of 9 of the 18 TF winding packs and their integration in the casings, an industrial contract was signed with Alstom (France). After the first phases, including manufacturing flow definition, manufacturing drawings and QA documentation, critical processes qualification, as well as procurement and commissioning of the tooling, the production was started in January 2014. After solving the setting issues linked to industrialization of complex tooling such as the winding machine or impregnation mold, the winding operations, as well as the joint area manufacture and the impregnations, were led. The first winding pack insertion process, as well as casing welding using an industrial robot, was performed during the year 2015 with the target to deliver the first coil to the cold test facility before the end of the year. Of course, in parallel, the successive coils manufacturing was engaged. This paper presents the status of the production with a focus on the issues encountered on the first coil and the production objectives.

Experimental and Analytical Approaches on JT—60SA TF Strand and TF Conductor Quality Control During Qualification and Production Manufacture Stages

In the framework of the JT-60SA project, aiming at upgrading the present JT-60U tokamak, Europe as part of its in-kind contribution within the Broader Approach agreement, will provide the full Toroidal Field (TF) magnet system. For this purpose, Fusion for Energy is committed to procure about 27 km of TF conductor. The TF conductor is cable-in-conduit type and includes 486 strands (2/3 NbTi-1/3 copper) wrapped with a thin stainless steel foil and embedded into a rectangular stainless steel jacket. The procurement is split into two main contracts: one for strand manufacturing and the other for cabling and jacketing. After having successfully passed the qualification stage, strand and conductor are now in mass production stage. In the present paper, we emphasize on the scientific and quality control approaches of both TF strand and TF conductor productions. For the NbTi strand, the focus is on design and measurement of performance critical aspects (CuNi barrier design optimization, AC losses, TCS performance, etc.). For Cu strand the main issue is RRR. For the TF conductor, the topics presented cover hydraulic correlation validation measurements (pressure drop tests), geometrical advanced controls, and electromagnetic behavior (hotspot checks and mainly the experimental results of the TF conductor full-size samples tested in SULTAN facility). The article presents comparisons of numerical simulation and experimental results used to confirm the design and the manufacturing processes.

Conceptual design and integration of a diagnostic neutral beam in ITER

In ITER, one of the key issues to achieve 400 s driven-burn operation at Q about 10 (Technical Basis for the ITER-FEAT Outline Design-Section I-3.2.2.IAEA) is helium ash accumulation. As a result, the real-time measurement of the thermalised helium density profile in the confinement region is of fundamental importance. This paper outlines the design of a modulated, 100 keV, hydrogen, diagnostic neutral beam (DNB), together with preliminary calculations of the performance for the measurement of helium ash by charge exchange recombination spectroscopy (CXRS). The DNB uses a negative ion beam as the primary beam to achieve a required performance with acceptable system efficiency (a higher neutralisation efficiency and smaller beam divergence). This uses the same negative ion source as the ITER HC as a consequence, the DNB has to share the vacuum vessel access through a single port with one of the H&CD injectors

The high heat flux components for ITER neutral beam system

Abstract All the alternatives considered for the reduced technical objectives/reduced cost international thermonuclear experimental reactor (RTO/RC ITER) foresee, as one of the methods for additional heating and current drive, the use of neutral injection. Two atomic deuterium beams, with 1 MeV energy and each with at least 16.7 MW power, are the reference for the engineering design of the system. The design of NB system developed for the ITER final design report (FDR) [1] is being modified to comply with the reduced size of the machine and to incorporate, if possible, some design improvements and simplifications. The analyses of overall power distribution and the maximum power densities on each of the beamline components have been updated. To obtain 16.7 MW neutral beam power, 35 MW ion beam power at the exit of the accelerator is necessary. Therefore, about 18 MW are deposited on the beamline components. Moreover, during the commissioning of the injectors, and the high voltage conditioning of the beam source, the full beam power is dumped and measured on a calorimeter. The beamline requires actively cooled high heat flux (HHF) components to exhaust high power density (up to 15 MW/m2) in steady state conditions. The operative life of this components (30 000 beam-on/beam off cycles, without replacement) requires, primarily, the verification of thermal fatigue safety margins. This paper describes the most relevant aspects of the mechanical design of the HHF components, focusing on the thermal and mechanical verifications.

Status of the JT-60SA magnet system

The JT-60SA experiment will be the worlds largest superconducting tokamak when it is assembled in 2019 in Naka, Japan (R=3m, a=1.2m). The superconducting magnet system includes 18 D-shaped toroidal field coils, each 7m high and 4.5m wide, 6 pulsed equilibrium field coils up to 12m in diameter and 4 central solenoid modules. Manufacturing of the superconducting magnets for JT-60SA is well established in Japan and in Europe. Conductor manufacturing is almost complete, half of the superconducting coils have been wound and the first cold test results for production coils will be available later in 2015. Challenges remain to integrate the coils with their mechanical structures and to assembly them into the tokamak.