V.M. Kulygin

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.

The EPSILON experimental pseudo-symmetric trap

Within the framework of the Adaptive Plasma Experiment (APEX) conceptual project, a trap with closed magnetic field lines, the Experimental Pseudo-Symmetric Closed Trap (EPSILON), is examined. The APEX project is aimed at theoretical and experimental development of the physical foundations for a steady state thermonuclear reactor designed on the basis of an alternative magnetic trap with tokamak-like large β plasma confinement. A discussion is given of the fundamental principle of pseudo-symmetry, which a magnetic configuration with tokamak-like plasma confinement should satisfy. Examples are given of calculations in the paraxial approximation of pseudo-symmetric curvilinear elements with a poloidal modulus B isoline. The EPSILON trap, consisting of two direct axisymmetric mirrors linked by two curvilinear pseudo-symmetric elements, is considered. To increase the equilibrium β, the plasma currents are short-circuited within curvilinear equilibrium elements. An untraditional scheme of MHD stabilization for a trap with closed field lines by use of axisymmetric mirrors with a divertor is analysed. The experimental installation EPSILON-One Mirror Element (OME), which is under construction for experimental investigation of stabilization by divertor, is discussed. The opportunity for applying the ECR method of plasma production in EPSILON-OME in conditions of high density and low magnetic field is examined.

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).

Neutral beams for the International Thermonuclear Experimental Reactor

Receiving higher emphasis on the neutral beam (NB) off-axis current drive, the NB system is being highlighted for the steady state operation of the International Thermonuclear Experimental Reactor (ITER). To fulfill the physics requirement of heating and current drive, the NB system delivers ∼50 MW of D0 beams at 1 MeV into the ITER plasmas. The NB injector was designed so as to minimize the axial length, to avoid cost impact on the building. It was estimated by nuclear analyses that the insulation gas around the beam source would cause radiation induced conductivity, which would result in a power dissipation of >100 kW in the gas itself. As a result the present design utilizes vacuum insulation around the beam source. Since the vacuum pressure inside/outside the beam source ranges 10−1–10−2 Pa, both gas (glow) and vacuum arc discharges are taken into account in the design.

The next step in the development of a negative ion beam plasma neutralizer for ITER NBI

Deuterium atom beam injectors developed for ITER plasma heating and current drive are based on negative ion acceleration and further neutralization with a gas target. The maximal efficiency of the gas stripping process is 60%. The replacement of the gas neutralizer by a plasma one must increase the neutral yield to 80%. An overview of experimental studies of microwave discharges in a multicusp magnetic system chosen as the base device for plasma neutralizer (PN) realization and design development of PNs for ITER neutral beam injectors (NBIs) is presented. The experimental results achieved with the PN model PNX-U are discussed. Plasma confinement, gas flows and ionization degree were investigated. High density plasmas with ne ~1018 m-3 with low electron and ion temperatures ( ≈ 5-6 eV) and a high ionization degree (not less than 40%) at its centre have been generated in operation with argon.

General Design of the Neutral Beam Injection System and Integration with ITER

The main technological aspects of the neutral beam (NB) injection system design are described and the constraints and requirements that derive from the integration in the overall reactor design are discussed.

The design of the high heat flux components of the ITER neutral beam injection system

Abstract The ITER neutral beam injection system is designed to deliver a total of 50 MW additional heating power to plasma. The system consists of three units, each delWering 16.7 MW. The power balance of each injector shows that to obtain a useful plasma heating of 16.7 MW an electrical input power of approximately 50 MW is used, the balance being deposited on the various components of the beam source and beam line. The paper describes the design solution utilized for the high heat ftux components and discusses the most relevant aspects of the thermo-mechanical and thermo-hydraulic design.