Overview of the iter negative-ion-based neutral beam injector and its development

The ITER fusion device is intended to demonstrate the viability of magnetically confined deuterium-tritium plasma as an energy source. One of the principal methods of heating and driving current in the plasma will be energetic beams of neutral atoms of D° at up to 1 MeV and of H° at up to 870 KeV, with a total injectable neutral beam power after transit through the neutralizer and downstream beamline elements of 16.5 MW from each ion source for pulse lengths of up to 3600 seconds. These requirements far exceed those of any previous positive or negative ion source, thus spawning a substantial development program to ensure that they will be met with a robustly reliable system. The ion source will consist of a large plasma expansion region fed by 8 RF driver units, and will be cesiated to enhance surface production of negative ions, followed by a multi-aperture multi-grid extractor and electrostatic accelerator. The plasma portion of the source is derived from a line of RF sources developed at IPP Garching, 1 and the extractor/accelerator from development work at JAEA 2 , with the integrated design of the ITER source being done at Consorzio RFX. Unlike the first generation of high power negative hydrogen ion isotope sources, the ITER source will have the major advantage of a succession of progressively more comprehensive test facilities, culminating in a full power and pulse length test bed at Consorzio RFX. This talk will discuss the major beamline components, including the ion source and accelerator, the neutralizer cell that converts a portion of the negative ions to neutrals, the residual ion deflection system, and the tokamak field compensation system. Some remaining physics and engineering issues, along with their expected resolution, will be discussed, as well as the development and testing strategy.