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.

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The 500-keV negative-ion based neutral beam injector for JT-60U started operation in 1996. The availability of the N-NBI system has been improved gradually through modifying ion source and optimizing its operation parameters. Recently, the extension of the pulse duration up to 30 sec has been intended to study quasi-steady state plasma on JT-60U. The most serious issue is to reduce the heat load on the grids for long pulse operation. Two modifications have been proposed to reduce the heat load. One is to suppress the beam spread which may caused by beamlet-beamlet interaction in the multi-aperture grid due to the space charge force. Indeed, the investigation of the beam deflection, which was measured by the infrared camera on the target plate set 3.5 m away from the grid, indicates the beam spread is in proportion to the current density. Thin plates were attached on the extraction grid to modify the local electric field. The plate thickness was optimized to steer the beamlet deflection. The other is to reduce the stripping loss, where the electron of the negative ion beam is stripped and accelerated in the ion source and then collides with the grids. The ion source was modified to reduce the pressure in the accelerator column to suppress the beam-ion stripping loss. Up to now, long pulse injection of 17 sec for 1.6 MW and 25 sec for ~1 MW has been obtained by one ion source with these modifications.

Hydrogen negative ion beams of 507 keV, 1 A and 486 keV, 2.8 A have been successfully produced in the JT-60U negative ion source with a three-stage accelerator by overcoming a poor voltage holding of the accelerator with large-size grids of ∼2 m2. This is the first result of H− beam acceleration up to 500 keV at a high current of over 1 A. In order to improve the voltage holding capability, the breakdown voltages of the large-size grids and small-size electrodes with uniform and locally strong electric fields were examined by changing the gap length. It was found that the voltage holding of the large-size grids was below half of that of the small-size electrodes with a uniform electric field which was used in the design of the accelerator. This degradation was found to be caused by the local electric field concentrations in addition to the size. Based on the results of the voltage holding tests and beam optics calculations, the gap lengths of the large-size grids were tuned to have a capability to sustain 600 kV. As a result, the gap tuning realized stable voltage holding during beam accelerations without significant degradations of the beam optics and stripping loss. These results indicated that stable 500 keV beam accelerations required for JT-60SA are feasible and this gap tuning is also applicable for the design of ITER accelerator.

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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 Nuclear Fusion Progress in the realization of the PRIMA neutral beam test facility

Major issues of NBI power supplies are a high-speed switching, regulation and transmission of dc ultra high voltage, and suppression of surge energy input to the beam source at breakdown. A GTO(gate turn off thyristor) inverter type power supply where the control is performed at low voltage ac side was designed for the ITER NB. Based on the remarkable progress of a high power IEGT (injection enhanced gate transistor), the design of the inverter has been modified to increase an efficiency and compactness using such new elements. A power loss in the inverter is reduced to be 30% of the GTO inverter system. For the transmission line of the dc UHV with intermediate voltages, a disk shape multi-conductor bushing with a transmission line test chamber has been developed. Dimensions of the bushing are 1.8 m in diameter and 140 mm in thickness at the edge. Electric fields at the conductor surface and insulator surface were designed to be lower than 5 kV/mm and 7 kV/mm, respectively. An electric field at the bottom of the ground potential outer conductor was designed to be lower than 1.2 kV/mm to prevent particle levitation which triggers breakdowns. The prototype transmission line has passed the standard impulse test up to 1,300 kV. A dc UHV up to 1,175 kV was successfully sustained for 300 s. To prevent the electric damage of the beam source at the breakdown, core snubbers using Fe-based nanocrystalline soft magnetic materials are adopted to dissipate the surge energy. 1) Hitachi Ltd. , 2) Toshiba Co.