The heating neutral beam (HNB) systems at ITER are designed to inject a total of 33 MW of either 1 MeV D0 or 870 keV H0 beams into the ITER plasma using two injectors with a possible addition of a third injector later to increase the injected power to ∼50 MW. The injectors become radioactive due to the neutron flux from ITER and, in order to avoid the resulting complex remote maintenance, the design, choice of materials and the manufacturing process of each component of the injector is, wherever possible, such that they survive the life time of ITER. To ensure a smooth operational phase of neutral beams at ITER a neutral beam test facility (NBTF) is under construction at Consorzio RFX, Padova, (hereinafter referred to as RFX), which consists of 2 test beds, the 100 kV “SPIDER”, and a 1 MV “MITICA” facilities, which will be used to optimize the source operation for H and D beams. MITICA is essentially a full scale ITER prototype injector for the ITER beam parameters. The manufacturing and operation of the facility will allow validation of the operational space of the injectors and provide valuable information about the manufacturing processes applicable to HNB components. Operation of the two facilities is expected to begin in 2016 and 2019 respectively. Currently experiments on the ELISE facility with a half ITER sized RF beam source are underway. ITER relevant parameters for the H beams have almost been achieved. Efforts are underway to optimise the same with D beams. The experimental database from ELISE will be an important input for establishing the ITER relevant parameter space on the SPIDER source. This paper discusses the present status of the design and development of the injectors for ITER and the progress on the test facilities.

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 HV bushing has a function of a bulk head between insulation gas (SF6) of 0.6 MPa and vacuum, which is needed to insulate - 1 MV in vacuum in ITER heating neutral beam (NB) injection system. Mechanical analyses were carried out to verify the mechanical integrity of the HV bushing composed of five-stage double-layered insulator columns with large brazed ceramic ring and fiber reinforced plastic (FRP) ring of which non-metallic material is required to satisfy a safety factor of >= 3.5. As for the FRP ring required sustaining the pressure load, seismic load and dead weight, it was confirmed that isotropic fiber cloth FRP rings having sufficient strength against shear stress should be used. The shape and fixation area of the Kovar sleeve were modified to lower the stress at the brazing area having vacuum leak tightness and pressure tightness against the air filled at 0.6 MPa. As a result, a design of the insulator for the HV bushing was established satisfying the requirement. (C) 2013 Elsevier B.V. All rights reserved.

The ITER is an international project which aims to develop an experimental reactor as a step to realize fusion energy. To inject the 1 MeV D 0 neutral beams of 33 MW for heating and current drive in ITER plasma, D - ions are accelerated by dc -1 MV. The accelerator consists of five acceleration stages, and each acceleration gap has to withstand 200 kV. Gaps between the accelerator and the vacuum vessel whose potential is ground must also sustain voltages up to 1 MV. Key issues are high voltage vacuum insulation and design studies have been performed. A minimum gap length of > 900 mm was selected as 1 MV insulation design criteria. Development of a high voltage bushing (HVB) which is a bulkhead and a feed-through between gas insulated HV transmission line and the beam source in vacuum is ongoing. The HVB consists of a stack of five large bore ceramic rings (1.6 m diameter), each 0.29 m in height. Five stage insulation concept was applied to both the acceleration gap and the HVB for better voltage holding with multi shorter gaps. R&D for insulation of one stage ceramic with screens was conducted and 5 hr voltage holding of dc 200kV (rated voltage) was confirmed.

Research activities of the Japanese tokamaks JT-60U, JFT-2M, and TRIAM-1M are described. The recent JT-60 program is focused on the establishment of a scientific basis of advanced steady-state operation. Plasma performance in transient and quasi steady states has been significantly improved, utilizing reversed shear and weak shear (high-beta(p)) ELMy H-modes characterized by both internal and edge transport barriers and high bootstrap current fractions. Development of each key issue for advanced steady-state operation has also been advanced. Advanced and basic research of JFT-2M has been performed to develop high-performance tokamak plasma as well as the structural material for a fusion reactor. Toroidal field ripple reduction with ferritic steel plates outside the vacuum vessel is successfully demonstrated. No adverse effects to the plasma were observed with poloidal fields inside the vacuum vessel (partial covering). Preparation is in progress for full-scale testing of the compatibility of the ferritic steel wall (full covering) with plasma. A heavy ion beam probe has been installed to study H-mode plasmas. Compact toroid (CT) injection experiments are performed, showing deep CT-penetration into the core region of the H-mode. The TRIAM project has investigated steady-state operation and high-performance plasma of a tokamak with the high toroidal magnetic field superconducting tokamak. Four important contributions in the fields of fusion technology of superconducting tokamaks, steady-state operation, high-performance plasma, and startup of plasma current without the assistance of center solenoid coils have been achieved on TRIAM-1M, especially regarding steady-state operation by realization of a discharge for >3 h.

The performance of a 500 keV negative-ion based neutral beam injection system for the JT-60U has been enhanced significantly. By improving the beam convergence, a beam injection pulse duration ofup to 10 seconds has been attained. This is the maximum duration defined by the system design specification. Small peaks were observed in the beam current density profile measured in the near field. These peaks are clearly resulted from the deflection of beamlets due to an unwanted electric field generated at an extractor. Correction ofthe beamlet deflection improved the beam convergence, resulting in a reduced temperature rise of the beam limiter at the port of the JT-60U by more than 5O Vo. A Doppler shifted spectrum of the Da line radiated from the negative-ion beam extracted from the ion source was measured. The Doppler shifted spectrum indicated that electron stripping of the negative ion beam occurs primarily inside the extraction grid and in the first acceleration gap, as expected from theoretical considerations. Moreover, the electron stripping contributed to half of the heat load on the grounded grid at the operating pressure of the ion source.

The negative-ion based NBI for JT-60U has been making efforts to increase beam power and beam energy since the operation started in March 1996. The NBI system has already operated negative ion beams of 14.3 A at 380 keV with deuterium and 18.4 A at 350 keV with hydrogen. In the beam injection into JT-60U, a deuterium neutral beam power of 3.6 MW at 350 keV has been injected for 0.9 sec, and plasma reactions against the high energy beam injection with the N-NBI have been confirmed to be in agreement with a theoretical prediction

Research & Development for both positive- and negative-ion based NBI systems is now in progress at JAERI. The positive-ion based NBI system, which consists of ten quasi-perpendicular beamlines and four tangential ones, is in operation with deuterium beams on the JT-60U device. The total injection power achieved with the quasi-perpendicular and the tangential NBI systems is 32 MW at a beam energy of 90-95 keV. R&D work and design studies for the negative-ion based NBI system have been carried out for JT-60U and for ITER. On the basis of the recent progress in high power negative ion source development, a 500keV/10MW negative-ion based NBI system for JT-60U is about to be constructed for demonstrating mega-ampere level NB current drive at high plasma density and core heating in reactor grade plasmas. The construction of the system starts in 1992, and the current drive experiment in JT-60U will start in 1995.

また,ク ライオ トラッピングに よるHe排 気 も報告 さ れてい るが,実 績 のあるArト ラ ッピングに しても排 気 しよ うとす るHeに 対 し約100倍 のArをHe排 気時 に導 入 しなければHeを 安定に排気で きない ことか ら,真 空 のベース圧 力が高 くな るとい う欠点が指摘 されてお りプ ラズマへの影響が予想 され る2).こ れ に対 して,ク ライ オ ソープシ ョンポ ンプの場 合,He排 気前に予め凝縮層 を クライオパネルへ形成 させ る方式 であるためHe排 気 時に吸着媒導入に よる真空環境への影響はない.し か し, Ar,N2凝 縮層 に よる クライオ ソープ シ ョンポ ンプに よ るHe排 気の報告 は数少な くデ ータも少ない3). そ こで,こ れ ら吸着媒 のHe排 気特性 を試験用 クライ オポ ンプを用いて測定 し,JT-60NBI用 クライオ ソープ シ ョンポ ンプの吸着媒の選定 を行 う.

The J1-60 neutral beam system has been successfully operated for 4 years under a wide range of operation conditions: beam energy of 30-75 keV, beam pulse up to 6 s, injection power up to 26 MW with hydrogen beams. The maximum injection power of 26 MW was obtained at 73 keV with a two-stage accelerator. In a lower energy beam injection with a single-stage accelerator, the beam power at 38 keV reached 18 MW. The system could routinely inject a nominal power of 20 MW with high reliability. The beam energy could be changed during a beam pulse, e.g. from 40 keV to 70 keV for 1.5 s. Helium beams were injected with one of the beamlines for a simulation experiment of helium ash, and the injected power was 0.4 MW at 31 keV. The helium beams could deposit in the vicinity of the plasma center column. Helium gas in the beamline was evacuated by SF6 gas condensed cryo-sorption pumps whose pumping speed was about 800 m3/s.

The two-electron transfer reaction was applied to the detection of helium ions in the thermal and epithermal energy range in JT-60. Using this technique, the helium ion density and the energy distribution were determined. The ion density and energy distribution obtained were consistent with the results from other diagnostics and from theory. The potential of this technique for measurements of the helium ash density and the energy distribution in fusion plasmas has been demonstrated.

High‐power long pulse ion sources were fabricated and tested at a prototype injector unit for JT‐60. Ion beams of 70 A at an energy of 75 keV were extracted repeatedly for up to 10 s. The heat loadings to each grid were within our design values and each grid turned out to be thermally stable during 10 s pulse. The neutral beam power deposited to the beam target was over 1.43 MW, which corresponds to the design value of the JT‐60 neutral beam injector. The e‐folding half‐width beam divergence angle was about 1.0° at optimum beam current and a proton ratio of about 80% was obtained. It was also confirmed that other beam line components, such as the ion beam dump and the cryopump, were sufficiently reliable.