Commissioning of the first KSTAR neutral beam injection system and beam experiments

The neutral beam injection (NBI) system was designed to provide plasma heating and current drive for high performance and long pulse operation of the Korean Superconducting Tokamak Advanced Research (KSTAR) device using two co-current beam injection systems. Each neutral beam injection system was designed to inject three beams using three ion sources and each ion source has been designed to deliver more than 2.0 MW of deuterium neutral beam power for the 100-keV beam energy. Consequently, the final goal of the KSTAR NBI system aims to inject more than 12 MW of deuterium beam power with the two NBI for the long pulse operation of the KSTAR. As an initial step toward the long pulse (～300 s) KSTAR NBI system development, the first neutral beam injection system equipped with one ion source was constructed for the KSTAR 2010 campaign and successfully commissioned. During the KSTAR 2010 campaign, a MW-deuterium neutral beam was successfully injected to the KSTAR plasma with maximum beam energy of 90 keV and the L-H transition was observed with neutral beam heating. In recent 2011 campaign, the beam power of 1.5 MW is injected with the beam energy of 95 keV. With the beam injection, the ion and electron temperatures increased significantly, and increase of the toroidal rotation speed of the plasma was observed as well. This paper describes the design, construction, commissioning results of the first NBI system leading the successful heating experiments carried in the KSTAR 2010 and 2011 campaign and the trial of 300-s long pulse beam extraction.     

Design of neutral beam injection system for KSTAR tokamak

The neutral beam injection (NBI-1) system has been designed for providing a 300 s deuterium beam of 120 kV/65 A as an auxiliary heating and current drive system of the KSTAR (Korea Superconducting Tokamak Advanced Research) tokamak. The deuterium beam is produced from a long pulse ion source composed of a bucket-type plasma generator and a multi-aperture tetrode accelerator with the help of discharge power supplies and high voltage (HV) power supplies. The beamline components (BLCs) include a neutralizer with an optical multi-channel analyzer (OMA) section, a bending magnet (BM), an ion dump assembly, a movable calorimeter, beam scrapers, and a cryo-sorption pump system in a rectangular vacuum tank. A beam duct equipped with bellows and a voltage break is placed between the NBI vacuum tank and the KSTAR vacuum vessel. All data and parameters of the NBI system are controlled by a control and data acquisition (CODAQ) system through the EPICS based Ethernet interface.     

Performance of upgraded JET neutral beam injectors

The JET neutral beam injection (NBI) system is undergoing an upgrade of both beam power and pulse duration, which will be completed in 2011. In order to obtain an early assessment of the performance of the upgraded injectors, two positive ion neutral injectors (PINIs) with modified ion source and accelerator configuration were installed on Octant 8 Neutral Injector Box and successfully commissioned in summer 2009. Both PINIs were routinely delivering ～2 MW of deuterium neutral beam power during the JET experimental campaign in autumn 2009. These early tests allowed us to predict with confidence that the JET NBI upgrade objective of injecting 34 MW of total deuterium neutral beam power into the JET plasma will be achieved.     

A scoping study of the application of neutral beam heating on the TCV tokamak

The TCV tokamak contributes to the physics understanding of fusion plasmas, broadening the parameter range of reactor relevant regimes, by investigations based on an extensive use of the existing main experimental tools: flexible shaping and high power real time-controllable electron cyclotron heating (ECH) and current drive (ECCD) systems. A proposed implementation of direct ion heating on the TCV by the installation of a 20–35 keV neutral beam injection (NBI) with a total power of 1–3 MW would permit an extension of the accessible range of ion to electron temperatures (T_i/T_e ～ 0.1–0.8) to well beyond unity, depending on the NBI/ECH mix and the plasma density. A NBI system would provide TCV with a tool for plasma study at reactor relevant T_i/T_e ratios ～1 and in investigating fast ion and MHD physics together with the effects of plasma rotation and high plasma β scenarios. The feasibility studies for a NBI heating on TCV presented in this paper were undertaken to construct a specification for the neutral beam injectors together with an experimental geometry for possible operational scenarios.     

The influence of neutral beam optimization for DEMO on injector design

The requirements for neutral beam injection (NBI) on DEMO are assessed and the consequences for the design of the injectors discussed. Optimization of current drive requires NBI within a 2 m × 2 m envelope at large tangency radii. This is compatible with beamlines of 20 m length and moderate high voltage stand-off distances between injectors. However, q-profile control will necessitate at least three beamlines of different injector types and may not be compatible with shinethrough. Material irradiation studies show that, with three exceptions, there is no significant design issue for distances greater than 3 m from the tokamak wall.     

W7-X neutral-beam-injection: Selection of the NBI source positions for experiment start-up

The stellarator W7-X will be equipped with two Neutral-Beam-Injector (NBI) boxes for balanced injection. Each NBI box has 2 tangential and 2 radial source positions. For the experimental start-up phase each NBI box will be only equipped with 2 ion sources. For the selection of the initial 2 NBI source positions per box three physical aspects were examined (transmission and duct power deposition, shine through and heating efficiency).Using hydrogen injection the heating power to the plasma under typically planned conditions should be 1.3 MW for the tangential sources and 1.1 MW for the radial sources (deuterium: 2 MW for the tangential sources, 1.8 MW for the radial sources). The tangential source positions all have similar heating efficiencies. One of them suffers from the lowest duct transmission (highest power-load to the duct). The same source hits a component with a low power-load capability. The W7-X inner wall design will be modified in order to enhance the maximum power-load capability of that component. For the radial source positions there is no clear physics advantage of one position over the other. Taking all aspects into consideration the decision was made to use one tangential source and one radial source per box during the experimental start-up phase.     

Engineering design for the HL-2M tokamak components

The HL-2A tokamak will be modified into HL-2M. The Bt at the plasma center (major radius R = 1.78 m) is 2.2 T, the minor radius is 0.65 m. The plasma current I_P of HL-2M will reach up to 2.5 MA, the elongation and triangularity is more than 1.8 and more than 0.5, respectively. The vacuum vessel torus consists of 20 sectors with “D” shaped cross-section and double wall structure. 20 toroidal field coil bundles comprise 140 turns which are designed with demountable joints, the poloidal field coils system consists of 25 coils. The engineering design and calculation for field coil system, vacuum vessel, support structure, etc. are finished, many key issues for manufacture process have been discussed with industry and the fabrication of main components of HL-2M tokamak will be carried out in factories.     

Commissioning and operation of 130 kV/130 A switched-mode HV power supplies with the upgraded JET neutral beam injectors

The design features, on-site testing, commissioning and operation are described of two new 130 kV/130 A HV power supply units serving four upgraded 130 kV/60 A positive ion neutral injectors (PINIs) on JET. Both units were factory tested at full power and pulse length into dummy resistive load. Following on-site installation, the factory tests were repeated. The transition from dummy-load testing to PINI operation required full integration of the HVPS within the overall JET control system, and rigorous testing of the co-ordinated actions and protections of all PINI power supplies (filament and arc for plasma source and negative suppression grid). The implementation of these functions is described. Extensive use was made of parasitic integrated test pulses, where the other PINIs could be operated normally, with the HVPS energised under full remote control together with the corresponding PINI plasma sources, but with the HVPS connected to dummy load. The amount of NB operation time dedicated to commissioning was thereby minimised, yet gave a high degree of confidence of readiness for HV energisation of the PINI, and first beam operation followed less than 24 h from HV connection to the PINI. The routine operating experience and performance, including load protection characteristics of the new HVPS units are also described.     

Integration design of TPE-RX Neutral Beam Injector on RFX-mod

The TPE-RX Neutral Beam Injector, which provides a 25 keV positive ion beam energy with a maximum current of 50 A for a pulse duration of 30 ms, will be installed on RFX-mod thanks to the agreement with the AIST Institute of Tsukuba (Japan). The main scientific objective is the study of the behavior of the fast ions, which in the RFP helical equilibrium have exhibited very long confinement times.The integration of TPE-RX NBI on RFX-mod requires the development of several new components: a mechanical interface between the RFX-mod vacuum vessel and neutralizer; a Magnetic Residual Ion Dump; a new vacuum pumping system designed to maximize pumping and minimize beam stopping due to reionization.As regards the power supplies the compliance of the Japanese equipment to the Italian safety rules has been considered and layout studies have been carried out; the integration of the NBI control system in the RFX timing sequence has been studied as well.     

Coupling of electron cyclotron waves to the desired mode in plasma of HL-2A tokamak

3 MW/68 GHz electron cyclotron resonance heating and current driving systems have been developed on HL-2A tokamak and two new units which can operate at two frequencies, 140 GHz and 105 GHz are on-going. Dual-polarizers for changing of the polarization of the wave beam have been designed for these systems based on the integral method of diffraction gratings. Finally, coupling between the electron cyclotron waves and the plasma has been discussed and the results indicate that the desired mode can be obtained for all possible experiments associated with the electron cyclotron resonance heating and current driving systems on HL-2A.     

The negative ion source test facility ELISE

The ITER neutral beam system is using inductively coupled radio frequency (RF) ion sources, that have demonstrated the required ITER parameters on (small) sources with extraction areas up to 200 cm². As a next step towards the full size ITER source IPP is presently constructing the test facility ELISE (“Extraction from a Large Ion Source Experiment”) operating with a “half-size” source which has approximately the width but only half the height of the ITER source. The modular driver concept is expected to allow a further extrapolation to the full size in one direction to be made. The main aim of this experiment is to demonstrate the production of a large uniform negative ion beam with ITER relevant parameters in stable conditions up to one hour.Plasma operation of the source is foreseen to be performed continuously for 1 h; extraction and acceleration of negative ions up to 60 kV is only possible in pulsed mode (10 s every 180 s) due to limitations of the existing IPP HV system. The design of the source and extraction system implements a high experimental flexibility and a good diagnostic access while still staying as close as possible to the ITER design. The main differences are the source operating in air and the use of a large gate valve between the source and the target chamber.ELISE is expected to start operation at the end of 2011 and is an important step for the development of the ITER NBI system; the experience gained early will support the design as well as the commissioning and operating phases of the PRIMA NBI test facilities and the ITER neutral beam system.     

Evaluation of tritium diffusion through the Neutral Beam Injector calorimeter panel

The Neutral Beam Test Facility (NBTF) to be realized in Padoa will test the Neutral Beam Injection (NBI), one of the Heating and Current Drive Systems foreseen for ITER. The NBI is based on the acceleration of hydrogen or deuterium negative ions up to 1 MeV. This work has been aimed at assessing the tritium release from the NBTF in order to provide data for the safety analysis. In particular, the diffusion of the tritium through the neutral beam target material (the CuCrZr alloy calorimeter panels) has been assessed by using literature data of the diffusion coefficient. The tritium generated inside the calorimeter panels moves into both the vacuum and water side: the tritium diffusion flux has been evaluated during the beam-on (200 °C) and the beam-off (20 °C) phases of the NBTF experiments consisting of an interim campaign and a final test. The penetration depth of the tritium through the 2 mm thick CuCrZr alloy material has been also evaluated by using a Monte-Carlo code. As main result, the assessed diffusion flux of tritium during both the beam-on and the beam-off phases are modest. In fact, at the end of the interim campaign (100 days), about the 96% of the all generated tritium (626.5 MBq) exits the calorimeter while the residual tritium inventory (25 MBq) leaves the copper alloy with a diffusion time of about 1 month. At the end of the final test (14 days) about the 99% of the total generated tritium (1.023 × 10⁴ MBq) leaves the copper alloy and the remaining tritium inventory (152.2 MBq) is released by about 32 days. In both the interim campaign and the final test, more than the 99% of the total tritium is transferred into the vacuum side of the calorimeter panel while negligible tritium amounts enter the cooling water system thus showing a very low impact on the environment.     

10 years of engineering and physics achievements by the Large Helical Device project

This article reviews 10 years of engineering and physics achievements by the Large Helical Device (LHD) project with emphasis on the latest results. The LHD is the largest magnetic confinement device among diversified helical systems and employs the world's largest superconducting coils. The cryogenic system has been operated for 50,000 h in total without any serious trouble and routinely provides a confining magnetic field up to 2.96 T in steady state. The heating capability to date is 23 MW of NBI, 2.9 MW of ICRF and 2.1 MW of ECH. Negative-ion-based ion sources with the accelerating voltage of 180 keV are used for a tangential NBI with the power of 16 MW. The ICRF system has full steady-state operational capability with 1.6 MW. In these 10 years, operational experience as well as a physics database have been accumulated and the advantages of stable and steady-state features have been demonstrated by the combination of advanced engineering and the intrinsic physical advantage of helical systems in LHD. Highlighted physical achievements are high beta (5% at the magnetic field of 0.425 T), high density (1.1 × 10²¹ m^?3 at the central temperature of 0.4 keV), high ion temperature (T_i of 5.2 keV at 1.5 × 10¹⁹ m^?3), and steady-state operation (3200 s with 490 kW). These physical parameters have elucidated the potential of net-current free helical plasmas for an attractive fusion reactor. It also should be pointed out that a major part of these engineering and physics achievements is complementary to the tokamak approach and even contributes directly to ITER.     

Li collection experiments on T-11M and T-10 in framework of Li closed loop concept

The concept of a steady state tokamak with plasma facing components (PFC) on the basis of liquid lithium circulation demands the decision of three tasks: lithium injection to the plasma, lithium ions collection before their deposition on the vacuum vessel and lithium returning to the injection zone. Main subject of paper is the investigations of Li collection by different types of limiters intersected the scrape-of-layer (SOL) in T-10 and T-11M tokamaks. For finding solution for this problem in T-11M and T-10, experiments have been applied with Li-, C-rail limiters and ring SS R-limiter-collector (T-11M). The efficiency of Li collection by limiters in T-11M and T-10 tokamaks was investigated by post mortem sample–witness analysis and (T-11M) by the use of the mobile graphite probe (limiter) as a recombination target in the stream of lithium ions. The characteristic depth of lithium penetration in the SOL area of T-11M is about 2 cm and 4 cm in SOL of T-10. The quantitative analysis of the sample–witnesses located on T-11M limiters showed that 60 ± 20% of the lithium injected during plasma operating of T-11M had been collected by limiters. It confirms an opportunity of the lithium ions collection by limiters in tokamak SOL.     

Lithium technologies for edge plasma control

We have investigated two new modes of operation been in T-10 limiter tokamak experiments with a novel rotary feeder of lithium dust. Quasi steady-state mode I and pulse mode II of dust delivery were realized in both OH and OH + ECRH disruption free plasmas at the lithium flow rate up to 2 × 10²¹ atoms/s. A higher flow rate in mode II with injection rate of ～5 × 10²¹ atoms/s caused a series of minor disruptions, which was completed by discharge termination after the major disruption. The observed decreases of bolometer and D_β signals, with increase of the electron density during the lithium dust injection, reveal the effects of the first wall conditioning. The lithium technology may provide inherent safety pathway for major disruption mitigation in a tokamak reactor, which requires demonstration in contemporary tokamak experiments.     

Simulation of EAST off-axis neutral beam heating and current drive

The capability of off-axis neutral beam heating and current drive has been investigated with NUBEAM for Experimental Advanced Superconducting Tokamak (EAST). Three different approaches to realize off-axis Neutral Beam Injection (NBI) have been studied. Simulation results for on- and off-axis NBI are reported. The effects of the alignment of NBI relative to the magnetic field pitch on off-axis neutral beam heating and current drive are observed and discussed qualitatively. By comparing the numerical results, a most favorable off-axis NBI configuration is recommended. The capability to control sawtooth is also investigated by comparing locations of the q = 1 rational surface and the peak of the fast ion density profile.     

Conceptual design of a high-voltage compact bushing for application to future N-NBI systems of fusion reactors

The energy of future neutral beam injector (NBI) heating systems of fusion power plants ranges from 1 to 2 MeV. They are based on powerful (several tens of MW) hydrogen negative ion electrostatic accelerators where electrodes are polarized by DC high-voltage. The beam line under vacuum is supplied by HV power supplies via a transmission line pressured under SF₆ and a high voltage feedthrough called bushing. The paper presents results obtained over experimental campaigns dedicated to high voltage vacuum insulation for future NBI systems (ITER). It addresses the problematic of the electron field emission and the high voltage breakdown limit under vacuum between large electrode surfaces. The paper highlights the dependence of the electron emission (dark current) with the voltage and the background tank pressure: at low pressure (～1E?3 Pa in hydrogen), an important dark current of I ～ 100 mA has been measured at 500 kV, while at higher pressure (～0.3 Pa in helium), the dark current has been nearly suppressed (less than 3 mA of dark current at 970 kV). The paper shows that a field induced gas adsorption process could occur on the emitting surfaces (cathode), and this process tends to lower the electron field emission current by increasing the work function of the electrode surface. The Fowler–Nordheim law applied to the measured dark current indicates about 70% of work function increase at 0.3 Pa in helium. Finally, a new high-voltage bushing concept relevant to the future NBI systems is presented; it is based on these experimental findings in high voltage vacuum insulation; the main feature of the new bushing concept is to take benefit of the field induced adsorption effect, i.e., the suppression of the dark current with helium gas, in the inner part of the bushing where the electric field intensity is highest.     

Design of the HL-2M power supply control system

The power supply control system is one of the most important local control systems of HL-2M tokamak. The power supply system of HL-2M is composed of four flywheel generator sets with total capacity of 600 MVA, all the magnetic field coil power supplies and high voltage power supplies of auxiliary heating system. The control system of power supply should make sure the large amount of on-site equipment and facilities to operate steadily and reliably.This paper presents the architecture of HL-2M control system and four kinds of established control networks. It describes how the power supply control system is built by using the established hardware structures and software agreements. Specially, it is focused on introducing an application of real-time technology based on the reflective shared memory and the fully digital compact solution for controlling the high power converters. Otherwise, a scheme which adopts PROFINET and Sequence of Events technology to carry out the intellect interlock control system is given.     

A bulk tungsten tile for JET: Heat flux tests in the MARION facility on the power-handling performance and validation of the thermal model

In the frame of the ITER-like Wall (ILW) for the JET tokamak, a divertor row made of bulk tungsten material has been developed for the position where the outer strike point is located in most of the foreseen plasma configurations. In the absence of active cooling, this represents a formidable challenge when one considers the temperature reached by tungsten (T_W,surf > 2000 °C) and the vertical gradient ?T/?z = 5 × 10⁴ K/m.As the development is drawing to an end and most components are in production, actual 1:1 prototypes are exposed to an ion beam with a power density around 7 MW/m² on the plasma-facing surface. Advantage is taken of the flexibility of the Marion facility to bombard the tungsten stack under shallow angles of incidence (～6°) with a powerful beam of ions and neutrals (>70 MW/m² on axis). The shallow angles are important, with respect to the toroidal wetted surface, for properly simulating the expected performance under actual tokamak conditions. The Marion tests have been used to validate for a few typical cases the thermal calculations that were steadily developed along with the tungsten tile and, at the same time, to gather information on the actual temperatures of individual components. The latter is an important factor to a finer estimation of the power handling capabilities.