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.     

Feeder components and instrumentation for the JT-60SA magnet system

The modifying of the JT-60U magnet system to the superconducting coils is progressing as a satellite facility for ITER by both parties of Japanese government and European commission in the Broader Approach agreement. The magnet system requires current supplies of 25.7 kA for 18 TF coils and of 20 kA for 4 CS modules and 6 EF coils. The magnet system generates an average heat load of 3.2 kW at 4 K to the cryogenic system. The feeder components connected to the power supply provide current supply. The cooling pipes connected to the cryogenic system provide coolant supply. The instrumentation of the JT-60SA magnet system is used for its operation.     

The ITER full size plasma source device design

In the framework of the strategy for the development and the procurement of the NB systems for ITER, it has been decided to build in Padova a test facility, including two experimental devices: a full size plasma source with low voltage extraction and a full size NB injector at full beam power (1 MV). These two different devices will separately address the main scientific and technological issues of the 17 MW NB injector for ITER. In particular the full size plasma source of negative ions will address the ITER performance requirements in terms of current density and uniformity, limitation of the electron/ion ratio and stationary operation at full current with high reliability and constant performances for the whole operating time up to 1 h. The required negative ion current density to be extracted from the plasma source ranges from 290 A/m² in D₂ (D^?) and 350 A/m² in H₂ (H^?) and these values should be obtained at the lowest admissible neutral pressure in the plasma source volume, nominally at 0.3 Pa. The electron to ion ratio should be limited to less than 1 and the admissible ion inhomogeneity extracted from the grids should be better than 10% on the whole plasma cross-section having a surface exposed to the extraction grid of the order of 1 m².The main design choices will be presented in the paper as well as an overview of the design of the main components and systems.     

Overview of recent developments in pellet injection for ITER

Pellet injection is the primary fueling technique planned for core fueling of ITER burning plasmas. Also, the injection of relatively small pellets to purposely trigger rapid small edge localized modes (ELMs) has been proposed as a possible solution to the heat flux damage from larger natural ELMs likely to be an issue on the ITER divertor surfaces. The ITER pellet injection system is designed to inject pellets into the plasma through both inner and outer wall guide tubes. The inner wall guide tubes will provide high throughput pellet fueling while the outer wall guide tubes will be used primarily to trigger ELMs at a high frequency (>15 Hz). The pellet fueling rate of each injector is to be up to 120 Pa m³/s, which will require the formation of solid D–T at a volumetric rate of ～1500 mm³/s. Two injectors are to be provided for ITER at the startup with a provision for up to six injectors during the D–T phase. The required throughput of each injector is greater than that of any injector built to date, and a novel twin-screw continuous extrusion system is being developed to meet the challenging design parameters. Status of the development activities is presented, highlighting recent progress.     

Structural analysis of the JT-60SA cryostat vessel body

The JT-60SA cryostat is a stainless steel vacuum vessel (14 m diameter, 16 m height) which encloses the Tokamak providing the vacuum environment (10^?3 Pa) necessary to limit the transmission of thermal loads to the components at cryogenic temperature. It must withstand both external atmospheric pressure during normal operation and internal overpressure in case of an accident.The paper summarizes the structural analyses performed in order to validate the JT-60SA cryostat vessel body design. It comprises several analyses: a buckling analysis to demonstrate stability under the external pressure; an elastic and an elastic–plastic stress analysis according to ASME VIII rules, to evaluate resistance to plastic collapse including localized stress concentrations; and, finally, a detailed analysis with bolted fasteners in order to evaluate the behavior of the flanges, assuring the integrity of the vacuum sealing welds of the cryostat vessel body.     

Overview of engineering design,manufacturing and assembly of JT-60SA machine

The JT-60SA experiment is one of the three projects to be undertaken in Japan as part of the Broader Approach Agreement, conducted jointly by Europe and Japan, and complementing the construction of ITER in Europe. The JT-60SA device is a fully superconducting tokamak capable of confining break-even equivalent deuterium plasmas with equilibria covering high plasma shaping with a low aspect ratio at a maximum plasma current of I_p = 5.5 MA. This makes JT-60SA capable to support and complement ITER in all the major areas of fusion plasma development necessary to decide DEMO reactor construction. After a complex start-up phase due to the necessity to carry out a re-baselining effort with the purpose to fit in the original budget while aiming to retain the machine mission, performance, and experimental flexibility, in 2009 detailed design could start. With the majority of time-critical industrial contracts in place, in 2012, it was possible to establish a credible time plan, and now, the project is progressing on schedule towards the first plasma in March 2019. After careful and focused R&D and qualification tests, the procurement of the major components and plant is now well advanced in manufacturing design and/or fabrication. In the meantime the disassembly of the JT-60U machine has been completed and the engineering of the JT-60SA assembly process has been developed. The actual assembly of JT-60SA started in January 2013 with the installation of the cryostat base. The paper gives an overview of the present status of the engineering design, manufacturing and assembly of the JT-60SA machine.     

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.     

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.     

Final design of the Switching Network Units for the JT-60SA Central Solenoid

This paper describes the approved detailed design of the four Switching Network Units (SNUs) of the superconducting Central Solenoid of JT-60SA, the satellite tokamak that will be built in Naka, Japan, in the framework of the “Broader Approach” cooperation agreement between Europe and Japan.The SNUs can interrupt a current of 20 kA DC in less than 1 ms in order to produce a voltage of 5 kV. Such performance is obtained by inserting an electronic static circuit breaker in parallel to an electromechanical contactor and by matching and coordinating their operations. Any undesired transient overvoltage is limited by an advanced snubber circuit optimized for this application. The SNU resistance values can be adapted to the specific operation scenario. In particular, after successful plasma breakdown, the SNU resistance can be reduced by a making switch.The design choices of the main SNU elements are justified by showing and discussing the performed calculations and simulations. In most cases, the developed design is expected to exceed the performances required by the JT-60SA project.     

Recent improvements to the ITER neutral beam system design

The ITER [1] fusion device is expected to demonstrate the feasibility of magnetically confined deuterium–tritium plasma as an energy source which might one day lead to practical power plants. Injection of energetic beams of neutral atoms (up to 1 MeV D⁰ or up to 870 keV H⁰) will be one of the primary methods used for heating the plasma, and for driving toroidal electrical current within it, the latter being essential in producing the required magnetic confinement field configuration. The design calls for each beamline to inject up to 16.5 MW of power through the duct into the tokamak, with an initial complement of two beamlines injecting parallel to the direction of the current arising from the tokamak transformer effect, and with the possibility of eventually adding a third beamline, also in the co-current direction. The general design of the beamlines has taken shape over the past 17 years [2], and is now predicated upon an RF-driven negative ion source based upon the line of sources developed by the Institute for Plasma Physics (IPP) at Garching during recent decades [3], [4], [5], and a multiple-aperture multiple-grid electrostatic accelerator derived from negative ion accelerators developed by the Japan Atomic Energy Agency (JAEA) across a similar span of time [6], [7], [8]. During the past years, the basic concept of the beam system has been further refined and developed, and assessment of suitable fabrication techniques has begun. While many design details which will be important to the installation and implementation of the ITER beams have been worked out during this time, this paper focuses upon those changes to the overall design concept which might be of general interest within the technical community.     

Empirical approach to the description of spectral performance degradation of silicon photodiodes used as particle detectors

The spectral deterioration of Hamamatsu S5821 silicon photodiodes for ion types and energies frequently used in Ion Beam Analysis was investigated. Focused proton beams with energies 430 keV and 2 MeV were applied to generate radiation damage via an area selective ion implantation in unbiased diodes at room temperature. The variations of spectroscopic features were measured “in situ” by Ion Beam Induced Current (IBIC) method as a function of fluence, within the 10⁹–5 × 10¹² ion/cm² range and diode bias voltages, between 0 and 100 V.An empirical model has been developed to describe the radiation damage. Equations are derived for the variations of the normalized peak position and peak width. The derived empirical equations are physically correct, as far as they account for the superposition of the influence of charge carrier trapping by native and radiation-induced defects and for the effect of charge carrier velocity saturation with electric field strength, as well.     

Development of Negative-Ion Based NBI System for JT-60

The negative-ion based neutral beam injector (N-NBI) for JT-60 has been developed for plasma core heating and neutral beam current drive in higher density plasmas. Construction of the N-NBI system was completed in 1996, and just after this completion, the efforts to increase beam power and beam energy started. The N-NBI system has already operated with negative ion beams with 14.3 A at 380 keV of deuterium and with 18.5A at 360 keV of hydrogen. During N-NBI experiments on JT-60, a deuterium neutral beam power of 5.2MW at 350keV has been injected for 0.7s stably, and the response of the JT-60 plasma to high energy beam injection with the N-NBI has been confirmed to be in agreement with a theoretical prediction.     

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.     

Annealing of ion implanted 4H–SiC in the temperature range of 100–800 °C analysed by ion beam techniques

Ion implantation induced damage formation and subsequent annealing in 4H–SiC in the temperature range of 100–800 °C has been investigated. Silicon Carbide was implanted at room temperature with 200 keV ⁴⁰Ar ions with two implantation fluences of 4 × 10¹⁴ and 2 × 10¹⁵ ions/cm². The samples were characterized by Rutherford backscattering and nuclear reaction analysis techniques in channeling mode using 2.00 and 4.30 MeV ⁴He ion beams for damage buildup and recovery in the Si and C sublattices, respectively. At low ion fluence, the restoration of the Si sublattice is evident already at 200 °C and a considerable annealing step occurs between 300 and 400 °C. Similar results have been obtained for the C sublattice using the nuclear resonance reaction for carbon, ¹²C(α,α)¹²C at 4.26 MeV. For samples implanted with the higher ion fluence, no significant recovery is observed at these temperatures.     

Architecture of SPIDER control and data acquisition system

The ITER Heating Neutral Beam injectors will be implemented in three steps: development of the ion source prototype, development of the full injector prototype, and, finally, construction of up to three ITER injectors. The first two steps will be carried out in the ITER neutral beam test facility under construction in Italy. The ion source prototype, referred to as SPIDER, which is currently in the development phase, is a complex experiment involving more than 20 plant units and operating with beam-on pulses lasting up to 1 h. As for control and data acquisition it requires fast and slow control (cycle time around 0.1 ms and 10 ms, respectively), synchronization (10 ns resolution), and data acquisition for about 1000 channels (analogue and images) with sampling frequencies up to tens of MS/s, data throughput up to 200 MB/s, and data storage volume of up to tens of TB/year. The paper describes the architecture of the SPIDER control and data acquisition system, discussing the SPIDER requirements and the ITER CODAC interfaces and specifications for plant system instrumentation and control.     

Gyrotron and power supply development for upgrading the electron cyclotron heating system on DIII-D

An upgrade of the electron cyclotron heating system on DIII-D to almost 15 MW is being planned which will expand it from a system with six 1 MW 110 GHz gyrotrons to one with ten gyrotrons. A depressed collector 1.2 MW 110 GHz gyrotron is being commissioned as the seventh gyrotron. A new 117.5 GHz 1.5 MW depressed collector gyrotron has been designed, and the first article will be the eighth gyrotron. Two more are planned, increasing the system to ten total gyrotrons, and the existing 1 MW gyrotrons will subsequently be replaced with 1.5 MW gyrotrons.Communications and Power Industries completed the design of the 117.5 GHz gyrotron, and are now fabricating the first article. The design was optimized for a nominal 1.5 MW at a beam voltage of 105 kV, collector potential depression of 30 kV, and beam current of 50 A, but can achieve 1.8 MW at 60 A. The design of the collector permits modulation above 100 Hz by either the body or the cathode power supply, or both, while modulation below 100 Hz must use only the cathode power supply.General Atomics is developing solid-state power supplies for this upgrade: a solid-state modulator for the cathode power supply and a linear high voltage amplifier for the body power supply. The solid-state modulator has series-connected insulated-gate bipolar transistors that are switched at a fixed frequency by a pulse-width modulation regulator to control the output voltage. The design of the linear high voltage amplifier has series-connected transistors to control the output voltage, which was successfully demonstrated in a proof-of-principle test at 2 kV. The designs of complete power supplies are progressing.The design features of the 117.5 GHz 1.5 MW gyrotron and the solid-state cathode and body power supplies will be described and the current status and plans are presented.     

Design and manufacturing of JT-60SA vacuum vessel

JT-60 is planned to be upgraded to JT-60SA tokamak machine with fully superconducting coils, which is a project of the JA-EU satellite tokamak program under both Broader Approach program and Japanese domestic program. The JT-60SA vacuum vessel (VV) has a D-shape poloidal cross section and a toroidal configuration with 10° facet segmented in toroidal direction. The material of the VV is 316L stainless steel with low cobalt content of <0.05 wt%. A double wall structure is adopted for the VV to ensure high rigidity and high toroidal one-turn resistance simultaneously.Fundamental welding R&D and a trial manufacturing of the 20° upper half of the VV have been performed to study the manufacturing procedure. After the confirmation of the quality of the mock-up, manufacturing of the actual VV started in November 2009.     

Soft X-ray fluorescence measurements of irradiated polymer films

Fluorescent soft X-ray carbon Kα emission spectra (XES) have been used to characterize the bonding of carbon atoms in polyimide (PI) and polycarbosilane (PCS) films. The PI films have been irradiated with 40 keV nitrogen or argon ions, at fluences ranging from 1 × 10¹⁴ to 1 × 10¹⁶ cm⁻². The PCS films have been irradiated with 5 × 10¹⁵ carbon ions cm⁻² of 500 keV and/or annealed at 1000°C. We find that the fine structure of the carbon XES of the PI films changes with implanted ion fluence above 1 × 10¹⁴ cm⁻² which we believe is due to the degradation of the PI into amorphous C:N:O. The width of the forbidden band as determined from the high-energy cut-off of the C Kα X-ray excitation decreases with the ion fluence. The bonding configuration of free carbon precipitates embedded in amorphous SiC which are formed in PCS after irradiation with C ions or combined treatments (irradiation and subsequent annealing) is close to either to that in diamond-like films or in silicidated graphite, respectively.     

Effect of the laser focus position on characteristics of X-ray and ion emission from gold plasmas generated by a sub-nanosecond laser

X-ray and ion emission from gold plasma produced by a sub-nanosecond Nd:glass laser has been studies as a function of distance of the target from the best focus position. Thermal ion (kinetic energy <19 keV) signals and soft X-ray flux (photon energy >0.7 keV) measurements decrease as the target is moved closer to the best focus position in spite of an increase in laser intensity. We observe simultaneously a strong correlation between the onset of this drop in the flux of soft X-ray and the growth of harder X-ray (photon energy 3–5 keV), alongside a growth in fast ion (energy >67 keV) numbers. This is indicative of the onset of non-linear processes at the higher irradiances (～10¹⁴ W/cm²) associated with the best focus position. Our results show that when using laser plasmas as X-ray or ion sources, X-ray and ion emission in a desired spectral range can be optimized by adjusting the focusing on the target.     

Status of design and procurement activities in JT-60SA

The JT-60SA experiment is one of the three projects to be undertaken in Japan as part of the Broader Approach Agreement, conducted jointly by Europe and Japan, and complementing the construction of ITER in Europe. It is a fully superconducting tokamak capable of confining break-even equivalent deuterium plasmas with equilibria covering high plasma shaping with a low aspect ratio at a maximum plasma current of I_p = 5.5 MA. In late 2007 the BA Parties, prompted by cost concerns, asked the JT-60SA Team to carry out a re-baselining effort with the purpose to fit in the original budget while aiming to retain the machine mission, performance, and experimental flexibility. Subsequently the Integrated Project Team has undertaken a machine re-optimization followed by engineering design activities aimed to reduce costs while maintaining the machine radius and plasma current. This effort led the Parties to the approval of the new design in late 2008 and hence final design and procurement activities have commenced. The paper will describe the process leading to the re-baselining, the resulting final design and technical solutions and the present status of procurement activities.