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.     

R&D progress of high power ion source on EAST-NBI

The neutral beam injector(NBI) system was designed and developed mainly for the plasma heating on the Experimental Advanced Superconducting Tokamak(EAST). The high power ion source is the key part of the NBI. A hot cathode ion source was used on the EAST-NBI. The ion source was conditioned on the ion source test bed with hydrogen gas and achieved the designed parameters. The deuterium gas was used when it moved to the EAST-NBI. The main performance of the ion source on EAST is presented in this paper. The highest beam power of 4.5 MW in NBI-1 and 2.75 MW in NBI-2 was achieved. The total neutral beam power is about 4.5 MW. The long pulse beam of 100 s is injected into the EAST plasma too.     

Long-pulse power-supply system for EAST neutral-beam injectors

The long-pulse power-supply system equipped for the 4 MW beam-power ion source is comprised of three units at ASIPP (Institute of Plasma Physics,Chinese Academy of Sciences):one for the neutralbeam test stand and two for the EAST neutral-beam injectors (NBI-1 and NBI-2,respectively).Each power supply system consists of two low voltage and high current DC power supplies for plasma generation of the ion source,and two high voltage and high current DC power supplies for the accelerator grid system.The operation range of the NB power supply is about 80 percent of the design value,which is the safe and stable operation range.At the neutral-beam test stand,a hydrogen ion beam with a beam pulse of 150 s,beam power of 1.5 MW and beam energy of 50 keV was achieved during the long-pulse testing experiments.The result shows that the power-supply system meets the requirements of the EAST-NBIs fully and lays a basis for achieving plasma heating.     

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.     

Development of KSTAR in-vessel components and heating systems

In-vessel components of the Korea Superconducting Tokamak Advanced Research (KSTAR) were developed for 2010 campaign to provide a crucial circumstance for achieving the strongly shaped and diverted plasma. Moreover, the in-vessel components such as limiter, divertor, passive stabilizer, in-vessel control coil (IVCC) system demonstrated good performances satisfying the original design concepts. In addition to the plasma facing components and the IVCC, in-vessel cryo-pump (IVCP) system was also installed to leverage divertor operation. Besides the in-vessel components, there have been substantial progresses in development of the heating and current drive system. The KSTAR heating and current drive system includes all kinds of the major heating systems such as neutral beam injection (NBI), ion cyclotron range of frequency (ICRF), electron cyclotron resonance heating and current drive (ECH and ECCD), lower hybrid current drive (LHCD) systems. As an initial stage for full equipment of the heating systems to total power of 26 MW, several key systems such as 1st NBI (called NBI-1), ICRF, and ECH-assisted startup system successfully demonstrated their excellent feasibilities in the design and performances for dedication to the 2010 campaign.     

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.     

Design and Development of a Power Supply System for NBI Test Stand of EAST

A neutral beam injector (NBI) test stand was constructed to develop a multi-megawatt prototype ion source as an auxiliary heating system on experimental advanced superconducting tokamak. A power supply system for the NBI test stand components such as a set of dc power supplies for plasma generator, a dc high voltage power supply of a tetrode accelerator, a transmission line and a surge energy suppressor. Stable arc discharges of the plasma generator with hydrogen gases for 100 s long pulse have been produced by six Langmuir probes feedback loop regulation mode to control the arc power supply. The 4 MW hydrogen ion beam of 1 s is extracted with beam energy of 80 keV and the beam current of 52 A. The dc high voltage power supply for the plasma grid of the prototype ion source was designed to contribute maximum voltage of 100 kV and current of 100 A. The high voltage power output is continuously adjustable to satisfy with plasma physics experiment in operation frequency of 10 Hz. To prevent damage of the beam source at high voltage breakdown, core snubber using deltamax soft magnetic materials have been adopted to satisfy the input energy into the accelerator from the power supply can be reduced to about 5 J in the case of breakdown at 80 kV. For the transmission line, a disc shape multi cable coaxial configuration was adopted and which the dimension of the diameter is 140 mm at the core snubber. The major issues of discharge characteristics with long pulse and beam extraction with high power for the prototype ion source were investigated on the NBI test stand.     

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.     

Criticality in the fabrication of ion extraction system for SST-1 neutral beam injector

For the heating of plasma in steady-state superconducting tokamak (SST-1) (Y.C. Saxena, SST-1 Team, Present status of the SST-1 project, Nucl. Fusion 40 (2000) 1069–1082; D. Bora, SST-1 Team, Test results on systems developed for the SST-1 tokamak, Nucl. Fusion 43 (2003) 1748–1758), a neutral beam injector is provided to raise the ion temperature to 1 keV. This injector has a capability of injecting hydrogen beam with the power of 0.5 MW at 30 keV. For the upgrade of SST-1, power of 1.7 MW at 55 KeV is required. Further, beam power is to be provided for a pulse length of 1000S. We have designed a neutral beam injector (S.K. Mattoo, A.K. Chakraborty, U.K. Baruah, P.K. Jayakumar, M. Bandyopadhyay, N. Bisai, Ch. Chakrapani, M.R. Jana, R. Onali, V. Prahlad, P.J. Patel, G.B. Patel, B. Prajapati, N.V.M. Rao, S. Rambabu, C. Rotti, S.K. Sharma, S. Shah, V. Sharma, M.J. Singh, Engineering design of the steady-state neutral beam injector for SST-1, Fusion Eng. Des. 56 (2001) 685–691; A.K. Chakraborty, N. Bisai, M.R. Jana, P.K. Jayakumar, U.K. Baruah, P.J. Patel, K. Rajasekar, S.K. Mattoo, Neutral beam injector for steady-state superconducting tokamak, Fusion Technol. (1996) 657–660; P.K. Jayakumar, M.R. Jana, N. Bisai, M. Bajpai, N.P. Singh, U.K. Baruah, A.K. Chakraborty, M. Bandyopadhyay, C. Chrakrapani, D. Patel, G.B. Patel, P. Patel, V. Prahlad, N.V.M. Rao, C. Rotti, V. Sreedhar, S.K. Mattoo, Engineering issues of a 1000S neutral beam ion source, Fusion Technol. 1 (1998) 419–422) satisfying the requirements for both SST-1 and its upgrade. Since intense power is to be transported to SST-1 situated at a distance of several meters from the ion source, the optical quality of the beam becomes a primary concern. This in turn, is determined by the uniformity of the ion source plasma and the extractor geometry. To obtain the desired optical quality of the beam, stringent tolerances are to be met during the fabrication of ion extractor system.

SST-1 neutral beam injector is based on positive ion source. The extraction system consists of three grids, each having extraction area of (width) 230 mm × (height) 480 mm and 774-shaped apertures of 8-mm diameter. To obtain horizontal focal length of 5.4 m and vertical of 7 m, each grid consists of two halves with 387 apertures. Two halves are inclined at an angle of 1.07 ± 0.01°. For long pulse operation, active water cooling is provided by in-laid down of dense network of 22 wavy semicircular (r = 1.1 ± 0.05 mm) cooling channels in the space available between the apertures. The required flatness of the copper plate is 100 μm and positioning tolerance of aperture is ±60 μm. The measurement obtained after fabrication is compared with the specifications. It is pointed out that fabrication within set tolerance limit could be achieved only through process of fabrication and high-resolution measurements.     

The upgrade of KSTAR timing system to support long-pulse operation and high-speed data acquisition

Since the first campaign of KSTAR in 2008, the home-made timing system had run for the synchronized operation of tokamak. The timing board which featured PMC-form factor, giga-bit optical communication, home-made protocol, multi-triggering capability, using GPS time and being integrated to EPICS (Experimental Physics and Industrial Control System), had advantages of compactness, modularity, platform independency and full functionality for the synchronized tokamak operation. However, there was deficiency in timing accuracy resulting from the engagement of software in realization of timing function and timing jitter due to poor isolation in output ports. Moreover, new requirements were on the rise as the plasma pulse length was getting longer and diagnostics operating at the higher frequency were newly installed.In order to meet new requirements and overcome the problems, the new timing board has been developed. As a result, the performance is remarkably enhanced: timing accuracy less than 5 ns, jitter less than 100 ps, 8 configurable multi-triggering sections, provision of maximum 100 MHz sampling clock. The KSTAR timing system upgraded by introducing the new timing board is participating in the 2011 campaign after calibration and consolidating the established timing system.This paper describes design, development and commissioning results of the new KSTAR timing system.     

Thermal analysis of EAST neutral beam injectors for long-pulse beam operation

Two sets of neutral beam injectors(NBI-1 and NBI-2) have been mounted on the EAST tokamak since 2014. NBI-1 and NBI-2 are co-direction and counter-direction, respectively. As with indepth physics and engineering study of EAST, the ability of long pulse beam injection should be required in the NBI system. For NBIs, the most important and difficult thing that should be overcome is heat removal capacity of heat loaded components for long-pulse beam extraction. In this article, the thermal state of the components of EAST NBI is investigated using water flow calorimetry and thermocouple temperatures. Results show that(1) operation parameters have an obvious influence on the heat deposited on the inner components of the beamline,(2) a suitable operation parameter can decrease the heat loading effectively and obtain longer beam pulse length, and(3) under the cooling water pressure of 0.25 MPa, the predicted maximum beam pulse length will be up to 260 s with 50 keV beam energy by a duty factor of 0.5. The results present that, in this regard, the EAST NBI-1 system has the ability of long-pulse beam injection.     

Overview of Development Status for EAST-NBI System

The neutral beam injection(NBI) system was developed on the Experimental Advanced Superconducting Tokamak(EAST) for plasma heating and current driving. This paper presents the brief history, design, development, and the main experimental results of the RD of neutral beam injector on the test bed and on EAST. In particular, it will describe:(1) how the two beamlines with a total beam power of 8 MW were developed;(2) the design of the EAST-NBI system including the high power ion source, main vacuum chamber, inner components, beam diagnostic system and sub-system;(3) the experimental results of beamline-1 on the summer campaign of EAST in 2014 and,(4) the status of beamline-2 and the future plan of EAST-NBIs.     

Commissioning the ICRF system and an ICRF assisted discharge cleaning at the KSTAR

Korean superconducting tokamak advanced research (KSTAR) is a national superconducting tokamak with the aim of a high beta operation based on advanced tokamak (AT) scenarios, and an ion cyclotron ranges of frequency (ICRF) heating is one of the essential tools to achieve this goal. The fabrication and high voltage (HV) test of the antenna and the matching system were finished in 2006 and the installation of the antenna, matching system and the transmitter at the KSTAR site was completed in 2007. Antenna conditioning was carried out to improve the HV holding condition of the antenna installed on the KSTAR and to check on the electro-magnetic (EM) interference with other equipments such as the superconducting magnet monitoring system and other machine and/or plasma diagnostic systems. The first KSTAR tokamak experimental campaign started by a vacuum pumping, a cryostat cooling and an ICRF system contributed to the successful tokamak shots through an ICRF assisted discharge cleaning of the vacuum vessel. In this paper, the installation processes of the ICRF system (with an emphasis on the quality assurance procedures of KSTAR), as well as the results from the first RF discharge experiment for the discharge cleaning and FWEH (fast wave electron heating) experiment for the KSTAR 1st experimental campaign are outlined.     

Measurement of beam power and profile for DNB on HT-7 tokamak

In normal experimental operation, a diagnostic neutral beam (DNB) can produce 6 A of extracted beam current in hydrogen at an energy of 49 keV with a pulse length of 100 ms. Hydrogen and deuterium beams can be produced as well. The diagnostic neutral beam has been added to the diagnostic set so that charge-exchange recombination spectroscopy (CXRS) can be used to acquire ion temperature and rotation. The beam power and beam profile distribution of the DNB injection can be obtained with a thermocouple probe measurement system on the HT-7 superconducting tokamak. The thermocouple probe measurement system with 13 thermocouples crossly distributed on the probe plate was used to measure the temperature rise of each coppery target, so the profile distribution of the ion/neutral beam was obtained by calculation. In this paper, the structure of the probe plate on the DNB for HT-7 tokamak and some measurement results are presented.     

Experimental Results of Ion Beam Extraction of High-Current Ion Source for EAST Neutral Beam Injector

The first campaign of ion beam extraction tests are completed for EAST NBI high current ion source. The hydrogen ion beam with beam voltage of 80 keV is extracted from tetrode accelerator system. During the experiment, the characteristics of arc discharge and ion beam extraction are studied. The arc power reaches 120 kW and the extracted ion beam power reaches 3 MW. The relationship of some key parameters of arc and beam are investigated, and the details are described in this paper.     

Conceptual design of 5 MW-NBI injector for HL-2 M tokamak

The HL-2 M tokamak is now under construction in Southwestern Institute of Physics in China. As one of the main auxiliary heating systems for HL-2 M tokamak, a new NBI beam line with 5 MW NBI power, 42° injection angle, based on 4 sets of 80 kV/45 A/5 s bucket ion sources with geometrical beam focus, is conceptually designed with geometrical calculation and engineering simulations. The preliminary structure and layout of key components including ion sources, neutralizers, ion dumps, deflection magnet, beam edge scraper, long pulse calorimeter target, short pulse calorimeter target, injection port and beam drift duct are determined. The magnetic shielding of the stray field of HL-2 M tokamak is analyzed. Beam power transmission efficiency is calculated with geometrical algorithm. The ratio of neutral beam injection power to ion beam power is as high as 48%.     

Determination of Average Ion Temperature of the TVD Tokamak by a Simple Charge Exchange Analyzer Detector

In this paper the average ion temperature of the TVD tokamak plasma is determined by a low cost charge exchange analyzer. When the neutral beam energy was equal to 3 keV, the stripping efficiency was 0.065. Estimated average ion temperatures for cold and hot regions of the plasma were 100 and 650 eV, respectively. Unlike the cold region of the plasma in which ion temperature was not related to the time, ion temperature in the hot region was maximum when the plasma current reached its maximum value.     

Performance of Diagnostic Neutral Beam for the HT-7 Tokamak

Diagnostic neutral beam (DNB), combined with spectral diagnostics, is employed to measure the ion temperature in HT-7. The factors affecting the extracted beam are studied in the experiment for the high performance diagnostic neutral beam. A 6.5 A extracted hydrogen current at 43 keV of 100 ms was obtained after optimization. The extracted beam has a proton ratio as high as 40%, and can penetrate into the core plasma after neutralization to measure the ion temperature effectively.     

Numerical prediction of pellet injection effects on core plasma fueling in the KSTAR tokamak

Since pellet injection into tokamak plasmas has been found to be an effective method for fueling and profile modification of core plasmas in tokamak experiments, a hypothetical injection of deutrium pellets into the KSTAR tokamak is numerically simulated in this work to investigate its influences on the fueling and transport of the core plasma depending on pellet parameters. A neutral gas shielding model and a pellet drift displacement model are used to describe the ablation and mass deposition from pellets on core plasma profiles. These models are coupled with a 1.5-dimensional (1.5D) core transport code to calculate the plasma density and temperature profiles responding to pellets injected into the target plasma. The simulation results indicate that a HFS (high field side) injection achieves more effective fueling due to a deeper pellet penetration into the core plasma, compared with a LFS (low field side) injection. The plasma density is found to increase during sequential pellet injections from both HFS and LFS, but the HFS case shows better fueling performance owing to a drift of the pellet ablatant in the major radius direction resulting in the deeper pellet penetration. Increasing the size and injection velocity of the pellet contributes to enhance the fueling efficiency. However, raising the power of neutral beam injection heating reduces the fueling efficiency because the pellet mass deposition is shifted toward the edge region in high temperature plasmas. It is concluded that the pellet size and injection direction among pellet and plasma parameters have the most dominant effects on fueling performance while the pellet velocity and heating power have relatively small influences on fueling.     

Analysis of charge-exchanged neutral particles during IBW heating in the HT-7 tokamak

The charge-exchange neutral particles fluxes and energy distribution in IBW heated plasma were investigated in the HT-7 tokamak. The RF frequency was 30 MHz and with an injecting power up to 200 kW. It is observed that the plasma performance is obviously enhanced by IBW heating. The electron temperature was increased by 0.5 keV and the central line averaged electron density was doubled. The neutral particle fluxes of high-energy increased and the bulk ions were heated during IBW heating. The ion temperature was increased by 0.3 keV and the ion heating efficiency of (2–3) eV kW⁻¹ × 10¹³ cm⁻³ was achieved. The velocity distribution of charge-exchanged neutral particles appears to be Maxwellian without high-energy tail ions up to the maximum RF power.