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.     

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.     

Engineering design of the double neutral beam injection system for MAST Upgrade

The objectives of the upcoming upgrade to the Mega-Ampere Spherical Tokamak (MAST) require the construction of a new neutral beam injection system. Although the principle of operation is similar to other beamlines including those on JET, the design process is complicated by strict requirements such as the need for intensive ion source commissioning. A double neutral beamline has been designed which satisfies the requirements. This paper reports the engineering design of the MAST Upgrade double beamline, describing the key challenges and unusual solutions adopted.     

Commissioning and initial operation of KSTAR superconducting tokamak

The commissioning and the initial operation for the first plasma in the KSTAR device have been accomplished successfully without any severe failure preventing the device operation and plasma experiments. The commissioning is classified into four steps: vacuum commissioning, cryogenic cool-down commissioning, magnet system commissioning, and plasma discharge.Vacuum commissioning commenced after completion of the tokamak and basic ancillary systems construction. Base pressure of the vacuum vessel was about 3 × 10^?6 Pa and that of the cryostat about 2.7 × 10^?4 Pa, and both levels meet the KSTAR requirements to start the cool-down operation. All the SC magnets were cooled down by a 9 kW rated cryogenic helium facility and reached the base temperature of 4.5 K in a month. The performance test of the superconducting magnet showed that the joint resistances were below 3 nΩ and the resistance to ground after cool-down was over 1 GΩ. An ac loss test of each PF coil made by applying a dc biased sinusoidal current showed that the coupling loss was within the KSTAR requirement with the coupling loss time constant less than 35 ms for both Nb₃Sn and NbTi magnets. All the superconducting magnets operated in stable without quench for long-time dc operation and with synchronized pulse operation by the plasma control system (PCS). By using an 84 GHz ECH system, second harmonic ECH assisted plasma discharges were produced successfully with loop voltage of less than 3 V. By the real-time feedback control, operation of 100 kA plasma current with pulse length up to 865 ms was achieved, which also meet the first plasma target of 100 kA and 100 ms. The KSTAR device will be operated to meet the missions of steady-state and high-beta achievement by system upgrades and collaborative researches.     

EAST托卡马克的中性束注入方案

高能中性束注入(Neutral beam injection,NBI)是核聚变装置托卡马克采用的芯部辅助加热和非感应电流驱动主要手段之一.本文介绍了国家大科学工程全超导托卡马克实验装置(Experimental advanced super-conductingtokamak,EAST)上的高能NBI加热方案及注入器的工程要求,并讨论了中性束在EAST等离子体中的传输等相关问题.     

诊断中性束在合肥托卡马克等离子体中的衰减计算

考虑诊断中性束在HT-7等离子体中的电离，复合和电荷交换，对不同靶等离子体的等离子体温度、密度和杂质含量，计算了中性束在等离子体中的衰减。分析结果表明，等离子体的密度对中性束的衰减影响较大，而温度的变化对中性束衰减影响较小。当增大时，中性束衰减率随等离子体密度显著增大；当中性束的能量增大时，中性束也衰减较甚。含杂质中性束的衰减率也增大。最后，根据诊断中性束在EAST托卡马克中等离子体衰减的计算结果，给出了EAST托卡马克中等离子体所要求的诊断中性束参数。     

MHD stability analysis for advanced tokamak modes in the KSTAR device

The Korea Superconducting Tokamak Advanced Research (KSTAR) device aims to demonstrating the steady-state operation of high-performance advanced tokamak (AT) modes. In order to meet this research goal it is critical to have a good magnetohydrodynamic (MHD) stability, so that KSTAR adopted a strong plasma shape and a conducting wall close to plasma for such stability. An early calculation during the KSTAR design phase had shown that a target AT mode stable up to β_N above 5 can be then obtained. A recent work by Katsuro-Hopkins et al. [O. Katsuro-Hopkins, S.A. Sabbagh, J.M. Bialek, H.K. Park, J.G. Bak, J. Chung, et al., Equilibrium and global MHD stability study of KSTAR high beta plasmas under passive and active mode control, Nucl. Fusion 50(2010) 025019] showed, however, that the maximum β_N value can be substantially lower than 5, unlike the earlier result. In this work, we present a more detailed study on the MHD stability limit of the KSTAR target AT mode and try to clarify the discrepancy observed in the previous two works. It is shown that in the reverse-shear plasma the target mode with β_N above 5 can be obtained if the pressure profile is relatively peaked, but the maximum β_N value is substantially reduced below 5 if the pressure profile becomes broader. This result suggests the importance of a proper control of the pressure profile to get the high-beta AT mode in KSTAR.     

Mechanical design for modification of a neutral beam for off-axis injection

DIII-D is planning to implement off-axis neutral beam current drive by neutral beam injection through a midplane port at angles up to 15° from horizontal. To accommodate the beam-line tilting, the following modifications are planned: (1) move the beam line away from the tokamak by 0.39 m to allow for a 0.68 m inside diameter welded bellows of necessary length to provide 15° of vertical motion between the vessel port and the beam line; (2) reduce the vertical height of the injected beam from 0.48 m to 0.43 m to provide clearance for the inclined beam as it passes through the length of the vessel port; (3) add a linkage system between the front of the beam line and the tokamak to restrain the NB against the vacuum loading from the bellows while maintaining zero roll about the axis of the beam line as it is moved about a virtual pivot axis; (4) add a forward and two rear vertical actuators for raising and lowering the beam line (These actuators require coordinated position control to rotate the NB about a virtual pivot axis.); (5) incorporate lateral restraint to comply with seismic requirements.     

Design and test of the KSTAR vacuum pumping system

The Korea Superconductor Tokamak Advanced Research (KSTAR) device is a tokamak mainly composed of a vacuum vessel, superconducting magnets, and cryostat. The internal volume of the vacuum vessel is about 110 m³ with a target pressure of 1 × 10⁻⁶ Pa, while the volume of the cryostat is 450 m³ with a target pressure of 5 × 10⁻³ Pa. To attain these target pressures, two identical vacuum pumping systems consisting of dry pumps, mechanical booster pumps, turbo-molecular pumps, and cryopumps were installed. The control system of the vacuum pumping systems was built using the experimental physics and industrial control system (EPICS), which has various merits such as easy access, convenient extension and flexible integration. The pump-down test of the pumping ducts was successfully executed under the control of the EPICS system.     

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.     

Optic diagnosis of neutral beam injection on HL—1M

During the operation of a high-power neutral beam injection(NBI)system on the HL-1M tokamak,an optical diagnostic means using CCD camera was developed to characterize the NBI performance.The vacuum valve opening process and NBI period in the HL-1M experiment were displayed by a lot of photos taken with this means.Thus,the Hα emission profiles of the neutral beam(NB) and its interaction with plasma were given.Finally,the reason possible for plasma breakdown during NB1 model II discharge was investigated.this in-situ diagnosis can provide more information of the NB1.     

Real-time plasma boundary reconstruction in the KSTAR tokamak using finite element method

A study is carried out on the real-time plasma shape identification in the KSTAR device. An improved form of the finite current element (FCE) method is utilized in this study. Results are shown that the plasma boundary can be reproduced in 7 mm accuracy for any plasma configuration in ideal cases without invoking measurement errors. A design guideline for magnetic diagnostics (MD) is established when the measurement signals are subject to Gaussian noise. It is found that the measurement errors in poloidal field (PF) coil currents have substantial influence on the determination of the plasma shape.     

Plasma-wall interaction in ATC during high power neutral beam injection

Measurements of the elemental composition of samples of the vacuum vessel wall surface and of relative impurity influx rates into ATC during high power beam-heated discharges are combined with a computer simulation of the plasma and previous measurements of power balance and scaling laws to give a model of the main plasma-wall interactions in ATC. It is shown that plasma and beam charge exchange neutrals are the primary causes of impurity influx during neutral beam injection.     

Development of the pellet injection system on the J-TEXT tokamak

Pellet injection is an attractive technology for core-fueling and magnetohydrodynamic study in magnetic-confinement fusion devices like tokamaks and stellarators. It can inject solid hydrogen/deuterium pellets into the plasma with deeper density deposition compared with other fueling methods, such as gas puffing. A three-barrel H₂ pellet injection system was installed on the J-TEXT tokamak and experiments were carried out. The pellets are formed in three barrels cooled by a cryocooler and compressor system at around 9 K, and are 0.8 mm/1 mm diameter and 0.8 mm length. The pellet is launched by helium propellant gas and injected from the low-field side of the plasma. The normal range of pellet speed is 210–310 m s⁻¹ for different propellant gas pressures. Due to the three-barrel structure, the number of injected pellets can be adjusted between one and three. Pellets can be launched sequentially with arbitrary time intervals, which enables flexible applications. The results of the experiments show that pellet fueling efficiency can reach 50%. The energy confinement time increased by about 7.5‒10 ms after pellet injection.     

Design and operation results of nitrogen gas baking system for KSTAR plasma facing components

A baking system for the Korea Superconducting Tokamak Advanced Research (KSTAR) plasma facing components (PFCs) is designed and operated to achieve vacuum pressure below 5 × 10^?7 mbar in vacuum vessel with removing impurities. The purpose of this research is to prevent the fracture of PFC because of thermal stress during baking the PFC, and to accomplish stable operation of the baking system with the minimum life cycle cost. The uniformity of PFC temperature in each sector was investigated, when the supply gas temperature was varied by 5 °C per hour using a heater and the three-way valve at the outlet of a compressor. The alternative of the pipe expansion owing to hot gas and the cage configuration of the three-way valve were also studied. During the fourth campaign of the KSTAR in 2011, nitrogen gas temperature rose up to 300 °C, PFC temperature reached at 250 °C, the temperature difference among PFCs was maintained at below 8.3 °C, and vacuum pressure of up to 7.24 × 10^?8 mbar was achieved inside the vacuum vessel.     

Design of the in-wall shielding for the ITER neutral beam port

The ITER neutral beam port is composed of connecting duct, port extension and port stub extension. The spaces between inner and outer shells of the port extension and port stub extension are filled with pre-assembled blocks, called in-wall shielding. The main purpose of IWS is to provide neutron shielding for the superconducting magnet, thermal shield and cryostat from the main vessel during plasma operation. In order to provide effective neutron shielding capability with the cooling water, 40 mm thick flat plates (steel type 304B4) are used in almost all areas of the volume between port shells. The IWS is composed of shield plates, upper/lower brackets and bolt/nut/washers. Major activities during design work are to develop installation concept of the IWS blocks for easy assembly into port structures and to perform structural analysis to assess sufficient strength, fabrication feasibility study and 3D modeling including drawing works.In this paper, major results of mechanical design are introduced. First, the design requirements for IWS and the developed IWS designs for easy assembly into the port structure are introduced. Second, is introduced the engineering analysis results to assess structural integrity. And then the fabrication feasibility study results are presented for major fabrication processes. Lastly, conclusion and future works are mentioned.     

Design of an ion extractor system for a prototype ion source experiment for SST-1 neutral beam injector

An ion extractor system has been designed for the steady state superconducting tokamak (SST-1) neutral beam injector (NBI) for an experiment using a prototype ion source with fully integrated regulated high voltage power supply (RHVPS) and data acquisition and control system (DACS) developed at Institute for Plasma Research (IPR) to obtain experience of NB operation. The extractor system is capable of extracting positive hydrogen ion beam of ∼10 A current at ∼20 kV. This paper presents the beam optics study for detailed design of an ion extraction system which could meet this requirement. It consists of 3 grid accel-decel system, each of the grid has 217 straight cylindrical holes of 8 mm diameter. Grids are placed on a specially designed G-10 block; a fiber reinforc plastic (FRP) isolator of outer diameter of 820 mm and 50 mm thickness. Provisions are made for supplying high voltage to the grid system through the embedded feed-throughs. Extractor system has been fabricated, mounted on the SST-1 neutral beam injector and has extracted positive hydrogen ion beam of 4 A at 20 kV till now.     

A repetitive pellet injection system for steady state fuelling in EAST superconducting tokamak

A new hydrogen/deuterium pellet injector has been developed for Experimental Advanced Superconducting Tokamak (EAST). The pellet injector based on a screw extruder is able to fire pellets (∅2 mm × 2 mm; frequency 1–10 Hz and velocity 150–300 m/s) in steady state mode with reliability greater than 95%. An injection line was designed for pumping propellant gas and for diagnostic purpose also. A guide tube for magnetic high-field side (HFS) injection was developed and theoretical calculation has been done. After successful engineering commissioning, the injection system served at EAST 2012 campaign and first experimental results were obtained.