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.     

Simulation of Toroidal Rotation Effect on Heat Flux Transport in the Edge Plasma of Small Size Divertor Tokamak

The effect of toroidal rotation on heat flux transport in the edge plasma of small size divertor was simulated by B2SOLP0.5.2D transport code. The main results of simulation shows that, the following: (1) the radial heat flux is strongly influenced by toroidal rotation. (2) The amplification of conduction part of radial heat flux imposes nonresilient profile of ion temperature, under which the effect of toroidal rotation on ion temperature profile is strong. (3) The ion distribution and its gradients are lower for counter-injection neutral beam than for co-injection neutral beam. (4) Reversal of toroidal rotation during using neutral beam injection result in reverses of radial electric field and E × B drift velocity. (5) The toroidal rotation strong influence on the ion temperature scale length of the ion temperature gradient (ITG). (6) Switch on and off all drifts leads to higher change in the ion density distribution in edge plasma of small size divertor tokamak when the unbalance neutral beam injection are considered (7) the comparison between radial heat flux at different momentum input shows that, the radial ion heat flux with larger ion temperature scale length in the case of co-injection neutral beam is larger than the ion heat flux with smaller ion temperature scale length in the case of counter-injection neutral beam.     

Ions Measurement at the Edge of HT-7 Tokamak

A reliable method of measuring ions and ion temperature in tokamak plasma is necessary, for which an omegatron-like instrument has been developed on the HT-7 tokamak. The basic layout of the omegatron-like instrument is shown in this article. The measurement of working gas ion has been performed in the last experimental campaign on HT-7 tokamak. The relations among ion current, the electron repeller voltage and trap voltage have been investigated. This omegatron-like instrument has also provided the edge-plasma ion temperature.     

The Beam Profile Calculation for Diagnostic Neutral Beam on HT-7 Tokamak

A new calculation method is introduced for convergence beam intensity. The program based on this method is prepared for beam intensity distribution and beam power calculation. Taking the HT-7 DNB as a reference, the beam profile calculated by program is similar with that of Gaussian fit of experimental data.     

Primary research of heat flux deposition pattern on the surface of HT-7 movable limiter

This paper presents a method of characterizing the heat flux deposition pattern on HT-7 movable limiter, a new component in the spring experimental campaign of year 2006. A new modeling establishment combines with heat transfer simulation with ANSYS code, and a shot of long pulse discharge is chosen with small radius 265 mm of movable limiter, which is 5 mm smaller than that of the main toroidal limiters. Both parallel and perpendicular heat flux are taken into account and different ratios of them are also tested in simulations. The simulation temperature values match well with the results of the IR-camera. Temperature distribution shows that ?_||/?_⊥ = 5 is quite suitable in HT-7 device, and the maximum heat flux is about 10 MW/m² and average heat flux is about 5 MW/m² on the movable limiter. This method can be also used in other tokamak devices with limiter configuration and circular cross-section shape.     

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.     

The Power Supply System of Ion Source for NBI

The power supply system of ion source for the Neutral Beam Injector (NBI) in the HT-7 superconducting tokamak is based on a single injector with one ion source that can deliver 700 kW of neutral beam power. Experiments and a discharges test on the ion source were successfully performed. In this paper, the circuit structures and features of every power supply are described and the results of the discharges test are presented.     

Calculation of Beam Extraction Pulse Duration for Beam Angular Divergence Measurement of EAST Neutral Beam Injection

Distributed thermocouple measurement will be used on EAST neutral beam injection beam angular divergence measurement. In order to determine the pulse duration of neutral beam extraction, the plates of calorimeter are analyzed by using heat transfer science. Mathematic derivations and the results are given according to the installation location of thermocouple in the plate. Those results could point the way to beam angular divergence measurement of EAST neutral beam injection.     

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.     

Advances in optical thermometry for the ITER divertor

Thermography will be an important diagnostic on the ITER tokamak, but the inclusion of reflective materials such as tungsten in the design for ITER's first wall and divertor region presents problems for optical temperature measurement. The ongoing testing of ITER plasma facing components (PFCs) provides an excellent opportunity to resolve such problems. This has focused on the variation of PFC emissivity with temperature and time, as well as environmental influence on thermography. The sensitivity of these systems to ambient temperature, due primarily to modification of the transmission of the optical path, has been established and minimised. The accuracy of the system is then sufficient to measure the variation of emissivity in heated material samples, by comparing its front-face luminance measured with an infrared camera to the temperature given by an implanted thermocouple. Measurements on both tungsten and carbon fibre composite are in broad agreement with theory, and thus give the material's function of emissivity with temperature at the start of its life. To determine its evolution, a bicolour pyroreflectometer was then installed. This uses two lasers to measure the reflectivity in addition to the luminance at two wavelengths, and thus the true temperature can be calculated. This was validated against the instrumented sample, then used along with the camera to observe an ITER mock-up during ∼50,000 s of 5 MW/m² testing. Emissivity was seen to vary little in the 500 °C region. Higher temperature tests are ongoing.     

Oxygen removal with D2-ICR cleanings after oxidation experiment in HT-7

The effectiveness of D₂-ICR cleaning with a pressure up to 0.18 Pa was surveyed on hot walls of 400-470 K for oxygen removal after oxidation experiment in the HT-7 superconducting tokamak. The oxygen removal rate in D₂-ICR cleaning was about (7-9) × 10²¹ O-atoms/h, about 5-10 times higher that that He-ICR cleanings before and after the D₂-ICR cleaning. In about 130 min He-ICR and D₂-ICR cleanings, about 8 × 10²¹ O-atoms were removed. After the cleanup, a lot of water still retained in HT-7 vessel and 67 disruptive plasmas were required before obtaining normal plasma. In present experiment, the recovered plasmas were easily controlled and much better than previous report in HT-7 oxidation experiments.     

A Calculation Model of Beam Power Based on Thermocouple for EAST Neutral Beam Injector

Beam power measurement usually adopts water flow calorimetry (WFC) for neutral beam injector. In this article a calculation model of beam power for EAST neutral beam injector was given basing on thermocouples installed in the calorimeter plate. In order to obtain calculation model: (1) the heat flux of the surface of beam stopping elements was calculated using heat transfer science, (2) the processing method of data obtained from the thermocouple was given, (3) mathematic derivations of beam power is given basing on theory of Gaussian distribution. The model could point another way to beam power measurement of EAST neutral beam injector besides WFC.     

Ion source plasma parameters measurement based on Langmuir probe with commercial frequency sweep

Langmuir probe is one of the main diagnostic tools to measure the plasma parameters in the ion source. In this article, the commercial frequency power, which is sine wave of 50 Hz, was supplied on the Langmuir probe to measure the plasma parameters. The best feature of this probe sweep voltage is that it does not need extra design. The probe I-V characteristic curve can be got in less than 5 ms and the plasma parameters, the electron temperature and the electron density, varying with the time can be got in one plasma discharge of 400 ms.     

HT-7 Multipoint Nd Laser Thomson Scattering Apparatus

1. IlltroductionThomson scattering diagnostic has been an important and standard method for measuring temperature and density profiles on all modern toka-maks, such as the TV Thotnson Scattering systemon TFTR [1], the LIDAR system on JET [2] and theNd:YAG laser Thomson scattering system on Dlll--D[3]. It has the attractive characteristics of not per-turbillg the plasma to be investigated and of derivingthe absolute values of electronic temperature Ti anddensity ne' From the knot'.led…     

A Calculation Model of Beam Angular Divergence for EAST Neutral Beam Injector

Distributed thermocouple measurement will be used on EAST neutral beam injection beam angular divergence measurement. In order to obtain calculation model of beam angular divergence: (1) the structure of accelerating electrodes are analyzed, (2) mathematic derivations of beam intensity distribution for neutral beam is given basing on theory of Gaussian distribution, (3) the processing method of experiment data and selection of algorithm are given. The model could point the way to beam angular divergence measurement of EAST neutral beam injection.     

He-ICR cleanings on full metallic walls in EAST full superconducting tokamak

He-ICR cleanings were successfully carried out in 2006-2007 on the full superconducting tokamak (EAST), which is an ITER-relevant experimental tokamak. Factors influence on He-ICR efficiency, such as magnetic field, ICRF power and working pressure were investigated. In EAST, the breakdown pressure for He-ICR cleanings could reach 10 Pa. The removal efficiency for H during 20 kW 4.5 × 10⁻³ Pa He-ICR cleaning was same as that in He-GDC cleaning (2 Pa, 4A). Highest H removal rate in EAST with full metallic material walls, 1.7 × 10²² H/h in 20 kW 3 × 10⁻² Pa He-ICR, was higher by a factor of four than that in HT-7 with carbon limiter configuration.     

An Indian test facility to characterise diagnostic neutral beam for ITER

The diagnostic neutral beam (DNB) line shall be used to diagnose the He ash content in the D–T phase of the ITER machine using the charge exchange recombination spectroscopy (CXRS). Implementation of a successful DNB at ITER requires several challenges related to the production, neutralization and transport of the neutral beam over path lengths of 20.665 m, to be overcome. The delivery is aided if the above effects are tested prior to onsite commissioning. As DNB is a procurement package for INDIA, an ITER approved Indian test facility, INTF, is under construction at Institute for Plasma Research (IPR), India and is envisaged to be operational in 2015. The timeline for this facility is synchronized with the RADI, ELISE (IPP, Garching), SPIDER (RFX, Padova) in a manner that best utilization of configurational inputs available from them are incorporated in the design. This paper describes the facility in detail and discusses the experiments planned to optimise the beam transmission and testing of the beam line components using various diagnostics.     

HT-7超导托卡马克上聚变中子产额和中子注量率分布计算

利用所编制的三维托卡马克中子计算程序,计算了HT-7装置上中子产额随等离子体中心离子温度与密度的变化.同时计算了中子注量率及其在空间不同位置与方位角的分布特性.考虑了离子温度与密度分布的形状因子不同所带来的误差.计算结果为中子诊断和中子辐射防护等提供了依据.     

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.