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.     

Operational experience with forced cooled superconducting magnets

Force-cooled concept has been chosen for ITER superconducting magnet to get reliable coil insulation using vacuum-pressure impregnation (VPI) technology. However 17 breakdowns occurred during operation of six magnets of this type or their single coil tests at operating voltage < 3 kV, while ITER needs 12 kV. All the breakdowns started on electric, cryogenic and diagnostic communications (ECDCs) by the high voltage induced at fast current variations in magnets concurrently with vacuum deterioration, but never on the coils, though sometimes the latter were damaged too. It suggests that simple wrap insulation currently employed on ECDCs and planned to be used in ITER is unacceptable. Upgrade of the ECDC insulation to the same level as on the coils is evidently needed. This could be done by covering each one from ECDCs with vacuum-tight grounded stainless steel casings filled up with solid insulator using VPI-technology. Such an insulation will be insensitive to in-cryostat conditions, excluding helium leaks and considerably simplifying the tests thus allowing saving time and cost. However it is not accepted in ITER design yet. So guarantee of breakdown prevention is not available.     

Design issues of the High Voltage platform and feedthrough for the ITER NBI Ion Source

In the ITER heating Neutral Beam Injector (NBI), a High Voltage air-insulated platform (named High Voltage Deck, HVD) will be installed to host the Ion Source and Extractor Power supply system and associated diagnostics referred to ?1 MV DC potential. All power and control cables are routed from the HVD via a feedthrough (HV bushing) to the gas insulated transmission line which feeds the Injector.The paper focuses on insulation and mechanical issues for both HVD and HV bushing which are very special components, far from the present industrial standards as far as voltage (?1 MV DC) and dimensions are concerned. For this purpose, a preliminary design of the HVD has been carried out as concerns the mechanical structure and external shield. Then, the structure has been verified with a seismic analysis applying the seismic load excitation specified for the ITER construction site (Cadarache) and carrying out verifications according to relevant international standards. As regards the HV bushing design, proposals for the complex inner conductor structure and for interfaces to the HVD and transmission line are outlined; alternative installation layouts (aside or underneath the HVD) are compared from both mechanical and electrical points of view.     

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.     

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.     

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.     

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.     

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.     

Study and optimization of boronization in Alcator C-Mod using the Surface Science Station (S3)

A Surface Science Station (S³) on the Alcator C-Mod tokamak is used to study and optimize the location and rate of boron film deposition in situ during electron cyclotron (EC) discharge plasmas using 2.45 GHz radio-frequency (RF) heating and a mixture of helium and diborane (B₂D₆) gasses. The radial profile of boron deposition is measured with a pair of quartz microbalances (QMB) on S³, the faces of which can be rotated 360° including orientations parallel and perpendicular to the toroidal magnetic field B_T ～0.1 T. The plasma electron density is measured with a Langmuir probe, also on S³ in the vicinity of the QMBs, and typical values are ～1 × 10¹⁶ m^?3. A maximum boron deposition rate of 0.82 μg/cm²/min is obtained, which corresponds to 3.5 nm/min if the film density is that of solid boron. These deposition rates are sufficient for boron film applications between tokamak discharges. However the deposition does not peak at the EC resonance as previously assumed. Rather, deposition peaks near the upper hybrid (UH) resonance, ～5 cm outboard of the EC resonance. This has implications for RF absorption, with the RF waves being no longer damped on the electrons at the EC resonance. The previously inferred radial locations of critical erosion zones in Alcator C-Mod also need to be re-evaluated. The boron deposition profile versus major radius follows the ion flux/density profile, implying that the boron deposition is primarily ionic. The application of a vertical magnetic field (B_V ～0.01 T) was found to narrow the plasma density and boron deposition profiles near the UH resonance, thus better localizing the deposition. A Monte Carlo simulation is developed to model the boron deposition on the different QMB/tokamak surfaces. The model requires a relatively high boron ion gyroradius of ～5 mm, indicating a B⁺¹ ion temperature of ～2 eV, to match the deposition on QMB surfaces with different orientation to B_T. Additionally, the boron ion trajectories become de-magnetized at high neutral gas throughput (～0.5 Pa m³ s^?1) and pressure (～2 Pa) when the largest absolute deposition rates are measured, resulting in deposition patterns, which are independent of surface orientation to B_T in optimized conditions.     

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.     

AC operation of large titanium sublimation pumps in a magnetic field: Results of the test stand for the W7-X neutral beam injectors

A neutral beam injection (NBI) system is being built for the Stellarator experiment Wendelstein 7-X (W7-X) currently under construction at IPP Greifswald. The NBI system consists of two injectors which are essentially a replica of the system present in the Tokamak experiment ASDEX-Upgrade at IPP Garching. A vacuum system with high pumping speed and large capacity is required to ensure proper vacuum conditions in the neutral beam line. For this purpose, large titanium sublimation pumps (TSP) are installed inside the NBI boxes, consisting of 4 m long hanging wires containing Ti and the surrounding condensation walls. The wires are DC ohmically heated up with 142 A to Ti sublimation temperature. A TSP system has been operated since many years in the AUG-NBI system, sublimating Ti in the pauses between the plasma discharges, when no magnetic field is present. However, at W7-X the superconducting coils generate a magnetic field permanently during experimental campaigns, whose stray B field with a maximum of 30 mT, affects the TSPs. Operated with DC, the wires would be deflected against the surrounding panels due to the Lorentz force. A simple possible solution is heating with AC, which reduces the wire deflection amplitude, inducing a risky wire oscillation. The feasibility of the AC operation in an equivalently strong B field such as the stray B field around W7-X has been demonstrated in a test stand for different AC waveforms and frequencies. Several test campaigns have shown no qualitative difference in the pumping properties between AC and DC operation of the TSP and no critical dynamic behaviour of the wires.     

Development of high voltage power supply for the KSTAR 170 GHz ECH and CD system

A 3.6 MW (66 kV/55 A) DC power supply system was developed for the 170 GHz EC H&CD system in KSTAR. The power supply system consists of a cathode power supply (CPS), an anode power supply (APS) and a body power supply (BPS). The cathode power supply is capable of supplying a maximum voltage of ?66 kV and a current of 55 A to the cathode with respect to the collector using pulse step modulation (PSM). The high voltage switching system for the cathode is made by a fast MOS-FET solid-state switch which can turn off the high voltage to the cathode within 3 μs in the occurrence of gyrotron faults. The APS is a voltage divider system consisting of a fixed resistor and zener diode units with the capability of 60 kV stand-off voltage. The anode voltage with respect to the cathode is controlled in a range of 0–60 kV by turning the MOS-FET switches connected in parallel to each zener diode on and off. For high frequency current modulation of the gyrotron, the parallel discharge switch is introduced between the cathode and anode in order to clamp the charged voltage in the stray capacitance. The BPS is a DC power supply with the capability of 50 kV/160 mA. The nominal operation parameter of BPS was 23 kV and 10 mA, respectively, and the voltage output is regulated with a stability of 0.025% of the rated voltage. The series MOS-FET solid-state switch is used for on/off modulation in the body voltage sychronizing with anode voltage. The parallel discharge switch is also introduced between the body and collector for high frequency RF modulation. This paper describes the key features of the high voltage power supply system of the KSTAR 170 GHz gyrotron as well as the test results of the power supply.     

Initial plasma start-up using partial solenoid coils in Versatile Experiment Spherical Torus (VEST)

Initial plasma start-up experiments based on ohmic discharge using partial solenoid coils located at both vertical ends of a center stack have been carried out in Versatile Experiment Spherical Torus (VEST) at Seoul National University. Ohmic discharges with the help of microwave pre-ionization have been performed according to the pre-programed start-up scenario which was experimentally verified by a series of vacuum field measurements using an internal magnetic probe array. A plasma current of around 0.4 kA has been achieved by ohmic discharge using partial solenoid coils, under the toroidal magnetic field of 0.1 T. The vacuum field calculation and fast camera image have revealed that the small plasma current even with significant amount of loop voltage up to 9.7 V is attributed to the imbalance of poloidal field for equilibrium. Modification of the start-up scenario and upgrade of power supplies are proposed to be carried out in order to achieve higher plasma current in the future experiments.     

Ion implantation studies on VOx films prepared by pulsed dc reactive sputtering

Vanadium oxide (VO_x) thin films find extensive use in room-temperature bolometers for IR imaging. It is desirable to control and modify the electronic properties of this temperature-sensitive material with treatments such as ion implantation and thermal annealing. In this work, we report on the modification of structural and electrical properties of VO_x thin films of varying compositions, deposited by pulsed dc reactive sputtering using a vanadium target under different oxygen flow rates. The as-deposited resistivities of the films ranged from 0.1 Ω cm to 100 Ω cm and the temperature coefficient of resistance (TCR) values varied from ?1.1% to ?2.7%. VO_x films used in microbolometers need to have a high TCR (>2%) and low resistivity values (1–10 Ω cm) in order to maximize sensitivity in conjunction with the read-out integrated circuit (ROIC). However, one usually finds a high TCR associated with high resistivity. Hence ion implantation followed by annealing was performed with the goal of improving the trade-off between TCR and resistivity. Two species – hydrogen (active) and helium (inert) – were chosen for implantation. Hydrogen is strongly electroactive and is well known for passivating defect states in a wide variety of electronic materials. As inert species, helium was chosen mainly to study the effects of bombardment on the film. The implanted films were annealed in an inert atmosphere to allow defect control and redistribution of atoms, and then characterized by current–voltage measurements over a wide temperature range. An order of magnitude change in resistance, and significant variations in TCR were observed. Further characterization has been done by X-ray diffraction (XRD) and field emission scanning electron microscopy (FESEM) to correlate these resistivity changes with the structure of the films.     

Vacuum system of SST-1 Tokamak

Vacuum chambers of Steady State Superconducting (SST-1) Tokamak comprises of the vacuum vessel and the cryostat. The plasma will be confined inside the vacuum vessel while the cryostat houses the superconducting magnet systems (TF and PF coils), LN₂ cooled thermal shields and hydraulics for these circuits. The vacuum vessel is an ultra-high (UHV) vacuum chamber while the cryostat is a high-vacuum (HV) chamber. In order to achieve UHV inside the vacuum vessel, it would be baked at 150 °C for longer duration. For this purpose, U-shaped baking channels are welded inside the vacuum vessel. The baking will be carried out by flowing hot nitrogen gas through these channels at 250 °C at 4.5 bar gauge pressure. During plasma operation, the pressure inside the vacuum vessel will be raised between 1.0 × 10^?4 mbar and 1.0 × 10^?5 mbar using piezoelectric valves and control system. An ultimate pressure of 4.78 × 10^?6 mbar is achieved inside the vacuum vessel after 100 h of pumping. The limitation is due to the development of few leaks of the order of 10^?5 mbar l/s at the critical locations of the vacuum vessel during baking which was confirmed with the presence of nitrogen gas and oxygen gas with the ratio of ～3.81:1 indicating air leak. Similarly an ultimate vacuum of 2.24 × 10^?5 mbar is achieved inside the cryostat. Baking of the vacuum vessel up to 110 °C with ±10 °C deviation was achieved with a net mass flow rate of 0.8 kg/s at 1.5 bar gauge inlet pressure and supply temperature of 230 °C at the heater end. Also during gas feed system installation, the pressure inside the VV was raised from 3.01 × 10^?5 mbar to 1.72 × 10^?4 mbar by triggering a pulse of lower amplitude of 25 voltage direct current (VDC) for 100 s to piezoelectric valve. This paper describes in detail the design and implementation of the various vacuum subsystems including relevant experimental results.     

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.     

Negative ion beam extraction in ROBIN

The RF based single driver ?ve ion source experiment test bed ROBIN (Replica Of BATMAN like source in INDIA) has been set up at Institute for Plasma Research (IPR), India in a technical collaboration with IPP, Garching, Germany. A hydrogen plasma of density 5 × 10¹² cm^?3 is expected in driver region of ROBIN by launching 100 kW RF power into the driver by 1 MHz RF generator. The cesiated source is expected to deliver a hydrogen negative ion beam of 10 A at 35 kV with a current density of 35 mA/cm² as observed in BATMAN.In first phase operation of the ROBIN ion source, a hydrogen plasma has been successfully generated (without extraction system) by coupling 80 kW RF input power through a matching network with high power factor (cos θ > 0.8) and different plasma parameters have been measured using Langmuir probes and emission spectroscopy. The plasma density of 2.5 × 10¹¹ cm^?3 has been measured in the extraction region of ROBIN. For negative hydrogen ion beam extraction in second phase operation, extraction system has been assembled and installed with ion source on the vacuum vessel. The source shall be first operated in volume mode for negative ion beam extraction. The commissioning of the source with high voltage power supply has been initiated.     

Assessing the feasibility of a high-temperature,helium-cooled vacuum vessel and first wall for the Vulcan tokamak conceptual design

The Vulcan conceptual design (R = 1.2 m, a = 0.3 m, B₀ = 7 T), a compact, steady-state tokamak for plasma–material interaction (PMI) science, must incorporate a vacuum vessel capable of operating at 1000 K in order to replicate the temperature-dependent physical chemistry that will govern PMI in a reactor. In addition, the Vulcan divertor must be capable of handling steady-state heat fluxes up to 10 MW m^?2 so that integrated materials testing can be performed under reactor-relevant conditions. A conceptual design scoping study has been performed to assess the challenges involved in achieving such a configuration. The Vulcan vacuum system comprises an inner, primary vacuum vessel that is thermally and mechanically isolated from the outer, secondary vacuum vessel by a 10 cm vacuum gap. The thermal isolation minimizes heat conduction between the high-temperature helium-cooled primary vessel and the water-cooled secondary vessel. The mechanical isolation allows for thermal expansion and enables vertical removal of the primary vessel for maintenance or replacement. Access to the primary vessel for diagnostics, lower hybrid waveguides, and helium coolant is achieved through ～1 m long intra-vessel pipes to minimize temperature gradients and is shown to be commensurate with the available port space in Vulcan. The isolated primary vacuum vessel is shown to be mechanically feasible and robust to plasma disruptions with analytic calculations and finite element analyses. Heat removal in the first wall and divertor, coupled with the ability to perform in situ maintenance and replacement of divertor components for scientific purposes, is achieved by combining existing helium-cooled techniques with innovative mechanical attachments of plasma facing components, either in plate-type helium-cooled modules or independently bolted, helium-jet impingement-cooled tiles. The vacuum vessel and first wall design enables a wide range of potential PFC materials and configurations to be tested with relative ease, providing a new approach to reactor-relevant PMI science.     

Analysis of dark current images of a CMOS camera during gamma irradiation

We report on the behaviour of the dark current images of the Event Detection Intelligent Camera (EDICAM) when placed into an irradiation field of gamma rays. EDICAM is an intelligent fast framing CMOS camera operating in the visible spectral range, which is designed for the video diagnostic system of the Wendelstein 7-X (W7-X) stellarator. Monte Carlo calculations were carried out in order to estimate the expected gamma spectrum and dose for an entire year of operation in W7-X. EDICAM was irradiated in a pure gamma field in the Training Reactor of BME with a dose of approximately 23.5 Gy in 1.16 h. During the irradiation, numerous frame series were taken with the camera with exposure times 20 μs, 50 μs, 100 μs, 1 ms, 10 ms, 100 ms. EDICAM withstood the irradiation, but suffered some dynamic range degradation. The behaviour of the dark current images during irradiation is described in detail. We found that the average brightness of dark current images depends on the total ionising dose that the camera is exposed to and the dose rate as well as on the applied exposure times.     

Development of a plasma driven permeation experiment for TPE

Experiments on retention of hydrogen isotopes (including tritium) at temperatures less than 800 °C have been carried out in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory [1], [2]. To provide a direct measurement of plasma driven permeation in plasma facing materials at temperatures reaching 1000 °C, a new TPE membrane holder has been built to hold test specimens (≤1 mm in thickness) at high temperature while measuring tritium permeating through the membrane from the plasma facing side. This measurement is accomplished by employing a carrier gas that transports the permeating tritium from the backside of the membrane to ion chambers giving a direct measurement of the plasma driven tritium permeation rate. Isolation of the membrane cooling and sweep gases from TPE's vacuum chamber has been demonstrated by sealing tests performed up to 1000 °C of a membrane holder design that provides easy change out of membrane specimens between tests. Simulations of the helium carrier gas which transports tritium to the ion chamber indicate a very small pressure drop (∼700 Pa) with good flow uniformity (at 1000 sccm). Thermal transport simulations indicate that temperatures up to 1000 °C are expected at the highest TPE fluxes.