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%.     

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.     

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.     

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.     

A small,novel, remote in-vessel inspection system for the Alcator C-Mod tokamak

A small robust system has been constructed for in-situ visual inspection of the Alcator C-Mod tokamak. The system consists of a small, light, wide-angle high definition camera and LED package housed in a nacelle on the end of thin, rigid, 3.5 m long support pole. The nacelle has two actuated degrees of freedom allowing the camera to observe nearly 4π steradians. The support pole has a specific slight curve that allows it to pass to either side of the center column of the tokamak to observe the entirety of the vessel interior, while still fitting through the small aspect ratio Alcator C-Mod vacuum port structure. The support pole and camera can enter the vessel through any horizontal vacuum port with an inner diameter greater than 4 cm, thus a dedicated port is not required. The inspection is typically undertaken during maintenance periods when the vessel is filled with a noble gas near atmospheric pressure thus minimizing the influx of water vapor and the concomitant loss of wall conditioning. The system is operated manually, producing photos and video which are reviewed in near real-time. Nearly the entire vessel, including the plasma facing components, can be carefully inspected in 3–5 h. The system provides improved characterization of the interior components and surfaces of the tokamak with a modest engineering and operational effort. Information gathered from the system has identified damage to plasma facing components that were interfering with tokamak operation, as well as damage to mechanical components which were redesigned during the remainder of the campaign, thereby enhancing program planning.     

A bulk tungsten tile for JET: Heat flux tests in the MARION facility on the power-handling performance and validation of the thermal model

In the frame of the ITER-like Wall (ILW) for the JET tokamak, a divertor row made of bulk tungsten material has been developed for the position where the outer strike point is located in most of the foreseen plasma configurations. In the absence of active cooling, this represents a formidable challenge when one considers the temperature reached by tungsten (T_W,surf > 2000 °C) and the vertical gradient ?T/?z = 5 × 10⁴ K/m.As the development is drawing to an end and most components are in production, actual 1:1 prototypes are exposed to an ion beam with a power density around 7 MW/m² on the plasma-facing surface. Advantage is taken of the flexibility of the Marion facility to bombard the tungsten stack under shallow angles of incidence (～6°) with a powerful beam of ions and neutrals (>70 MW/m² on axis). The shallow angles are important, with respect to the toroidal wetted surface, for properly simulating the expected performance under actual tokamak conditions. The Marion tests have been used to validate for a few typical cases the thermal calculations that were steadily developed along with the tungsten tile and, at the same time, to gather information on the actual temperatures of individual components. The latter is an important factor to a finer estimation of the power handling capabilities.     

Engineering design for the HL-2M tokamak components

The HL-2A tokamak will be modified into HL-2M. The Bt at the plasma center (major radius R = 1.78 m) is 2.2 T, the minor radius is 0.65 m. The plasma current I_P of HL-2M will reach up to 2.5 MA, the elongation and triangularity is more than 1.8 and more than 0.5, respectively. The vacuum vessel torus consists of 20 sectors with “D” shaped cross-section and double wall structure. 20 toroidal field coil bundles comprise 140 turns which are designed with demountable joints, the poloidal field coils system consists of 25 coils. The engineering design and calculation for field coil system, vacuum vessel, support structure, etc. are finished, many key issues for manufacture process have been discussed with industry and the fabrication of main components of HL-2M tokamak will be carried out in factories.     

An initial study of demountable high-temperature superconducting toroidal field magnets for the Vulcan tokamak conceptual design

Recent developments have made it possible to consider high-temperature superconductor (HTS) for the design of tokamak toroidal field (TF) magnet systems, potentially influencing the overall design and maintenance scheme of magnetic fusion energy devices. Initial assessments of the engineering challenges and cryogenic-dependent cost and parameters of a demountable, HTS TF magnet system have been carried out using the Vulcan tokamak conceptual design (R = 1.2 m, a = 0.3 m, B₀ = 7 T) as a baseline. Jointed at the midplane to allow vertical removal of the primary vacuum vessel and routine maintenance of core components, structural D-shaped steel support cases provide cryogenic cooling for internally routed YBCO superconducting cables. The cables are constructed by layering ～50 μm thick commercially available YBCO tape, and the interlocking steel support cases self align during assembly to form internal resistive joints between YBCO cables. It is found that designing the TF magnet system for operation between 10 K and 20 K minimizes the total capital and operating cost. Since YBCO is radiation-sensitive, Monte Carlo simulation is used to study advanced shielding materials compatible with the small size of Vulcan. An adequate shield is determined to be 10 cm of zirconium borohydride, which reduces the nuclear heating of the TF coils by a factor of 11.5 and increases the YBCO tape lifetime from two calendar years in the unshielded case to 42 calendar years in the shielded case. Although this initial study presents a plausible conceptual design, future engineering work will be required to develop realistic design solutions for the TF joints, support structure, and cryogenic system.     

Design,integration and feasibility studies of the Tore-Supra West divertor structure

Actively cooled tungsten plasma facing components will be used in the ITER divertor. In order to fully validate such a technology (industrial manufacturing, operation with long plasma duration), the implementation of a tungsten axis symmetric divertor in the tokamak Tore-Supra is studied. With this major upgrade, so called WEST (Tungsten Environment in Steady state), Tore-Supra will be the only European tokamak able to address the problematic of long plasma discharges with an actively cooled metallic divertor.To do so, it is planned to install two symmetric divertor coils inside the vacuum vessel. This assembly, called divertor structure, is made up of two stainless steel casings containing a copper winding pack cooled by hot pressurized water (200 °C, 4 MPa). These two casings are located at the top and bottom of the vacuum vessel in order to create two magnetic X-point areas, which are protected by W-PFCs (Tungsten Plasma Facing Components) in order to extract the thermal loads. The two casing are robustly maintained together by 18 brackets in order to constitute a rigid assembly attached thanks to 12 legs (one per lower vertical port) outside the Tore_Supra vacuum vessel.The paper will illustrate the technical developments performed during 2011 in order to produce a preliminary design of the Tore-Supra WEST divertor structure with a particular focus on: the mechanical design of this major component and its integration in the Tokamak, the manufacturing issues and the technical results of the feasibility studies done with industry as well as the design of a scale one coil mock up.     

Progress in development and design of the neutral beam injector for JT-60SA

Neutral beam (NB) injectors for JT-60 Super Advanced (JT-60SA) have been designed and developed. Twelve positive-ion-based and one negative-ion-based NB injectors are allocated to inject 30 MW D⁰ beams in total for 100 s. Each of the positive-ion-based NB injector is designed to inject 1.7 MW for 100 s at 85 keV. A part of the power supplies and magnetic shield utilized on JT-60U are upgraded and reused on JT-60SA. To realize the negative-ion-based NB injector for JT-60SA where the injection of 500 keV, 10 MW D⁰ beams for 100 s is required, R&Ds of the negative ion source have been carried out. High-energy negative ion beams of 490–500 keV have been successfully produced at a beam current of 1–2.8 A through 20% of the total ion extraction area, by improving voltage holding capability of the ion source. This is the first demonstration of a high-current negative ion acceleration of >1 A to 500 keV. The design of the power supplies and the beamline is also in progress. The procurement of the acceleration power supply starts in 2010.     

Manufacturing and testing of W/Cu mono-block small scale mock-up for EAST by HIP and HRP technologies

ITER-like W/Cu mono-block plasma-facing components (PFCs) will be used in vertical target regions of the experimental advanced superconducting tokamak (EAST) divertor. The first W/Cu mono-block small scale mock-up with five W mono-blocks has been manufactured successfully by technological combination of hot isostatic pressing (HIP) and hot radial pressing (HRP). The joining of a W mono-block and a pure copper interlayer was achieved by means of HIP technology and the bonding strength was over 150 MPa. The good bonding between the pure copper interlayer and a CuCrZr cooling tube was obtained by means of HRP technology. In order to understand deeply the process of HRP, the stress distribution of the mock-up during HRP process was simulated using ANSYS code. Ultrasonic Nondestructive Testing (NDT) of the W/Cu and Cu/CuCrZr interfaces was performed, showing that excellent bonding of the W/Cu and Cu/CuCrZr interfaces. The thermal cycle fatigue testing of the mock-up has been carried out by means of an e-beam device in Southwest Institute of Physics, Chengdu (SWIP) and the mock-up withstood 1000 cycles of heat loads up to 8.4 MW/m² with the cooling water of 2 m/s, 20 °C, 0.2 MPa.     

Design and construction of Alborz tokamak vacuum vessel system

The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. At the heart of the tokamak is the vacuum vessel and limiter which collectively are referred to as the vacuum vessel system. As one of the key components for the device, the vacuum vessel can provide ultra-high vacuum and clean environment for the plasma operation. The VV systems need upper and lower vertical ports, horizontal ports and oblique ports for diagnostics, vacuum pumping, gas puffing, and maintenance accesses. A limiter is a solid surface which defines the edge of the plasma and designed to protect the wall from the plasma, localizes the plasma–surface interaction and localizes the particle recycling. Basic structure analyses were confirmed by FEM model for dead weight, vacuum pressure and plasma disruptions loads. Stresses at general part of the VV body are lower than the structure material allowable stress (117 MPa) and this analysis show that the maximum stresses occur near the gravity support, and is about 98 MPa.     

Applications to cultural heritage diagnostics at the new nuclear microprobe beam line at CEDAD

A nuclear microprobe beam line has been installed at CEDAD (Centre for Dating and Diagnostics), University of Salento, Lecce, Italy. The beam line is connected to the ?30° port of the high energy switching magnet of a 3 MV HVEE 4130HC Tandetron accelerator. It is based on an Oxford Microbeam magnetic quadrupole triplet and its general features are presented. The results of functional tests are presented showing how a lateral spatial resolution as low as ～2 μm has been achieved in vacuum by analysing standard reference material. The results obtained in the analysis of ancient radiocarbon dated biological tissues are presented for the identification and distribution of toxic elements such as Pb.     

RF calculation and thermal-stress analysis of the ECRH launcher on EAST tokamak

EAST is a medium sized superconducting tokamak with major radius R = 1.8 m, minor radius a = 0.45 m, plasma current I_p ≤ 1 MA, toroidal field B_T ≤ 3.5 T and expected plasma pulse length up to 1000 s. An electron cyclotron resonance heating (ECRH) launcher for four-beam injection is being installed on EAST tokamak. Four electron cyclotron wave beams which are generated from four sets of 140 GHz/1 MW/1000 s gyrotrons will be injected into the plasma by the spherical focusing mirrors and plane mobile mirrors. The focusing mirrors are spherical to focus Gaussian beams after reflection. Four plane mobile mirrors independently steer continuously in the poloidal and toroidal direction controlled by motors. With the suitable distance between mirrors and appropriate focal length of focusing mirror, the beam radius in the resonance layer of plasma is 31.145 mm. The heat from plasma radiation and metal losses is loaded on the mobile mirror. In order to decrease the temperature and thermal stress, the inner equivalent diameter of water channels is 8 mm and the suggested water velocity is 4 m/s.     

Nuclear micro-beam analysis of deuterium distribution in carbon fibre composites for controlled fusion devices

Probes made of carbon fibre composite NB41 were exposed to deuterium plasmas in the TEXTOR tokamak and in a simulator of plasma–wall interactions, PISCES. The aim was to assess the deuterium retention and its lateral and depth distribution. The analysis was performed by means of D(³He, p)⁴He and ¹²C(³He, p)¹⁴N nuclear reactions analysis using a standard (1 mm spot) and micro-beam (20 μm resolution). The measurements have revealed non uniform distribution of deuterium atoms in micro-regions: differences by a factor of 3 between the maximum and minimum deuterium concentrations. The differences were associated with the orientation and type of fibres for samples exposed in PICSES. For surface structure in the erosion zone of samples exposed to a tokamak plasma the micro-regions were more complex. Depth profiling has indicated migration of fuel into the bulk of materials.     

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.     

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.     

Introduction to resonant magnetic perturbation coils of the J-TEXT Tokamak

To investigate the interactions between both the static and rotating resonant magnetic perturbations (RMP) and the tokamak plasma, two sets of coils, namely static RMP (SRMP) and dynamic RMP (DRMP), are constructed on the J-TEXT tokamak. SRMP is reconstructed from TEXT-U and mainly produces static m/n = 1/1, 2/1 and 3/1 resonant perturbation field, where m and n are the poloidal and toroidal mode numbers, respectively. DRMP, newly designed and installed inside the vacuum vessel, can generate pure 2/1 RMP. DRMP is also designed to operate in the AC mode and can produce rotating 2/1 RMP which will be used to study the tearing mode control. Due to the effect of the eddy current in the vacuum vessel wall, the amplitudes of the 2/1 component will be attenuated to about 1/3.6 of the DC value when the operation frequency is larger than 500 Hz. However, DRMP can still provide sufficient large rotating 2/1 perturbation for tearing mode related studies.