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.     

EAST ion cyclotron resonance heating system for long pulse operation

Radio frequency (RF) power in the ion cyclotron range of frequencies (ICRF) is one of the primary auxiliary heating techniques for Experimental Advanced Superconducting Tokamak (EAST). The ICRF system for EAST has been developed to support long-pulse high-β advanced tokamak fusion physics experiments. The ICRF system is capable of delivering 12 MW 1000-s RF power to the plasma through two antennas. The phasing between current straps of the antennas can be adjusted to optimize the RF power spectrum. The main technical features of the ICRF system are described. Each of the 8 ICRF transmitters has been successfully tested to 1.5 MW for a wide range of frequency (25–70 MHz) on a dummy load. Part of the ICRF system was in operation during the EAST 2012 spring experimental campaign and a maximum power of 800 kW (at 27 MHz) lasting for 30 s has been coupled for long pulse H mode operation.     

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.     

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.     

The lower hybrid current drive system for steady-state operation of the Vulcan tokamak conceptual design

The steady-state current drive system for the Vulcan tokamak concept has been designed, taking into account requirements of high field, small size, and high operational wall temperature (B₀ = 7 T, R₀ = 1.2 m, T_wall > 800 K). This lower hybrid current drive system allows steady-state operation by utilizing high field side launch, high RF source frequency (8 GHz), and dedicated current drive ports. An iterative MHD and current drive solver is used to determine the ideal launching spectra and location to assure strong single pass absorption. It is found that with nominal Vulcan operational parameters (n_e ≈ 4 × 10²⁰ m^?3, T_e ≈ 2.8 keV, I_p = 1.7 MA, P_LHCD = 19.8 MW) bootstrap currents of ～70% and lower hybrid current drive efficiencies of 1.16 × 10¹⁹ A W m^?2 could be achieved. The optimized solution yielded advanced tokamak profiles with q values on-axis above 2. A conceptual design of the system is presented, which takes into account space, power, cooling, and launched spectrum requirements. The system is found to be compatible with the vacuum vessel design and requires cooling power of <1 MW per waveguide bundle.     

W7-X neutral-beam-injection: Selection of the NBI source positions for experiment start-up

The stellarator W7-X will be equipped with two Neutral-Beam-Injector (NBI) boxes for balanced injection. Each NBI box has 2 tangential and 2 radial source positions. For the experimental start-up phase each NBI box will be only equipped with 2 ion sources. For the selection of the initial 2 NBI source positions per box three physical aspects were examined (transmission and duct power deposition, shine through and heating efficiency).Using hydrogen injection the heating power to the plasma under typically planned conditions should be 1.3 MW for the tangential sources and 1.1 MW for the radial sources (deuterium: 2 MW for the tangential sources, 1.8 MW for the radial sources). The tangential source positions all have similar heating efficiencies. One of them suffers from the lowest duct transmission (highest power-load to the duct). The same source hits a component with a low power-load capability. The W7-X inner wall design will be modified in order to enhance the maximum power-load capability of that component. For the radial source positions there is no clear physics advantage of one position over the other. Taking all aspects into consideration the decision was made to use one tangential source and one radial source per box during the experimental start-up phase.     

Design and thermal performance of an improved mechanically attached module for divertor plate of LHD

A mechanically attached divertor module with improved performance has been developed and partially installed in the large helical device (LHD). The advantage of the new design is to eliminate any metal bolts on the armor tile surface which leads to high Z impurity emission. The new module consists of a couple of armor tiles made of iso-graphite and a thin graphite sheet, which are tightened by two stainless steel (SS) or TZM bolts horizontally sandwiching a SS cooling pipe. The heat flows directly from the tile surface to the cooling pipe. The previous design used a copper heat sink has been used. Steady heat flux tests up to 1.2 MW/m² have been carried out for the new module without any trouble using a electron gun ACT while the allowable heat flux of the previous module was limited below 0.3 MW/m² due to deformation of the copper heat sink. The outgassing from the new module during heat flux tests up to 0.5 MW/m² is decreased to about one-third. Thermal fatigue test up to 115 cycles under a steady heat flux of 1 MW/m² for the new module has been performed without any troubles. Comparisons between the previous and new modules for the structure, thermal performance, and outgassing are made.     

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%.     

FAST: A European ITER satellite experiment in the view of DEMO

Fusion Advanced Studies Torus (FAST) aims to contribute to the exploitation of ITER and to explore innovative DEMO technology. FAST has been designed to study, in an integrated scenario: (a) relevant plasma-wall interaction problems, with a large power load (P/R ～ 22 MW/m; P/R² ～ 12 MW/m²) and with a full metallic wall; (b) to tackle operational problems in regimes with relevant fusion parameters; (c) to investigate the non-linear dynamics of fast particles (alpha like) in burning plasmas. FAST will operate on a wide parameters range, namely in high performance H-mode (BT ～ 8.5 T; IP ～ 8 MA) as well as in advanced Tokamak operation up to full non-inductive current scenario (IP ～ 2 MA). The main heating is based on 30 MW ICRH, but the ports have been designed to allocate up to 20 MW of 1 MeV NNBI. Helium gas at 30 K is used for cooling of the full machine, a preliminary analysis shows the possibility of realizing FAST with a complete superconductor set of coils. An innovative active system is under development to reduce and to control the magnetic ripple. Tungsten (W) or liquid lithium (L–Li) has been chosen for the divertor material plates and the code EDGE2D has been used to optimize the divertor geometry.     

High heat flux testing of mock-ups for a full tungsten ITER divertor

In order to evaluate the option to start the ITER operation with a full tungsten (W) divertor, the EU-DA launched an extensive R&D program. It consisted in its initial phase in the high heat flux (HHF) testing of W mock-ups and medium scale prototypes up to 20 MW/m² in the AREVA FE 200 facility (F). Critical heat flux (CHF) experiments were carried out on the items which survived the above thermal fatigue testing.After 1000 cycles at 10 MW/m², the full W Plasma Facing Components (PFCs) mock-ups successfully sustained either 1000 cycles at 15 MW/m² or 500 cycles at 20 MW/m².However, some significant surface melting, as well as the complete melting of a few monoblocks, occurred during the HHF thermal fatigue testing program representative of the present ITER requirements for the strike point region, namely 1000 cycles at 10 MW/m² followed by 1000 cycles at 20 MW/m².The results of the CHF experiments were also rather encouraging, since the tested items sustained heat fluxes in the range of 30 MW/m² in steady-state conditions.     

ICRF physics aspects of wall conditioning plasma characterization in TEXTOR

This paper focuses on encouraging results obtained on the characterization of RF produced plasmas during pulsed-mode wall conditioning discharges in ion cyclotron resonance frequency (ICRF) regime in the limiter tokamak TEXTOR. Recent Ion Cyclotron Wall Conditioning (ICWC) experiment carried out in TEXTOR tokamak, lead to the identification of various dependences of the antenna-plasma coupling efficiency on the plasma parameters for possible ICWC-discharge cleaning in ITER at half field. Our ICWC experiments emphasize on (i) study of antenna coupling during the mode conversion scenario, (ii) reproducible generation of ICRF plasmas for wall conditioning, by coupling RF power from one or two ICRF antennas and (iii) effect of application of an additional (along with toroidal magnetic field) stationary vertical (B_V ? B_T) or oscillating poloidal magnetic field (B_p ? B_T) on antenna coupling and relevant plasma parameters.     

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.     

Determination of differential cross-sections for the natK(p, p0) and 39K(p, α0) reactions in the backscattering geometry

In the present work, new, differential cross-section values are presented for the ^natK(p, p₀) reaction in the energy range E_lab = 3000–5000 keV (with an energy step of 25 keV) and for detector angles between 140° and 170° (with an angular step of 10°). A qualitative discussion of the observed cross-section variations through the influence of strong, closely spaced resonances in the p + ³⁹K system is also presented. Information has also been extracted concerning the ³⁹K(p,α₀) reaction for E_lab = 4000–5000 keV in the same angular range. As a result, more than ～500 data points will soon be available to the scientific community through IBANDL (Ion Beam Analysis Nuclear Data Library – http://www-nds.iaea.org/ibandl/) and could thus be incorporated in widely used IBA algorithms (e.g. SIMNRA, WINDF, etc.) for potassium depth profiling at relatively high proton beam energies.     

Effect of 10 MW/m2 plasma exposure on W-SiC/SiC dual layer tiles

W-SiC/SiC dual layer tile has many advantages as a high heat flux component (HHFC) material for fusion, in theory. However, due to insufficient data known, its high potentiality and near term availability has not been well recognized. This work provides the recent materials R&D status and the first plasma exposure test result from the world largest helical device, large helical device of National Institute for Fusion Science in Japan. Tungsten armor with SiC/SiC substrate layer survived during the LHD plasma exposure with 10 MW/m² maximum heat load for the 5.3-s operation cycle. The macro and microstructure evolution, including crack and pore formation, was analyzed and an excellent high heat load resistance was demonstrated.     

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.     

Thermal effects and component cooling in Magnum-PSI

Magnum-PSI is a linear plasma generator, built at the FOM-Institute for Plasma Physics Rijnhuizen. Subject of study will be the interaction of plasma with a diversity of surface materials. The machine is designed to provide an environment with a steady state high-flux plasma (up to 10²⁴ H⁺ ions/m² s) in a 3 T magnetic field with an exposed surface of 80 cm² up to 10 MW/m². Magnum-PSI will provide new insights in the complex physics and chemistry that will occur in the divertor region of the future experimental fusion reactor ITER and reactors beyond ITER. The conditions at the surface of the sample can be varied over a wide range, such as plasma temperature, beam diameter, particle flux, inclination angle of the target, background pressure and magnetic field. An important subject of attention in the design of the machine was thermal effects originating in the excess heat and gas flow from the plasma source and radiation from the target.     

Design and analysis of the W7-X divertor scraper element

Thehigh heat-flux divertor of the Wendelstein 7-X large stellarator experiment consists of 10 divertor units which are designed to carry a steady-state heat flux of 10 MW/m². However, the edge elements of this divertor are limited to only 5 MW/m², and may be overloaded in certain plasma scenarios. It is proposed to reduce this heat by placing an additional “scraper element” in each of the ten divertor locations. It will be constructed using carbon fiber composite (CFC) monoblock technology. The design of the monoblocks and the path of the cooling tubes must be optimized in order to survive the significant steady-state heat loads, provide adequate coverage for the existing divertor, be located within sub-millimeter accuracy, and take into account the boundaries to other in vessel components, all at a minimum cost. Computational fluid dynamics modeling has been performed to examine the thermal transfer through the monoblock swirl tube channels for the design of the monoblock orientation. An iterative physics modeling and computer aided design process is being performed to optimize the placement of the scraper element within the severe spatial restrictions.     

Vulcan: A steady-state tokamak for reactor-relevant plasma–material interaction science

An economically viable magnetic-confinement fusion reactor will require steady-state operation and high areal power density for sufficient energy output, and elevated wall/blanket temperatures for efficient energy conversion. These three requirements frame, and couple to, the challenge of plasma–material interaction (PMI) for fusion energy sciences. Present and planned tokamaks are not designed to simultaneously meet these criteria. A new and expanded set of dimensionless figures of merit for PMI have been developed. The key feature of the scaling is that the power flux across the last closed flux surface P/S ? 1 MW m^?2 is to be held constant, while scaling the core volume-averaged density weakly with major radius, n ～ R^?2/7. While complete similarity is not possible, this new “P/S” or “PMI” scaling provides similarity for the most critical reactor PMI issues, compatible with sufficient current drive efficiency for non-inductive steady-state core scenarios. A conceptual design is developed for Vulcan, a compact steady-state deuterium main-ion tokamak which implements the P/S scaling rules. A zero-dimensional core analysis is used to determine R = 1.2 m, with a conventional reactor aspect ratio R/a = 4.0, as the minimum feasible size for Vulcan. Scoping studies of innovative fusion technologies to support the Vulcan PMI mission were carried out for three critical areas: a high-temperature, helium-cooled vacuum vessel and divertor design; a demountable superconducting toroidal field magnet system; and a steady-state lower hybrid current drive system utilizing a high-field-side launch position.     

Mechanical design of the second ICRF antenna for EAST

In order to satisfy the requirements of heating plasma on EAST project, 3 MW ion cyclotron range of frequency (ICRF) heating system will be available at the second stage. Based on this requirement, the second ICRF antenna, has been designed for EAST. The antenna which is planned to operate with a frequency ranging from 30 MHz to 110 MHz, comprises four poloidal current straps. The antenna has many cooling channels inside the current straps, faraday shield and baffle to remove the dissipated RF loss power and incoming plasma heat loads. The antenna is supported via a cantilever support box to the external support structure. Its assembly is plugged in the port and fixed on the support box. External slideway and bellows allow the antenna to be able to move in the radial direction. The key components of the second ICRF antenna has been designed together with structural and thermal analysis presented.     

PIGE analysis of magnesium and beryllium

In this work, we present an alternative method for PIGE analysis of magnesium and beryllium in thick samples. This method is based on the ERYA – Emitted Radiation Yield Analysis – code, which integrates the nuclear reaction excitation function along the depth of the sample. For this purpose, the excitations functions of the ²⁵Mg(p,p′γ)²⁵Mg (E_γ = 585 keV) and ⁹Be(p,γ)¹⁰B (E_γ = 718 keV) reactions were employed. Calculated gamma-ray yields were compared, at several proton energy values, with experimental yields for thick samples made of inorganic compounds containing magnesium or beryllium. The agreement is better than 5%. Taking into consideration the experimental uncertainty of the measured yields and the errors related to the stopping power values, this agreement shows that effects as the beam energy straggling, ignored in the calculation, seem to play a minor role.