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.     

Direct numerical simulation of turbulent mixed convection in a vertical channel in a wall-normal magnetic field

Liquid metal coolants have a significant role in the design of advanced fusion reactors. There is a need for an investigation of the thermal behavior of the liquid metal in working reactor environment, such as when fluid flow at low Prandtl number (Pr) with a buoyancy effect, is subjected to a magnetic field. In the present study, a direct numerical simulation (DNS) for a low Pr number fluid flow resulting in turbulent heat transfer with buoyancy effect under a magnetic field has been carried out between two vertical plates kept at different temperatures. In this simulation, the values of the Hartmann number (Ha) were 0 and 6, Pr number was 0.06 and Grashof numbers were 6.4 × 10⁵, 9.6 × 10⁵, and 1.6 × 10⁶. The turbulent quantities of the parameters such as the mean temperature, turbulent heat flux, and temperature variance were obtained by direct numerical simulation (DNS). The Reynolds number (Re) for channel flow based on friction velocity averaged by both walls, viscosity, and channel half-width was set to be constant as Re_τ* = 150. A uniform magnetic field was applied in a direction perpendicular to the walls of the channel. The profiles of mean velocity and velocity fluctuations became asymmetric, and the tendency was enhanced with the increasing buoyancy effect. However, by the application of a magnetic field the tendency decreased. In other words, thermal transport between the walls became weak due to the magnetic effect.     

High current and low q95 scenario studies for FAST in the view of ITER and DEMO

The Fusion Advanced Study Torus (FAST) has been proposed as a possible European satellite, in view of ITER and DEMO, in order to: (a) explore plasma wall interaction in reactor relevant conditions, (b) test tools and scenarios for safe and reliable tokamak operation up to the border of stability, and (c) address fusion plasmas with a significant population of fast particles. A new FAST scenario has been designed focusing on low-q operation, at plasma current I_P = 10 MA, toroidal field B_T = 8.5 T, with a q₉₅ ≈ 2.3 that would correspond to I_P ≈ 20 MA in ITER. The flat-top of the discharge can last a couple of seconds (i.e. half the diffusive resistive time and twice the energy confinement time), and is limited by the heating of the toroidal field coils. A preliminary evaluation of the end-of-pulse temperatures and of the electromagnetic forces acting on the central solenoid pack and poloidal field coils has been performed. Moreover, a VDE plasma disruption has been simulated and the maximum total vertical force applied on the vacuum vessel has been estimated.     

Revisiting the poloidal magnetic confinement

This note proposes a closed poloidal magnetic configuration with an in-vessel coil system held by shielded supports. A dipole field is bounded by external coils and constrained into a hollow torus aiming at uniform intensity. In the horizontal mid-plane region the external coils and the dipole outer coils are broken in four arcs and bridged by couple of straight branches. Arcs and straight branches build a set of four side coils. In the clearance between their straight branches four tunnels in the poloidal magnetic field are achieved, to pass the supports and the feeders of the in-vessel coil system.A poloidal machine with a plasma thick as those of present large experiments is outlined. The dipole radius is 5.4 m, the plasma about it has a constant poloidal cross-section about 40 m², a volume about 1300 m³ and a minimum thickness 1 m in the outboard. The magnetic field ranges from 1.4 to 1.8 T.     

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.     

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.     

Direct numerical simulation of turbulent liquid metal flow entering a magnetic field

This paper presents direct numerical simulations (DNS) of fully developed turbulent liquid-metal flow in a circular duct entering a magnetic field. The case of a magnetohydrodynamic flow leaving a strong magnetic field has been extensively studied experimentally and numerically owing to its similarity to typical flow configurations appearing in liquid metal blankets of nuclear fusion reactors. Although also relevant to the design of fusion reactor blankets, the flow entering the fringing field of a magnet remains unexplored because its high intricacy precludes any simplification of the governing equations. Indeed, the complexity of the magnetohydrodynamic–turbulence interaction can only be analysed by direct numerical simulations or experiments. With that purpose, this paper addresses the case of a fully developed turbulent flow (Re_τ ≈ 520) entering low, intermediate and strong magnetic fields under electrically insulating and poorly conducting walls by means of three-dimensional direct numerical simulations. Purely hydrodynamic computations (without the effect of the magnetic field) reveal an excellent agreement against previous experimental and numerical results. Current MHD results provide a very detailed information of the turbulence decay and reveal new three-dimensional features related to liquid-metal flow entering strong increasing magnetic fields, such as flow instabilities due to the effect of the Lorentz forces within the fringing region at high Ha numbers.     

Simulation of runaway electron generation and diffusion during major disruptions in the HL-2A tokamak

The generation and diffusion of runaway electrons (REs) during major disruptions in the HL-2A tokamak has been studied numerically. The diffusion caused by the magnetic perturbation is especially addressed. The simulation results show that the strong magnetic perturbation (δB/B ∼ 1.0 × 10⁻³) can cause a significant loss of REs due to the radial diffusion and restrain the RE avalanche effectively. The results also indicate that the REs are generated initially in the plasma core during disruptions, and that the toroidal electric field does not exhibit a centrally hollow phenomenon. In addition, it is found that the toroidal effects have little impact on the generation of RE and the evolution of toroidal electric field.     

Numerical study of corrosion of ferritic/martensitic steels in the flowing PbLi with and without a magnetic field

A computational suite called TRANSMAG has been developed to address corrosion of ferritic/martensitic steels and associated transport of corrosion products in the eutectic alloy PbLi as applied to blankets of a fusion power reactor. The computational approach is based on simultaneous solution of flow, energy and mass transfer equations with or without a magnetic field, assuming mass transfer controlled corrosion and uniform dissolution of iron in the flowing PbLi. First, the new tool is applied to solve an inverse mass transfer problem, where the saturation concentration of iron in PbLi at temperatures up to 550 °C is reconstructed from the experimental data on corrosion in turbulent flows without a magnetic field. As a result, a new correlation for the saturation concentration C^S has been obtained in the form C^S = e^{13.604–12975/T}, where T is the temperature of PbLi in K and C^S is in wppm. Second, the new correlation is used in the computations of corrosion in laminar flows in a rectangular duct in the presence of a strong transverse magnetic field. As shown, the mass loss increases with the magnetic field such that the corrosion rate in the presence of a magnetic field can be a few times higher compared to purely hydrodynamic flows. In addition, the corrosion behavior was found to be different between the side wall of the duct (parallel to the magnetic field) and the Hartmann wall (perpendicular to the magnetic field) due to formation of high-velocity jets at the side walls. The side walls experience a stronger corrosion attack demonstrating a mass loss up to 2–3 times higher compared to the Hartmann walls. Also, computations of the mass loss are performed to characterize the effect of the temperature (400–550 °C) and the flow velocity (0.1–1 m/s) on corrosion in the presence of a strong 5 T magnetic field prototypic to the outboard blanket conditions.     

Preliminary design of the ITER ECH Upper Launcher

The design of the ITER electron cyclotron launchers recently reached the preliminary design level - the last major milestone before design finalization. The ITER ECH system contains 24 installed gyrotrons providing a maximum ECH injected power of 20 MW through transmission lines towards the tokamak. There are two EC launcher types both using a front steering mirror; one equatorial launcher (EL) for plasma heating and four upper launchers (UL) for plasma mode stabilization (neoclassical tearing modes and the sawtooth instability). A wide steering angle range of the ULs allows focusing of the beam on magnetic islands which are expected on the rational magnetic flux surfaces q = 1 (sawtooth instability), q = 3/2 and q = 2 (NTMs).In this paper the preliminary design of the ITER ECH UL is presented, including the optical system and the structural components. Highlights of the design include the torus CVD-diamond windows, the frictionless, front steering mechanism and the plasma facing blanket shield module (BSM). Numerical simulations as well as prototype tests are used to verify the design     

Study on flow instability for feasibility of a thin liquid film first wall

This study proposes a probability of the evaporated gas that agitates a growing instability wave in a thin liquid film first wall. The liquid first wall was considered to be in vacuum and the effect of the ambient gas was neglected but the evaporated gas by the high energy fluxes is a probable cause of unstable wave agitation. The criterion is approximately expressed by the density ratio (Q₂) and the Weber number (We) as Q₂ × We^0.5 ≈ 5 × 10⁻⁴. Performed indirect experimental supported this criterion. For a case study of liquid Pb-17Li film with a velocity of 10 m/s, the evaporated gas pressure must be below 6.2 × 10³ Pa to maintain stable conditions. By recent study, this pressure is generated at 1600 K temperature and it is believed to be attainable by the energy fluxes on the first wall. This result is so far not confirmed so the full verification by experimental is to be performed.     

Computational studies of thermoelectric MHD driven liquid lithium flow in metal trenches

The LiMIT system (Lithium/Metal Infused Trenches) is an innovative plasma-facing component for tokamak divertors, recently proposed at the University of Illinois. Thanks to the coupling of two metals having different Seebeck coefficients, the device is able to generate internal thermoelectric currents as a response to an incoming heat flux from the plasma. One of the two metals is liquid lithium and the second metal is a solid composing the trenches (tungsten, or molybdenum, or stainless steel, etc.). Together with the high toroidal magnetic field, the liquid lithium is propelled by a JxB electrodynamic force inside the solid trenches. In the present work we present a numerical characterization of the device. The diffusion–advection of heat is solved together with the Navier–Stokes equations forced by the JxB electrodynamic force, comprising the thermoelectric contribution. We report parametric plots to show the influence of the toroidal magnetic field and of the plasma heat flux. It is found that the average flow velocity of the liquid lithium peaks at a critical magnetic field, always lower than 1.0 T, and then decreases with an inverse law in the range of tokamak-relevant fields. The flow velocity of the lithium increases with a square-root law versus an increasing heat flux. The heat transfer coefficient of the cooling channels is parametrically investigated, revealing that coefficients higher than >4000 W/m² K are needed for the device in order to withstand heat fluxes of 10 MW/m².     

Effect of magnetic field on the growth of Be films prepared by thermal evaporation

Grain refinement of beryllium deposits is studied as a significant subject for beryllium capsule in the Inertial Confinement Fusion project. The Be films were prepared on the Si (1 0 0) substrates by thermal evaporation with and without a magnetic field, respectively. The two separate groups of prepared Be films were characterized. The results showed the grain diameter in the Be film transited from 300 nm to 18 nm and the surface roughness of the Be film decreased from 61 nm to 3 nm by application of the magnetic field during the deposition process of Be coating. However, the Be film grown with the magnetic field was easily oxidized in comparison with that grown without magnetic field due to the refined grains, and the oxidation was gradually decreased with the increase of etching depth in the Be film. The reason for grain refinement of Be film was also qualitatively described.     

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.     

Carbon film growth and hydrogenic retention of tungsten exposed to carbon-seeded high density deuterium plasmas

Tungsten (W) targets have been exposed to high density (n_e ? 4 × 10¹⁹ m^?3), low temperature (T_e ? 3 eV) CH₄-seeded deuterium (D) plasma in Pilot-PSI. The surface temperature of the target was ～1220 K at the center and decreased radially to ～650 K at the edges. Carbon film growth was found to only occur in regions where there was a clear CII emission line, corresponding to regions in the plasma with T_e ? 2 eV. The maximum film thickness was ～2.1 μm after a plasma exposure time of 120 s. ³He nuclear reaction (NRA) analysis and thermal desorption spectroscopy (TDS) determine that the presence of a thin carbon film dominates the hydrogenic retention properties of the W substrate. Thermal desorption spectroscopy analysis shows retention increasing roughly linearly with incident plasma fluence. NRA measures a C/D ratio of ～0.002 in these films deposited at high surface temperatures.     

Boron lattice site location in (BGa)As and (BGa)P thin films studied using RBS and NRA with a channeled He+ ion beam

In this study the boron lattice site location in ternary B_xGa_1?xAs and B_xGa_1?xP thin films grown on (0 0 1) GaAs and (0 0 1) GaP, respectively, using low pressure metal-organic vapour-phase epitaxy (MOVPE) with boron concentrations between x = 0.8% and x = 3.2% was investigated with RBS and the ¹⁰B(α,p)¹³C nuclear reaction using a 2.3 MeV He⁺ ion beam. For this purpose, the ion beam was aligned with the [0 0 1], [0 1 1] and [1 1 1] axis and the RBS and proton yield from the nuclear reaction compared with random ion incidence. For comparison, theoretical proton yields which assume boron to be located on substitutional lattice sites only were calculated for each sample/axis combination and compared with the experimental yields. The RBS/channeling measurements show a very good crystal quality of the films with χ_min being in the range of 3–5% for the [0 1 1] axis. The best crystal qualities, i.e. the lowest χ_min values and dechanneling rates, are achieved for low boron concentrations. From NRA/channeling it can be deduced that in the B_xGa_1?xAs films the fraction of interstitial boron is approximately 5% for low boron concentrations of x = 1% and 6–10% for concentrations up to x = 3.2%, whereas the fraction of interstitial boron is less than 3% in the B_xGa_1?xP film studied despite a concentration of x = 2.0%. This indicates that antisite effects of the boron incorporation are more likely in GaAs compared to GaP.     

Experimental study of onset of subcooled annular flow boiling

An experimental study on the onset of nucleate boiling (ONB) is performed for water annular flow to provide a systematic database for low pressure and velocity conditions. A parametric study has been conducted to investigate the effect of pressure, inlet subcooling, heat and mass flux on flow boiling. The test section includes a Pyrex tube with 21 mm inner diameter and a stainless steel (SS-304) rod with outer diameter of 6 mm. Pressure, heat and mass flux are in the range of 1.73 < P < 3.82 bar, 40 < q < 450 kW/m² and 70 < G < 620 kg/m² s, respectively. The results illustrate that inception heat flux is extremely dependent on pressure, inlet subcooling temperature and mass flux; for example in pressure, velocity and inlet subcooling as 3.27 bar, 230 kg/m² s and 41.3 °C; consequently q_w,ONB is 177.3 kW/m². In other case with higher inlet temperature of 71.5 °C and with P, 3.13 bar and G, 232 kg/m² s the inception heat flux reached to 101.6 kW/m². The data of ONB heat flux are over estimated from the existing correlation, and maximum deviation of wall superheat (ΔT_w,ONB) from correlations is 30%. Experimental data of inception heat flux are within 22% of that predicted from the correlation.     

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.     

Influence of thermal performance on design parameters of a He/LiPb dual coolant DEMO concept blanket design

Spanish Breeding Blanket Technology Programme TECNO_FUS is exploring the technological capabilities of a Dual-Coolant He/Pb15.7Li breeding blanket for DEMO and studying new breeding blanket design specifications. The progress of the channel conceptual design is being conducted in parallel with the extension of MHD computational capabilities of CFD tools and the underlying physics of MHD models. A qualification of MHD effects under present blanket design specifications and some approaches to their modelling were proposed by the authors in [1]. The analysis was accomplished with the 2D transient algorithm from Sommeria and Moreau [2] and implemented in the OpenFOAM CFD toolbox [3]. The thermal coupling was implemented by means of the Boussinesq hypothesis. Previous analyses showed the need of improvement of FCI thickness and thermal properties in order to obtain a desirable liquid metal temperature gain of 300 °C. In the present study, an assessment through sensitivity and parametric analyses of the required FCI thickness is performed.Numerical simulations have been carried out considering a Robin-type thermal boundary condition which assumes 1D steady state thermal balance across the solid FCI and Eurofer layers. Such boundary condition has been validated with a fluid–solid coupled domain analysis.Results for the studied flow conditions and channel dimensions show that, in order to obtain a liquid metal temperature gain of about 300 °C, the required FCI material should have a very small effective heat transfer coefficient ((k/δ) ≤ 1 W/m²K) and fluid velocities should be about 0.2 m/s or less. Moreover, special attention has to be placed on the temperature difference across the FCI layer. However, for a maximised liquid metal thermal gain, higher velocities would be preferable, what would also imply a reduced temperature difference across the FCI layer.     

A Systematic study of modular coil characteristics for 2-field periods quasi-axisymmetric stellarator QAS-LA

Modular coil characteristics of a 2-field periods quasi-axisymmetric stellarator QAS-LA configuration with an aspect ratio A_p = 3, magnetic pressure ∼4% and rotational transform ι ∼ 0.15 per field period supplied by its own shaping have been detailed studied. In addition, the characteristics of modular coils for QAS-LA were compared with those of an intermediate QA configuration QAS-LAx and a tokamak based on the same center magnet field B₀, aspect ratio and number of coils. As expected, the B_max/B₀, force F and overturning moment M, increase with the increased complexity of the coil shape. The relationships between the modular coils’ parameters (such as radius curvature ρ, distance from coil to coil Δ_c–c and the cross-section of coils) and the electromagnetic characteristics have been systematically summarized. The approximate formula for the maximum magnetic field in the coil body as functions of modular coil parameters (Δ_c–c, ρ) was derived for a simple two wire system which will be useful when optimizations of coil properties are called for.