Tokamak active laser pyrometry for tungsten deposited layer characterisation

In modern fusion reactors, the erosion of plasma facing surface results in layers deposition on tokamak “cold” surfaces. To provide efficient operation of tokamaks, it is essential to characterise the deposited layer with high tritium content. In situ rapid surface characterisation without reactor components disassembly is required. Active laser pyrometry together with a repetition rate Nd–YAG laser (1 Hz–1 kHz repetition rate frequency) applied for surface heating can be used to characterise some thermo-physical properties (thermal capacity, thermal contact, and conductivity) of a micrometric layer. The pyrometer system was developed and applied to characterise some properties of a W-layer (140 μm) on a CFC-substrate. The numerical code developed for 3-D simulation of LH of a surface with the deposited layer was applied to simulate the experimental heating temperatures. The experimental and simulation results were compared. W-layer characterisation was performed by fitting the experimental and theoretical heating temperatures.     

Nanosecond laser ablation of Thoria fuel pellets for microstructural study

A two-dimensional numerical model has been developed simulating the process of laser based surface etching of Thoria targets via pulsed laser ablation enabling their surface preparation for subsequent metallographic investigation. The heat conduction equation solved by an explicit finite difference method provides simulated data on the temperature distribution at the surface and within the target, melt depth and evaporation rate from the target as a function of time, during and after the laser pulse. Calculations have been performed for laser and target parameters corresponding to experimental conditions matching our reported experimental observations on pulsed laser etching of Thoria pellets via laser ablation. The calculated maximum surface temperature reached by the laser treated Thoria target exceeds the estimated value of thermodynamic critical temperature of Thoria. Thus, our results on simulation of pulsed laser ablation for an average laser flux of 10 J/cm² delivered by a 10 ns Nd:YAG laser pulse corresponding to a peak laser intensity of 3.87 × 10⁹ W/cm² suggest, that explosive boiling could probably be an additional material-removal mechanism other than normal boiling and evaporation when surface etching Thoria with such intense laser radiation. Since explosive boiling is usually accompanied by intense material ejection, this mechanism of material-removal should be avoided to ensure minimum induced target damage associated with the technique of laser based etching. Our calculations thus help us to make a proper choice of laser parameters facilitating subsequent metallographic investigation of laser etched Thoria fuel pellets, at the same time, minimizing unwanted associated thermal effects such as target damage through crater formation, as has been experimentally observed.     

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%.     

Application of laser-induced breakdown spectroscopy for characterization of material deposits and tritium retention in fusion devices

Laser-induced breakdown spectroscopy (LIBS) is discussed as a possible method to characterize the composition, tritium retention and amount of material deposits on the first wall of fusion devices. The principle of the technique is the ablation of the co-deposited layer by a laser pulse with P (power density) ≥ 0.5 GW/cm² and the spectroscopic analysis of the light emitted by the laser induced plasma. The typical spatial extension of the laser plasma plume is in the order of 1 cm with typical plasma parameters of n_e ≈ 3 × 10²² m^?3 and T_e ≈ 1–2 eV averaged over the plasma lifetime which is below 1 μs. In this study “ITER-Like” mixed deposits with a thickness of about 2 μm and consisting of a mixture of W/Al/C and D on bulk tungsten substrates have been analyzed by LIBS to measure the composition and hydrogen isotopes content at different laser energies, ranging from about 2 J/cm² (0.3 GW/cm²) to about 17 J/cm² (2.4 GW/cm²) for 7 ns laser pulses. It is found that the laser energies above about 7 J/cm² (1 GW/cm²) are needed to achieve the full removal of the deposit layer and identify a clear interface between the deposit and the bulk tungsten substrate by applying 15–20 laser pulses while hydrogen isotopes decrease strongly after the first laser pulse. Under these conditions, the evolution of the spectral line intensities of W/Al/C/hydrogen can be used to evaluate the layer composition.     

Synergistic effects of hydrogen plasma exposure,pulsed laser heating and temperature on rhodium surfaces

The combined effect of hydrogen plasma exposure and surface heating, either continuous or by short laser pulses (5 ns), on the surface morphology of rhodium layers has been studied. Investigations were performed by reflectivity measurements, scanning electron microscopy (SEM), X-ray photoelectron spectroscopy (XPS) and atomic force microscope (AFM). While surfaces exposed at room temperature exhibit little modifications, strong surface changes are observed for surface temperatures higher than 250 °C. At 500 °C, the plasma exposed surface exhibits a nanoscale structure (50–100 nm) with a high level of porosity and a low reflectivity. Additional laser irradiation of the surface strongly enhances the observed surface damage. Localized surface melting is observed with craters extending deep into the substrate together with a dense network of voids.     

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.     

Aerosol formation and hydrogen co-deposition by colliding ablation plasma plumes of lithium and lead

In a high-repetition inertial fusion reactor, along with pellet implosions, the interior of target chamber is to be exposed to high-energy, short pulses of X-ray, unburned DT and He ash particles and pellet debris. As a result, wall materials will be subjected to ablation, ejecting particles in the plasma state to collide with each other in the center of volume. The interaction dynamics of ablation plasmas of lithium and lead, candidate first wall materials, has been investigated in the deposited energy density range from 3 to 10 J/cm²/pulse at a repetition rate of 10 Hz, each 6 ns long. The plasma density and electron temperature of colliding ablation plumes have been found to vary from the order of 10⁸–10¹³ 1/cm³ and from 0.7 to 1.5 eV, respectively. The formation of aerosol in the form of droplet has been observed with diameters between 100 nm and 10 μm. Also, hydrogen co-deposition has been found to occur particularly for colliding plumes of lithium, resulting in the H/Li atomic ratio from 0.15 to 0.27 in the hydrogen partial pressure range from 10 to 50 Pa.     

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.     

Nanostructured rhodium films for advanced mirrors produced by Pulsed Laser Deposition

In this paper advantages in the production by Pulsed Laser Deposition (PLD) of nanostructured, nanoengineered rhodium films to be used in tokamak First Mirrors (FMs) are shown. The peculiar PLD capability to tailor film structure at the nanoscale gives the possibility to deposit low roughness Rh films with a wide variety of structures and morphologies. By a proper movimentation of the substrate and using high fluence (10–19 J/cm²) infrared laser pulses, it has been possible to deposit planar and homogeneous Rh films effectively suppressing surface defects on areas of the order of 10 cm² with a satisfactory specular reflectivity. Multilayer deposition has been exploited to produce coatings with high adhesion and good mechanical properties. Finally, an estimation of the requirements to produce by PLD rhodium films suitable for the requests of ITER is provided.     

Ge and Ti post-ion acceleration from laser ion source

Laser ion sources (LIS) are employed with success to generate, in vacuum, Ge and Ti ion beams with high current, ion energy, charge states and directivity.Nanoseconds infrared laser pulses, with intensities of the order of 10¹⁰ W/cm², induce high ablation in Ge and Ti targets. Ions are produced in vacuum with energy distribution following the Coulomb–Boltzmann-shifted distribution and they are ejected mainly along the normal to the target surface. The free ion expansion process occurs in a constant-potential chamber placed at 30 kV positive voltage. An electric field of 5 kV/cm was used to accelerate the ions emitted from the plasma at INFN-LNS laser facility. Time-of-flight technique is employed to measure the mean ion energies of the post-accelerated particles. Ion charge states and energy distributions were measured through an ion energy spectrometer.     

Recent contribution of JET to the ITER physics

In recent years the JET scientific programme has focussed on addressing physics issues essential for the consolidation of design choices and the efficient exploitation of ITER in parallel to qualifying ITER operating scenarios and developing advanced control tools. This paper reports on recent achievements in the following areas: mitigation of edge localised modes (ELMs), effects of toroidal field (TF) ripple, advanced tokamak scenarios, material migration and fuel retention. Active methods have been developed to mitigate ELMs without adversely affecting confinement. A systematic characterisation of the edge plasma, pedestal energy and ELMs, and their impact on plasma-facing components as well as their compatibility with material limits has been performed. The unique JET capability of varying the TF ripple from its normal low value δB_T = 0.08% up to δB_T = 1% has been used to elucidate the role of TF ripple on confinement and ELMs. Increased TF ripple in ELMy H-mode plasmas is found to have a detrimental effect on plasma stored energy and density, especially at low collisionality. The development of ITER advanced tokamak scenarios has been pursued. In particular, β_N values above the ‘no-wall limit’ (β_N ～ 3.0) have been sustained for a resistive time. Gas balance studies combined with shot-resolved measurements from deposition monitors and divertor spectroscopy have confirmed the strong role of fuel co-deposition with carbon in the retention mechanism through long-range migration and also provided further evidence for the important role of ELMs in the material migration process within the JET inner divertor leg.     

Lithium technologies for edge plasma control

We have investigated two new modes of operation been in T-10 limiter tokamak experiments with a novel rotary feeder of lithium dust. Quasi steady-state mode I and pulse mode II of dust delivery were realized in both OH and OH + ECRH disruption free plasmas at the lithium flow rate up to 2 × 10²¹ atoms/s. A higher flow rate in mode II with injection rate of ～5 × 10²¹ atoms/s caused a series of minor disruptions, which was completed by discharge termination after the major disruption. The observed decreases of bolometer and D_β signals, with increase of the electron density during the lithium dust injection, reveal the effects of the first wall conditioning. The lithium technology may provide inherent safety pathway for major disruption mitigation in a tokamak reactor, which requires demonstration in contemporary tokamak experiments.     

Operational characteristics of the high flux plasma generator Magnum-PSI

In Magnum-PSI (MAgnetized plasma Generator and NUMerical modeling for Plasma Surface Interactions), the high density, low temperature plasma of a wall stabilized dc cascaded arc is confined to a magnetized plasma beam by a quasi-steady state axial magnetic field up to 1.3 T. It aims at conditions that enable fundamental studies of plasma–surface interactions in the regime relevant for fusion reactors such as ITER: 10²³–10²⁵ m⁻² s⁻¹ hydrogen plasma flux densities at 1–5 eV. To study the effects of transient heat loads on a plasma-facing surface, a high power pulsed magnetized arc discharge has been developed. Additionally, the target surface can be transiently heated with a pulsed laser system during plasma exposure. In this contribution, the current status, capabilities and performance of Magnum-PSI are presented.     

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.     

The shaping of a national ignition campaign pulsed waveform

The National Ignition Facility (NIF) at Lawrence Livermore National Laboratory is a stadium-sized facility containing a 192 beam, 1.8 MJ, 500 TW ultraviolet laser system used for inertial confinement fusion research. For each experimental shot, NIF must deliver a precise amount of laser power on the target for successful and efficient target ignition, and these characteristics vary depending on the physics of the particular campaign. The precise temporal shape, energy and timing characteristics of a pulsed waveform target interaction are key components in meeting the experimental goals. Each NIF pulse is generated in the Master Oscillator Room (MOR) using an electro-optic modulator to vary the intensity of light in response to an electrical input. The electrical drive signal to the modulator is produced using a unique, high-performance arbitrary waveform generator (AWG). This AWG sums the output of 140 electrical impulse generators, each producing a 300 ps pulse width Gaussian signal separated in time by 250 ps. By adjusting the amplitudes and summing the 140 impulses, a pulsed waveform can be sculpted from a seed 45 ns square pulse. Using software algorithms written for NIF's Integrated Computer Control System (ICCS), the system is capable of autonomously shaping 48 unique experimental pulsed waveforms for each shot that have demonstrated up to 275:1 contrast ratio with ±3% absolute error averaged over any 2 ns interval, meeting the stringent pulse requirements needed to achieve ignition. In this paper, we provide an overview of the pulse shaping system, software algorithms and associated challenges that have been overcome throughout the evolution of the controls.     

Manufacturing of self-passivating W-Cr-Si alloys by mechanical alloying and HIP

Self-passivating tungsten-based alloys may provide a major safety advantage in comparison with pure tungsten, which is presently the main candidate material for the plasma-facing protection of future fusion power reactors. WCrSi alloys were manufactured by mechanical alloying (MA) and HIP at 1300 °C and 200 MPa for 1 h. Different MA conditions were investigated to obtain powders with lowest possible amount of contaminants and small and homogeneous particle and crystallite size. Milling in WC vials under Ar without process control agent provided best results. After HIP densities close to 100% were obtained. First oxidation tests on preliminary alloys showed self-passivating behavior with rates comparable to WCrSi thin films at 800 °C but worse performance at 1000 °C. In all cases a Cr₂WO₆ protective layer is formed at the surface.     

The JET programme in support of ITER

As part of its mission to prepare the operation of ITER, a major programme of enhancements has just been completed on the JET tokamak. These enhancements include a complete replacement of the plasma-facing components in JET, from carbon-based to the combination of beryllium and tungsten foreseen for ITER, an upgrade of the neutral beam heating available on JET from 20 MW/short pulse to 30 MW/long pulse operation, the installation of a high frequency pellet injection system for plasma fuelling and ELM control studies, an upgrade to the JET vertical stability system and a suite of new diagnostics.The future JET programme is foreseen to proceed progressively from a test of fuel retention in the standard regimes of ITER operation towards more aggressive, high performance experiments that will demonstrate the operating space limits with the new wall. Depending on the results of the earlier experiments, the exploitation of the enhancements is foreseen to be completed with a deuterium-tritium experiment. This would represent the most integrated test of ITER operational scenarios possible before ITER itself.JET is a cooperative programme funded and exploited in collaboration by all of the European fusion laboratories. As such, JET is a test bed for multi-national use of a single fusion facility, as is foreseen for ITER. Opportunities for broadening the participation in JET to other ITER Parties are presently being explored. If these opportunities can be implemented, JET would provide not only an integrated test of ITER regimes of operation but also a demonstration of how ITER will be operated, even to the extent of including significant numbers of the same team who will eventually operate ITER.     

Laser damage thresholds of ITER mirror materials and first results on in situ laser cleaning of stainless steel mirrors

A laser ablation system has been constructed and used to determine the damage threshold of stainless steel, rhodium and single-, poly- and nanocrystalline molybdenum in vacuum, at a number of wavelengths between 220 nm and 1064 nm using 5 ns pulses. All materials show an increase of the damage threshold with decreasing wavelength below 400 nm. Tests in a nitrogen atmosphere showed a decrease of the damage threshold by a factor of 2–3. Cleaning tests have been performed in vacuum on stainless steel samples after applying mixed Al/W/C/D coatings using magnetron sputtering. In situ XPS analysis during the cleaning process as well ex situ reflectivity measurements demonstrate near complete removal of the coating and a substantial recovery of the reflectivity. The first results also show that the reflectivity obtained through cleaning at 532 nm may be further increased by additional exposure to UV light, in this case 230 nm, an effect which is attributed to the removal of tungsten dust from the surface.     

Coupling of electron cyclotron waves to the desired mode in plasma of HL-2A tokamak

3 MW/68 GHz electron cyclotron resonance heating and current driving systems have been developed on HL-2A tokamak and two new units which can operate at two frequencies, 140 GHz and 105 GHz are on-going. Dual-polarizers for changing of the polarization of the wave beam have been designed for these systems based on the integral method of diffraction gratings. Finally, coupling between the electron cyclotron waves and the plasma has been discussed and the results indicate that the desired mode can be obtained for all possible experiments associated with the electron cyclotron resonance heating and current driving systems on HL-2A.     

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.