Surface Of Plasma Facing Components Research Articles

Diagnostic options for radiative divertor feedback control on NSTX-U

A radiative divertor technique is used in present tokamak experiments and planned for ITER to mitigate high heat loads on divertor plasma-facing components (PFCs) to prevent excessive material erosion and thermal damage. In NSTX, a large spherical tokamak with lithium-coated graphite PFCs and high divertor heat flux (q(peak) ≤ 15 MW/m(2)), radiative divertor experiments have demonstrated a significant reduction of divertor peak heat flux simultaneously with good core H-mode confinement using pre-programmed D(2) or CD(4) gas injections. In this work diagnostic options for a new real-time feedback control system for active radiative divertor detachment control in NSTX-U, where steady-state peak divertor heat fluxes are projected to reach 20-30 MW/m(2), are discussed. Based on the NSTX divertor detachment measurements and analysis, the control diagnostic signals available for NSTX-U include divertor radiated power, neutral pressure, spectroscopic deuterium recombination signatures, infrared thermography of PFC surfaces, and thermoelectric scrape-off layer current. In addition, spectroscopic "security" monitoring of possible confinement or pedestal degradation is recommended. These signals would be implemented in a digital plasma control system to manage the divertor detachment process via an actuator (impurity gas seeding rate).

The Materials Analysis Particle Probe (MAPP) Diagnostic System in NSTX

Lithium conditioning of plasma-facing surfaces has been implemented in National Spherical Torus Experiment (NSTX) leading to improvements in plasma performance such as reduced D recycling and a reduction in edge localized modes. Analysis of postmortem tiles and offline experiments along with atomistic modeling has identified interactions between Li-O-D and Li-C-D as chemical channels for deuterium retention in ATJ graphite. However, previous surface chemistry analysis of NSTX tiles were conducted postmortem (i.e., after a completed annual campaign), and it was not possible to correlate the performance of particular discharges with the state of the material surface at the time. Materials Analysis Particle Probe (MAPP) is the first in-vacuo surface analysis diagnostic directly integrated into a tokamak and capable of chemical surface analysis of plasma facing samples retrieved from the vessel in between discharges. It uses X-ray photoelectron spectroscopy, direct recoil spectroscopy, low energy ion surface spectroscopy, and thermal desorption spectroscopy to investigate the chemical functionalities between D and lithiated graphite at both the near surface (5-10 nm) and top surface layer (0.3-0.6 nm), respectively. MAPP will correlate plasma facing component surface chemistry with plasma performance and lead the way to improved understanding of plasma-surface interactions and their effect on global plasma performance. Remote operation and data acquisition, integrated into NSTX diagnostic and interlocks, make MAPP an advanced PMI diagnostic with stringent engineering constraints.

Lithium wall conditioning and surface dust detection on NSTX, and dust removal

Lithium evaporation onto National Spherical Torus Experiment (NSTX) plasma-facing components (PFCs) has resulted in improved energy confinement, and reductions in the number and amplitude of edge-localized modes (ELMs) up to the point of complete ELM suppression. The associated PFC surface chemistry has been investigated with a novel plasma–material interface probe connected to an in-vacuo surface analysis station. Analysis has demonstrated that the binding of D atoms to the polycrystalline graphite material of PFCs is fundamentally changed by lithium—in particular, deuterium atoms become weakly bonded near lithium atoms themselves bound to either oxygen or the carbon from the underlying material. Surface dust inside NSTX has been detected in real time using a highly sensitive electrostatic dust detector. In a separate experiment, electrostatic removal of dust via three concentric spiral-shaped electrodes covered by a dielectric and driven by a high-voltage three-phase waveform was evaluated for its potential application to fusion reactors.

Development of laser lock-in thermography for plasma facing component surface characterisation

Infrared (IR) photothermal techniques are candidates for in situ characterisation of tokamak plasma facing components (PFC) surfaces, by means of an external thermal excitation coupled with an IR temperature measurement. Among these techniques, the laser lock-in thermography (LLIT) uses a modulated laser excitation which gives 2 major advantages: enhancement of signal to noise ratio and emissivity independence, which is a plus when the components have various and unpredictable surface quality. With this method, it is possible to develop a process, which could be used remotely, either mounted onto an in situ inspection device (articulated arm) or in a PFC test bed. This paper presents the results obtained with a continuous modulated laser heat source on particular samples (W coating on CFC substrate, C layer on graphite substrate). The identification of the experimental data with a theoretical model allows a quantitative characterisation of the layers.

Theoretical studies of macroscopic erosion mechanisms of melt layers developed on plasma facing components

During plasma instabilities in tokamak devices, metallic plasma facing components (PFC) undergo surface vaporization and melting. Macroscopic losses of melt layers are of a serious concern to the lifetime of PFC, the damage of nearby components, and potential core plasma contamination. A normal or inclined plasma stream flowing at the melt layer surface of PFC at very high velocities (∼10 5 m/s) can induce Kelvin–Helmholtz (K–H) instabilities. We present an extensive linear stability theory and capillary droplet ejection model adapted to the problem of melt layer erosion and splashing. Based on this linear analysis, the stability criterion is established accounting the influence of the thicknesses of both plasma stream and melt layer. The growth rate of the most unstable wave is investigated with respect to different parameters such as plasma density and velocity, material properties, and melt layer thickness. A capillary droplet ejection model is then developed and used to analytically estimate the erosion rate of the melt layer for tungsten and aluminum targets. The present work brings a detailed understanding of the onset of K–H instabilities developed in melt layers due to plasma stream impact and builds a theoretical basis to estimate a macroscopic erosion rate, material losses and lifetime for PFC.

Wall reflection issues for optical diagnostics in fusion devices

The problem of light reflection has been raised as a high priority issue for optical diagnostics in next step fusion devices where metallic wall environment will generate significant perturbations in the diagnostics measurements. Tore Supra is a large size tokamak equipped with water-cooled stainless-steel panels used to sustain the plasma long shot radiations. These panels are highly reflective and affect significantly optical systems. In particular, we show that the infrared imaging diagnostic, which surveys the plasma facing component surface temperature for safety purposes, can give incorrect information due to reflected light coming from the bottom limiter. In the visible range, motional Stark effect and Zeff measurements experience important drifts during the plasma heating phases due to parasitic light coming from the limiter, but also from the plasma itself when the viewing lines are facing the reflecting walls. In the next step fusion devices such as ITER, the possibility to use optical measurements needs to be accessed by a modeling of the diagnostic light in its machine environment and the development of new techniques of online correction.

Particle Balance and Fuel Retention in Tore Supra

Fuel retention in carbon plasma-facing components (PFCs) is such a major concern for next-step operation that it could prevent the use of this material in the D-T phase of ITER. Because of its complete set of actively cooled PFCs, Tore Supra offers a unique opportunity to study this phenomenon in conditions where the plasma exposure time is much longer than the thermal equilibration time of the PFCs. In addition to the main characteristics of permanent retention measured during long-discharge operation, this paper discusses the different mechanisms possibly at work in the continuous increase of the in-vessel inventory and describes the morphology and physical properties of the deposits found at several locations in the vacuum chamber. The main results are (1) that D retention mainly depends on the lower hybrid power coupled to the plasma and, to a lesser extent, on the edge temperature and fueling method, (2) that permanent D retention is mainly due to codeposition, and (3) that the hydrogenated carbon deposits present at the surface of the different PFCs are strongly disorganized graphite carbons when they are exposed to high heat fluxes, whose formation occurs through a heterogeneous growth involving both codeposition of nanoparticles and basic structural unit vapor condensates.

Preliminary assessment of possible carbide formation on Be and T contaminated CFC tiles from JET

The fusion plasma, with a typical temperature of 10 keV, has to be brought into contact with a physical wall in order to remove the helium produced and drain the excess energy in the fusion plasma. The fusion plasma is far too hot to be brought into direct contact with a physical wall. It would degrade the wall and the debris from the wall would extinguish the plasma. Therefore, schemes are developed to cool down the plasma locally before it impacts on a physical surface. The resulting plasma wall-interaction in ITER is facing several challenges including surface erosion, material redeposition and tritium retention. In JET facility both beryllium and carbon are used together as plasma facing materials, and the resultant surfaces show considerable mixing between the two elements. In this work a first XPS characterisation has been performed on the surface of several PFC (plasma facing components) samples obtained from various JET tiles from the Mk II GB divertor in JET. The substrates of the tiles are CFC, and the deposited films have various D/C ratios depending on the position in the tokamak. Ion beam analysis shows that the D/C ratio is ∼1 in the shadowed areas and Be/C ∼1–2 in the plasma-exposed regions of the inner divertor. These values could give rise to the formation of beryllium carbides, and indications of a carbide component are found in the XPS spectra. The processes for material migration and the morphology of the deposits are discussed.

A possible method of carbon deposit mapping on plasma facing components using infrared thermography

The material eroded from the surface of plasma facing components is redeposited partly close to high heat flux areas. At these locations, the deposit is heated by the plasma and the deposition pattern evolves depending on the operation parameters. The mapping of the deposit is still a matter of intense scientific activity, especially during the course of experimental campaigns. A method based on the comparison of surface temperature maps, obtained in situ by infrared cameras and by theoretical modelling is proposed. The difference between the two is attributed to the thermal resistance added by deposited material, and expressed as a deposit thickness. The method benefits of elaborated imaging techniques such as possibility theory and fuzzy logics. The results are consistent with deposit maps obtained by visual inspection during shutdowns.

Plasma/Liquid-Metal Interactions During Tokamak Operation

One of the critical technological challenges of future tokamak fusion devices is the ability for plasma-facing components to handle both normal and abnormal plasma/surface interaction events that compromise their lifetime and operation of the machine. Under normal operation plasma/surface interactions that are important include: sputtering, particle implantation and recycling, He pumping and ELM (edge localized modes)-induced erosion. In abnormal or off-normal operation: disruptions and vertical displacement events (VDEs) are important. To extend PFC lifetime under these conditions, liquid-metals have been considered as candidate PFCs (Plasma-Facing Components), including: liquid lithium, tin-lithium, gallium and tin.Liquid lithium has been measured to have nonlinear increase of physical sputtering with rise in temperature. Such increase can be a result of exposure to ELM-level particle fluxes. The significant increase in particle flux to the divertor and nearby PFCs can enhance sputtering erosion by an order of magnitude or more. In addition from the standpoint of hydrogen recycling and helium pumping liquid lithium appears to be a good candidate plasma-facing material (PFM). Advanced designs of first wall and divertor systems propose the application of liquid-metals as an alternate PFC to contend with high-heat flux constraints of large-scale tokamak devices. Additional issues include PFC operation under disruptions and long temporal instabilities such as VDEs. A comprehensive two-fluid model is developed to integrate core and SOL (scrape-off layer) parameters during ELMs with PFC surface evolution using the HEIGHTS package. Special emphasis is made on the application of lithium as a candidate plasma-facing liquid-metal.

Hydrogen and helium entrapment in flowing liquid metal plasma-facing surfaces

The ability to use liquids as plasma-facing component (PFCs) depends on their interaction with the plasma and the magnetic field. One important issue for the moving liquid is the ability to entrain particles that strike the PFC surface (helium and hydrogen isotopes) while accommodating high heat loads. To study this problem, a numerical model has been developed using the HEIGHTS computer simulation package. The model was used to investigate pumping of He particles by the flowing liquid rather than requiring a standard vacuum system. Hydrogen isotope (DT) particles are likely be trapped in the liquid metal surface (e.g., lithium) due to the high chemical solubility of hydrogen. The incident He particles in the established low-recycling regime at PFCs could be harder to pump using standard vacuum pumping techniques. The analysis results indicate, however, a reasonable chance of adequate helium self-trapping in flowing lithium as PFC without active pumping.

MHD problems in free liquid surfaces as plasma-facing materials in magnetically confined reactors

The development of solid plasma-facing components (PFCs) that can withstand high heat and particle fluxes during normal and abnormal events has proven to be a difficult task for future power-producing magnetically confined reactors. Solid PFC cannot be reliably used because of the large erosion losses during normal and off-normal events such as disruptions and their consequences. The use of liquid metal surfaces for protection of PFCs seems attractive but is not easily implemented. The ability to use liquids as PFC surfaces depends on their overall integrated interaction with the plasma as well as with the strong magnetic field in the reactor. The temperature of a flowing liquid surface governed by surface velocity should be low enough to avoid core plasma contamination by the metal vapor. A high liquid flow velocity, V, may also be necessary to overcome the force F=[ J× B], where J is current and B is magnetic field. In this study, MHD flow patterns of the liquid metal across the magnetic field ( B=5 T) are investigated. The retarding force F=[ J× B] results in a rather high pressure drop required to move the liquid metal across the magnetic field. New two-dimensional calculations of MHD effects and resulting pressure drop have been completed with the upgraded heights software package.

Tritium Decontamination of TFTR D-T Plasma Facing Components Using an Ultra Violet Laser

ABSTRACTTritium decontamination of the surface of plasma facing components used during the deuterium-tritium (D-T) phase of the Tokamak Fusion Test Reactor (TFTR) was investigated using an ultra violet (UV) laser with a wavelength of 193 nm, a pulse energy of 200 mJ, a pulse duration of 25 ns and a beam size of 2.3 cm by 0.7 cm. Tritium was released immediately after the samples were irradiated by the UV laser. An initial spike of tritium release was observed within 40 seconds for each of three types of TFTR D-T plasma facing components. Most of the decrease in surface tritium concentration occurred in the first minute of UV laser irradiation. In a second experiment, the UV laser was focused to irradiate the deposited layers on JT-60 graphite tile that had experienced hydrogen plasma operation. The effective absorption coefficient and the ablation threshold for the JT-60 codeposits irradiated by the UV laser were determined to be 1.9 µm−1 and 1.0 J/cm2, respectively. An erosion rate of 1.1 µm/pulse was reached at a laser energy density of 7.6 J/cm2.

Modeling hydrogen and helium entrapment in flowing liquid metal surfaces as plasma-facing components in fusion devices

In a fusion reactor, the ability to use liquids as plasma-facing components (PFCs) depends on their interaction with the plasma and the magnetic field. One important issue for the moving liquid is the ability to entrain particles that strike the PFC surface (helium and hydrogen isotopes) while accommodating high heat loads. To study this problem, an analytical model and a two-dimensional comprehensive numerical model have been developed and implemented in the HEIGHTS computer simulation package. The models take into account the kinetics of particle injection, motion and interactions with the liquid lattice, and the ultimate release from the surface. The models were used to investigate an important issue, whether He particles can be pumped by the PFC liquid rather than requiring a standard vacuum system. Hydrogen isotope (DT) particles that strike the surface will likely be trapped in the liquid–metal surface (e.g., lithium) due to the high chemical solubility of hydrogen. The impinging He particles in the established low-recycling regime at PFCs could be harder to pump using the standard vacuum pumping techniques. The analysis results indicate a reasonable chance of adequate helium self-trapping in flowing lithium as PFC without active pumping.

Physics basis for the fusion ignition research experiment plasma facing components

A design study of a fusion ignition research experiment (FIRE) is underway to investigate and assess near term opportunities for advancing the scientific understanding of self-heated fusion plasmas. The emphasis for the FIRE program is on understanding the behavior of plasmas dominated by alpha heating ( Q≥5). Study activities have focused on the technical evaluation of a compact, high field, highly shaped tokamak. One of the key issues for the design is to find suitable plasma facing components (PFCs). We have investigated a variety of plasma edge and divertor conditions ranging from reduced recycling high heat flux conditions (attached) to reduced heat flux detached operation. The inner divertor detaches easily while impurities must be added to the outer divertor to achieve detachment. The outer divertor and private space baffle will have to be actively cooled. The plasma-facing surface of the divertor is tungsten bonded to a CuCrZr heat sink. The remainder of the PFCs are beryllium coated copper attached to the vacuum vessel. Plasma current disruptions impose strong constraints on the design. Appreciable PFC surface melting and evaporation and onset of ‘plasma shielding’ are expected. The forces induced on the PFC due to disruptions determine the size of the attachment of the PFC to the vacuum vessel.

Design of the Fusion Ignition Research Experiment (FIRE) Plasma Facing Components

A design study of a Fusion Ignition Research Experiment (FIRE) is underway to investigate and assess near term opportunities for advancing the scientific understanding of self-heated fusion plasmas. The emphasis for the FIRE program is on understanding the behavior of plasmas dominated by alpha heating (Q ≥ 5). Study activities have focused on the technical evaluation of a compact, high field, highly shaped tokamak. One of the key issues for the design is to find suitable plasma facing components (PFCs). We have investigated a variety of plasma edge and divertor conditions ranging from reduced recycling high heat flux conditions (attached) to reduced heat flux detached operation. The inner divertor detaches easily while impurities must be added to the outer divertor to achieve detachment. The outer divertor and private space baffle will have to be actively cooled. The plasma-facing surface of the divertor is tungsten bonded to a CuCrZr heat sink. The remainder of the PFCs are beryllium coated copper attached to the vacuum vessel. Plasma current disruptions impose strong constraints on the design. Appreciable PFC surface melting and evaporation and onset of “plasma shielding” are expected. The forces induced on the PFC due to disruptions determine the size of the attachment of the PFC to the vacuum vessel.

減圧下での水蒸発と凝縮に関する実験と解析

When cooling tubes installed in plasma-facing components (PFCs) of a fusion reactor are damaged by some accidents, water may be discharged into a plasma chamber (PC) and vacuum vessel (VV) of the fusion reactor. In such a case, the discharged water will impinge on a surface of PFCs at high temperature and evaporate and then the VV may be broken due to pressurization by steam. This event is called an ingress-of-coolant event (ICE). The water-vapor two-phase flow behavior in the PC and VV during the ICE was investigated using an integrated ICE test facility. Furthermore, the condensation effect due to vapor inside a suppression tank which is installed to reduce the pressurization by the ICE was observed visually. In addition, the thermal-hydraulics during the ICE were validated numerically by the TRAC-PF1 code and analytical results on the pressure rise were in good agreement with the experimental results.

Analyses of Passive Plasma Shutdown during Ex-vessel Loss of Coolant Accident in the First Wall/Shield Blanket of Fusion Reactor

An ex-vessel loss of coolant accident (LOCA) in the first wall/shield blanket of a fusion reactor has been analyzed by a hybrid code consisting of plasma dynamics and heat transfer analysis of in-vessel components. We investigated possibility of passive plasma shutdown scenario during the accident in International Thermonuclear Experimental Reactor (ITER). The safety analysis code which we developed can treat impurity concentration from the first wall and the divertor with a transport probability into the main plasma and a time delay given as input. It was found that the plasma is passively shutdown by a density limit disruption due to beryllium release from heated first wall surfaces about 168 seconds after the LOCA, when the transport probability of beryllium from the first wall into the plasma and the time delay were assumed to be 10−2 and the energy confinement time, respectively. At that time, the surface temperature of the outboard center (plasma facing component (PFC) with beryllium) and the temperature of the coolant tube in the first wall (stainless steel 316) reach about 1,120°C and about 1,080°C, respectively. Although the coolant tube does not melt, the copper heat sink between the PFC and the coolant tube melts before the passive shutdown. The heat sink of copper in the outboard baffle also melts before the passive shutdown, though the PFC surface of tungsten does not melt. Consequently, we have a possibility of passive plasma shutdown before the cooling tubes melt during the ex-LOCA of the first walllshield blanket in ITER, however, further studies are needed on the effects on plasma burn control, impurity release and emission of implanted D-T fuel.

Plasma facing component design for steady state tokamak SST-1

A superconducting steady state tokamak is being designed to address some of the physics and technological issues of steady state operation. Circular plasma operation with poloidal limiter and elongated plasma operation with poloidal divertor are planned for a maximum steady state input power of 0.2 MW and 1 MW, respectively. The conceptual design of various plasma facing components is discussed. The poloidal limiter and divertor are actively cooled so as to keep the surface temperature of the plasma facing components less than 1000°C. In the present design the steady state heat removal capability is limited to a maximum of 1 MW/m 2 due to mechanical attachment scheme adopted for armor tile mounting to the actively cooled substrate. The plasma facing component surfaces have been profiled/shaped so that maximum steady state heat flux on the surface is ≤ 1 MW/m 2.

Proposal of High Efficient Discharge Cleaning Using Local Island Divertor in LHD

A new method of high efficient discharge cleaning using Local Island Divertor (LID) in the Large Helical Device (LHD) project is proposed as one of the optional merits of LID. This method drastically shortens the discharge cleaning time required in wall conditioning to remove main impurities such as oxygen on the surface of plasma facing components. The first estimation concludes very fast conditioning in less than one hour even in a large machine like LHD, which usually needs more than a few weeks with conventional discharge cleaning methods. Demonstration in actual plasma devices is proposed. 13 refs., 3 figs.