Main Plasma Facing Components Research Articles

Survey of tungsten gross erosion from main plasma facing components in WEST during a L-mode high fluence campaign

An initial high fluence campaign was performed in WEST, in 2023, on the newly installed actively cooled tungsten divertor composed of ITER-grade monoblocks. The campaign consisted in the repetition of a 60 s long Deuterium L-mode pulse in attached divertor conditions, cumulating over 10000s of plasma exposure. A maximum deuterium fluence of approximately 5⋅1026 m−2 was reached in the outer strike point region, representative of a few high performance ITER pulses. Gross tungsten erosion inferred from visible spectroscopy shows that the most eroded plasma facing component is the inner divertor target with rates ten times larger than on the outer divertor target. The outer midplane tungsten bumpers, located a few centimeters from the plasma, show gross erosion rates two times lower than at the outer divertor. We conclude that the outer midplane bumpers have a negligible contribution to the long range tungsten migration and deposition onto the lower divertor. The cumulated gross erosion rate on the inner divertor translates in an effective gross erosion thickness of about 20μm, while it is about 2μm for the outer divertor. Strikingly, these orderings coincide with the thickness of deposits found locally on the divertor: the exposed surfaces of high field side monoblocks are covered with several tens of μm tungsten deposits, while on the lower field side, few μm thin tungsten deposits are only found on the magnetically shadowed parts of monoblocks. The strong impact of those deposits on WEST operation, namely perturbation of surface temperature measurement with infra-red thermography, and the emission of flakes causing radiative perturbation of the confined plasma, calls for anticipating similar issues in ITER. In particular, the start of research operation shall consider the definition of a divertor erosion budget in order to anticipate the formation of deleterious deposits.

Interaction of iron melt with tungsten and WFe composite structure evolution

Development of new plasma facing components (PFC) draw attention of numerous scientific groups in fusion energy field. Tungsten as a main PFC material main have poor machinability, thus various designs were proposed to overcome this problem. However, such technologies as functional graded layers, sintering and additive technologies are limited in production size and have low cost-efficiency. This research considers the alternative approach of steel melt injection or melt infiltration of tungsten mesh. The results obtained establish the mechanism of phase evolution during interaction of solid W with liquid steel and iron. Energy dispersive spectroscopy, X-ray diffraction and Thermocalc calculations used to analyze the phase composition of tungsten wetted with various steels and pure iron. Results show that interaction rate significantly depends on melt temperature and overheating above its liquidus. Then the overheating exceeds 150 °C erosion of tungsten occur.

Dust generation and accumulation in JET-ILW: morphology and stability of co-deposits on main plasma-facing components and wall probes

Dust particles and co-deposits were sampled for the first time from beryllium limiters and bulk tungsten divertor (both after ILW-3), and test mirrors from the main chamber after ILW-2 and ILW-3. The focus was on the morphology of molten particles and metal outgrowths. In parallel, the stability of beryllium layers under the impact of hot water was examined on limiters and Be coatings. The study performed by several microscopy methods including SEM, AFM, FIB, TEM and Be-sensitive EDX has revealed: (i) an asymmetric distribution of Be particles with the largest objects (125-550 μm) on side surfaces of the Be tile: (ii) the presence of highly porous particles, resembling blisters; (iii) very few thin flakes of co-deposits on the divertor, on the shadowed edge of lamella; (iv) the elemental composition and internal structure of the needle-shaped outgrowths on the mirrors: MoO; (v) no detectable impact of water on the beryllium morphology.

First demonstration of radiative power exhaust with impurity seeding in the island divertor at Wendelstein 7-X

Radiative power exhaust by impurity seeding was demonstrated for the first time in island divertor configurations at the stellarator Wendelstein 7-X. Feasibility of stable plasma operation was shown during seeding with both neon (Ne) and nitrogen (N2). High radiative power losses (80%) were found to reduce the divertor heat loads globally by 2/3 with both seeding gases injected at a single toroidal location into one of five magnetic islands. Heat flux detachment was achieved for the price of a loss of (%) in the stored energy. Ne seeding allows for sustained enhancement of edge radiation with a very slow decay of line emission of several tens of seconds after the end of the injection indicating a high recycling of this noble gas at the carbon main plasma facing components. In N2 seeded discharges it is shown that a response of line emission and plasma parameters is in correlation to the puff duration which indicates a higher level of absorption of this seeding gas in the wall. Continuous N2 seeding results in global cooling of the scrape-off layer (SOL) and decay of radiation over several seconds after the injection. Damping of counter-streaming SOL flows, and divertor particle fluxes induced by Ne and N2 seeding have been measured and provide evidence for a reduction of the convective part of the divertor heat fluxes. Losses in density in response to seeding can be compensated by feedback controlled divertor fueling. The controlled reduction of heat fluxes within this complex 3D edge island geometry is a very promising finding concerning detachment control in a future all-metal divertor.

Full coverage infrared thermography diagnostic for WEST machine protection

The WEST platform aims at testing ITER like W divertor targets in an integrated tokamak environment. To operate long plasma discharges, IR thermography is required to monitor the main plasma facing components by means of real time surface temperature measurements, while providing essential data for various physics studies.To monitor the new divertor targets, the WEST IR thermography protection system has been deeply renewed, to match the new tokamak configuration. It consists of 7 endoscopes located in upper ports viewing the whole lower divertor and the 5 heating devices. Electronic devices and computers allow data storage of ≈3 Gb/s IR images and real time video frames processing at 50 Hz rate, to ensure the protection of the main plasma facing components during plasma discharges by a feedback control of the power injected by the heating systems.Each endoscope provides 2 views covering 2 divertor sectors of 30° (toroidally) and 1 view of a heating antenna. Each optical line is composed of a tight entrance window followed by a head objective which forms an image transported through the endoscope by a series of 4 optical relays and mirrors, up to a camera objective. Finally, 12 IR cameras specially developed for WEST environment capture the thermographic data, at the wavelength of 3.9 μm, with a 640 × 512 pixels frame size.The paper describes the design constraints and diagnostic technologies: optics, mechanics, electronics, hard & software, cameras. Tvhe laboratory characterization procedures (Modulation Transfer Function, slit response, calibration), and the measurement performance results are given (spatial resolution, temperature threshold). Finally, first results obtained during experimental campaigns in WEST are presented.

Infrared imaging systems for wall protection in the W7-X stellarator (invited).

Wendelstein 7-X aims at quasi-steady state operation with up to 10 MW of heating power for 30 min. Power exhaust will be handled predominantly via 10 actively water cooled CFC (carbon-fiber-reinforced carbon) based divertor units designed to withstand power loads of 10 MW/m2 locally in steady state. If local loads exceed this value, a risk of local delamination of the CFC and failure of entire divertor modules arises. Infrared endoscopes to monitor all main plasma facing components are being prepared, and near real time software tools are under development to identify areas of excessive temperature rise, to distinguish them from non-critical events, and to trigger alarms. Tests with different cameras were made in the recent campaign. Long pulse operation enforces additional diagnostic design constraints: for example, the optics need to be thermally decoupled from the endoscope housing. In the upcoming experimental campaign, a graphite scraper element, in front of the island divertor throat, will be tested as a possible means to protect the divertor pumping gap edges during the transient discharge evolution.

Ion Microbeam Analyses of Dust Particles and Codeposits from JET with the ITER-Like Wall.

Generation of metal dust in the JET tokamak with the ITER-like wall (ILW) is a topic of vital interest to next-step fusion devices because of safety issues with plasma operation. Simultaneous Nuclear Reaction Analysis (NRA) and Particle-Induced X-ray Emission (PIXE) with a focused four MeV 3He microbeam was used to determine the composition of dust particles related to the JET operation with the ILW. The focus was on "Be-rich particles" collected from the deposition zone on the inner divertor tile. The particles found are composed of a mix of codeposited species up to 120 μm in size with a thickness of 30-40 μm. The main constituents are D from the fusion fuel, Be and W from the main plasma-facing components, and Ni and Cr from the Inconel grills of the antennas for auxiliary plasma heating. Elemental concentrations were estimated by iterative NRA-PIXE analysis. Two types of dust particles were found: (i) larger Be-rich particles with Be concentrations above 90 at% with a deuterium presence of up to 3.4 at% and containing Ni (1-3 at%), Cr (0.4-0.8 at%), W (0.2-0.9 at%), Fe (0.3-0.6 at%), and Cu and Ti in lower concentrations and (ii) small particles rich in Al and/or Si that were in some cases accompanied by other elements, such as Fe, Cu, or Ti or W and Mo.

Whole-machine material migration studies in the TEXTOR tokamak with molybdenum

MoF6 injection from a localised source into plasma edge in the TEXTOR tokamak was the last experiment before the final shut-down of the TEXTOR machine. During decommissioning all plasma-facing components (PFCs) became available for surface studies. Detailed mapping of Mo deposition was performed in order to determine its migration on global scale. The concentration of Mo on PFC decays exponentially with distance from the source. The decay length is of the order of 0.1m on the main PFC and 1m on the receded components. Also the decay lengths modelled with the ERO code are between 0.15–1.3m, depending on the anomalous cross-field diffusion coefficient. The inner bumper limiter is found to be the major repository for Mo. Material balance measurements show that only up to 22% of the injected Mo was detected on all the PFCs thus indicating that a large fraction of injected Mo may have been pumped out before being deposited.

Vibration of fusion reactor components with magnetic damping

The aim of this paper is to assess the importance of the magnetic damping in the dynamic response of the main plasma facing components of fusion machines, under the strong Lorentz forces due to Vertical Displacement Events. The additional eddy currents due to the vibration of the conducting structures give rise to volume loads acting as damping forces, a kind of viscous damping, being these additional loads proportional to the vibration speed. This effect could play an important role when assessing, for instance, the inertial loads associated to VV movements in case of VDEs. In this paper, we present the results of a novel numerical formulation, in which the field equations are solved by adopting a very effective fully 3D integral formulation, not limited to the analysis of thin shell structures, as already successfully done in several approaches previously published.

A new radiation-hard endoscope for divertor spectroscopy on JET

In preparation for ITER, JET has been upgraded with a new ITER-like wall (ILW) whereby the main plasma-facing components, previously made of carbon, have been replaced by Be in the main chamber and W in the divertor. A new endoscope with optimised divertor view has been developed in order to survey and monitor the emission of specific impurities such as tungsten, beryllium and the possibly remaining carbon in the tungsten divertor of the JET-ILW. It operates in the wavelength range from 390nm to 2500nm with high optical transmittance (≥30%) as well as high spatial resolution, that is ≤2mm at the object plane and ≤3mm over the whole depth of field (±0.7m). The endoscope is a prototype for testing an ITER relevant design concept based on reflective optics only. The endoscope has an optimised observation in the near ultraviolet and in the blue spectral region to ensure the detection of the W I-emission line at 400.8nm.In parallel to the new optical design, a new type of ITER-like shutter system based on pneumatic techniques has been developed and integrated in the endoscope head. The new optical design includes options for an in situ calibration of the endoscope transmittance during the experimental campaign.

A new visible spectroscopy diagnostic for the JET ITER-like wall main chamber

In preparation for ITER, JET has been upgraded with a new ITER-like wall (ILW), whereby the main plasma facing components, previously of carbon, have been replaced by mainly Be in the main chamber and W in the divertor. As part of the many diagnostic enhancements, a new, survey, visible spectroscopy diagnostic has been installed for the characterization of the ILW. An array of eight lines-of-sight (LOS) view radially one of the two JET neutral beam shine through areas (W coated carbon fibre composite tiles) at the inner wall. In addition, one vertical LOS views the solid W tile at the outer divertor. The light emitted from the plasma is coupled to a series of compact overview spectrometers, with overall wavelength range of 380-960 nm and to one high resolution Echelle overview spectrometer covering the wavelength range 365-720 nm. The new survey diagnostic has been absolutely calibrated in situ by means of a radiometric light source placed inside the JET vessel in front of the whole optical path and operated by remote handling. The diagnostic is operated in every JET discharge, routinely monitoring photon fluxes from intrinsic and extrinsic impurities (e.g., Be, C, W, N, and Ne), molecules (e.g., BeD, D(2), ND) and main chamber and divertor recycling (typically Dα, Dβ, and Dγ). The paper presents a technical description of the diagnostic and first measurements during JET discharges.

Thermo-hydraulic and structural analysis for finger-based concept of ITER blanket first wall

The blanket first wall is one of the main plasma facing components in ITER tokamak. The finger-typed first wall was proposed through the current design progress by ITER organization. In this concept, each first wall module is composed of a beam and twenty fingers. The main function of the first wall is to remove efficiently the high heat flux loading from the fusion plasma during its operation. Therefore, the thermal and structural performance should be investigated for the proposed finger-based design concept of first wall. The various case studies were performed for a unit finger model considering different loading conditions. The finite element model was made for a half of a module using symmetric boundary conditions to reduce the computational effort. The thermo-hydraulic analysis was performed to obtain the pressure drop and temperature profiles. Then the structural analysis was carried out using the maximum temperature distribution obtained in thermo-hydraulic analysis. Finally, the transient thermo-hydraulic analysis was performed for the generic first wall module to obtain the temperature evolution history considering cyclic heat flux loading with nuclear heating. After that, the thermo-mechanical analysis was performed at the time step when the maximum temperature gradient was occurred. Also, the stress analysis was performed for the component with a finger and a beam to check the residual stress of the component after thermal shrinkage assembly.

Operational limits during high power long pulses with radiofrequency heating in Tore Supra

Issues related to the limitations and optimization of long pulse operation at high radiofrequency (RF) power levels in the Tore Supra tokamak are presented. An increasing operational limitation was encountered during the experimental campaigns in 2006–2007, affecting the high power and long pulse performance. This limitation was characterized by the sudden appearance of a multifaceted asymmetric radiation from the edge (MARFE), often followed by a disruption. The analyses revealed that the limitation could be linked to over-heating and flaking of the carbon re-deposition layers on the main plasma facing components (PFCs). The carbon deposits on all PFCs were therefore completely removed during the winter shutdown 2007–2008. Following this, a remarkable improvement in the injected power capability was observed, resulting in almost 12 MW of injected power during 10 s, without any of the previous signs of limitation (MARFE, disruption).Furthermore, the RF antennas are subject to localized heat loads due to RF sheath effects and interaction by fast particles, effects which need to be minimized in particular for long pulse operation. Experimental results concerning the heat load on the antennas, caused by fast ion losses in the presence of magnetic ripple, are presented in this paper.

Gas balance and fuel retention in fusion devices

The evaluation of hydrogenic retention in present tokamaks is of crucial importance to estimate the expected tritium (T) vessel inventory in ITER, limited from safety considerations to 350 g. In the framework of the European Task Force on Plasma Wall Interaction (EU TF on PWI) efforts are underway to investigate gas balance and fuel retention during discharges, and to compare the data obtained with those from post-mortem analysis of in-vessel components exposed over whole experimental campaigns. This paper summarizes the principal findings from coordinated studies on gas balance and fuel retention from a number of European tokamaks, namely, ASDEX-Upgrade (AUG), JET, TEXTOR and Tore Supra (TS). For most devices, the long-term retention fraction deduced from integrated particle balance is ∼10–20%. This is larger than the ∼3–4% deduced from post-mortem analysis of plasma facing components (PFCs). However, from the database available for tokamaks with their main PFCs made of carbon, the important conclusion is that the T inventory limit (set by the working guideline for operations) could be reached in ITER within fewer than 100 discharges. This, therefore, would seriously impact on operation of the device unless efficient T removal processes are developed.

Edge Physics, Divertors, Pump Limiters

The properties of the scrape-off layer (SOL) determine both the particle removal as well as the power exhaust. The particle flows in the SOL and their radial decay give constraints for the construction of the main plasma facing components, pump limiters and divertors. Both systems allow for an adequate particle and power removal option.

Divertor Concepts I: Edge Physics, Divertors, Pump Limiters

The properties of the scrape-off layer (SOL) determine both the particle removal as well as the power exhaust. The particle flows in the SOL and their radial decay give constraints for the construction of the main plasma facing components: pump limiters and divertors. Both systems allow for an adequate particle and power removal option.

Edge Physics, Divertors, Pump Limiters

The properties of the scrape-off layer (SOL) determine both the particle removal as well as the power exhaust. The particle flows in the SOL and their radial decay give constraints for the construction of the main plasma facing components, pump limiters and divertors. Both systems allow for an adequate particle and power removal option.