[Show abstract][Hide abstract] ABSTRACT: Abstract
In optical diagnostic systems of ITER, mirrors will be used to guide the light from plasma
towards detectors and cameras. The mirrors will be subjected to erosion due to fast particles
and to deposition of impurities from the plasma which will affect adversely the mirror
reflectivity and therefore must be suppressed or mitigated at the maximum possible extent.
Predictive modeling envisages the successful suppression of deposition in the diagnostic
ducts with fins trapping the impurities on their way towards mirrors located in the end
of these ducts. To validate modeling predictions, cylindrical and cone-shaped diagnostic
ducts were exposed in TEXTOR for 3960 s of plasma operation. After exposure, no drastic
suppression of deposition was observed in the cylindrical ducts with fins. At the same time,
no detectable deposition was found on the mirrors located at the end of cone-shaped ducts
outlining the advantages of the cone geometry. Analyses of exposure provide evidence that
the contamination of exposed mirrors was due to wall conditioning discharges and not due
to working plasma exposure. Cleaning by plasma sputtering was performed on molybdenum
mirrors pre-coated with a 100 nm thick aluminum film. Aluminum was used as a proxy of
beryllium. During exposure in electron cyclotron resonance-generated helium plasma, the
entire coating was sputtered within nine hours, leaving no trace of aluminum and leading to
the full recovery of the specular reflectivity without detrimental effects on the mirror surface.
[Show abstract][Hide abstract] ABSTRACT: High-density discharges on JET with ITER-like Wall (ILW) have been analysed with the aim of establishing a mechanism for the H-mode density limit (DL) and compared with experiments in the JET carbon material configuration. The density limit is up to 20% higher in the JET-ILW than in the JET-C machine. The observed H-mode density limit is found close to the Greenwald limit. It is sensitive to the main plasma shape and is almost independent of the heating power. It has been observed that the transition from H-mode to L-mode is not always an abrupt event but may exhibit a series of H–L–H transitions, the so-called “dithering H-mode”.
[Show abstract][Hide abstract] ABSTRACT: Dust and tritium inventories in the vacuum vessel have upper limits in ITER that are set by nuclear safety requirements. Erosion, migration and re-deposition of wall material together with fuel co-deposition will be largely responsible for these inventories. The diagnostic suite required to monitor these processes, along with the set of the corresponding measurement requirements is currently under review given the recent decision by the ITER Organization to eliminate the first carbon/tungsten (C/W) divertor and begin operations with a full-W variant . This paper presents the result of this review as well as the status of the chosen diagnostics.
[Show abstract][Hide abstract] ABSTRACT: Laser-induced breakdown spectroscopy (LIBS) is considered as a promising method for in-situ diagnostic of the co-deposition and fuel retention during and in between plasma discharges in fusion devices. LIBS has been investigated intensively under laboratory conditions, while the application of LIBS in fusion devices is still in early stages. Moreover, the LIB processes are influenced by additional conditions in fusion devices, particularly the magnetic field. The experiments in TEXTOR show a significant enhancement in the spectral line emission and a deeper penetration of the laser-produced plasma into the edge plasma in the presence of magnetic field. These effects can be attributed to an increased confinement of the plasma by the magnetic field. The interference of magnetic field may compromise the quantitative interpretation of LIB spectra. Therefore, quantitative analysis of ITER-like co-deposits was done in laboratory without magnetic field as well as in TEXTOR with a magnetic field of Bt ∼ 2.25 T.
[Show abstract][Hide abstract] ABSTRACT: Abstract This contribution reports on the concept of a circular self-rotating and temperature self-stabilising plasma-facing component (PFC), and test of a related prototype in TEXTOR tokamak. This PFC uses the Lorentz force induced by plasma current and magnet field (J × B) to create a torque applied on metallic discs which produce a rotational movement. Additional thermionic current, present at high operation temperatures, brings additional temperature stabilisation ability. This self-rotating disk limiter was exposed to plasma in the TEXTOR tokamak under different radial positions to vary the heat flux. This disk structure shows the interesting ability to stabilise its maximum temperature through the fact that the self-induced rotation is modulated by the thermal emission current. It was observed that the rotation speed increased following both the current collected by the limiter, and the temperature of the tungsten disks.
[Show abstract][Hide abstract] ABSTRACT: JET is used as a test bed for ITER, to investigate beryllium migration which connects the lifetime of first-wall components under erosion with tokamak safety, in relation to long-term fuel retention. The (i) limiter and the (ii) divertor configurations have been studied in JET-ILW (JET with a Be first wall and W divertor), and compared with those for the former JET-C (JET with carbon-based plasma-facing components (PFCs)). (i) For the limiter configuration, the Be gross erosion at the contact point was determined in situ by spectroscopy as between 4% (Ein = 35 eV) and more than 100%, caused by Be self-sputtering (Ein = 200 eV). Chemically assisted physical sputtering via BeD release has been identified to contribute to the effective Be sputtering yield, i.e. at Ein = 75 eV, erosion was enhanced by about 1/3 with respect to the bare physical sputtering case. An effective gross yield of 10% is on average representative for limiter plasma conditions, whereas a factor of 2 difference between the gross erosion and net erosion, determined by post-mortem analysis, was found. The primary impurity source in the limiter configuration in JET-ILW is only 25% higher (in weight) than that for the JET-C case. The main fraction of eroded Be stays within the main chamber. (ii) For the divertor configuration, neutral Be and BeD from physically and chemically assisted physical sputtering by charge exchange neutrals and residual ion flux at the recessed wall enter the plasma, ionize and are transported by scrape-off layer flows towards the inner divertor where significant net deposition takes place. The amount of Be eroded at the first wall (21 g) and the Be amount deposited in the inner divertor (28 g) are in fair agreement, though the balancing is as yet incomplete due to the limited analysis of PFCs. The primary impurity source in the JET-ILW is a factor of 5.3 less in comparison with that for JET-C, resulting in lower divertor material deposition, by more than one order of magnitude. Within the divertor, Be performs far fewer re-erosion and transport steps than C due to an energetic threshold for Be sputtering, and inhibits as a result of this the transport to the divertor floor and the pump duct entrance. The target plates in the JET-ILW inner divertor represent at the strike line a permanent net erosion zone, in contrast to the net deposition zone in JET-C with thick carbon deposits on the CFC (carbon-fibre composite) plates. The Be migration identified is consistent with the observed low long-term fuel retention and dust production with the JET-ILW.
[Show abstract][Hide abstract] ABSTRACT: In JET-ILW isotopic plasma wall changeover experiments have been carried out to determine the amount of particles accessible by changing the plasma from H to D and from D to H. The gas balance analysis integrated over the experimental sessions show that the total amount of H or D removed from the wall is in the range of (1-3) × 1022D. For both changeover experiments, the respective plasma isotopic ratio behaviour is exactly the same as a function of the pulse number. After only 80 s of plasma (4 pulses), the plasma isotopic ratio is lower than 10%, below 4.5% after 13 pulses and then saturates around ~2-3%. In these conditions, the removal efficiency through plasma operation becomes very poor. The saturation of the plasma isotopic ratio in the range of 10% is also observed for the JET-C configuration although the amount of tritium retained in the vessel after the DT pulses was more than one order of magnitude compared to the retention observed with the JET-ILW. This demonstrates that the amount of particle recovery through plasma changeover is independent from the long term retention. Since this long term reservoir results from codeposition, these experiments suggest that there is a limited access to these codeposited particles by plasma isotopic changeover. Finally, in ITER, change over from D/T to H at the end of the discharge for possibly reducing the long term retention does not appear as a good strategy.
[Show abstract][Hide abstract] ABSTRACT: Tungsten (W) specimens previously exposed to deuterium (D) plasmas both in the TEXTOR tokamak and high flux linear plasma generator (LPG) were subsequently loaded with tritium at 573 K for 3 h. Retention of tritium in the near-surface W layer was examined by imaging plate technique. On the TEXTOR-plasma-exposed W surface, tritium was mainly trapped in carbon deposits. For LPG-plasma-exposed W specimens, tritium was trapped in defects created in the near-surface layer during the course of D plasma exposure.
[Show abstract][Hide abstract] ABSTRACT: Ferromagnetic pebbles are investigated as high heat flux (q∥) plasma facing components in fusion devices with short power decay length (λq) on a conceptual level. The ability of a pebble concept to cope with high heat fluxes is retained and extended by the acceleration of ferromagnetic pebbles in magnetic fields. An alloying concept suited for fusion application is outlined and the compatibility of ferromagnetic pebbles with plasma operation is discussed.Steel grade 1.4510 is chosen as a well characterized candidate material to perform an analysis of the heating process. Scaling relationships as a function of q∥ for maximum and optimal pebble diameter, allowed exposure time, and removal time safety margin are obtained numerically for spherical pebble geometry. The acceleration of ferromagnetic pebbles in a tokamak resulting from magnetic gradients is studied and operation parameters for an ITER-based reactor are outlined. Counter-intuitively, it is found that ferromagnetic pebbles perform better for narrow λq profiles, making them an attractive heat exhaust concept for next step devices and thus an option to be investigated in detail.The key results of this study are that very high heat fluxes are accessible in the operation space of ferromagnetic pebbles, that ferromagnetic pebbles are compatible with tokamak operation and current divertor designs, that the heat removal capability of ferromagnetic pebbles increases as λq decreases and, finally, that for fusion relevant values of q∥ pebble diameters below 100μm are required.
[Show abstract][Hide abstract] ABSTRACT: Cracking thresholds and crack patterns in tungsten targets have been studied in recent experiments after repetitive ITER-like ELM heat pulses in combination with plasma exposure in PSI-2 (Γtarget = 2.5–4.0 × 1021 m−2 s−1, ion energy on surface Eion = 60 eV, Te ≈ 10 eV). The heat pulses were simulated by laser irradiation. A Nd:YAG laser with energy per pulse of up to 32 J and a duration of 1 ms at the fundamental wavelength (λ = 1064 nm, repetition rate 0.5 Hz) was used to irradiate ITER-grade W samples with repetitive heat loads.
Fusion Engineering and Design 02/2015; DOI:10.1016/j.fusengdes.2015.01.028 · 1.15 Impact Factor
[Show abstract][Hide abstract] ABSTRACT: The migration of beryllium, tungsten and carbon to remote areas of the inner JET-ILW divertor and the accompanying co-deposition of deuterium has been investigated using post-mortem analysis of the housings of quartz-micro balances (QMBs) and their quartz crystals. The analysis of the deposition provides that the rate of beryllium atoms is significantly reduced compared to the analogue deposition rate of carbon during the carbon wall conditions (JET-C) at the same locations of the QMBs. A reduction factor of 50 was found at the entrance gap to the cryo-pumps while it was 14 under tile 5, the semi-horizontal target plate. The deposits consist of C/Be atomic ratios of typically 0.1–0.5 showing an enrichment of carbon in remote areas compared to directly exposed areas with less carbon. The deuterium retention fraction D/Be is between 0.3 and 1 at these unheated locations in the divertor.
[Show abstract][Hide abstract] ABSTRACT: The analysis of the deposition of eroded wall material on the plasma-facing materials in fusion devices is one of the crucial issues to maintain the plasma performance and to fulfill safety requirements with respect to tritium retention by co-deposition. Laser ablation with minimal damage to the plasma facing material is a promising method for in situ monitoring and removal of the deposition, especially for plasma-shadowed areas which are difficult to reach by other cleaning methods like plasma discharge. It requires the information of ablation process and the ablation threshold for quantitative analysis and effective removal of the different deposits. This paper presents systemic laboratory experimental analysis of the behavior of the ITER relevant materials, graphite, tungsten, aluminum (as a substitution of beryllium) and mixed deposits ablated by a Nd:YAG laser (1064 nm) with different energy densities (1–27 J/cm2, power density 0.3–3.9 GW/cm2). The mixed deposits consisted of W–Al–C layer were deposited on W substrate by magnetron sputtering and arc plasma deposition. The aim was to select the proper parameters for the quantitative analysis and for laser removal of the deposits by investigating the ablation efficiency and ablation threshold for the bulk materials and deposits. The comparison of the ablation and saturation energy thresholds for pure and mixed materials shows that the ablation threshold of the mixed layer depends on the concentration of the components. We propose laser induced breakdown spectroscopy for determination of the elemental composition of deposits and then we select the laser parameters for the layer removal. Comparison of quantitative analysis results from laboratory to that from TEXTOR shows reasonable agreements. The dependence of the spectra on plasma parameters and ambient gas pressure is investigated.
[Show abstract][Hide abstract] ABSTRACT: The isotopic exchange efficiencies of JET Ion Cyclotron Wall Conditioning (ICWC) discharges produced at ITER half and full field conditions are compared for JET carbon (C) and ITER like wall (ILW). Besides an improved isotope exchange rate on the ILW providing cleaner plasma faster, the main advantage compared to C-wall is a reduction of the ratio of retained discharge gas to removed fuel. Complementing experimental data with discharge modeling shows that long pulses with high (∼240 kW coupled) ICRF power maximizes the wall isotope removal per ICWC pulse. In the pressure range 1–7.5 × 10−3 Pa, this removal reduces with increasing discharge pressure. As most of the wall-released isotopes are evacuated by vacuum pumps in the post discharge phase, duty cycle optimization studies for ICWC on JET-ILW need further consideration. The accessible reservoir by H2-ICWC at ITER half field conditions on the JET-ILW preloaded by D2 tokamak operation is estimated to be 7.3 × 1022 hydrogenic atoms, and may be exchanged within 400 s of cumulated ICWC discharge time.
[Show abstract][Hide abstract] ABSTRACT: Systematic study of deuterium irradiation effects on tungsten was done under ITER - relevant high particle flux density, scanning a broad surface temperature range. Polycrystalline ITER - like grade tungsten samples were exposed in linear plasma devices to two different ranges of deuterium ion flux densities (high: 3.5-7 · 1023 D+/m2 s and low: 9 · 1021 D+/m2 s). Particle fluence and ion energy, respectively 1026 D+/m2 and ∼38 eV were kept constant in all cases.The experiments were performed at three different surface temperatures 530 K, 630 K and 870 K. Experimental results concerning the deuterium retention and surface modifications of low flux exposure confirmed previous investigations. At temperatures 530 K and 630 K, deuterium retention was higher at lower flux density due to the longer exposure time (steady state plasma operation) and a consequently deeper diffusion range. At 870 K, deuterium retention was found to be higher at high flux density according to the thermal desorption spectroscopy (TDS) measurements. While blisters were completely absent at low flux density, small blisters of about 40-50 nm were formed at high flux density exposure. At the given conditions, a relation between deuterium retention and blister formation has been found which has to be considered in addition to deuterium trapping in defects populated by diffusion.
[Show abstract][Hide abstract] ABSTRACT: This review summarizes surface morphology changes of tungsten caused by heat and particle loadings from edge plasmas, and their effects on enhanced erosion and material lifetime in ITER and beyond. Pulsed heat loadings by transients (disruption and ELM) are the largest concerns due to surface melting, cracking, and dust formation. Hydrogen induced blistering is unlikely to be an issue of ITER. Helium bombardment would cause surface morphology changes such as W fuzz, He holes, and nanometric bubble layers, which could lead to enhanced erosion (e.g. unipolar arcing of W fuzz). Particle loadings could enhance pulsed heat effects (cracking and erosion) due to surface layer embrittlement by nanometric bubbles and solute atoms. But pulsed heat loadings alleviate surfaces morphology changes in some cases (He holes by ELM-like heat pulses). Effects of extremely high fluence (∼1030 m−2), mixed materials, and neutron irradiation are important issues to be pursued for ITER and beyond. In addition, surface refurbishment to prolong material lifetime is also an important issue.
[Show abstract][Hide abstract] ABSTRACT: Discharge wall conditioning is an effective tool to improve plasma performance by (i) reducing the
generation of plasma impurities liberated from the wall and (ii) controlling the recycling of hydrogenic fluxes.
On ITER discharge wall conditioning will be employed as well for (iii) mitigating the tritium inventory build-up,
for which one relies mostly on the removal of tritium-rich co-deposited layers. Ion cyclotron wall conditioning
(ICWC) is a well-studied discharge wall conditioning technique having the advantage over Glow Discharge
Conditioning (GDC) that it is applicable in the presence of magnetic fields. The ICWC mode of operation is
included in the functional requirements of the ITER ion cyclotron resonance heating and current drive system,
and is envisaged for use between ITER plasma pulses, in the presence of the toroidal magnetic field.
Ion Cyclotron Range of Frequencies (ICRF) plasma production employing ICRH&CD antennas designed for
Fast Waves excitation is studied extensively on JET in the frame of fuel removal experiments by isotopic
exchange aiming at the development of ICWC scenarios for ITER. This paper compares isotopic exchange
efficiencies of JET ICWC discharges produced at ITER half and full field conditions for the JET carbon (C) and ITER like wall (ILW). ICWC on the ILW is found to be more efficient providing cleaner plasma faster, and has as significant advantage compared to the C-wall: an improved ratio of retained discharge gas to removed fuel, mitigating permanent retention during conditioning. A close to complete isotopic change over of the JET-ILW by D2-ICWC alone, evidenced by sampling the plasma isotopic ratio in tokamak discharges, was achieved within 630s of cumulated ICWC discharge time. The accessible reservoir by H2-ICWC at ITER half field conditions on the JET-ILW preloaded by D2 tokamak operation is larger than 7.3 x1022 hydrogenic atoms. Conditioning efficiency optimization, ICRF discharge initiation and the characterization of the ICWC particle flux on the PFC are briefly addressed.
25th IAEA Fusion Energy Conference, St. Peterburg 2014; 10/2014