Deuterium trapping and impurity collection on a 1990 JET Be belt limiter
From a tile of the JET toroidal belt limiter installed during the 1990 operation period, samples were cut from the plasma-facing side along the poloidal direction and the surface layers were analysed by SEM and by the ion beam techniques NRA, PIXE, RBS and HIERD. The poloidal distribution of D is fairly uniform with values of up to 2×1017 cm−2. The contamination of the Be by metallic impurities from the vessel wall is typically up to 2×1016 Ni cm−2, 5×1015 Cr cm−2 and 2×1016 Fe cm−2, giving totally about 0.2 g of metals deposited on the upper limiter. Other impurities amount up to 3×1017 C cm−2, 7×1017 O cm−2, 2×1017 Cl cm−2, 2×1016 K cm−2, 4×1015 Ca cm−2, 3×1015 Ti cm−2 and 1×1015 Zn cm−2. The D deposition is about a factor of 4 smaller than on the C limiter of the previous discharge period, while the deposition of the Inconel components is about one order of magnitude smaller than on the C limiter. O is found to be trapped in the Be at depths up to several micrometers, which may be caused by segregation of pure Be through the oxide to the surface of oxide layers on the limiter. This would improve the O gettering of the Be. The lower D concentration is consistent with the observed degassing of D in the time between discharges.
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Cited by (16)
Dust in magnetic confinement fusion devices and its impact on plasma operation
1999, Journal of Nuclear MaterialsThe formation mechanisms of dust particles in fusion devices and the possible interactions of dust with fusion plasmas are discussed. The growth of small particles in the plasma edge of fusion discharges is suggested in analogy to the well-known growth in reactive process plasmas (dusty plasmas). Results from the analysis of collected dust from TEXTOR-94 and first in situ observations by laser light scattering are presented, giving indirect evidence for some of the discussed mechanisms. Dust formation is an important PSI issue with serious potential implications and deserves future in depth studies.
Analysis and oxidation of thick deposits on TEXTOR plasma facing components
1999, Journal of Nuclear MaterialsDeuterium and hydrogen containing co-deposited layers formed on plasma facing components at the TEXTOR tokamak were characterised by a number of techniques including surface analysis methods and thermal desorption spectrometry. The aim of the investigation was to determine the composition and structure of the layers prior and after their exposure to air at elevated temperatures of 300°C and 550°C. The impact of the oxidation on the deuterium content and distribution in the surface region and in the bulk of PFC was addressed. The initial deuterium-to-carbon concentration ratio was in the range 0.04–0.06 in the flaking layers on top of the PFC and up to 0.17 on their side surfaces. The oxidation at 300°C for 2 h resulted in partial removal of the deuterium, especially from thin or loosely bound layers, but the release was not accompanied by the change in the deposit structure. The content of plasma impurity atoms in the layers was also not influenced by the exposure to air. Following the oxidation at 550°C partial powderisation of thick, flaking films was observed.
The adsorption of H<inf>2</inf>O vs O<inf>2</inf> on beryllium
1997, Surface ScienceThe adsorption mechanism and initial oxidation of sputtered beryllium exposed to H2O and to O2 were studied using a combination of the DRS, AES and XPS techniques. For both cases, the “clustering” Langmuir type mechanism was found to fit the adsorption kinetics. The initial sticking coefficients, estimated from these fits, are S0(H2O) ≈ 1 and S0(O2) ≈ 0.07.
Oxide islands, ∼ 3 monolayers thick, are formed in both cases, spreading laterally till a full layer is formed. For O2 the oxidation stops at this stage, while for H2O it continues at a lower rate, reaching a saturation level of about six monolayers. Observation of significant broadening of the AES O KVV peak in the stage of thickness growth for the H2O exposure, and DRS H depth profiling indicate that hydrogen is trapped in the oxide matrix. Possible hydrogen-enhanced-diffusion of ions through the oxide seems to enable the further growth.
Ion-induced release of deuterium from co-deposits by high energy helium bombardment
1997, Journal of Nuclear MaterialsIon-induced release of deuterium from thick and thin co-deposits formed on plasma facing surfaces in a tokamak and from thick hydrogenated carbon films by MeV3He+ beams was studied. The layers of different thickness and D content, up to 2 × 1019 cm−2, built on graphite, CFC, Ni, Ti, V and Mo under different operation conditions (i.e. heating mode or wall composition in the machine) were irradiated with 3He ions in the MeV range, 0.8–1.5 MeV and a total fluence exceeding 1 × 1017 He+ cm−2. The D content was determined by nuclear reaction analysis (NRA) and the films, prior and after the irradiation, were also examined by RBS and EDS. The amount released was found to be dependent on the layer elemental composition: 55–60% from the carbon and boron containing films and only 10% from those containing silicon. Effective detrapping cross-sections have been calculated on the basis of the results obtained.
Composition of the plasma facing material Tokamakium
1996, Journal of Nuclear MaterialsIn experiments with magnetically confined hot plasmas in respect to controlled thermonuclear fusion, such as tokamaks or stellerators, the surface layers of the vessel walls are modified by the plasma by erosion, redeposition, hydrogen isotope implantation and heating due to the power load from the plasma. Thus, the composition and structure of the surface layers are finally different to those of the material initially installed. This new material at the surface layers of tokamak experiments is sometimes also named ‘Tokamakium’. Generally, all elements ever introduced into the vessel can be found in the surface layers of the plasma-facing wall tiles. At all areas of the plasma-facing components both erosion and deposition have been observed including the deposition of metal droplets. Some areas are erosion dominated, while at others deposition dominates with atomic depositions of up to several μm, which partly flake. The elements of dust particles introduced into the vessel by in-vessel works get mostly incorporated as impurities into the surface layers of the plasma facing material.
Status of beryllium development for fusion applications
1995, Fusion Engineering and DesignBeryllium is a leading candidate material for the neutron multiplier of tritium breeding blankets and the plasma-facing component of first-wall and divertor systems. Depending on the application, the fabrication methods proposed include hot-pressing, hot-isostatic-pressing, cold-isostatic-pressing/sintering, rotary electrode processing and plasma spraying. Product forms include blocks, tubes, pebbles, tiles and coatings. While, in general, beryllium is not a leading structural material candidate, its mechanical performance, as well as its performance with regard to sputtering, heat transport, tritium retention/release, helium-induced swelling and chemical compatibility, is an important consideration in first-wall/blanket design. Differential expansion within the beryllium causes internal stresses which may result in cracking, thereby affecting the heat transport and barrier performance of the material. Overall deformation can result in loading of neighboring structural material. Thus, in assessing the performance of beryllium for fusion applications, it is important to have a good database in all of these performance areas, as well as a set of properties correlations and models for the purpose of interpolation/extrapolation.
In this current work, the range of anticipated fusion operating conditions is reviewed. The thermal, mechanical, chemical compatibility, tritium retention/release, and helium retention/swelling databases are then reviewed for fabrication methods and fusion operating conditions of interest. Properties correlations and uncertainty ranges are also discussed. In the case of the more complex phenomena of tritium retention/release and helium-induced swelling, fundamental mechanisms and models are reviewed in more detail. Areas in which additional data are needed are highlighted, along with some trends which suggest ways of optimizing the performance of beryllium for fusion applications.