Fuel retention and erosion-deposition on inner wall cladding tiles in JET-ILW

Abstract

The morphology of beryllium coatings on the Inconel inner wall cladding tiles after JET-ILW campaigns was determined. The focus was on: (i) fuel retention and its share in the overall fuel inventory; (ii) the change of the layer structure and composition. The study is motivated in the view of planned D-T operation in JET. Four tiles were examined: the initial not exposed; one exposed to two campaigns (ILW1-2) and two facing the plasma during ILW1–3. As determined with ion beam and microscopy methods, the initial Be layer (9.0 μm thick) contained up to 4–5 at.% of impurities, mainly H, O, C, Ni. In the exposed tiles, the impurity content increases to 14–26 at.% (up to 20 at.% O, 1.7 at.% C, 1.0 at.% N, 1.3 at.% Ni and under 0.1 at.% W). The surface composition indicates gettering of O and a long-term retention of N. The Be thickness on the tile exposed to ILW1–2 was between 7.6 and 9.7 μm, thus indicating erosion in some areas, while the thickness after ILW1–3 increased to 10–12 μm. The D content was in the range 1.2–3.4×1017 cm−2 after ILW1–2 and 3.2–10×1017 cm−2 after ILW1–3 on most of the analyzed area, but in the limiter shadow values up to 58 ×1017 cm−2 were measured. Taking into account the total area of the Be-coated inner wall cladding tiles, the lower limit of D inventory amounts to 5.3×1022 atoms corresponding to about 176 mg, i.e. somewhat greater than the amount determined on Be limiters. The formation and spalling-off of Be-O particles was revealed.

Evaluation of tritium retention in plasma facing components during JET tritium operations

“…The global retention value has been calculated using the divertor data presented in figure 1(a) with additional data from main chamber beryllium (Be) limiter tiles [4], Be coated inconel (Be-inconel) recessed Inner Wall Cladding tiles (IWC) [16], Tungsten coated Carbon Fibre Composite (W-CFC) recessed inner limiter tiles, passive diagnostics from the outer recessed wall of the vacuum vessel [17] and remote divertor regions [17,18]. Data for these analyses include IBA techniques from VR, Sweden and University of Helsinki, Finland with facilities described in [11].…”

“…In the most part the same areas for each component type are used, except in the case of IWC which has been reduced from 11.2 m 2 to 7.2 m 2 to reflect a more recent assessment and data presented in [16]. The 2011-2016 ILW1-3 IWC data from [16] is scaled by the plasma operation time to provide a retention value for ILW3.…”

Application of Ion Beam Analysis in Studies of First Wall Materials in Controlled Fusion Devices

“…In the exposed plate, one detects gettered oxygen (20-40%) and co-deposited H, D, C, N. Hydrogen is clearly detected. Its content is greater than that of D, because the campaign was finished with 300 discharges fueled with H [59,61,62]. 3 He-based NRA is extremely efficient in D studies on PFC surfaces.…”

Characterization of Structural Stability of Magnetron Sputtered Tungsten-boron Thin Films at Elevated Temperatures

“…The mitigation process together with plasma reactions causes several processes such as formation of deposited layers [8,9], changes in surface chemical composition and structure [10], aging of divertor materials [11], affecting the lifetime of construction parts [12], and changing the properties of plasma facing surfaces [13]. Moreover, the deposition and formation of co-deposited layers is among the main processes leading to the retention and accumulation of fusion fuel atoms -deuterium and tritium [14].…”