[en] The aim of this section is to present work done by participants of R&D sub-topics on water chemistry and corrosion study under the EU DEMO Water Cooled Lithium Lead Breeding Blanket (WCLL BB) project. The selection of light water reactor (LWR) water chemistry would enable the exploitation of valuable existing knowledge from the global LWR fleet in terms of water quality conditioning, control, and purification technologies. However, some specific parameters of the DEMO plant, such as the water radiolysis induced by 14 MeV neutrons, the rapid plasma burning and dwell cycle, presence of tritium in the coolant, and an unknown effect of the very high magnetic fields (4–8 T) required to confine the plasma, may have significant effects on the water chemistry, and needs to be considered during the design phase. (author)

[en] Highlights: • A working water chemistry specification has been defined for the WCLL blanket. • Tests are being undertaken to expand the Eurofer-97/water corrosion database. • A test rig to investigate flow-accelerated corrosion of CuCrZr pipe has been set up. • Industry support has been utilised to exploit synergies with fission. -- Abstract: The European DEMO design will potentially use single-phase water cooling in various components that require protection against corrosion. Coolant conditions will be similar to those of fission plant but with additional considerations arising from materials choices (Eurofer-97, CuCrZr) and 14 MeV neutron irradiation. Presently, many aspects of the water chemistry and corrosion behaviour are not well defined, and several strands of work, reported here, are ongoing to address these challenges under the EUROfusion framework in collaboration with industrial partners to leverage knowledge and expertise from elsewhere in the nuclear industry. Starting with the water-cooled lithium-lead blanket concept and considering the interaction of water with Eurofer-97, the foundation of this work has been the definition of a working water chemistry specification, supported by a review of relevant operating experience from light water fission reactors to understand the potential for technology transfer. Radiolysis modelling has been used to assess options for suppression of oxidising species under high energy neutron irradiation as these can be corrosive to components within the plant. High temperature water corrosion testing facilities have also been employed to expand the corrosion database and supplement existing experimental activities in the EUROfusion programme. In-vessel cooling of the divertor will use CuCrZr tubes under lower-temperature, high flow velocity conditions, which will lead to different considerations compared to the blanket and the potential for flow-accelerated corrosion. Additionally, high, unidirectional, heat fluxes lead to a radial temperature profile and the possibility of sub-nucleate boiling. A separate test setup, currently under construction, to expand this corrosion database is described.

[en] A complete global balance for material transport in JET requires knowledge of the net erosion in the main chamber, net deposition in the divertor and the amount of dust in the divertor region. Following the end of the first JET ITER-like wall campaign a set of tiles has been removed from the main chamber and the divertor. This paper describes the initial tile surface profiling results for evaluating the erosion in the main chamber and deposition in the divertor. Tile profiling was performed on upper dump plate tiles and inner wall guard limiters made of beryllium and on inner divertor tiles made of tungsten coated carbon (C)-fibre composites. Additionally, the mass of dust collected from the JET divertor is also reported. Present results are compared with JET-C campaign results with the all-carbon C wall. (paper)

[en] Full text: Comprehensive and systematic surveys of dust particles were performed at JET with the ITER-like wall (JET-ILW) after two experimental campaigns, ∼19 h of plasma each. Thought the total amounts removed were small (around 1 g) the study of dust categories is crucial for ITER because these are unique data from a full metal-wall (beryllium and tungsten) machine. The identification of various categories of particles allows conclusions on mechanisms underlying their generation and mobilization. This work deals with dust collected with sticky pads from the divertor tiles and, with metal splashes on erosion-deposition probes in the divertor and the main chamber wall. The local sampling is essential for: i) finding a correlation between the type of dust and the deposition pattern, and ii) comparison of dust identified in a given location after consecutive campaigns The search has identified several forms. a) Flakes of Be-rich deposits (up to 800 μm) with embedded tiny metal particles: Ni, W. Irregular droplet-like W inclusions, up to 200 nm, are nearly uniformly distributed in the deposit with an exception of the bottom of the layer, i.e., film formed at the early commissioning phase without high power beam operation. There is also a significant content of nitrogen retained after plasma edge cooling. Films are crystalline, but the presence of amorphous regions cannot be fully excluded. b) Regular Be droplets, diameter 5-10 μm and - on probes - Be splashes with small bubbles thus indicating boiling of the droplet. c) Spherical W droplets, ∼100 μm diameter, which could be formed in the experiment on tungsten melting. d) Droplets of Inconel. e) Irregular debris or flakes up to 300 μm containing W, Mo-W. f) Ceramics containing boron nitride, zirconium oxide, alumina. In conclusion, the study clearly shows a correlation between the operation mode, material erosion, growth of codeposits and generation of dust. (author)

[en] Full text: Metallic first mirrors are essential plasma-facing components (PFC) in all optical spectroscopy and imaging systems used for plasma diagnosis. First mirror test (FMT) has been carried out at the JET tokamak with the ITER-like wall (JET-ILW). Over 120 test Mo mirrors were exposed in JET during the entire project. The aim is to provide an overview of results obtained for mirrors exposed during: i) the third ILW campaign, ILW3, 2015-2016, 23.6 h plasma; ii) all three campaigns, i.e., ILW 1-3: 2011-2016, 62 h in total; and iii) a comparison to results in JET-C. Examinations were done by optical, electron and ion beam techniques. The total reflectivity of all mirrors in the main chamber has decreased by 2-3% from the initial value. All of them are coated by a very thin codeposit (5-15 nm) containing D, Be, C and O. This has affected the optically active layer (15-20 nm on Mo) and led to increased diffuse reflectivity. No W and N have been found on the surface. All mirrors from the divertor lost reflectivity by 20-80%. There are significant differences in the surface state dependent on the location and exposure time. Reflectivity loss is connected predominantly with the codeposition of Be and some C species. The thickest layers have been found in the outer divertor: 850 nm after ILW1-3, indicating the average growth rate of 4 pm/s. The layers thickness is not directly proportional to the exposure time. Nitrogen, tungsten and nickel are on all mirrors from the divertor. The highest N and W contents are in the inner divertor: N reaches 1 x 10¹⁷/cm², W is up to 3.0 x 10¹⁶/cm², while the greatest Ni content is in the outer leg: 2.5 x 10¹⁷/cm². The results obtained for the main chamber mirrors allow some optimism regarding the diagnostics reliability in ITER. Tests done in JET-C and JET-ILW show that the degradation of optical properties in a machine with metal PFC is distinctly smaller than in the carbon surrounding. However, a long-term exposure and off-normal events may change surface properties of the mirrors. Laser- or plasma induced cleaning techniques of tokamak mirrors have not brought any positive results. There are some indications that single crystal mirrors may be cleaned more efficiently than polycrystalline. Search for engineering solutions for mirror exchange in a reactor should not be abandoned especially for the divertor mirrors. (author)

[en] A build-up of co-deposits in remote areas of the divertor can contribute significantly to the overall fuel retention. The control of plasma-material interactions via the study and understanding of erosion-deposition of PFCs provides vital information for the efficient future operation of ITER. The major aim of this work is to reveal details of beryllium deposition and fuel (deuterium) retention on divertor plasma-facing componentsremoved from the JET ITER-Like Wall divertor after cumulative exposure during the first two (ILW−1 + 2) and all three (ILW−1 + 2 + 3) campaigns. Ion beam analysis techniques such as Rutherford backscattering spectrometry, nuclear reaction analysis and proton induced x-ray emission have been extensively used for post-mortem analyses of selected tiles from JET following each campaign and can provide relevant information on plasma-surface interactions like tungsten erosion, beryllium deposition and plasma fuel retention with divertor tiles via implantation or co-deposition. The studied divertor tiles represent a unique set of samples, which have been exposed to plasmas since the beginning of the JET-ILW operation for three successive plasma campaigns. This is a comprehensive comparison of divertor components after these operation periods. The results presented summarise deposition and fuel retention on Tiles 4 (inner base) and 6 (outer base). Although the deposition pattern is similar to that determined after individual campaigns, D retention is not a cumulative process and is determined mainly by the last campaign, and the total Be deposit after the 3 campaigns (i.e. data 1 + 2 + 3 = tile exposed 2011–2016) is less than the sum of the deposits after each individual campaign (sum 1 + 2 + 3) for Tile 4 but greater for Tile 6. (topical issue article)

[en] A selected set of samples from JET-ILW divertor tiles exposed in 2011–2012 has been analysed using thermal desorption spectrometry (TDS). The highest amount of deuterium was found on the regions with the thickest deposited layers, i.e. on the horizontal (apron) part and on the top part of Tile 1, which resides deep in the scrape-off layer. Outer divertor Tiles 6, 7 and 8 had nearly an order of magnitude less deuterium. The co-deposited layers on the JET tiles and the W coatings contain C, O and Ni impurities which may change the desorption properties. The D_2 signals in the TDS spectra were convoluted and the positions of the peaks were compared with the Be and C amounts but no correlations between them were found. The remaining fractions of D in the analysed samples at ITER baking temperature 350 °C are rather high implying that co-deposited films may be difficult to be de-tritiated. (paper)

[en] Since August 2011 the JET tokamak has been operated with the ITER-Like Wall (JET-ILW): beryllium (Be) in the main chamber and tungsten (W) in the divertor. i.e. the material configuration recently decided for ITER. Material erosion and fuel inventory studies are among top priorities of the JET-ILW programme. Various types of diagnostic tools, i.e. marker tiles and wall probes including test mirrors, have been employed to assess the overall material migration pattern. The specific goals of this work were to determine: (i) fuel retention in the divertor; (ii) erosion-deposition pattern of beryllium and other species; (iii) the reflectivity and surface morphology of mirrors studied within the First Mirror Test at JET for ITER. Analyses of in-vessel components have shown the erosion of Be inner wall limiters and the deposition of material on the upper tiles in the inner divertor leg. The thickest deposits, up to 15 m, contain mainly beryllium with some minority species: carbon and also nitrogen from edge cooling. Their content is low: Be/C concentration ratio >16; Be/N >45. Also fuel inventory in JET-ILW is small, both relative: Be/D >10 in deposits and absolute being below 5x10¹⁸ cm^-2. This value is distinctly lower than in JET with carbon walls (JET-C) where layers of a few hundreds of micrometers were formed. The study has not identified on wall components the formation of flaking deposits which could contribute to the dust formation. It should also be stressed that the reflectivity of polycrystalline molybdenum mirrors tested on the main chamber wall was retained or even improved in some cases. This result may have a positive impact (e.g. cost reduction) on the planning and development of maintenance procedures for ITER diagnostic mirrors. The options will be presented. These findings indicate advantages of metal components in comparison to the carbon surrounding. (author)

[en] Post-mortem measurements of 2-dimensional tritium (T) distribution using an imaging plate (IP) technique were performed for tungsten (W) divertor tiles (W-coated CFC) used in JET-ITER like wall (ILW) project. The observed T distributions were clearly inhomogeneous, and there were band-like regions with high T concentrations that extended in the toroidal direction on tiles 1, 3, 4 and 6. The concentrations of T in the band-like regions were higher by an order of magnitude than the concentrations in other parts. The inhomogeneous T distributions were explained by non-uniform co-deposition with other elements such as beryllium. The concentrations of T on the outboard vertical tiles (tiles 7 and 8) were low and relatively uniform in comparison with other tiles

[en] Highlights: • D concentrations were determined for the JET ILW-2 divertor tiles with SIMS. • The highest concentrations (∼8 · 10¹⁸ D/cm²) were found on the upper inner divertor. • D retention is generally higher for ILW-2 divertor tiles than after ILW-1 campaign. • Difference is related to longer exposure of some tiles and changes in SP-distribution. - Abstract: Divertor tiles removed after the second JET ITER-Like Wall campaign 2013–2014 (ILW-2) were studied using Secondary Ion Mass Spectrometry (SIMS). Measurements show that the thickest beryllium (Be) dominated deposition layers are located at the upper part of the inner divertor and are up to ∼40 µm thick at the lower part of Tile 0 exposed in 2011–2014. The highest deuterium (D) amounts (>8 · 10¹⁸ at./cm²), in contrast, were found on the upper part of Tile 1 (2013–2014), where the Be deposits are ∼10 µm thick. D was mainly retained in the near-surface layer of the Be deposits but also deeper in tungsten (W) and molybdenum (Mo) layers of the marker coated tiles, especially at W–Mo layer interfaces. D retention for the ILW-2 divertor tiles is higher than for the first campaign 2011–2012 (ILW-1) and probable reasons for the difference are that SIMS measurements for the ILW-2 samples were done deeper than for the ILW-1 samples, some of the tiles were exposed during both ILW-1 and ILW-2 and therefore had a longer exposure time, and the differences between ILW-1 and ILW-2 campaigns e.g. in strike point distributions and injected powers.