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Tungsten-coated divertor files exposed during the third JET ITER-Like Wall (ILW) campaign in 2015-2016 (ILW-3) were studied with Secondary Ion Mass Spectrometry (SIMS). ILW-3 campaign contained more high-power plasma discharges and longer plasma time than the earlier ILW campaigns. Measurements showed increased beryllium (Be) deposition on the upper inner divertor, whereas on the outer divertor, Be deposition was lower than during the second campaign in 2013-2014 (ILW-2). Increased intensifies of nickel, molybdenum and tungsten were observed at the surface layer of the inner divertor Be dominated deposits. These layers are probably formed during the high-power plasma discharge phase near the end of the ILW-3 campaign. Compared to the earlier campaigns, D retention on the upper inner divertor was observed to on a similar level than after ILW-2, whereas at the lower inner divertor and most parts of the outer divertor, D retention was lower for ILW-3 than ILW-2. D retention was increased at lower part of outer divertor Tile 7, where Be deposition was slightly increased. Probable reason for the reduction is the higher surface temperature of the files due to higher powers used.

Fuel retention and material migration results from JET ITER-like wall beryllium limiter tiles are presented for three operating periods. Ion beam analysis results support the general picture of erosion during limiter configurations with local deposition on tile ends far into the scrape off layer. Similar trends of fuel concentrations are observed in all JET operating periods; (i) low on surfaces exposed to high heat flux and erosion and (ii) higher in deposits. The pattern of fuel retention and deposition correlates with heat flux and distribution of limiter plasmas touching inner and outer limiters. The D/Be ratio in the thickest deposit is similar to 0.01. Global fuel retention attributed to limiters is

We compare the deposition of material on poloidal sets of divertor tiles that had been exposed in JET in 1998–2009 and 1998–2007. Post mortem analyses suggest toroidally integrated deposition being increased by 197.5 cm 3 during 2007–2009. The analysis of dust collected from the divertor indicates the amount accumulated during the same period to be 248.4 g. Converting the weight of dust to volume, the fraction of material entering the divertor that was converted to dust and flakes is 43 ± 10%. The size of most dust particles ranged from 10 to 100 μm. The integrated amount of deposition on the “marker” tiles exposed in 2007–9 was found to be more than twice the amount expected from film growth on other tiles plus the dust because the plasma responds differently to the new tiles.

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 ...

The aim of this work was to provide a comprehensive surface characterization of rhodium and molybdenum mirrors exposed on the main chamber wall of JET during campaigns with carbon plasma-facing components. As determined using X-ray photoelectron spectroscopy, Secondary Ion Mass Spectrometry, Scanning Electron Microscopy and optical measurements the co-deposits formed on the mirrors were composed of C, Be, Ni, Mo and Fe. Most of the metallic elements on the surface were in the oxidized states. After removing a part of the film by sputtering with Ar+ ions, Mo carbide state was studied by XPS. The reflectivity measurements showed the dependence of the thickness of the deposited film on the optical properties of the mirror: a thinner film on the Mo mirrors results in an identical reduction in reflectivity as that seen on the exposed Rh mirror. © 2013 Euratom. Published by Elsevier B.V. All rights reserved.

In this study the tritium distribution and the effect of structural changes thereon have been analyzed in the bulk of the tile selected from the JET Mark II SRP divertor. Tritium content has been analyzed by the full combustion technique [1] . The structure has been investigated by the method of Scanning Electron Microscopy. Tritium depth profiles have been measured at different poloidal positions. A high specific activity of tritium (up to 156 MBq g −1 ) was found at the plasma-facing surface. At some tile positions up to 98–99% of the T can be in the surface slice of 1 mm thickness, whereas in other poloidal positions there can be more T in the bulk than at the surface. The structural changes of the tile material both at the surfaces and in the bulk have been measured.

Flash-lamp photonic cleaning has been tested in situ and at the Beryllium Handling Facility (BeHF) at JET. Two adjacent number 4 divertor tiles have been exposed to numerous pulses up to the nominal energy of 500 J. Starting the photon-cleaning process with tile G4A and using energies up to 350 J did not appear to be efficient for detritiation. Consequently, the untreated tile G4B has been treated at the maximum energy of 500 J. Combustion measurements confirmed that the photon-cleaning was partly efficient as about 74% of the initial tritium concentration has been released. The average tritium concentration on the surface of the tile after treatment was 2.45 × 10 8 Bq/cm 3 which is only four times lower than the initial activity. However, it is remarkable to notice that the bulk activity of the tiles remains constant indicating that during the detritiation treatment there is no tritium diffusing into the bulk of the tile.

The fusion plasma, with a typical temperature of 10 keV, has to be brought into contact with a physical wall in order to remove the helium produced and drain the excess energy in the fusion plasma. The fusion plasma is far too hot to be brought into direct contact with a physical wall. It would degrade the wall and the debris from the wall would extinguish the plasma. Therefore, schemes are developed to cool down the plasma locally before it impacts on a physical surface. The resulting plasma wall-interaction in ITER is facing several challenges including surface erosion, material redeposition and tritium retention. In JET facility both beryllium and carbon are used together as plasma facing materials, and the resultant surfaces show considerable mixing between the two elements. In this work a first XPS characterisation has been performed on the surface of several PFC (plasma facing components) samples obtained from various JET tiles from the Mk II GB divertor in JET. The substrates of the tiles are CFC, and the deposited films have various D/C ratios depending on the position in the tokamak. Ion beam analysis shows that the D/C ratio is ∼1 in the shadowed areas and Be/C ∼1–2 in the plasma-exposed regions of the inner divertor. These values could give rise to the formation of beryllium carbides, and indications of a carbide component are found in the XPS spectra. The processes for material migration and the morphology of the deposits are discussed.

Beryllium has been chosen as the first wall material for ITER. In order to understand the issues of material migration and tritium retention associated with the use of beryllium, a largely beryllium first wall will be installed in JET. As part of the JET ITER-like wall, beryllium tiles with marker coatings are proposed as a diagnostic tool for studying the erosion and deposition of beryllium around the vessel. The nominal structure for these coatings is a similar to 10 mu m beryllium surface layer separated from the beryllium tile by a 2-3 mu m metallic inter-layer. Two types of coatings are tested here; one with a nickel inter-layer anti one with a copper/beryllium mixed inter-layer. The coating samples were deposited by DC magnetron Sputtering at General Atomics and were exposed to deuterium plasma in PISCES-B. The results of this testing show that the beryllium/nickel marker coating would be suitable for installation in JET.

Morphology of castellated Be tiles from the belt limiter exposed to the JET plasma for 56,000 s was examined on both sides of castellated grooves, on plasma-facing and side surfaces of the tiles. The essential results are (i) deuterium retention in the castellated grooves and in other locations is associated with co-deposition of carbon; (ii) the decay length of deposition in the castellation is around 1.5 mm; (iii) no deuterium is detected in bulk Be; (iv) bridging of gaps by molten beryllium occurred but gaps were not filled with Be; (v) on side surfaces of the tiles the formation of BeO layer was detected at a distance of 20 mm and more from the plasma-facing surface. The consequences for a long-term operation of a reactor-class device with several different plasma-facing materials are addressed.

Tritium co-deposition with carbon in the colder remote areas of the JET vessel immobilises a substantial amount of the fuel and increases the tritium inventory of the machine. A well defined physical or chemical description of these co-deposits is still not available. Therefore they are reported in the literature as a carbonaceous amorphous C:H layer, and unfortunately sometimes as a ‘diamond-like film’ referring to their sp 3 hybridisation. From the present study it appears that the gas–solid reactions taking place during the thermal treatment of carbon samples involves various tritiated hydrocarbon species having one or two carbon atoms in their structure like methyl ( CH 3 ) or ethyl ( CH 2 CH 3 ), having practically all bonds saturated with heavier hydrogen isotopes. The thermal response of the samples was also investigated by laser irradiation. During the laser scans the co-deposited layers reach high temperatures, sometimes exceeding 2300 °C, releasing all co-deposited tritium to the gas phase.

Tritium surface distribution on the JET Mark II Septum Replacement Plate (Mk II SRP) divertor tiles was measured by imaging plate technique. It was observed that areal tritium concentration was higher at the entrance of inner/outer pumping slots (so called ‘shadowed area’). The tritium distribution profiles were similar to those obtained in the Mk IIA divertor which was exposed to a series of D–T plasma operation (DTE1). Tritium concentration of the plasma facing surface was lower compared to that of the shadowed area. Particularly, it was very low at the outer divertor surface. The inner divertor surface also showed low level of tritium retention, though it was covered by the thick carbon deposition on that. This could be caused by tritium release due to the temperature rise when the inner strike point was on the tiles. On the plasma shadowed area like tile gaps, high tritium retention owing to the codeposition was observed.

In the year 2001, a quartz microbalance system (QMB) was installed in the remote area of the inner JET divertor to investigate in situ material erosion and redeposition processes. When removed in 2004, the system was found to be coated all over with carbon deposits. The deposit on the quartz oscillator and the outer and inner housing was analysed by various methods, as SIMS (secondary ion mass spectroscopy), stylus depth profilometry, EPMA (electron probe microanalysis), TIPT (Tritium imaging plate technique) and colorimetry and compared to the frequency change of the quartz. The layer thickness was determined to 1.85 ± 0.1 μm in average on an area of 0.95 cm 2 which has to be related to the equivalent of 1.77 × 10 −4 g measured from the frequency change of 23 640 Hz. This corresponds to a carbon areal density of 9.3 × 10 18 C atoms/cm 2 . Significant deposition was found also on the surfaces inside the QMB housing which can only be understood if reflection and low sticking is assumed for a high fraction of particles.

Ion micro beam analysis has been applied to the investigation of plasma deposited layers covering the divertor tiles in the JET tokamak. Since the layers are about 100 μm thick they are too thick to be completely investigated by ordinary ion beam analysis. Cross sections of the layers were prepared by cutting and polishing. Elemental depth profiles were determined from the two dimensional images that could be derived by nuclear reaction analysis and resonant backscattering spectrometry, using ion beams focused to a few μm spot size. A combination of analysis methods are shown, which allow measurements of the concentration profiles of carbon, beryllium, deuterium, oxygen and stainless steel components at levels of a few percent, with an accuracy better than 10%.

In order to perform a fully integrated material test, JET has launched the ITER-like wall project with the aim of installing a full metal wall during the next major shutdown. The material foreseen for the main chamber wall is bulk Be at the limiters and Be coatings on inconel tiles elsewhere. R&D process comprises global characterization (structure, purity etc) of the evaporated films and testing of their performance under heat loads. The major results are (i) the layers have survived energy loads of 20 MJ m−2 which is significantly above the required level of 5–10 MJ m−2, (ii) melting limit of beryllium coating would be at the energy level of 30 MJ m−2, (iii) cyclic thermal load of 10 MJ m−2 for up to 50 cycles have not induced any noticeable damage such as flaking or detachment.

Migration of beryllium into the divertor and deposition on tungsten in the final phase of the first ITER-like-wall campaign of JET are modelled with the 3D Monte Carlo impurity transport code ERO. The simulation covers the inner wall and the inner divertor. To generate the plasma background for Monte Carlo tracing of impurity particles, we use the EDGE2D/EIRENE code set. At the relevant regions of the wall, the estimated plasma conditions vary around Te≈5eV and ne≈2×1017m-3 (far-scrape-off layer; more than 10 cm away from the LCFS). We calculate impurity distributions in the plasma using the main chamber source as a free parameter in modelling and attempt to reproduce inter-ELM spectroscopic Be II line (527 nm) profiles at the divertor. The present model reproduces the level of emission close to the inner wall, but further work is needed to match also the measured emission peak values and ultimately link the modelled poloidal net deposition profiles of beryllium to post mortem data.

Tritium surface distributions on the plasma-facing surface and four sides of JET Mk IIA divertor tiles employed in the D-T operation phase of JET were measured by Tritium Imaging Plate Technique (TIPT). Tritium distribution on the plasma-facing surface was consistent with carbon deposition profiles and asymmetric in both poloidal and toroidal directions. The toroidal asymmetry was attributed to the alignment of the tiles preventing direct impact of flux lines to tile edges. Accordingly, no significant carbon deposition or tritium accumulation was observed on two sides facing the toroidal direction. As already reported, heavy codeposition retaining high levels of tritium was observed on the plasma-shadow area of the horizontal target tile surface and the bottom side of the vertical target tile of the inner divertor region where it was kept relatively cool by water coolant. In addition, TIPT has clearly distinguished at least two different carbon deposition layers with different tritium retention in poloidal direction, showing that the poloidal asymmetry on the horizontal target tiles is due to the different carbon deposition properties in the poloidal direction. All the results suggest that tritium retention in the divertor area, which was determined by the carbon/hydrocarbon distribution, correlates closely with divertor geometry andmore » surface temperature.« less

First wall components in controlled fusion devices undergo severe modification by various physical and chemical processes arising from plasma–material interactions: material erosion, its migration in the plasma and re-deposition. Ion beam analysis techniques play a prominent role in the studies of wall components exposed to hot plasmas. The intention of this work is to give a concise overview of methods used in the determination of surface composition of materials facing the plasma in the JET tokamak. The aim was to determine the amount and distribution (lateral and in-depth) of several elements or isotopes on large areas, i.e. mainly on high heat flux components such as limiter and divertor tiles. The major interest was in the quantification of deuterium co-deposited together with plasma impurity species: beryllium, boron, carbon (12 and 13) originating either from the wall erosion or deliberately introduced to the plasma as material transport markers. Special instrumentation, analytical approach and highlights of results obtained in studies of material transport by means of tracer techniques are presented.

The deposition of carbon in the louvre area of the inner divertor of JET was measured by means of a Quartz Microbalance diagnostic (QMB) during 806 exposures (total of about 6479 s) in various divertor conditions. Exposure time was controlled by means of a shutter. The overall integrated frequency shift of the deposition crystal from the start of the measurements to the end was 23 640 Hz, corresponding to an average carbon deposition flux of 5.5 × 10−8 g/cm2 s or 2.8 × 1015 C/cm2 s. Extrapolating this to the total time of 26.4 hours of divertor plasmas with the MKII GB SRP divertor (gas box divertor with septum replacement plate) yields a total amount of 35.4 g of carbon layers deposited on the louvre region. It was found that the deposition increases significantly with decreasing distance of the strike point position to the louvre entrance. Elmy-H-mode discharges with the strike point on the horizontal target dominate the carbon layer formation on the QMB.

Approximately 1 g of the 150 g collected at the JET was send to the Tritium Laboratory Karlsruhe (TLK) for physico-chemical characterisation. The flakes were extensively analysed using several experimental tools, such as calorimetry, X-ray-diffraction, thermal desorption, helium pycnometry, scanning electron microscopy, etc. The specific surface area measurements according to the Brunauer-Emmett-Teller (BET) method, showed a value of about (4.7 ± 0.3) m 2 g -1 while their average tritium activity was estimated to be (1.17 ± 0.1) TBq g -1 (i.e. 3-3.3 mg of tritium per gram of flakes) and their average real density (1.69 ± 0.02) g cm -3 . Based on these results we estimate that at the end of the remote tile exchange (RTE) at JET the 3.0 g of tritium which did not return to the AGHS after the DTEI remained in the vessel mainly in form of flakes, i.e. approximately (950 ± 80) g of flakes are still inside the machine.

The paper presents new Monte-Carlo transport simulations of methane 13CH4 injected through a hole in a testlimiter and exposed to the edge plasma of TEXTOR. The results show that the spatial distribution of 13C re-deposited locally on the testlimiter surface can be modelled if the parameter S for the sticking of returning hydrocarbons 13CHy is set to zero or almost zero. This is interpreted as a negligible effective sticking of the returning hydrocarbon radicals due to the instantaneous re-erosion caused by the hydrogen carried with the CHy radicals (`self re-erosion'). However, the calculated local deposition efficiency of 13C, remains too high compared with the observed value. Therefore, in addition an enhanced yield for chemical erosion caused by the background hydrogen for the fresh re-deposits has to be assumed. Similar assumptions can reproduce also the high amount of carbon deposition found on the inner louvers in the MkIIa divertor configuration of JET and on the plasma-shadowed areas of the MkIIGB divertor.

After the D-T operation (DTE-1 campaign) at JET a large number of limiter and divertor tiles were dismounted from the torus for ex situ examination. The relative distributions of deuterium, tritium, beryllium and carbon are presented and discussed. Significant asymmetry observed in the distribution of erosion and deposition zones indicates preferential flow of the deuterium background plasma and impurities towards the inner divertor leg. The comparison of the beryllium content on the limiter tiles from the main chamber and the content of this element on the inner divertor tiles clearly proves the beryllium erosion from the main chamber wall and its transport to the divertor. However, no beryllium is detected in the shadowed regions of the divertor where the formation of thick and fuel-rich carbon films occurs. This is interpreted in terms of different mechanisms governing the erosion and transport of Be and C. The results allow a conclusion that the operation with a full beryllium wall would lead to a significantly decreased fuel inventory due to removal of the carbon source.

Fast and efficient tritium removal is needed for future D-T machines with carbon plasma-facing components. A novel method for tritium release has been demonstrated on co-deposited layers on tiles retrieved from the Tokamak Fusion Test Reactor (TFTR) and from the Joint European Torus (JET). A scanning continuous wave neodymium laser beam was focused to =100 W/mm2 and scanned at high speed over the co-deposits, heating them to temperatures =2000 C for about 10 ms in either air or argon atmospheres. Fiber optic coupling between the laser and scanner was implemented. Up to 87% of the co-deposited tritium was thermally desorbed from the JET and TFTR samples. This technique appears to be a promising in-situ method for tritium removal in a next-step D-T device as it avoids oxidation, the associated de-conditioning of the plasma-facing surfaces, and the expense of processing large quantities of tritium oxide.

A set of Be and W tiles removed after the first ITER-like wall campaigns (JET-ILW) from 2011 to 2012 has been analysed. The results indicate that the primary erosion site is in the main chamber (Be) as in previous carbon campaigns (JET-C). In particular the limiters tiles near the mid-plane are eroded probably during the limiter phases of discharges. W is found at low concentrations on all plasma-facing surfaces of the vessel indicating deposition via plasma transport initially from the W divertor and from main chamber W-coated tiles; there are also traces of Mo (used as an interlayer for these coatings). Deposited films in the inner divertor have a layered structure, and every layer is dominated by Be with some W and O content.

In 1998, 100 g of highly tritiated carbon dust and flakes containing an estimated 1 g of Tritium were removed from the JET machine using a vacuum cleaner with a cyclone dust separator during the Remote Exchange of the divertor tiles. Two pots of flakes were filled and transferred, in drums, to the Active Gas Handling System where they were stored and ventilated to the Exhaust Detritiation System (EDS). A ventilated glovebox allowed the material to be handled safely and undergo calorimetric tritium assay. The glovebox had an air atmosphere with a continuous purge to the EDS with a facility for argon purging of isolators within the glovebox. Design features of the glovebox, operational experiences in handling tritiated carbon with a specific activity in excess of 1 TBq/g and results of the tests are discussed.

Recent results on measurements of tritium and other hydrogen isotopes in first wall materials of large tokamaks are discussed and evaluated. Data on the in situ and ex situ release of tritium from plasma facing components under different conditions are assessed.

The retention of hydrogen isotopes, particularly tritium, in the plasma-facing walls of fusion devices presents a serious difficulty for the exploitation of future machines. Investigation of techniques to reduce this effect is hampered by the difficulty in making measurements. Current methods rely on analysis of samples removed from machines after venting to atmosphere (J.P. Coad, et al., J. Nucl. Mater. 226 (1995) 156). Conclusions have, therefore, been drawn on a global basis, since it is not possible to assess the effects of individual plasma configurations. Detailed studies require a more prompt means of measuring the hydrogen content of samples, ideally on a shot-by-shot basis and without the necessity of breaking the machine vacuum (A. Huber, et al., this Conference). A new technique is described that meets these criteria and has been demonstrated at JET.

Deposition and H-isotope retention in JET is highly asymmetric, with deposition predominantly in the inner divertor, where flaking deposits form on water-cooled louvres shadowed from the plasma. The asymmetry implies drift in the SOL of JET from outboard to inboard under normal operating conditions, which has been measured. In order to model the amounts of deposition, assumptions have to be made about the transport at the inner target. Analysis of divertor tiles shows that material from the main chamber travels along the SOL to the inner divertor wall, from where carbon is preferentially removed leaving a beryllium-rich film. The carbon travels to shadowed areas (such as the louvres) where deposits with high H-isotope content accrue. The analysis indicates that chemical processes must be important.

During previous Joint European Torus (JET) deuterium-to-tritium change-over experiments, subdivertor tritium concentrations were compared with those measured at the strike point region and found to differ significantly during the first few discharges, which was correlated with wall saturation. New deuterium-to-hydrogen fueling experiments in JET have been made and are compared to these previous experiments. Rates of hydrogenic species exchange are similar to those found in previous tritium experiments, granting differences in divertor configuration and mass ratio. In the new experiments, measurements of the CD and CH molecular band intensities near the divertor strike point monitor an intermediate stage of particle exchange between the plasma and wall. The CD/CH ratio correlates well with both the plasma and subdivertor concentration. The neutral transport code EIRENE and the wall hydrogen trapping and diffusion code WDIFFUSE have been used to evaluate the wall saturation. It appears that chemically-related processes play a role in mediating the plasma–wall exchange.

The neutral tritium concentration in the subdivertor region of JET [P.-H. Rebut, R. J. Bickerton, and B. E. Keen, Nucl. Fusion 25, 1011 (1985)] is measured during deuterium-to-tritium changeover experiments with a novel species-selective Penning gauge coupled to a high-resolution spectrometer. The subdivertor measurements, when compared with edge and strike point values, are a sensitive characterization of the status of the wall saturation. The neutral transport code (EIRENE) [D. Reiter, Forschungszentrum, Juelich: Report Juel-2599 (1992)] and a wall hydrogen trapping and diffusion code (WDIFFUSE) [J. Hogan, R. Maingi, P. Mioduszewski et al., J. Nucl. Mater. 241, 612 (1997)] evaluate the wall tritium recycling coefficients (RT) and compare them with quantitative, testable models for the dynamic exchange between recycling surfaces and the edge/pedestal region. Using the dynamic inventory model, tritium transport following the injection of trace amounts of tritium has been investigated for JET high-confinem...

Erosion and deposition at the vessel walls of the main chamber of JET were measured with long term samples during the whole operation period of the Mark I carbon divertor from April 1994 until March 1995. Assuming toroidal symmetry, about 70 g Ni + Cr + Fe was found to be eroded from the inner torus wall and about 55 g Be from the outer torus wall due to sputtering by energetic charge-exchange neutrals. Deposition has been measured on a poloidal section of the Mark I C divertor and on tiles from the inner and outer wall limiters. Eroded material is redeposited in the divertor and in the main chamber where it is found predominantly at the sides of the poloidal limiters and at the outer vessel wall. At these deposition-dominated areas in total about 25 g Ni + Cr + Fe, 28 g Be and 370 g C are found. The carbon originates mostly from the carbon limiters; additionally some may be originating from the divertor plates.

The power deposited in the JET divertor during ELMs has been evaluated using an infrared camera specifically designed for fast measurements. The first results [E. Gauthier, A. Charkin, S. Clement et al., Proc. 24th Euro. conf. on contr. Fusion and Plasma Phys., Berchtesgaden, 1997 (European Physical Society, 1998), vol. 21A, p. 61.] indicated that during type I ELMs, surface temperatures in excess of 2000°C were measured, leading to peak power fluxes in the order of 4 GW/m2. The time integrated power flux exceeded the measured plasma energy loss per ELM by a factor of four. The reasons for this discrepancy are studied in this paper. Redeposited carbon layers of up to 40 μm have been found on the divertor surface in the places where the highest temperatures are measured. The impact of such layers on the power flux evaluation has been studied with numerical calculations, and a controlled simulation of ELM heating has been performed in the JET neutral beam test facility. It is found that neglecting the existence of layers on the surface in a 2D calculation can lead to overestimating the power by a factor of 3, whereas the error in the calculation of the energy is much smaller. An energy based calculation reduces the peak power during type I ELMs to values around 1.2 GW/m2.

After operation of the JET machine with the MkIIa divertor, samples of dust and flakes/deposits have been removed from the divertor region for analysis. Flakes/deposits up to 40 μm thick have been found in the inner leg of the divertor. The flakes/deposits formed on cold surfaces shadowed from the plasma. The flakes were found to be mostly carbon saturated with deuterium to a D/C ratio of 0.4. Analysis of the material removed has shown that it contains significant fractions (∼4%) of the tritium (produced by D–D reactions) and deuterium introduced in the JET machine during this period of operation. Dust material has also been collected. Vacuum cleaning collected very little material but a significant quantity of material was collected by smears. The analysis of this smeared material has shown it to be of a very similar chemical composition to that of flakes but to contain a significantly lower inventory of tritium. The dust has an average particle size of 27 μm and a specific surface area of 4 m 2 /g. The vast majority of this material is associated with co-deposited layers in the divertor region. The dust is on average present at a level of 120 μg/cm 2 .

Following a description of the state of knowledge concerning cleaning of the vacuum vessel of fusion reactors from Tritium by controlled oxidation of codeposited layers of carbon and hydrogen isotopes the experimental facility Deuterium Release Experiment (DEREX) is explained. In this facility a codeposited a-C:D layer from TEXTOR (1 μm thickness, D/C=0.03) has been oxidised in oxygen (300–600 Pa) at 523–673 K. The results indicate, that substantial release rates, leading to complete D-release within some days, are obtained at T ⩾623 K. Surprisingly, an over proportional dependence of release rate on oxygen partial pressure was observed; this result seems worth to be examined in more detail in order to optimise the cleaning method.

Pieces of beryllium tile exposed to a D-D plasma in JET have been studied by thermal desorption spectroscopy. These tiles have a thick layer of redeposited Be-C-O with considerable hydrogen and deuterium present. The samples were heated at a constant rate of 2°C/min, from 100°C to 900°C. Desorption peaks occurred in the range of 140-480°C. There was no significant desorption at temperatures above 600°C. The amount of deuterium detected varied from a low of 8 × 10 21 /m 2 to a high of 2.1 × 10 23 /m 2 . In one case, the amount of deuterium in a tile piece was seven times greater than the amount in a neighboring tile piece. Some of the tile pieces in the plasma-exposed region showed surface melting. Despite this, the deuterium yield from one of these pieces is >10 23 /m 2 .

New data on hydrogen plasma isotopes retention in beryllium and tungsten are now becoming available from various laboratories for conditions similar to those expected in the International Thermonuclear Experimental Reactor (ITER) where previous data were either missing or largely scattered. Together with a significant advancement in understanding, they have warranted a revisitation of the previous estimates of tritium inventory in ITER, with beryllium as the plasma facing material for the first-wall components, and tungsten in the divertor with some carbon-fibre-composites clad areas, near the strike points. Based on these analyses, it is shown that the area of primary concern, with respect to tritium inventory, remains codeposition with carbon and possibly beryllium on the divertor surfaces. Here, modelling of ITER divertor conditions continues to show potentially large codeposition rates which are confirmed by tokamak findings. Contrary to the tritium residing deep in the bulk of materials, this surface tritium represents a safety hazard as it can be easily mobilised in the event of an accident. It could, however, be possibly removed and recovered. It is concluded that active and efficient methods to remove the codeposited layers are needed in ITER and periodic conditioning/cleaning would be required to control the tritium inventory and avoid exhausting the available fuel supply. Some methods which could possibly be used for in-situ cleaning are briefly discussed in conjunction with the research and development work required to extrapolate their applicability to ITER.

Analytical results of a complete JET cryopump regeneration, including the nitrogen panel, following the first ITER-Like Wall campaign are presented along with the in-situ analyses of residual gas. H/D mixtures and impurities such as nitrogen and neon were injected during plasma operation in the vessel to study radiation cooling in the scrape-off-layer and divertor region. The global gas inventory over the campaign is incomplete, suggesting residual volatile impurities are remaining on the cryogenic panel. This paper presents results on a) residual deuterium on the panel which is related to the campaign very low, b) impurities like nitrogen which sticks on the panel and c) the ammonia production which can be observed in the RGA spectrum., 7 pages, 3 figures. This is an author-created, un-copyedited version of an article accepted for publication in Physica Scripta. IOP Publishing Ltd and IAEA are not responsible for any errors or omissions in this version of the manuscript or any version derived from it

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.

The DIVIMP/NIMBUS codes are employed to simulate divertor carbon profiles and the net erosion measured by colorimetry using the most recent chemical sputtering data and models. The relative roles of physical and chemical sputtering are assessed in terms of erosion and impurity contamination. Beryllium photon fluxes are also modelled, showing much less sputtering than predicted by the pure beryllium sputtering data. In addition, the erosion of fixed divertor target Langmuir probe tips has been measured photographically and gives an effective sputtering yield of

Erosion from and deposition on JET divertor tiles used during the 2007–2009 campaign and on inner wall guard limiter (IWGL) tiles used during 2005–2009 are studied. The tungsten coating on the divertor tiles was mostly intact with the largest erosion ∼30% in a small local area. Locally high erosion areas were observed on the load bearing divertor tile 5 and on the horizontal surface of the divertor tile 8. The IWGL tiles show a complicated distribution of erosion and deposition areas. The total amount of carbon deposited on the all IWGL tiles during the campaign 2005–2009 is estimated to be 65 g. The density of carbon deposits is estimated to be 0.67–0.83 g/cm 3 .

The amounts of deuterium retained within the JET vessel after campaigns in the years 1985 to 1989 have been reevaluated and shown to indicate that codeposition of D and wall material is the principal retention mechanism. During the period covered Be was introduced into JET by evaporating thin layers over the first wall and as a limiter material: the addition of Be is shown to have little effect on the rate of accumulation of the D inventory. The amount retained per pulse ( / 4 × 10 20 D atoms per pulse on average) is shown to fit well with a model wherein the amount of carbon (or other wall material) passing through the plasma during the discharge determines the maximum retention (which can only be realised if first wall temperatures do not exceed 270°C). Such a model may be adapted to predict hydrogen isotope retention in other large machines.

Carbon-based materials are being widely used in present tokamaks, because of their favourable properties, especially with respect to their thermal shock resistance. However, the significant effects resulting from neutron irradiation on the thermal-physical, mechanical properties and on the tritium trapping capacity could preclude the utilization of carbon as a plasma facing material in forthcoming devices. Therefore, the effects of neutron irradiation are of primary concern in the use of carbon materials in the next step devices. This paper presents the latest results of experimental investigations on the effects of neutron irradiation on the thermal-physical, mechanical properties and tritium retention behaviour of various graphites and carbon/carbon fiber composites for the doses of 10 −3 –30 dpa and irradiation temperatures of 200–1500°C. The state of experimental investigation of the behaviour of plasma hydrogen isotopes in carbon materials is given. In particular, the issue of neutron-induced changes in tritium behaviour is critically analysed.

A set of stainless steel (SS) and molybdenum mirror samples located in the divertor and at the outer mid-plane of the vessel were exposed in JET from 2005 to 2007. A selection of these mirror samples with well adhered deposits (i.e. not flaking) of up to a few hundred nanometers in thickness and with Be/C ratios ranging from 0 to similar to 1 have been cleaned using a laser system developed at CEA, Saclay. Following laser cleaning the recovered reflectivity was generally better in the infrared than the visible spectrum, with recovery of up to 90% of the initial reflectivity being obtained at 1600 nm for both Mo and SS mirrors falling as low as 20-30% of initial reflectivity at a wavelength of 400 nm for some SS mirrors, rising to similar to 80% for Mo mirrors. Some deposit remained on the mirrors after the cleaning trials. (C) 2010 EURATOM. Published by Elsevier B.V. All rights reserved.

Experiments in JET with carbon-based plasma-facing components have been carried out in preparation of the ITER-Like Wall with beryllium main chamber and full tungsten divertor. The preparatory work was twofold: (i) development of techniques, which ensure safe operation with the new wall and (ii) provision of reference plasmas, which allow a comparison of operation with carbon and metallic wall. (i) Compatibility with the W divertor with respect to energy loads could be achieved in N-2 seeded plasmas at high densities and low temperatures, finally approaching partial detachment, with only moderate confinement reduction of 10%. Strike-point sweeping increases the operational space further by re-distributing the load over several components. (ii) Be and C migration to the divertor has been documented with spectroscopy and QMBs under different plasma conditions providing a database which will allow a comparison of the material transport to remote areas with metallic walls. Fuel retention rates of 1.0-2.0 x 10(21) D s(-1) were obtained as references in accompanied gas balance studies. (C) 2010 Published by Elsevier B.V.

During detachment experiments of JET in 2007 (Brezinsek et al., 2009 [1]), 1 × 1022 molecules of 12CD4 were injected at a rate of about 5 × 1020 s−1 into L-mode plasma. The injection was done at the outer strike point in the centre of the horizontal target, just before opening the machine. The deposited layers were analyzed for deuterium post mortem after the removal of tiles. The heaviest local D deposition density is found immediately upstream of the gas inlet, but downstream the deposition is found over a larger area. In total, 3.7 × 1020 deuterium atoms were found locally deposited; if the mean D/C in the deposits is 0.4, then about 10% of the injected carbon was locally deposited. Transport and local deposition of the injected carbon was modeled with the 3D Monte Carlo impurity transport code ERO (Kirschner et al., 2000 [2]). The plasma background was generated with the onion-skin solver of the DIVIMP code (Stangeby and Elder, 1992 [3]). Locally deposited fraction is reproduced with ERO but the distribution not satisfactorily. In particular the poloidal transport of carbon is several times weaker than measured.

During the JET Preliminary Tritium Experiment (PTE), an estimated 2 × 1012 (1.1 × 1021 atoms) of tritium were injected into the JET vacuum vessel. A series of experiments was performed whose purpose was to deplete the torus of tritium, to compare the effectiveness of different methods of tritium removal and to obtain a quantitative understanding of the processes involved. The effectiveness of the cleaning procedures was such that the normal tokamak programme was resumed one week after the PTE and the routing of exhaust gases to the atmosphere after two weeks. The release of tritium from the vessel was found to scale with the deuterium release from the vessel, suggesting that dilution and mixing of the hydrogen isotopes in the vessel walls is important. High density, disruptive tokamak discharges were found to be the most successful plasma pulses for tritium removal. Purges with deuterium gas were also effective and have the advantage of operational simplicity. Helium discharges, on the other hand, resulted in low tritium release from the vessel walls. It was demonstrated that the tritium release rate could be predicted using data from hydrogen to deuterium changeover experiments. Using the superior quality of data available from the tritium cleanup experiment, the physical mechanisms necessary to describe the hydrogenic uptake and release from the JET torus were identified. Tile release of tritium is reproduced using a model that incorporates implantation into a thin surface layer as well as diffusion of tritium into and out of the bulk material

Three graphite tiles coated with B 4 C were mounted in the X-point strike region of JET during operations in 1991. The coatings remained adherent, though areas reached temperatures at which local melting occurred. Boron and deuterium analyses in the area surrounding the tiles suggest sublimation of boron was an important release mechanism, and that the region between the strike zones (the private flux region) is a major redeposition area.

The material transport by erosion and redeposition at the plasma facing wall areas in high temperature plasma experiments has been studied using limiter-like carbon probes with well defined surface deposits and depth markers. The probes have been exposed in the scrape-off layer (SOL) of the Joint European Torus (JET) during single discharges. For the evaluation of these experiments a computer program, ERO, has been developed. The calculated erosion-deposition rates for carbon as a function of the distance to the last closed flux surface (LCFS) agree well with the experimental results. For a single 4He JET discharge, erosion yields of 530 A for the silicon deposit and 80 A for vanadium have been measured near the LCFS. A large amount of redeposited silicon (about 17% of the sputtered atoms) has been found on the probe surface in co-deposition with carbon on an area not favoured by the proposed model. This observation can be explained by an additional force on the impurity ions (e.g. the existence of a local electric field), which may cause the deposition

Cross sections of deposited layers in the Joint European Torus (JET) were analysed using the nuclear micro analysis at the Tandem Laboratory, Uppsala University. For deuterium and beryllium the nuclear reactions 2D(3He,p)4He and 9Be(3He,pn)11B were exploited for analysis. Typically the analyses have been made with 10 μm spatial resolution and a sensitivity of better than one atomic percent for beryllium or deuterium in carbon matrix. Comparing several different surface treatment techniques shows that polishing the sample surface give very good optical surface information but that some amount of deuterium and beryllium probably is removed. For good quantitative results the measurement can either be done on a rough surface or the top of the polished surface can be cut off.

The impact of edge localized modes (ELMs) carrying energies of up to 450 kJ on carbon erosion in the JET inner divertor is assessed by means of time resolved measurements using an in situ quartz microbalance diagnostic. The inner target erosion is strongly nonlinearly dependent on the ELM energy: a single 400 kJ ELM produces the same carbon erosion as ten 150 kJ events. The ELM-induced enhanced erosion is attributed to the presence of codeposited carbon-deuterium layers on the inner divertor target, which are thermally decomposed under the impact of ELMs.