K. Heinola

Diffusion of hydrogen in bcc tungsten studied with first principle calculations

First principle calculations were used to study the hydrogen migration properties in bulk bcc tungsten. Hydrogen has low solubility in tungsten and occupies the tetrahedral interstitial site with an energy difference of 0.38 eV compared to the octahedral interstitial site. The hydrogen diffusion coefficient was evaluated using the harmonic transition state theory and was found to agree with the experimental results at temperatures above 1500 K. The height of the migration barrier between two adjacent tetrahedral sites was found to be 0.21 eV, which is lower than the value 0.39 eV obtained for the migration barrier from degassing measurements in the temperature range between 1100 and 2400 K. The tunneling correction to the diffusion rate provides much better agreement with the experimental result at 29 K than the extrapolated experimental D from high temperature measurements.

Bond-order potential for point and extended defect simulations in tungsten

A reactive interatomic bond-order potential for bcc tungsten is presented. Special attention in the potential development was given for obtaining accurate formation and migration energies for point defects, making the potential useful in atomic scale simulations of point and extended defects. The potential was used to calculate binding energies and trapping distances for vacancies in vacancy clusters and the recombination radius for self-interstitial atom and monovacancy.

Material migration patterns and overview of first surface analysis of the JET ITER-like wall

Following the first JET ITER-like wall operations a detailed in situ photographic survey of the main chamber and divertor was completed. In addition, a selection of tiles and passive diagnostics were removed from the vessel and made available for post mortem analysis. From the photographic survey and results from initial analysis, the first conclusions regarding erosion, deposition, fuel retention and material transport during divertor and limiter phases have been drawn. The rate of deposition on inner and outer base divertor tiles and remote divertor corners was more than an order of magnitude less than during the preceding carbon wall operations, as was the concomitant deuterium retention. There was however beryllium deposition at the top of the inner divertor. The net beryllium erosion rate from the mid-plane inner limiters was found to be higher than for the previous carbon wall campaign although further analysis is required to determine the overall material balance due to erosion and re-deposition.

Beryllium Migration in JET ITER-like Wall Plasmas

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.

Erosion and deposition in the JET divertor during the first ILW campaign

Erosion and deposition were studied in the JET divertor during the first JET ITER-like wall campaign 2011 to 2012 using marker tiles. An almost complete poloidal section consisting of tiles 0, 1, 3 ...

Surface analysis of tiles and samples exposed to the first JET campaigns with the ITER-like wall

This paper reports on the first post-mortem analyses of tiles removed from JET after the first campaigns with the ITER-like wall (ILW) during 2011?12?[1]. Tiles from the divertor have been analysed by ion beam analysis techniques and by secondary ion mass spectrometry to determine the amount of beryllium deposition and deuterium retention in the tiles exposed to the scrape-off layer. Films 10?20??m thick were present at the top of tile 1, but only very thin films (<?1??m) were found in the shadowed areas and on other divertor tiles. The total amount of Be found in the divertor following the ILW campaign was a factor of ???9 less than the material deposited in the 2007?09 carbon campaign, after allowing for the longer operations in 2007?09.

Long-term fuel retention in JET ITER-like wall

Post-mortem studies with ion beam analysis, thermal desorption, and secondary ion mass spectrometry have been applied for investigating the long-term fuel retention in the JET ITERlike wall components. The retention takes place via implantation and co-deposition, and the highest retention values were found to correlate with the thickness of the deposited impurity layers. From the total amount of retained D fuel over half was detected in the divertor region. The majority of the retained D is on the top surface of the inner divertor, whereas the least retention was measured in the main chamber on the mid-plane of the inner wall limiter. The recessed areas of the inner wall showed significant contribution to the main chamber total retention. Thermal desorption spectroscopy analysis revealed the energetic T from DD reactions being implanted in the divertor. The total T inventory was assessed to be >0.3 mg.

First dust study in JET with the ITER-like wall: sampling, analysis and classification

Results of the first dust survey in JET with the ITER-Like Wall (JET-ILW) are presented. The sampling was performed using adhesive stickers from the divertor tiles where the greatest material depos ...

Tile profiling analysis of samples from the JET ITER-like wall and carbon wall

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.

Deuterium irradiation-induced defect concentrations in tungsten

Deuterium retention in the implantation-induced defects in polycrystalline tungsten has been studied. Deuterium was implanted with different energies and concentrations of retained D were analysed with secondary ion mass spectrometry and nuclear reaction analysis. Annealings were carried out at four pre-determined temperatures corresponding to four different defect types that can trap deuterium. A quantitative number of each defect type produced by 5, 15 and 30 keV D implantation with a dose of 5.8◊ 10 16 cm 2 was obtained.