Fusion Technology Activities at JET in Support of the ITER Program

AbstrAct Among the technological activities performed at JET in support of the scientific objectives of both JET and ITER, a significant effort is devoted to the investigation of the erosion, transport and deposition of wall materials, and of their fuel retention properties. With the analysis of wall tiles retrieved in the 2010 shutdown, the full characterization of the previous JET carbon wall is obtained. In order to confirm the expectations on properties of the new ITER-Like Wall (ILW) installed in 2011, a large number of marker tiles and profiled tiles have been prepared and installed both in the main wall and in the divertor. These will be retrieved from the vessel during a short shutdown at the end of 2012 and analysed. The major changes introduced by the new ILW materials in JET required also a new nuclear characterization of the machine. Neutronics measurements have been performed to obtain the neutron/g-ray field changes inside and outside the JET machine. The experimental data are also used to validate neutronics codes used in ITER design. A new calibration of neutron detectors, scheduled in the 2012 shutdown and adopting the same procedure as in ITER, has been prepared based on extensive neutronics calculations. IntroductIon The JET research program includes technological activities in support of the scientific objectives of both JET and ITER. To this purpose, in 2010-11 the JET machine has undergone a major change to replace the previous carbon wall with a new ITER-Like Wall (ILW) making use of beryllium and tungsten in plasma facing components The erosion/deposition of wall materials in JET is characterized by net erosion on the main chamber wall and outer divertor, and migration of eroded material mainly to the inner divertor. During the 2009-10 shutdown phase, dust was collected from JET vessel [2] and several removed tiles were selected for analysis of deposits and surface. In this paper, first results of analyses on tiles exposed in 2007-2009 are presented. Considering also all previous results of erosion and deposition studies, a full characterization of the C wall in JET will be derived for comparison with the new ILW. Presently, one of the major objectives of the JET program is the investigation of the wall material transport, erosion/deposition and the fuel retention properties in the ILW. It is expected from laboratory experiments that the main mechanism for the fuel retention in the ILW is co-deposition in Be-layers. Implantation is the main mechanism for the retention in W, but it is expected to play a negligible role in comparison with the Be co-deposition. In order to confirm these expectations, and the results obtained during first ILW plasmas from gas balance method The major changes introduced by the new ILW materials in JET required also a new nuclear characterization of the machine. Neutronics analyses have been performed to calculate the neutron/gray field changes inside and outside the machine, the material activation and the shutdown doserat

Material migration and fuel retention studies during the JET carbon divertor campaigns

The first divertor was installed in the JET machine between 1992 and 1994 and was operated with carbon tiles and then beryllium tiles in 1994-5. Post-mortem studies after these first experiments demonstrated that most of the impurities deposited in the divertor originate in the main chamber, and that asymmetric deposition patterns generally favouring the inner divertor region result from drift in the scrape-off layer. A new monolithic divertor structure was installed in 1996 which produced heavy deposition at shadowed areas in the inner divertor corner, which is where the majority of the tritium was trapped by co-deposition during the deuterium-tritium experiment in 1997. Different divertor geometries have been tested since such as the Gas-Box and High-Delta divertors; a principle objective has been to predict plasma behaviour, transport and tritium retention in ITER. Transport modelling experiments were carried out at the end of four campaigns by puffing C-13-labelled methane, and a range of diagnostics such as quartz-microbalance and rotating collectors have been installed to add time resolution to the post-mortem analyses. The study of material migration after D-D and D-T campaigns clearly revealed important consequences of fuel retention in the presence of carbon walls. They gave a strong impulse to make a fundamental change of wall materials. In 2010 the carbon divertor and wall tiles were removed and replaced with tiles with Be or W surfaces for the ITER-Like Wall Project.EURATOM 633053RCUK Energy Programme P012450/

Tritium distributions on W-coated divertor tiles used in the third JET ITER-like wall campaign

Tritium (T) distributions on tungsten (W)-coated plasma-facing tiles used in the third ITER-like wall campaign (2015-2016) of the Joint European Torus (JET) were examined by means of an imaging plate technique and beta-ray induced x-ray spectrometry, and they were compared with the distributions after the second (2013-2014) campaign. Strong enrichment of T in beryllium (Be) deposition layers was observed after the second campaign. In contrast, T distributions after the third campaign was more uniform though Be deposition layers were visually recognized. The one of the possible explanations is enhanced desorption of T from Be deposition layers due to higher tile temperatures caused by higher energy input in the third campaign.EURATOM 633053Japan Society for the Promotion of Science JP2628935

Long-term fuel retention and release in JET ITER-Like Wall at ITER-relevant baking temperatures

The fuel outgassing efficiency from plasma-facing components exposed in JET-ILW has been studied at ITER-relevant baking temperatures. Samples retrieved from the W divertor and Be main chamber were annealed at 350 and 240 °C, respectively. Annealing was performed with thermal desoprtion spectrometry (TDS) for 0, 5 and 15 h to study the deuterium removal effectiveness at the nominal baking temperatures. The remained fraction was determined by emptying the samples fully of deuterium by heating W and Be samples up to 1000 and 775 °C, respectively. Results showed the deposits in the divertor having an increasing effect to the remaining retention at temperatures above baking. Highest remaining fractions 54 and 87% were observed with deposit thicknesses of 10 and 40 μm, respectively. Substantially high fractions were obtained in the main chamber samples from the deposit-free erosion zone of the limiter midplane, in which the dominant fuel retention mechanism is via implantation: 15 h annealing resulted in retained deuterium higher than 90%. TDS results from the divertor were simulated with TMAP7 calculations. The spectra were modelled with three deuterium activation energies resulting in good agreement with the experimentsEURATOM 63305

Material migration and fuel retention studies during the JET carbon divertor campaigns

The first divertor was installed in the JET machine between 1992 and 1994 and was operated with carbon tiles and then beryllium tiles in 1994–5. Post-mortem studies after these first experiments demonstrated that most of the impurities deposited in the divertor originate in the main chamber, and that asymmetric deposition patterns generally favouring the inner divertor region result from drift in the scrape-off layer. A new monolithic divertor structure was installed in 1996 which produced heavy deposition at shadowed areas in the inner divertor corner, which is where the majority of the tritium was trapped by co-deposition during the deuterium-tritium experiment in 1997. Different divertor geometries have been tested since such as the Gas-Box and High-Delta divertors; a principle objective has been to predict plasma behaviour, transport and tritium retention in ITER. Transport modelling experiments were carried out at the end of four campaigns by puffing 13C-labelled methane, and a range of diagnostics such as quartz-microbalance and rotating collectors have been installed to add time resolution to the post-mortem analyses. The study of material migration after D-D and D-T campaigns clearly revealed important consequences of fuel retention in the presence of carbon walls. They gave a strong impulse to make a fundamental change of wall materials. In 2010 the carbon divertor and wall tiles were removed and replaced with tiles with Be or W surfaces for the ITER-Like Wall Project

The influence of carbon impurities on the formation of loops in tungsten irradiated with self-ions

The microstructure changes taking place in W under irradiation are governed by many factors, amongst which C impurities and their interactions with self-interstitial atoms (SIA). In this work, we specifically study this effect by conducting a dedicated 2-MeV self-ions irradiation experiment, at room temperature. Samples were irradiated up to 0.02, 0.15 and 1.2 dpa. Transmission electron microscopy (TEM) expectedly revealed a large density of SIA loops at all these doses. Surprisingly, however, the loop number density increased in a non-monotonous manner with the received dose. Performing chemical analysis with secondary ion spectroscopy measurements (SIMS), we find that our samples were likely contaminated by C injection during the irradiation. Employing an object kinetic Monte Carlo (OKMC) model for microstructure evolution, we demonstrate that the C injection is the likely factor explaining the evolution of loops number density. Our findings highlight the importance of the well-known issue of C injection during ion irradiation experiments, and demonstrate how OKMC models can help to rationalize this effect.Peer reviewe