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  • 1. Coenen, J. W.
    et al.
    Matthews, G. F.
    Krieger, K.
    Iglesias, D.
    Bunting, P.
    Corre, Y.
    Silburn, S.
    Balboa, I.
    Bazylevs, B.
    Conway, N.
    Coffey, I.
    Dejarnac, R.
    Gauthier, E.
    Gaspar, J.
    Jachmich, S.
    Jepu, I.
    Makepeace, C.
    Scannell, R.
    Stamp, M.
    Petersson, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Pitts, R. A.
    Wiesen, S.
    Widdowson, A.
    Heinola, K.
    Baron-Wiechec, A.
    Transient induced tungsten melting at the Joint European Torus (JET)2017In: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T170, article id 014013Article in journal (Refereed)
    Abstract [en]

    Melting is one of the major risks associated with tungsten (W) plasma-facing components (PFCs) in tokamaks like JET or ITER. These components are designed such that leading edges and hence excessive plasma heat loads deposited at near normal incidence are avoided. Due to the high stored energies in ITER discharges, shallow surface melting can occur under insufficiently mitigated plasma disruption and so-called edge localised modes-power load transients. A dedicated program was carried out at the JET to study the physics and consequences of W transient melting. Following initial exposures in 2013 (ILW-1) of a W-lamella with leading edge, new experiments have been performed on a sloped surface (15 degrees slope) during the 2015/2016 (ILW-3) campaign. This new experiment allows significantly improved infrared thermography measurements and thus resolved important issue of power loading in the context of the previous leading edge exposures. The new lamella was monitored by local diagnostics: spectroscopy, thermography and high-resolution photography in between discharges. No impact on the main plasma was observed despite a strong increase of the local W source consistent with evaporation. In contrast to the earlier exposure, no droplet emission was observed from the sloped surface. Topological modifications resulting from the melting are clearly visible between discharges on the photographic images. Melt damage can be clearly linked to the infrared measurements: the emissivity drops in zones where melting occurs. In comparison with the previous leading edge experiment, no runaway melt motion is observed, consistent with the hypothesis that the escape of thermionic electrons emitted from the melt zone is largely suppressed in this geometry, where the magnetic field intersects the surface at lower angles than in the case of perpendicular impact on a leading edge. Utilising both exposures allows us to further test the model of the forces driving melt motion that successfully reproduced the findings from the original leading edge exposure. Since the ILW-1 experiments, the exposed misaligned lamella has now been retrieved from the JET machine and post mortem analysis has been performed. No obvious mass loss is observed. Profilometry of the ILW-1 lamella shows the structure of the melt damage which is in line with the modell predictions thus allowing further model validation. Nuclear reaction analysis shows a tenfold reduction in surface deuterium concentration in the molten surface in comparison to the non-molten part of the lamella.

  • 2.
    Fazinic, Stjepko
    et al.
    Rudjer Boskovic Inst, Bijenicka 54, Zagreb 10000, Croatia..
    Tadic, Tonic
    Rudjer Boskovic Inst, Bijenicka 54, Zagreb 10000, Croatia..
    Vuksic, Marin
    Rudjer Boskovic Inst, Bijenicka 54, Zagreb 10000, Croatia..
    Rubel, Marek
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Petersson, Per
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Fortuna-Zalesna, Elibieta
    Warsaw Univ Technol, Fac Mat Sci & Technol, Woloska 141, PL-02507 Warsaw, Poland..
    Widdowson, Anna
    Culham Sci Ctr, Culham Ctr Fus Energy, Abingdon OX14 3DB, Oxon, England..
    Ion Microbeam Analyses of Dust Particles and Codeposits from JET with the ITER-Like Wall2018In: Analytical Chemistry, ISSN 0003-2700, E-ISSN 1520-6882, Vol. 90, no 9, p. 5744-5752Article in journal (Refereed)
    Abstract [en]

    Generation of metal dust in the JET tokamak with the ITER-like wall (ILW) is a topic of vital interest to next-step fusion devices because of safety issues with plasma operation. Simultaneous Nuclear Reaction Analysis (NRA) and Particle Induced X-ray Emission (PIXE) with a focused four MeV He-3 microbeam was used to determine the composition of dust particles related to the JET operation with the ILW. The focus was on "Be-rich particles" collected from the deposition zone on the inner divertor tile. The particles found are composed of a mix of codeposited species up to 120 m in size with a thickness of 30-40 mu m, The main constituents are D from the fusion fuel, Be and W from the main plasma-facing components, and Ni and Cr from the Inconel grills of the antennas for auxiliary plasma heating. Elemental concentrations were estimated by iterative NRA-PIXE analysis. Two types of dust particles were found: (i) larger Be-rich particles with Be concentrations above 90 at% with a deuterium presence of up to 3.4 at% and containing Ni (1-3 at%), Cr (0.4-0.8 at%), W (0.2-0.9 at%), Fe (0.3-0.6 at%), and Cu and Ti in lower concentrations and (ii) small particles rich in Al and/or Si that were in some cases accompanied by other elements, such as Fe, Cu, or Ti or W and Mo.

  • 3. Fortuna-Zalesna, E.
    et al.
    Grzonka, J.
    Moon, Sunwoo
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Petersson, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Widdowson, A.
    Fine metal dust particles on the wall probes from JET-ILW2017In: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T170, article id 014038Article in journal (Refereed)
    Abstract [en]

    Collection and ex situ studies of dust generated in controlled fusion devices during plasma operation are regularly carried out after experimental campaigns. Herewith results of the dust survey performed in JET after the second phase of operation with the metal ITER-like wall (2013-2014) are presented. For the first-time-ever particles deposited on silicon plates acting as dust collectors installed in the inner and outer divertor have been examined. The emphasis is on analysing metal particles (Be and W) with the aim to determine their composition, size and surface topography. The most important is the identification of beryllium dust in the form of droplets (both splashes and spherical particles), flakes of co-deposits and small fragments of Be tiles. Tungsten and nickel rich (from Inconel) particles are also identified. Nitrogen from plasma edge cooling has been detected in all types of particles. They are categorized and the origin of various constituents is discussed.

  • 4.
    Garcia Carrasco, Alvaro
    et al.
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Petersson, Per
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Widdowson, A.
    Fortuna-Zalesna, E.
    Jachmich, S.
    Brix, M.
    Marot, L.
    Plasma impact on diagnostic mirrors in JET2017In: NUCLEAR MATERIALS AND ENERGY, ISSN 2352-1791, Vol. 12, p. 506-512Article in journal (Refereed)
    Abstract [en]

    Metallic mirrors will be essential components of all optical systems for plasma diagnosis in ITER. This contribution provides a comprehensive account on plasma impact on diagnostic mirrors in JET with the ITER-Like Wall. Specimens from the First Mirror Test and the lithium-beam diagnostic have been studied by spectrophotometry, ion beam analysis and electron microscopy. Test mirrors made of molybdenum were retrieved from the main chamber and the divertor after exposure to the 2013-2014 experimental campaign. In the main chamber, only mirrors located at the entrance of the carrier lost reflectivity (Be deposition), while those located deeper in the carrier were only slightly affected. The performance of mirrors in the JET divertor was strongly degraded by deposition of beryllium, tungsten and other species. Mirrors from the lithium-beam diagnostic have been studied for the first time. Gold coatings were severely damaged by intense arcing. As a consequence, material mixing of the gold layer with the stainless steel substrate occurred. Total reflectivity dropped from over 90% to less than 60%, i.e. to the level typical for stainless steel.

  • 5.
    Garcia Carrasco, Alvaro
    et al.
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Petersson, Per
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Schwarz-Selinger, T.
    Wauters, T.
    Douai, D.
    Bobkov, V.
    Cavazzana, R.
    Krieger, K.
    Lyssoivan, A.
    Moeller, S.
    Spolaore, M.
    Rohde, V.
    Rubel, M.
    Investigation of probe surfaces after ion cyclotron wall conditioning in ASDEX upgrade2017In: NUCLEAR MATERIALS AND ENERGY, ISSN 2352-1791, Vol. 12, p. 733-735Article in journal (Refereed)
    Abstract [en]

    For the first time, material analysis techniques have been applied to study the effect of ion cyclotron wall conditioning (ICWC) on probe surfaces in a metal-wall machine. ICWC is a technique envisaged to contribute to the removal of fuel and impurities from the first wall of ITER. The objective of this work was to assess impurity migration under ICWC operation. Tungsten probes were exposed in ASDEX Upgrade to discharges in helium. After wall conditioning, the probes were covered with a co-deposited layer containing D, B, C, N, O and relatively high amount of He. The concentration ratio He/C+B was 0.7. The formation of the co-deposited layer indicates that a fraction of the impurities desorbed from the wall under ICWC operation are transported by plasma and deposited away from their original location.

  • 6. Hakola, A.
    et al.
    Brezinsek, S.
    Douai, D.
    Balden, M.
    Bobkov, V.
    Carralero, D.
    Greuner, H.
    Elgeti, S.
    Kallenbach, A.
    Krieger, K.
    Meisl, G.
    Oberkofler, M.
    Rohde, V.
    Schneider, P.
    Schwarz-Selinger, T.
    Lahtinen, A.
    De Temmerman, G.
    Caniello, R.
    Ghezzi, F.
    Wauters, T.
    Garcia Carrasco, Alvaro
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Petersson, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Radovic, I. Bogdanovic
    Siketic, Z.
    Plasma-wall interaction studies in the full-W ASDEX upgrade during helium plasma discharges2017In: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 57, no 6, article id 066015Article in journal (Refereed)
    Abstract [en]

    Plasma-wall interactions have been studied in the full-W ASDEX Upgrade during its dedicated helium campaign. Relatively clean plasmas with a He content of > 80% could be obtained by applying ion cyclotron wall conditioning (ICWC) discharges upon changeover from D to He. However, co-deposited layers with significant amounts of He and D were measured on W samples exposed to ICWC plasmas at the low-field side (outer) midplane. This is a sign of local migration and accumulation of materials and residual fuel in regions shadowed from direct plasma exposure albeit globally D was removed from the vessel. When exposing W samples to ELMy H-mode helium plasmas in the outer strike-point region, no net erosion was observed but the surfaces had been covered with co-deposited layers mainly consisting of W, B, C, and D and being the thickest on rough and modified surfaces. This is different from the typical erosion-deposition patterns in D plasmas, where usually sharp net-erosion peaks surrounded by prominent net-deposition maxima for W are observed close to the strike point. Moreover, no clear signs of W nanostructure growth or destruction could be seen. The growth of deposited layers may impact the operation of future fusion reactors and is attributed to strong sources in the main chamber that under suitable conditions may switch the balance from net erosion into net deposition, even close to the strike points. In addition, the absence of noticeable chemical erosion in helium plasmas may have affected the thickness of the deposited layers. Retention of He, for its part, remained small and uniform throughout the strike-point region although our results indicate that samples with smooth surfaces can contain an order of magnitude less He than their rough counterparts.

  • 7. Louche, F.
    et al.
    Wauters, T.
    Ragona, R.
    Moeller, S.
    Durodie, F.
    Litnovsky, A.
    Lyssoivan, A.
    Messiaen, A.
    Ongena, J.
    Petersson, Per
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Brezinsek, S.
    Linsmeier, Ch.
    Van Schoor, M.
    Design of an ICRF system for plasma-wall interactions and RF plasma production studies on TOMAS2017In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 123, p. 317-320Article in journal (Refereed)
    Abstract [en]

    Ion cyclotron wall conditioning (ICWC) is being developed for ITER and W7-X as a baseline conditioning technique in which the ion cyclotron heating and current drive system will be employed to produce and sustain the currentless conditioning plasma. The TOMAS project (TOroidal MAgnetized System, operated at the FZ-juelich, Germany) proposes to explore several key aspects of ICWC. For this purpose we have designed an ICRF system made of a single strap antenna within a metallic box, connected to a feeding port and a pre-matching system. We discuss the design work of the antenna system with the help of the commercial electromagnetic software CST Microwave Studio (R). The simulation results for a given geometry provide input impedance matrices for the two-port system. These matrices are afterwards inserted into various circuit models to assess the accessibility of the required frequency range. The sensitivity of the matching system to uncertainties on plasma loading and capacitance values is notably addressed. With a choice of three variable capacitors we show that the system can cope with such uncertainties. We also demonstrate that the system can cope as well with the high reflected power levels during the short breakdown phase of the RF discharge, but at the cost of a significantly reduced coupled power.

  • 8. Matejicek, Jiri
    et al.
    Weinzettl, Vladimir
    Mackova, Anna
    Malinsky, Petr
    Havranek, Vladimir
    Naydenkova, Diana
    Klevarova, Veronika
    Petersson, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Gasior, Pawel
    Hakola, Antti
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Fortuna, Elzbieta
    Kolehmainen, Jukka
    Tervakangas, Sanna
    Interaction of candidate plasma facing materials with tokamak plasma in COMPASS2017In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 493, p. 102-119Article in journal (Refereed)
    Abstract [en]

    The interaction of tokamak plasma with several materials considered for the plasma facing components of future fusion devices was studied in a small-size COMPASS tokamak. These included mainly tungsten as the prime candidate and chromium steel as an alternative whose suitability was to be assessed. For the experiments, thin coatings of tungsten, P92 steel and nickel on graphite substrates were prepared by arc-discharge sputtering. The samples were exposed to hydrogen and deuterium plasma discharges in the COMPASS tokamak in two modes: a) short exposure (several discharges) on a manipulator in the proximity of the separatrix, close to the central column, and b) long exposure (several months) at the central column, aligned with the other graphite tiles. During the discharges, standard plasma diagnostics were used and a local emission of spectral lines in the visible near ultraviolet regions, corresponding to the material erosion, was monitored. Before and after the plasma exposures, the sample surfaces were observed using scanning electron microscopy, the coatings thickness was measured using Rutherford backscattering spectroscopy, and the concentration profiles of hydrogen and deuterium were measured by elastic recoil detection analysis. The uniformity of the coatings and their thickness was verified before the exposure. After the exposure, no reduction of the thickness was observed, indicating the absence of 'global' erosion. Erosion was observed only in isolated spots, and attributed to unipolar arcing. Slightly larger erosion was found on the steel coatings compared to the tungsten ones. Incorporation of deuterium in a thin surface layer was observed, in dependence on the exposure mode. Additionally, boron enrichment of the long-exposure samples was observed, as a result of the tokamak chamber boronization.

  • 9.
    Rubel, Marek
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Moon, Soonwoo
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Petersson, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Garcia Carrasco, Alvaro
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Hallén, Anders
    KTH, School of Information and Communication Technology (ICT), Electronics.
    Krawczynska, A.
    Fortuna-Zalesna, E.
    Gilbert, M.
    Plocinski, T.
    Widdowson, A.
    Metallic mirrors for plasma diagnosis in current and future reactors: tests for ITER and DEMO2017In: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T170, article id 014061Article in journal (Refereed)
    Abstract [en]

    Optical spectroscopy and imaging diagnostics in next-step fusion devices will rely on metallic mirrors. The performance of mirrors is studied in present-day tokamaks and in laboratory systems. This work deals with comprehensive tests of mirrors: (a) exposed in JET with the ITER-like wall (JET-ILW); (b) irradiated by hydrogen, helium and heavy ions to simulate transmutation effects and damage which may be induced by neutrons under reactor conditions. The emphasis has been on surface modification: deposited layers on JET mirrors from the divertor and on near-surface damage in ion-irradiated targets. Analyses performed with ion beams, microscopy and spectro-photometry techniques have revealed: (i) the formation of multiple co-deposited layers; (ii) flaking-off of the layers already in the tokamak, despite the small thickness (130-200 nm) of the granular deposits; (iii) deposition of dust particles (0.2-5 mu m, 300-400 mm(-2)) composed mainly of tungsten and nickel; (iv) that the stepwise irradiation of up to 30 dpa by heavy ions (Mo, Zr or Nb) caused only small changes in the optical performance, in some cases even improving reflectivity due to the removal of the surface oxide layer; (v) significant reflectivity degradation related to bubble formation caused by the irradiation with He and H ions.

  • 10.
    Rubel, Marek
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Petersson, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Zhou, Yushan
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Coad, J. P.
    Lungu, C.
    Jepu, I.
    Porosnicu, C.
    Matveev, D.
    Kirschner, A.
    Brezinsek, S.
    Widdowson, A.
    Alves, E.
    Fuel inventory and deposition in castellated structures in JET-ILW2017In: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 57, no 6, article id 066027Article in journal (Refereed)
    Abstract [en]

    Since 2011 the JET tokamak has been operated with a metal ITER-like wall (JET-ILW) including castellated beryllium limiters and lamellae-type bulk tungsten tiles in the divertor. This has allowed for a large scale test of castellated plasma-facing components (PFC). Procedures for sectioning the limiters into single blocks of castellation have been developed. This facilitated morphology studies of morphology of surfaces inside the grooves for limiters after experimental campaigns 2011-2012 and 2013-2014. The deposition in the 0.4-0.5 mm wide grooves of the castellation is 'shallow'. It reaches 1-2 mm into the 12 mm deep gap. Deuterium concentrations are small (mostly below 1 × 1018 cm-2). The estimated total amount of deuterium in all the castellated limiters does not exceed the inventory of the plasma-facing surfaces (PFS) of the limiters. There are only traces of Ni, Cr and Fe deposited in the castellation gaps. The same applies to the carbon content. Also low deposition of D, Be and C has been measured on the sides of the bulk tungsten lamellae pieces. Modelling clearly reflects: (a) a sharp decrease in the measured deposition profiles and(b) an increase in deposition with the gap width. Both experimental and modelling data give a strong indication and information to ITER that narrow gaps in the castellated PFC are essential. X-ray diffraction on PFS has clearly shown two distinct composition patterns: Be with an admixture of Be-W intermetallic compounds (e.g. Be22W) in the deposition zone, whilst only pure Be has been detected in the erosion zone. The lack of compound formation in the erosion zone indicates that no distinct changes in the thermo-mechanical properties of the Be PFC might be expected.

  • 11.
    Rubel, Marek
    et al.
    KTH.
    Widdowson, A.
    Culham Sci Ctr, Culham Ctr Fus Energy, Abingdon OX14 3DB, Oxon, England..
    Grzonka, J.
    Warsaw Univ Technol, PL-02507 Warsaw, Poland.;Inst Elect Mat Technol, PL-01919 Warsaw, Poland..
    Fortuna-Zalesna, E.
    Warsaw Univ Technol, PL-02507 Warsaw, Poland..
    Moon, Sunwoo
    KTH.
    Petersson, Per
    KTH.
    Ashikawa, N.
    Natl Inst Fus Sci, Toki, Gifu 5095292, Japan..
    Asakura, N.
    Natl Inst Quantum Radiol Sci & Technol, Rokkasho 0393212, Japan..
    Hamaguchi, D.
    Natl Inst Quantum Radiol Sci & Technol, Rokkasho 0393212, Japan..
    Hatano, Y.
    Toyama Univ, Hydrogen Isotope Res Ctr, Gofuku 3190, Toyama 9308555, Japan..
    Isobe, K.
    Natl Inst Quantum Radiol Sci & Technol, Rokkasho 0393212, Japan..
    Masuzaki, S.
    Natl Inst Fus Sci, Toki, Gifu 5095292, Japan..
    Kurotaki, H.
    Natl Inst Quantum Radiol Sci & Technol, Rokkasho 0393212, Japan..
    Oya, Y.
    Shizuoka Univ, Suruga Ku, 836 Ohya, Shizuoka 4228529, Japan..
    Oyaidzu, M.
    Natl Inst Quantum Radiol Sci & Technol, Rokkasho 0393212, Japan..
    Tokitani, M.
    Natl Inst Fus Sci, Toki, Gifu 5095292, Japan..
    Dust generation in tokamaks: Overview of beryllium and tungsten dust characterisation in JET with the ITER-like wall2018In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 136, p. 579-586Article in journal (Refereed)
    Abstract [en]

    Operation of the JET tokamak with beryllium and tungsten ITER-like wall provides unique opportunity for detailed studies on dust generation: quantity, morphology, location, etc. The programme carried out in response to ITER needs for safety assessment comprises: (i) remotely controlled vacuum cleaning of the divertor; (ii) local sampling of loosely bound matter from plasma-facing components (PFC); (iii) collection of mobilized dust on various erosion-deposition probes located in the divertor and in the main chamber. Results of comprehensive analyses performed by a number of complementary techniques, e.g. a range of microscopy methods, electron and ion spectroscopy, liquid scintillography and thermal desorption, are summarized by following points: (a) Total amount of dust collected by vacuum cleaning after three campaigns is about 1-1.4 g per campaign (19.1-23.5 h plasma operation), i.e. over 100 times smaller than in JET operated with carbon walls (i.e. in JET-C). (b) Two major categories of Be dust are identified: flakes of co-deposits formed on PFC and droplets (2-10 mu m in diameter). Small quantifies, below 1 g, of Be droplets and splashes are associated mainly with melting of beryllium limiters.

  • 12.
    Ström, Petter
    et al.
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Petersson, Per
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Arredondo Parra, R.
    Oberkofler, M.
    Schwarz-Selinger, T.
    Primetzhofer, D.
    Sputtering of polished EUROFER97 steel: Surface structure modification and enrichment with tungsten and tantalum2018In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 508, p. 139-146Article in journal (Refereed)
    Abstract [en]

    Surface structure modification and enrichment with tungsten and tantalum were measured for polished EUROFER97 samples after exposure to a deuterium ion beam. Time-of-flight medium energy ion scattering and time-of-flight elastic recoil detection analysis were implemented for measuring atomic composition profiles. Atomic force microscopy and optical microscopy were used to investigate surface morphology. The deuterium particle fluence was varied between 1021 D/m2 and 1024 D/m2, projectile energy was 200 eV/D and exposure temperatures up to 1050 K were applied. The average fraction of tungsten plus tantalum to total metal content in the 2 nm closest to the sample surface was increased from an initial 0.0046 to 0.12 for the sample exposed to the highest fluence at room temperature. The enrichment was accompanied by an increase in surface roughness of one order of magnitude and grain dependent erosion of the material. The appearance of protrusions with heights up to approximately 40 nm after ion beam exposure at room temperature was observed on individual grains. Samples exposed to 1023 D/m2 at temperatures of 900 K and 1050 K displayed recrystallization and cracking while changes to the total surface fraction of tungsten and tantalum were limited to less than a factor of two compared to the sample exposed to the same fluence at room temperature.

  • 13.
    Ström, Petter
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Petersson, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Possnert, Goran
    Uppsala Univ, Dept Phys & Astron, Tandem Lab, Box 529, SE-75120 Uppsala, Sweden..
    Erratum: "A combined segmented anode gas ionization chamber and time-of-flight detector for heavy ion elastic recoil detection analysis" [Rev. Sci. Instrum. 87, 103303 (2016)]2018In: Review of Scientific Instruments, ISSN 0034-6748, E-ISSN 1089-7623, Vol. 89, no 4, article id 049901Article in journal (Refereed)
  • 14. Tsavalas, P.
    et al.
    Lagoyannis, A.
    Mergia, K.
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Triantou, K.
    Harissopulos, S.
    Kokkoris, M.
    Petersson, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Be ITER-like wall at the JET tokamak under plasma2017In: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T170, article id 014049Article in journal (Refereed)
    Abstract [en]

    The JET tokamak is operated with beryllium and tungsten plasma-facing components to prepare for the exploitation of ITER. To determine beryllium erosion and migration in JET a set of markers were installed. Specimens from different beryllium marker tiles of the main wall of the ITER-like wall (ILW) JET tokamak from the first and the second D-D campaign were analyzed with nuclear reaction analysis, x-ray fluorescence spectroscopy, scanning electron microscopy and x-ray diffraction (XRD). Emphasis was on the determination of carbon plasma impurities deposited on beryllium surfaces. The C-12(d, p(0))C-13 reaction was used to quantify carbon deposition and to determine depth profiles. Carbon quantities on the surface of the Be tiles are low, varying from (0.35 +/- 0.07) x 10(17) to (11.8 +/- 0.6) x 10(17) at cm(-2) in the deposition depth from 0.4 to 6.7 mu m, respectively. In the 0.4-0.5 mm wide grooves of castellation sides the carbon content is found up to (14.3 +/- 2.5) x 10(17) at cm(-2) while it is higher (up to (38 +/- 4) x 10(17) at cm(-2)) in wider gaps (0.8 mm) separating tile segments. Oxygen (O), titanium (Ti), chromium (Cr), manganese (Mn), iron (Fe), nickel (Ni) and tungsten (W) were detected in all samples exposed to plasma and the reference one but at lower quantities at the latter. In the central part of the Inner Wall Guard Limiter from the first ILW campaign and in the Outer Poloidal Limiter from the second ILW campaign the Ni interlayer has been completely eroded. XRD shows the formation of BeNi in most specimens.

  • 15. Weckmann, Armin
    et al.
    Petersson, Per
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Kirschner, A.
    Wienhold, P.
    Brezinsek, S.
    Kreter, A.
    Pospieszczyk, A.
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Whole-machine material migration studies in the TEXTOR tokamak with molybdenum2017In: NUCLEAR MATERIALS AND ENERGY, ISSN 2352-1791, Vol. 12, p. 518-523Article in journal (Refereed)
    Abstract [en]

    MoF6 injection from a localised source into plasma edge in the TEXTOR tokamak was the last experiment before the final shut-down of the TEXTOR machine. During decommissioning all plasma-facing components (PFCs) became available for surface studies. Detailed mapping of Mo deposition was performed in order to determine its migration on global scale. The concentration of Mo on PFC decays exponentially with distance from the source. The decay length is of the order of 0.1 m on the main PFC and 1 m on the receded components. Also the decay lengths modelled with the ERO code are between 0.15-1.3 m, depending on the anomalous cross-field diffusion coefficient. The inner bumper limiter is found to be the major repository for Mo. Material balance measurements show that only up to 22% of the injected Mo was detected on all the PFCs thus indicating that a large fraction of injected Mo may have been pumped out before being deposited.

  • 16.
    Weckmann, Armin
    et al.
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Petersson, Per
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Rubel, Marek
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Ström, Petter
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Kurki-Suonio, T.
    Aalto Univ, Dept Appl Phys, Aalto 00076, Finland..
    Sarkimaki, K.
    Aalto Univ, Dept Appl Phys, Aalto 00076, Finland..
    Kirschner, A.
    Forschungszentrum Julich, Inst Energie & Klimaforsch Plasmaphys, D-52425 Julich, Germany..
    Kreter, A.
    Forschungszentrum Julich, Inst Energie & Klimaforsch Plasmaphys, D-52425 Julich, Germany..
    Brezinsek, S.
    Forschungszentrum Julich, Inst Energie & Klimaforsch Plasmaphys, D-52425 Julich, Germany..
    Romazanov, J.
    Forschungszentrum Julich, Inst Energie & Klimaforsch Plasmaphys, D-52425 Julich, Germany..
    Wienhold, P.
    Forschungszentrum Julich, Inst Energie & Klimaforsch Plasmaphys, D-52425 Julich, Germany..
    Pospieszczyk, A.
    Forschungszentrum Julich, Inst Energie & Klimaforsch Plasmaphys, D-52425 Julich, Germany..
    Hakola, A.
    VTT Tech Res Ctr Finland Ltd, Espoo 02044, Finland..
    Airila, M.
    VTT Tech Res Ctr Finland Ltd, Espoo 02044, Finland..
    Review on global migration, fuel retention and modelling after TEXTOR decommission2018In: NUCLEAR MATERIALS AND ENERGY, Vol. 17, p. 83-112Article, review/survey (Refereed)
    Abstract [en]

    Before decommissioning of the TEXTOR tokamak in 2013, the machine was conditioned with a comprehensive migration experiment where MoF6 and N-15(2) were injected on the very last operation day. Thereafter, all plasmafacing components (PFCs) were available for extensive studies of both local and global migration of impurities - Mo, W, Inconel alloy constituents, 15 N, F - and fuel retention studies. Measurements were performed on 140 limiter tiles out of 864 throughout the whole machine to map global transport. One fifth of the introduced molybdenum could be found. Wherever possible, the findings are compared to results obtained previously in other machines. This review incorporates both published and unpublished results from this TEXTOR study and combines findings with analytical methods as well as modelling results from two codes, ERO and ASCOT. The main findings are: Both local and global molybdenum transport can be explained by toroidal plasma flow and (sic) x (sic) drift. The suggested transport scheme for molybdenum holds also for other analysed species, namely tungsten from previous experiments and medium-Z metals (Cr-Cu) introduced on various occasions. Analytical interpretation of several deposition profile features is possible with basic geometrical and plasma physics considerations. These are deposition profiles on the collector probe, the lower part of the inner bumper limiter, the poloidal cross-section of the inner bumper limiter, and the poloidal limiter. Any deposition pattern found in this TEXTOR study, including fuel retention, has neither poloidal nor toroidal symmetry, which is often assumed when determining deposition profiles on global scale. Fuel retention is highly inhomogeneous due to local variation of plasma parameters - by auxiliary heating systems and impurity injection - and PFC temperature. Local modelling with ERO yields good qualitative agreement but too high local deposition efficiency. Global modelling with ASCOT shows that the radial electric field and source form have a high impact on global deposition patterns, while toroidal flow has little influence. Some of the experimental findings could be reproduced. Still, qualitative differences between simulated and experimental global deposition patterns remain. The review closes with lessons learnt during this extensive TEXTOR study which might be helpful for future scientific exploitation of other tokamaks to be decommissioned.

  • 17. Widdowson, A.
    et al.
    Alves, E.
    Baron-Wiechec, A.
    Barradas, N. P.
    Catarino, N.
    Coad, J. P.
    Corregidor, V.
    Garcia-Carrasco, A.
    Heinola, K.
    Koivuranta, S.
    Krat, S.
    Lahtinen, A.
    Likonen, J.
    Mayer, M.
    Petersson, Per
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Van Boxel, S.
    Overview of the JET ITER-like wall divertor2017In: NUCLEAR MATERIALS AND ENERGY, ISSN 2352-1791, Vol. 12, p. 499-505Article in journal (Refereed)
    Abstract [en]

    The work presented draws on new analysis of components removed following the second JET ITER-like wall campaign 2013-14 concentrating on the upper inner divertor, inner and outer divertor corners, lifetime issues relating to tungsten coatings on JET carbon fibre composite divertor tiles and dust/particulate generation. The results show that the upper inner divertor remains the region of highest deposition in the JET-ILW. Variations in plasma configurations between the first and second campaign have altered material migration to the corners of the inner and outer divertor. Net deposition is shown to be beneficial in the sense that it reduces W coating erosion, covers small areas of exposed carbon surfaces and even encapsulates particles.

  • 18. Widdowson, A.
    et al.
    Coad, J. P.
    Alves, E.
    Baron-Wiechec, A.
    Barradas, N. P.
    Brezinsek, S.
    Catarino, N.
    Corregidor, V.
    Heinola, K.
    Koivuranta, S.
    Krat, S.
    Lahtinen, A.
    Likonen, J.
    Matthews, G. F.
    Mayer, M.
    Petersson, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. EUROfus Consortium JET, England.
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. EUROfus Consortium JET, England.
    Overview of fuel inventory in JET with the ITER-like wall2017In: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 57, no 8, article id 086045Article in journal (Refereed)
    Abstract [en]

    Post mortem analyses of JET ITER-Like-Wall tiles and passive diagnostics have been completed after each of the first two campaigns (ILW-1 and ILW-2). They show that the global fuel inventory is still dominated by co-deposition; hence plasma parameters and sputtering processes affecting material migration influence the distribution of retained fuel. In particular, differences between results from the two campaigns may be attributed to a greater proportion of pulses run with strike points in the divertor corners, and having about 300 discharges in hydrogen at the end of ILW-2. Recessed and remote areas can contribute to fuel retention due to the larger areas involved, e.g. recessed main chamber walls, gaps in castellated Be main chamber tiles and material migration to remote divertor areas. The fuel retention and material migration due to the bulk W Tile 5 during ILW-1 are presented. Overall these tiles account for only a small percentage of the global accountancy for ILW-1.

  • 19. Widdowson, A.
    et al.
    Coad, J. P.
    Alves, E.
    Baron-Wiechec, A.
    Barradas, N. P.
    Catarino, N.
    Corregidor, V.
    Heinola, K.
    Krat, S.
    Likonen, J.
    Matthews, G. F.
    Mayer, M.
    Petersson, Per
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Rubel, Marek
    KTH.
    Impurity re-distribution in the corner regions of the JET divertor2017In: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T170, article id 014060Article in journal (Refereed)
    Abstract [en]

    The International Thermonuclear Experimental Reactor (ITER) will use a mixture of deuterium (D) and tritium (T) as the fuel to generate power. Since T is both radioactive and expensive the Joint European Torus (JET) has been at the forefront of research to discover how much T is used and where it may be retained within the main reaction chamber. Until the year 2010 the JET plasma facing components were constructed of carbon fibre composites. During the JET carbon (C) phases impurities accumulated at the corners of the divertor located towards the bottom of the chamber in regions shadowed from the plasma where they are very difficult to reach and remove. This build-up of C and the associated H-isotope (including T) retention were of particular concern for future fusion reactors therefore, in 2010 JET changed the wall protection to (mainly) Be and the divertor to tungsten (W)-the JET ITER-like wall (ILW)-the choice of materials for ITER. This paper reveals that with the JET ILW impurities are still accumulating in the shadowed regions, with Be being the majority element, though the overall quantities are very much reduced from those in the C phases. Material will be transported into the shadowed regions principally when the plasma strike points are on the corner tiles, but particles typically have about a 75% probability of reflection from line-of sight surfaces, and multiple reflection/scattering results in deposition over all surfaces.

  • 20.
    Zhou, Yushan
    et al.
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Bergsåker, Henric
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Bykov, Igor
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Petersson, Per
    KTH, School of Electrical Engineering and Computer Science (EECS), Fusion Plasma Physics.
    Possnert, G.
    Likonen, J.
    Pettersson, J.
    Koivuranta, S.
    Widdowson, A. M.
    Microanalysis of deposited layers in the inner divertor of JET with ITER-like wall2017In: NUCLEAR MATERIALS AND ENERGY, ISSN 2352-1791, Vol. 12, p. 412-417Article in journal (Refereed)
    Abstract [en]

    In JET with ITER-like wall, beryllium eroded in the main chamber is transported to the divertor and deposited mainly at the horizontal surfaces of tiles 1 and 0 (high field gap closure, HFGC). These surfaces are tungsten coated carbon fibre composite (CFC). Surface sampleswere collected following the plasma operations in 2011-2012 and 2013-2014 respectively. The surfaces, as well as polished cross sections of the deposited layers at the surfaces have been studied with micro ion beam analysis methods (mu-IBA). Deposition of Beand other impurities, and retention of D is microscopically inhomogeneous. Impurities and trapped deuterium accumulate preferentially in cracks, pits and depressed regions, and at the sides of large pits in the substrate (e.g. arc tracks where the W coating has been removed). With careful overlaying of mu-NRA elemental maps with optical microscopy images, it is possible to separate surface roughness effects from depth profiles at microscopically flat surface regions.

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