V. Philipps

Forschungszentrum Jülich, Jülich, North Rhine-Westphalia, Germany

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Publications (531)729.09 Total impact

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    ABSTRACT: The impact on the deuterium retention of simultaneous exposure of tungsten to a steady-state plasma and transient cyclic heat loads has been studied in the linear PSI-2 facility with the main objective of qualifying tungsten (W) as plasma-facing material. The transient heat loads were applied by a high-energy laser, a Nd:YAG laser ( λ = 1064 nm) with an energy per pulse of up to 32 J and a duration of 1 ms. A pronounced increase in the D retention by a factor of 13 has been observed during the simultaneous transient heat loads and plasma exposure. These data indicate that the hydrogen clustering is enhanced by the thermal shock exposures, as seen on the increased blister size due to mobilization and thermal production of defects during transients. In addition, the significant increase of the D retention during the simultaneous loads could be explained by an increased diffusion of D atoms into the W material due to strong temperature gradients during the laser pulse exposure and to an increased mobility of D atoms along the shock-induced cracks. Only 24% of the retained deuterium is located inside the near-surface layer ( d <4 μ m). Enhanced blister formation has been observed under combined loading conditions at power densities close to the threshold for damaging. Blisters are not mainly responsible for the pronounced increase of the D retention.
    No preview · Article · Feb 2016 · Physica Scripta
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    Full-text · Article · Feb 2016 · Physica Scripta
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    ABSTRACT: Experiments in the JET tokamak equipped with the ITER-like wall (ILW) revealed that the inner and outer target plate at the location of the strike points represent after one year of operation intact tungsten (W) surfaces without any beryllium (Be) surface coverage. The dynamics of near-surface retention, implantation, desorption and recycling of deuterium (D) in the divertor of plasma discharges are determined by W target plates. As the W plasma-facing components (PFCs) are not actively cooled, the surface temperature (T surface) is increasing with plasma exposure, varying the balance between these processes in addition to the impinging deuteron fluxes and energies. The dynamic behaviour on a slow time scale of seconds was quantified in a series of identical L-mode discharges (JET Pulse Number (JPN)) by intra-shot gas analysis providing the reduction of deuterium retention in W PFCs by 1/3 at a base temperature (T base) range at the outer target plate between 65 °C and 150 °C equivalent to a T surface span of 150 °C and 420 °C. The associated recycling and molecular D desorption during the discharge varies only at lowest temperatures moderately, whereas desorption between discharges rises significantly with increasing T base. The retention measurements represent the sum of inner and outer divertor interaction at comparable T surface. The dynamic behaviour on a fast time scale of ms was studied in a series of identical H-mode discharges (JPN ) and coherent edge-localized mode (ELM) averaging. High energetic ELMs of about 3 keV are impacting on the W PFCs with fluxes of which is about four times higher than inter-ELM ion fluxes with an impact energy of about E im = 200 eV. This intra-ELM ion flux is associated with a high heat flux of about 60 MW m−2 to the outer target plate which causes T surface rise by Δ T = 100 K per ELM covering finally the range between 160 °C and 1400 °C during the flat-top phase. ELM-induced desorption from saturated near-surface implantation regions as well as deep ELM-induced deuterium implantation areas under varying baseline temperature takes place. Subsequent refuelling by intra-ELM deuteron fluxes occurs and a complex interplay between deuterium fuelling and desorption can be observed in the temporal ELM footprint of the surface temperature (IR thermography), the impinging deuteron flux (Langmuir probes), and the Balmer radiation (emission spectroscopy) as representative for the deuterium recycling flux. In contrast to JET-C, a pronounced second peak, 8 ms delayed with respect to the initial ELM crash, in the D α radiation and the ion flux has been observed. The peak can be related to desorption of implanted energetic intra-ELM D+ diffusing to the W surface, and performing local recycling.
    Full-text · Article · Feb 2016 · Physica Scripta
  • M.Z. Tokar · N. Gierse · V. Philipps · U. Samm
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    ABSTRACT: For the interpretation of the line radiation observed from laser induced ablation spectroscopy (LIAS) such parameters as the density and temperature of electrons within very compact clouds of atoms and singly charged ions of ablated material have to be known. Compared to the local plasma conditions prior to the laser pulse, these can be strongly changed during LIAS since new electrons are generated by the ionisation of particles ejected from the irradiated target. Because of their transience and spatial inhomogeneity it is technically difficult to measure disturbances induced in the plasma by LIAS. To overcome this uncertainty a numerical model has been elaborated, providing a self-consistent description for the spreading of ablated particles and accompanying modifications in the plasma. The results of calculations for LIAS performed on carbon-containing targets in Ohmic and additionally heated discharges in the tokamak TEXTOR are presented. Due to the increase in the electron density the 'ionisation per photon' ratio, S/XB factor, is significantly enhanced compared to unperturbed plasma conditions. The impact of the amount of material ablated and of the plasma conditions before LIAS on the level of the S/XB-enhancement is investigated.
    No preview · Article · Sep 2015 · Nuclear Fusion
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    ABSTRACT: Abstract In optical diagnostic systems of ITER, mirrors will be used to guide the light from plasma towards detectors and cameras. The mirrors will be subjected to erosion due to fast particles and to deposition of impurities from the plasma which will affect adversely the mirror reflectivity and therefore must be suppressed or mitigated at the maximum possible extent. Predictive modeling envisages the successful suppression of deposition in the diagnostic ducts with fins trapping the impurities on their way towards mirrors located in the end of these ducts. To validate modeling predictions, cylindrical and cone-shaped diagnostic ducts were exposed in TEXTOR for 3960 s of plasma operation. After exposure, no drastic suppression of deposition was observed in the cylindrical ducts with fins. At the same time, no detectable deposition was found on the mirrors located at the end of cone-shaped ducts outlining the advantages of the cone geometry. Analyses of exposure provide evidence that the contamination of exposed mirrors was due to wall conditioning discharges and not due to working plasma exposure. Cleaning by plasma sputtering was performed on molybdenum mirrors pre-coated with a 100 nm thick aluminum film. Aluminum was used as a proxy of beryllium. During exposure in electron cyclotron resonance-generated helium plasma, the entire coating was sputtered within nine hours, leaving no trace of aluminum and leading to the full recovery of the specular reflectivity without detrimental effects on the mirror surface.
    Full-text · Article · Aug 2015 · Nuclear Fusion
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    ABSTRACT: High-density discharges on JET with ITER-like Wall (ILW) have been analysed with the aim of establishing a mechanism for the H-mode density limit (DL) and compared with experiments in the JET carbon material configuration. The density limit is up to 20% higher in the JET-ILW than in the JET-C machine. The observed H-mode density limit is found close to the Greenwald limit. It is sensitive to the main plasma shape and is almost independent of the heating power. It has been observed that the transition from H-mode to L-mode is not always an abrupt event but may exhibit a series of H–L–H transitions, the so-called “dithering H-mode”.
    No preview · Article · Aug 2015 · Journal of Nuclear Materials
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    ABSTRACT: Dust and tritium inventories in the vacuum vessel have upper limits in ITER that are set by nuclear safety requirements. Erosion, migration and re-deposition of wall material together with fuel co-deposition will be largely responsible for these inventories. The diagnostic suite required to monitor these processes, along with the set of the corresponding measurement requirements is currently under review given the recent decision by the ITER Organization to eliminate the first carbon/tungsten (C/W) divertor and begin operations with a full-W variant [1]. This paper presents the result of this review as well as the status of the chosen diagnostics.
    No preview · Article · Aug 2015 · Journal of Nuclear Materials
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    Q. Xiao · R. Hai · H. Ding · A. Huber · V. Philipps · N. Gierse · G. Sergienko
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    ABSTRACT: Laser-induced breakdown spectroscopy (LIBS) is considered as a promising method for in-situ diagnostic of the co-deposition and fuel retention during and in between plasma discharges in fusion devices. LIBS has been investigated intensively under laboratory conditions, while the application of LIBS in fusion devices is still in early stages. Moreover, the LIB processes are influenced by additional conditions in fusion devices, particularly the magnetic field. The experiments in TEXTOR show a significant enhancement in the spectral line emission and a deeper penetration of the laser-produced plasma into the edge plasma in the presence of magnetic field. These effects can be attributed to an increased confinement of the plasma by the magnetic field. The interference of magnetic field may compromise the quantitative interpretation of LIB spectra. Therefore, quantitative analysis of ITER-like co-deposits was done in laboratory without magnetic field as well as in TEXTOR with a magnetic field of Bt ∼ 2.25 T.
    Full-text · Article · Aug 2015 · Journal of Nuclear Materials
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    Full-text · Dataset · Jul 2015
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    A. Terra · G. Sergienko · M. Hubeny · A. Huber · Ph. Mertens · V. Philipps
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    ABSTRACT: Abstract This contribution reports on the concept of a circular self-rotating and temperature self-stabilising plasma-facing component (PFC), and test of a related prototype in TEXTOR tokamak. This PFC uses the Lorentz force induced by plasma current and magnet field (J × B) to create a torque applied on metallic discs which produce a rotational movement. Additional thermionic current, present at high operation temperatures, brings additional temperature stabilisation ability. This self-rotating disk limiter was exposed to plasma in the TEXTOR tokamak under different radial positions to vary the heat flux. This disk structure shows the interesting ability to stabilise its maximum temperature through the fact that the self-induced rotation is modulated by the thermal emission current. It was observed that the rotation speed increased following both the current collected by the limiter, and the temperature of the tungsten disks.
    Full-text · Article · Jul 2015
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    ABSTRACT: 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.
    No preview · Article · Jun 2015 · Nuclear Fusion
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    ABSTRACT: In JET-ILW isotopic plasma wall changeover experiments have been carried out to determine the amount of particles accessible by changing the plasma from H to D and from D to H. The gas balance analysis integrated over the experimental sessions show that the total amount of H or D removed from the wall is in the range of (1-3) × 1022D. For both changeover experiments, the respective plasma isotopic ratio behaviour is exactly the same as a function of the pulse number. After only 80 s of plasma (4 pulses), the plasma isotopic ratio is lower than 10%, below 4.5% after 13 pulses and then saturates around ~2-3%. In these conditions, the removal efficiency through plasma operation becomes very poor. The saturation of the plasma isotopic ratio in the range of 10% is also observed for the JET-C configuration although the amount of tritium retained in the vessel after the DT pulses was more than one order of magnitude compared to the retention observed with the JET-ILW. This demonstrates that the amount of particle recovery through plasma changeover is independent from the long term retention. Since this long term reservoir results from codeposition, these experiments suggest that there is a limited access to these codeposited particles by plasma isotopic changeover. Finally, in ITER, change over from D/T to H at the end of the discharge for possibly reducing the long term retention does not appear as a good strategy.
    No preview · Article · Apr 2015 · Nuclear Fusion
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    ABSTRACT: Tungsten (W) specimens previously exposed to deuterium (D) plasmas both in the TEXTOR tokamak and high flux linear plasma generator (LPG) were subsequently loaded with tritium at 573 K for 3 h. Retention of tritium in the near-surface W layer was examined by imaging plate technique. On the TEXTOR-plasma-exposed W surface, tritium was mainly trapped in carbon deposits. For LPG-plasma-exposed W specimens, tritium was trapped in defects created in the near-surface layer during the course of D plasma exposure.
    No preview · Article · Apr 2015 · Fusion Science and Technology
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    ABSTRACT: Ferromagnetic pebbles are investigated as high heat flux (q∥) plasma facing components in fusion devices with short power decay length (λq) on a conceptual level. The ability of a pebble concept to cope with high heat fluxes is retained and extended by the acceleration of ferromagnetic pebbles in magnetic fields. An alloying concept suited for fusion application is outlined and the compatibility of ferromagnetic pebbles with plasma operation is discussed.Steel grade 1.4510 is chosen as a well characterized candidate material to perform an analysis of the heating process. Scaling relationships as a function of q∥ for maximum and optimal pebble diameter, allowed exposure time, and removal time safety margin are obtained numerically for spherical pebble geometry. The acceleration of ferromagnetic pebbles in a tokamak resulting from magnetic gradients is studied and operation parameters for an ITER-based reactor are outlined. Counter-intuitively, it is found that ferromagnetic pebbles perform better for narrow λq profiles, making them an attractive heat exhaust concept for next step devices and thus an option to be investigated in detail.The key results of this study are that very high heat fluxes are accessible in the operation space of ferromagnetic pebbles, that ferromagnetic pebbles are compatible with tokamak operation and current divertor designs, that the heat removal capability of ferromagnetic pebbles increases as λq decreases and, finally, that for fusion relevant values of q∥ pebble diameters below 100μm are required.
    Full-text · Article · Mar 2015
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    ABSTRACT: Cracking thresholds and crack patterns in tungsten targets have been studied in recent experiments after repetitive ITER-like ELM heat pulses in combination with plasma exposure in PSI-2 (Γtarget = 2.5–4.0 × 1021 m−2 s−1, ion energy on surface Eion = 60 eV, Te ≈ 10 eV). The heat pulses were simulated by laser irradiation. A Nd:YAG laser with energy per pulse of up to 32 J and a duration of 1 ms at the fundamental wavelength (λ = 1064 nm, repetition rate 0.5 Hz) was used to irradiate ITER-grade W samples with repetitive heat loads.
    Full-text · Article · Feb 2015 · Fusion Engineering and Design

  • No preview · Article · Feb 2015
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    ABSTRACT: The migration of beryllium, tungsten and carbon to remote areas of the inner JET-ILW divertor and the accompanying co-deposition of deuterium has been investigated using post-mortem analysis of the housings of quartz-micro balances (QMBs) and their quartz crystals. The analysis of the deposition provides that the rate of beryllium atoms is significantly reduced compared to the analogue deposition rate of carbon during the carbon wall conditions (JET-C) at the same locations of the QMBs. A reduction factor of 50 was found at the entrance gap to the cryo-pumps while it was 14 under tile 5, the semi-horizontal target plate. The deposits consist of C/Be atomic ratios of typically 0.1–0.5 showing an enrichment of carbon in remote areas compared to directly exposed areas with less carbon. The deuterium retention fraction D/Be is between 0.3 and 1 at these unheated locations in the divertor.
    No preview · Article · Jan 2015 · Journal of Nuclear Materials
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    ABSTRACT: The analysis of the deposition of eroded wall material on the plasma-facing materials in fusion devices is one of the crucial issues to maintain the plasma performance and to fulfill safety requirements with respect to tritium retention by co-deposition. Laser ablation with minimal damage to the plasma facing material is a promising method for in situ monitoring and removal of the deposition, especially for plasma-shadowed areas which are difficult to reach by other cleaning methods like plasma discharge. It requires the information of ablation process and the ablation threshold for quantitative analysis and effective removal of the different deposits. This paper presents systemic laboratory experimental analysis of the behavior of the ITER relevant materials, graphite, tungsten, aluminum (as a substitution of beryllium) and mixed deposits ablated by a Nd:YAG laser (1064 nm) with different energy densities (1–27 J/cm2, power density 0.3–3.9 GW/cm2). The mixed deposits consisted of W–Al–C layer were deposited on W substrate by magnetron sputtering and arc plasma deposition. The aim was to select the proper parameters for the quantitative analysis and for laser removal of the deposits by investigating the ablation efficiency and ablation threshold for the bulk materials and deposits. The comparison of the ablation and saturation energy thresholds for pure and mixed materials shows that the ablation threshold of the mixed layer depends on the concentration of the components. We propose laser induced breakdown spectroscopy for determination of the elemental composition of deposits and then we select the laser parameters for the layer removal. Comparison of quantitative analysis results from laboratory to that from TEXTOR shows reasonable agreements. The dependence of the spectra on plasma parameters and ambient gas pressure is investigated.
    Full-text · Article · Dec 2014 · Journal of Nuclear Materials
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    ABSTRACT: The isotopic exchange efficiencies of JET Ion Cyclotron Wall Conditioning (ICWC) discharges produced at ITER half and full field conditions are compared for JET carbon (C) and ITER like wall (ILW). Besides an improved isotope exchange rate on the ILW providing cleaner plasma faster, the main advantage compared to C-wall is a reduction of the ratio of retained discharge gas to removed fuel. Complementing experimental data with discharge modeling shows that long pulses with high (∼240 kW coupled) ICRF power maximizes the wall isotope removal per ICWC pulse. In the pressure range 1–7.5 × 10−3 Pa, this removal reduces with increasing discharge pressure. As most of the wall-released isotopes are evacuated by vacuum pumps in the post discharge phase, duty cycle optimization studies for ICWC on JET-ILW need further consideration. The accessible reservoir by H2-ICWC at ITER half field conditions on the JET-ILW preloaded by D2 tokamak operation is estimated to be 7.3 × 1022 hydrogenic atoms, and may be exchanged within 400 s of cumulated ICWC discharge time.
    Full-text · Article · Dec 2014 · Journal of Nuclear Materials
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    ABSTRACT: Systematic study of deuterium irradiation effects on tungsten was done under ITER - relevant high particle flux density, scanning a broad surface temperature range. Polycrystalline ITER - like grade tungsten samples were exposed in linear plasma devices to two different ranges of deuterium ion flux densities (high: 3.5-7 · 1023 D+/m2 s and low: 9 · 1021 D+/m2 s). Particle fluence and ion energy, respectively 1026 D+/m2 and ∼38 eV were kept constant in all cases.The experiments were performed at three different surface temperatures 530 K, 630 K and 870 K. Experimental results concerning the deuterium retention and surface modifications of low flux exposure confirmed previous investigations. At temperatures 530 K and 630 K, deuterium retention was higher at lower flux density due to the longer exposure time (steady state plasma operation) and a consequently deeper diffusion range. At 870 K, deuterium retention was found to be higher at high flux density according to the thermal desorption spectroscopy (TDS) measurements. While blisters were completely absent at low flux density, small blisters of about 40-50 nm were formed at high flux density exposure. At the given conditions, a relation between deuterium retention and blister formation has been found which has to be considered in addition to deuterium trapping in defects populated by diffusion.
    Full-text · Article · Nov 2014 · Journal of Nuclear Materials

Publication Stats

9k Citations
729.09 Total Impact Points

Institutions

  • 1987-2015
    • Forschungszentrum Jülich
      • • Institute of Energy and Climate Research (IEK)
      • • Zentralabteilung für Chemische Analysen (ZCH)
      Jülich, North Rhine-Westphalia, Germany
  • 2011
    • Culham Centre for Fusion Energy
      Abingdon-on-Thames, England, United Kingdom
  • 2002
    • Nagoya University
      • Center for Integrated Research in Science and Engineering (CIRSE)
      Nagoya, Aichi, Japan