J. P. Gunn

Cea Leti, Grenoble, Rhône-Alpes, France

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Publications (126)164.83 Total impact

  • [Show abstract] [Hide abstract]
    ABSTRACT: Recently, the surface of carbon fibre composite tiles of the toroidal pump limiter of Tore Supra has been analysed by scanning electron and atomic force microscopies. In regions where fibres are perpendicular to the surface, a specific erosion pattern has been observed. It is constituted of a striation oriented with an angle oblique with respect to the magnetic field. The characteristic wavelength of this structure is micrometric, and similar to the fibre size. Modelling has been undertaken to reproduce this micrometric pattern. It is shown to originate from the carbon composite structure, for which it has been found by measurement using laboratory plasma that the erosion rate of the fibres is different from that of the surrounding matrix. Modelling emphasizes the effect of the impinging flux angle distributions of deuterium ions and carbon impurities that are preliminarily determined from computation of the magnetic sheath. In the case of deuterium the sheath is shown to have little effect on the particle trajectories for the simulation parameters considered here, although when impurities are included the sheath deflection is significant. Furthermore this study shows how the fibre organization in the composite influences the striation direction and points out the importance of the angular dependence of the sputtering yield.
    Nuclear Fusion 11/2014; 54(12):123006. · 2.73 Impact Factor
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    ABSTRACT: In this paper, we report results of a dedicated experiment that gives the plasma penetration profiles inside a gap of a tokamak castellated plasma-facing component. A specially designed probe that recreates a gap between two tiles has been built for the purpose of this study. It allows to measure ion saturation profiles along the 2 sides and at the bottom of the gap for both poloidal and toroidal orientations. The novelty of such experiment is the real time measurement of the plasma flux inside the gap during a tokamak D-shaped discharge compared to previous experimental studies which were mainly post-mortem. This experiment was performed in the COMPASS tokamak and results are compared with particle-in-cell simulations. The plasma deposition is found to be asymmetric in both orientations with a stronger effect in poloidal gaps. The Larmor radius of the incoming ions plays a role in the plasma penetration only in poloidal gaps but seems to have little impact in toroidal gaps. Profiles are qualitatively well reproduced by simulations. Ion current is recorded at the bottom of a toroidal gap under certain conditions.
    Fusion Engineering and Design 10/2014; · 0.84 Impact Factor
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    ABSTRACT: The EDGE2D-EIRENE code is applied for simulation of divertor detachment during matched density ramp experiments in high triangularity, L-mode plasmas in both JET-Carbon (JET-C) and JET-ITER-Like Wall (JET-ILW). The code runs without drifts and includes either C or Be as impurity, but not W, assuming that the W targets have been coated with Be via main chamber migration. The simulations reproduce reasonably well the observed particle flux detachment as density is raised in both JET-C and JET-ILW experiments and can better match the experimental in-out divertor target power asymmetry if the heat flux entering the outer divertor is artificially set at around 2–3 times that entering the inner divertor. A careful comparison between different sets of atomic physics processes used in EIRENE shows that the detachment modelled by EDGE2D-EIRENE relies only on an increase of the particle sinks and not on a decrease of the ionization source. For the rollover and the beginning of the partially detached phase, the particle losses by perpendicular transport and the molecular activated recombination processes are mainly involved. For a deeper detachment with significant target ion flux reduction, volume recombination appears to be the main contributor. The elastic molecule-ion collisions are also important to provide good neutral confinement in the divertor and thus stabilize the simulations at low electron temperature (T e), when the sink terms are strong. Comparison between EDGE2D-EIRENE and SOLPS4.3 simulations of the density ramp in C shows similar detachment trends, but the importance of the elastic ion-molecule collisions is reduced in SOLPS4.3. Both codes suggest that any process capable of increasing the neutral confinement in the divertor should help to improve the modelling of the detachment. A further outcome of this work has been to demonstrate that key JET divertor diagnostic signals—Langmuir probe T e and bolometric tomographic reconstructions—are running beyond the limit of validity in high recycling and detached conditions and cannot be reliably used for code validation. The simulations do, however, reproduce the trend of the evolution of the line integrated bolometer chord measurements. The comparison between the code results and high-n Balmer line radiation intensity profiles confirms that a strong volume recombination is present during the experimental detachment and may play a role in this process, as suggested by the code.
    Nuclear Fusion 09/2014; 54(9). · 2.73 Impact Factor
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    ABSTRACT: The upgrade of its ion cyclotron resonance (ICRH) and lower hybrid current drive (LHCD) heating systems makes the Tore Supra (TS) tokamak particularly well suited to address the physics and technology of high-power and steady-state plasma–surface interactions. High radio frequency (RF) heating powers have been successfully applied up to 12.2MW coupled to the plasma, in which about 7.85MW flows through the scrape-off layer. Thermal calculation based on thermography measurements gives the heat flux density distribution on the TS toroidal limiter located at the bottom of the machine. The target heat flux densities are divided by the incidence angle of the field lines with the surface and mapped to the magnetic flux surface to evaluate the power flowing in the scrape-off layer (SOL). The power profile shows a narrow component near the last closed flux surface and a wide component in the rest of the SOL. The narrow component is attributed to significant cross-field heat flux density around the plasma contact point, about 0.8% of the parallel heat flux density in the SOL, when incident angles are nearly tangential to the surface. The wide component is used to derive the experimental heat flux decay length (λq ) and parallel heat flux in the SOL. The power widths are measured for a series of 1 MA/3.8 T discharges involving a scan of RF injected power 3.5 � Ptot � 12.2MW. Independently of the heating power, we measured λq,OMP = 14.5±1.5mm at the outer mid-plane and parallel heat flux in the SOL in the range 130 � QLCFS 490MWm−2. TS values obtained with L-mode limiter plasmas are broader than those derived from L-mode divertor plasmas, confirming earlier results obtained with an ohmically heated plasma leaning on the inboard wall of TS.
    Nuclear Fusion 01/2014; 54(1):013013. · 2.73 Impact Factor
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    ABSTRACT: A theory-based scaling for the characteristic length of a circular, limited tokamak scrape-off layer (SOL) is obtained by considering the balance between parallel losses and non-linearly saturated resistive ballooning mode turbulence driving anomalous perpendicular transport. The SOL size increases with plasma size, resistivity, and safety factor q. The scaling is verified against flux-driven non-linear turbulence simulations, which reveal good agreement within a wide range of dimensionless parameters, including parameters closely matching the TCV tokamak. An initial comparison of the theory against experimental data from several tokamaks also yields good agreement.
    Nuclear Fusion 12/2013; 53(12):2001-. · 2.73 Impact Factor
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    ABSTRACT: To support the design of an ITER ion-cyclotron range of frequency heating (ICRH) system and to mitigate risks of operation in ITER, CEA has initiated an ambitious Research & Development program accompanied by experiments on Tore Supra or test-bed facility together with a significant modelling effort. The paper summarizes the recent results in the following areas:
    Nuclear Fusion 07/2013; 53(8):083012. · 2.73 Impact Factor
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    ABSTRACT: We report on the first full 3D EMC3-Eirene simulations of the bulk plasma, neutral gas and W impurity transport in Tore Supra (TS). Three configurations are addressed: (A) the high field side-, (B) bottom- and (C) low field side limited case and compared to Mach probe measurements at the top of the device. As in previous 2D simulations the 3D modelling also showed a strong discrepancy between the measured and simulated Mach number in case B. In order to investigate the finite size of the Mach probe we included this object as a plasma limiting structure in the simulation and found that the probe indeed perturbs the plasma, which explains the discrepancy only partly. The W transport simulation was found to have a rather weak dependence on the magnetic configuration and the simulated tungsten confinement time shows a similar behavior as those measured recently by Meyer et al. The absolute value, however, differs by a factor of 10.
    Journal of Nuclear Materials 07/2013; 438:S254–S257. · 2.02 Impact Factor
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    ABSTRACT: The influence of RF electric fields on retarding field analyzer (RFA) measurements of sheath potential, Vsh is investigated. One-dimensional particle-in-cell simulations show that the RFA is able to measure reliably the rectified dc sheath potential only for ion plasma frequencies ωpi similar to the rf wave frequency ωrf, while for real SOL conditions (ωpi > ωrf), when the RFA is magnetically connected to an RF antenna, it is strongly underestimated. An alternative method to investigate RF sheaths effects is proposed that uses broadening of the ion distribution function as evidence of the rf electric fields in the sheath. RFA measurements in Tore Supra indicate that the average effects of rf potentials do indeed propagate from the antenna 12 m along magnetic field lines.
    Journal of Nuclear Materials 07/2013; · 2.02 Impact Factor
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    ABSTRACT: This paper describes the design and operation of a new tungsten (W) injection system for impurity transport experiments in the Tore Supra tokamak. The system is mounted on a reciprocating manipulator and injects a controlled amount of gaseous tungsten hexacarbonyl, W(CO)6 at arbitrary depth in the scrape-off layer, using an inertially activated valve. Injected W(CO)6 is dissociated in the plasma, forming a radially localized plume of W atoms. The injector does not require an external gas feed and can perform a large number of injections from an on-board reservoir of W(CO)6. Some examples of W injections in Tore Supra are included, demonstrating successful operation and discussing some technical issues of the injector prototype.
    The Review of scientific instruments 07/2013; 84(7):073501. · 1.52 Impact Factor
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    ABSTRACT: Using high-resolution diagnostics, this paper investigates experimentally two ways of influencing the radiofrequency (RF) sheath patterns (magnitude and spatial distribution) developing around active Tore Supra ion cyclotron antennae. Firstly two types of Faraday screen electrical designs were compared. The new design aimed at reducing the parallel RF electric field integrated along “long field lines” passing in front of the antenna, by interrupting all parallel RF current paths on its front face. This proved inefficient for attenuating the RF-sheaths on the screen itself, and only weakly modified their distribution. Secondly the ratio of strap RF voltage amplitudes was varied, either during spontaneous bifurcations of the antenna electrical operational point or through the power balance between RF generators. This affected the left–right symmetry of screen heat loads: the local fluxes tended to evolve as the RF voltages on the nearest strap. Both experiments question the method used to optimize the new screen.
    Journal of Nuclear Materials 07/2013; 438:S330–S333. · 2.02 Impact Factor
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    ABSTRACT: This paper presents an analysis of the carbon–deuterium circulation and the resulting balance in Tore Supra over the period 2002–2007. Carbon balance combines the estimation of carbon gross erosion from spectroscopy, net erosion and deposition using confocal microscopy, lock-in thermography and SEM, and a measure of the amount of deposits collected in the vacuum chamber. Fuel retention is determined from post-mortem (PM) analyses and gas balance (GB) measurements. Special attention was paid to the deuterium outgassed during the nights and weekends of the experimental campaign (vessel under vacuum, Plasma Facing Components at 120 °C) and during vents (vessel at atmospheric pressure, PFCs at room temperature). It is shown that this outgassing is the main process reconciling the PM and GB estimations of fuel retention, closing the coupled carbon–deuterium balance. In particular, it explains why the deuterium concentration in deposits decreases with increasing depth.
    Journal of Nuclear Materials 07/2013; 438(Supplement):S120-S125. · 2.02 Impact Factor
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    ABSTRACT: A new method to measure power flux in strongly magnetized plasmas is described, while the flaws inherent in standard Langmuir probe techniques are demonstrated. Instead of small cylindrical pins which overestimate the ion current density by several 100%, a concave probe has been developed which is immune to sheath expansion, and which inherently provides accurate measurements. A retarding field analyzer directly measures the ion component of the power flux by means of an integral method that eliminates the need to calculate the heat transmission factor. Evidence shows that strong secondary electron emission from surfaces with non-oblique magnetic field incidence angles is ubiquitous in the scrape-off layer of the Tore Supra tokamak. This results in sheath collapse, causing the power flux to be dominated by the electrons. The radially integrated power flux measured by the probes agrees well with the power convected to the limiter.
    Journal of Nuclear Materials 07/2013; 438:S184–S188. · 2.02 Impact Factor
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    ABSTRACT: The ion velocity distribution parallel to magnetic field lines, f is extracted from retarding field analyzer (RFA) measurements in the ASDEX Upgrade scrape-off layer. The RFA ion current-voltage characteristic is transformed into a system of linear equations, from which f is unfolded using a standard regularization method. The algorithm is validated on numerically generated data. The experimentally measured f compares favorably with the ion velocity distributions calculated from a quasi-neutral kinetic simulation of the plasma pre-sheath. The technique can be used to measure f by RFAs in plasma processing, high-energy particle accelerators or by satellite-born RFAs.
    Journal of Nuclear Materials 07/2013; · 2.02 Impact Factor
  • J P Gunn, V Fuchs, M Kočan
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    ABSTRACT: Interpretation of retarding field analyser (RFA) measurements in flowing tokamak scrape-off layer plasmas requires a model of the perturbation introduced by the wake, or presheath, of the instrument. We extend a collisionless kinetic Mach probe theory to include a simple model of ion–ion collisions. Collisions lead to significant modifications in the downstream presheath. Depending on the ion-to-electron temperature ratio, flow speed and collisionality, the measurement obtained by averaging the apparent temperatures on each side of a bi-directional RFA is predicted to overestimate the unperturbed ion temperature, but with errors not exceeding 50% most of the time. The error increases with Mach number, and decreases with the ion-to-electron temperature ratio.
    Plasma Physics and Controlled Fusion 03/2013; 55(4):045012. · 2.37 Impact Factor
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    ABSTRACT: Tritium (T) retention constitutes an outstanding constraint for ITER. It has been proposed that the end of the discharge could be used for reducing the amount of tritium trapped in the device by switching to He or H2 injection during the ~200 s of plasmas following the burning phase (power and plasma current ramp down). Thanks to the long discharge capabilities of Tore Supra, long pulse experiments (> mn) have been carried out to evaluate the effectiveness of such a scenario in reducing the tritium inventory during plasma operations. Starting with the device operated only in D2, series of changeover experiments from D2 to He and from D2 to H2 have been carried out in Tore Supra. The results demonstrate that with He the amount of D recovered after 130 s is limited to 0.8 × 1022 D whilst no further gain is foreseen. From these experiments, it is demonstrated that He injection will not contribute to the drop of the tritium inventory in the vessel. In contrast, with H2 injection the amount of D recovered after 250 s is ~4.2 × 1022 D with no limitation observed in the amount that could be removed from the vessel. The higher efficiency in removing D from the vessel by H2 injection compared to He is attributed to the H charge-exchange (CX) flux (four to six times larger than the He CX flux) allowing for a significantly stronger plasma wall interaction with carbon deposition and layer areas. In Tore Supra, since most of the D retention through co-deposition with eroded material (C) takes place in these areas, H plasmas result in a better removal efficiency of D(T) from these regions. These experimental observations are supported by the results obtained using the EIRENE code for evaluating both the ion and CX fluxes for He and H plasmas. Finally, the consequences of removing D(T) from the vessel for the next discharges are unfavourable for both the He and H2 removal methods. Indeed, in both cases, twice the amount of D(T) removed through the isotope exchange has to be re-injected since co-deposition of the re-injected D(T) will also take place in addition to the plasma wall isotope exchange. In these conditions, the low efficiency of the H2 gas injection for controlling the plasma isotopic ratio inhibits a recovery of the initial plasma isotopic ratio over a time scale in the range of 200 s.
    Nuclear Fusion 02/2013; 53(3):033003. · 2.73 Impact Factor
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    ABSTRACT: This paper presents the first three-dimensional (3D) particle-in-cell (PIC) simulations of castellated plasma-facing components (PFCs) in tokamaks. Special focus is given to crossings between poloidal and toroidal gaps where elevated heat loads are expected to occur. Moreover, the crossings may affect the plasma penetration into the gaps between tiles. Both of these problems are of high importance for ITER when estimating the lifetime of its PFCs. Localized heat loads can potentially lead to damage of the tiles, while the plasma penetration is related to fuel retention in the gaps due to redeposition of eroded wall material. This problem has previously been targeted by 2D PIC simulations using our in-house code SPICE2, where toroidal and poloidal gaps (PGs) had to be simulated separately. This paper presents the results of a full 3D3V code SPICE3, which allows us to simulate a more realistic geometry of the tiles including the gap crossings and includes the complete E × B drift, which could not be simulated in 2D. The results of self-consistent simulations show that the crossing acts as a transport channel for electrons, allowing them to enter the plasma shadowed region in PGs. As a consequence, the potential near the gap entrance is modified allowing more ions to flow deep inside the gap. The combination of the plasma flow and an E × B drift in the crossing directs ions onto one tile corner, which receives elevated heat load.
    Plasma Physics and Controlled Fusion 01/2013; 55(2):025006. · 2.37 Impact Factor
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    ABSTRACT: Since spring 2011, one of the three ion cyclotron reconance heating (ICRH) antennas in the Tore Supra (TS) tokamak is equipped with a new type of Faraday screen (FS). Results from Radio Frequency (RF) simulations of the new Faraday screen suggest the innovative structure with cantilevered bars and ‘shark tooth’ openings significantly changes the current flow pattern on the front of the antenna which in turn reduces the RF potential and RF electrical field in particular parallel to the magnetic field lines which contributes to generating RF sheaths. Effects of the new FS operation on RF-induced scrape-off layer (SOL) modifications are studied for different plasma and antenna configurations — scans of strap power ratio imbalance, phasing, injected power and SOL density.
    20th International Conference on Nuclear Engineering and the ASME 2012 Power Conference, Anaheim, California, USA; 07/2012
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    ABSTRACT: Particle balance is investigated using a Mach probe at the top of the scrape-off layer of circular ohmically heated L-mode plasmas in the Tore Supra tokamak [G. Giruzzi etal., Nucl. Fusion 49, 104010 (2009)]. Contributions from both poloidal EXB flows and ionization sources are found to be small. As a result the local parallel flow is a response of the radial flux distribution between the two strike points of open field lines, and the density profile is determined by the field-line-integrated radial flux. By scanning the poloidal position of the strike point on a secondary limiter situated at the outboard midplane, an indirect poloidal mapping of the radial flux distribution is obtained. The radial flux is centered at the outboard midplane and is relatively well described by a Gaussian distribution of half poloidal width of about 50 Degree-Sign at the last closed flux surface, decaying to about 30 Degree-Sign in the far scrape-off layer. The turbulent radial flux measured locally with a rake probe shows a reasonable agreement with the poloidal mapping obtained by the Mach probe. It is shown than the radial convective velocity decays along radius at the plasma top but should increase with radius at the outboard midplane.
    Physics of Plasmas 07/2012; 19(7). · 2.38 Impact Factor
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    ABSTRACT: New measurements of ion energies and currents in type I and mitigated ELMs have been carried out in the ASDEX Upgrade far scrape-off layer using a retarding field analyser (RFA). The ion temperature averaged over an ELM, Ti ELM measured 35–60 mm outside the separatrix (i.e. 15–25 mm in front of the outboard limiter) is in the range 20–200 eV, which is 5–50% of the ion temperature at the pedestal top. Ti ELM decreases with the separatrix distance with the e-folding length of ~10 mm measured in the far SOL for a particular set of conditions, and increases with the ELM energy WELM. Lowest Ti ELM is measured during mitigated type I ELMs. Likewise, the ELM-averaged ion current e-folding length increases with WELM, similar to the e-folding length of the heat flux density at the RFA probe head during an ELM, monitored by a fast IR camera. The most plausible explanation of observed trends is that on average the filaments of larger ELMs travel faster radially and have less time to dilute by parallel losses along field lines before reaching the far SOL. These observations provide further evidence that the fraction of the ELM energy deposited on the main chamber plasma-facing components increases with WELM.
    Nuclear Fusion 01/2012; 52(2):023016. · 2.73 Impact Factor
  • Antennas and Propagation (EUCAP), 2012 6th European Conference on; 01/2012

Publication Stats

536 Citations
164.83 Total Impact Points

Institutions

  • 2009–2014
    • Cea Leti
      Grenoble, Rhône-Alpes, France
  • 2013
    • Max Planck Institute for Plasma Physics
      • Max Planck Institute for Plasma Physics, Greifswald
      Garching bei München, Bavaria, Germany
  • 2012
    • Atomic Energy and Alternative Energies Commission
      Fontenay, Île-de-France, France
  • 2010
    • Acadia University
      Wolfville, Nova Scotia, Canada
    • The Police Academy of the Czech Republic in Prague
      Praha, Praha, Czech Republic