J. P. Gunn

Cea Leti, Grenoble, Rhône-Alpes, France

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Publications (131)201.8 Total impact

  • [Show abstract] [Hide abstract]
    ABSTRACT: The WEST project (W-Environment in Steady-State Tokamak) aims to transform the Tore Supra limiter configuration to an x-point divertor, providing a test bed for ITER-like plasma-facing components (actively cooled W monoblocs) under high heat flux, steady-state plasma irradiation. The lower divertor includes an actively cooled, W-coated CuCrZr baffle to provide neutral compression and improve particle exhaust. As part of the new diagnostic equipment of Tore Supra within WEST project, a set of Langmuir probes will find place on the baffle in order to provide plasma flux and electron temperature measurements for physics studies and real-time machine protection functions during steady-state discharges. On the baffle top surface, irradiation coming from the plasma, energetic ripple-ions losses, photons and energetic neutrals from charge exchange reactions produce power fluxes up to 3 MW/m2, representing a challenge for the Langmuir probes operating conditions. In this paper Copper–Chrome–Zirconium (CuCrZr) cylindrical probe concept design is proposed. Finite element thermo-mechanical analysis (FEA) confirmed the consistency of this solution under the steady-state plasma condition in the worst case (highest thermal load).
    Fusion Engineering and Design 03/2015; 93. DOI:10.1016/j.fusengdes.2015.02.009 · 1.15 Impact Factor
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    ABSTRACT: Higher than expected electron temperatures (Te) are often measured by divertor Langmuir probes (LP) in high recycling and detached regimes in JET and other tokamaks. As a possible mechanism to explain this discrepancy, we investigate the effect of penetration of fast, almost collisionless electrons connecting the hot upstream scrape-off layer (SOL) region to the divertor targets in JET. We simulate the electron velocity distribution function (EVDF) near the divertor targets using a simple 1D kinetic model using parallel SOL profiles from EDGE2D-EIRENE simulations. The resulting EVDF is used to construct synthetic LP current–voltage (IV) characteristics and evaluation of Te is performed in the same way as for experimental data. Results indicate that the process does not explain the anomalously high Te values estimated from the target probe measurements if the EDGE2D-EIRENE simulated parallel profiles are a good representation of reality.
    Journal of Nuclear Materials 01/2015; 463. DOI:10.1016/j.jnucmat.2015.01.051 · 2.02 Impact Factor
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    ABSTRACT: A series of experiments has been performed on JET to investigate the dynamics of transient melting due to edge localized modes (ELMs). The experiment employs a deliberately misaligned lamella in one module of the JET bulk tungsten outer divertor, allowing the combination of stationary power flux and ELMs to transiently melt the misaligned edge. During the design of the experiment a number of calculations were performed using 2D particle-in-cell simulations and a heat transfer code to investigate the influence on the deposited power flux of finite Larmor radius effects associated with the energetic ELM ions. This has been performed using parameter scans inside a range of pedestal temperatures and densities to scope different experimentally expected ELM energies. On the one hand, we observe optimistic results, with smoothing of the heat flux due to the Larmor gyration on the protruding side of the lamella which sees the direct parallel flux—the deposited power tends to be lower than the nominal value expected from geometric magnetic field line impact over a distance smaller than 2 Larmor radii, a finding which is always valid during ELMs for such a geometry. On the other hand, the fraction of the flux not reaching the directly wetted side is transferred and spread to the top surface of the lamella. The hottest point of the lamella (corner side/top) does not always benefit from the gain from the Larmor smoothing effect because of an enhanced power deposition from the second contribution.
    Nuclear Fusion 12/2014; 54(12). DOI:10.1088/0029-5515/54/12/123011 · 3.24 Impact Factor
  • M. Kočan · R.A. Pitts · S.W. Lisgo · A. Loarte · J.P. Gunn · V. Fuchs
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    ABSTRACT: The interaction of ELM filaments with the ITER beryllium first wall panels (FWPs) is studied using a simple ad-hoc fluid model of the filament parallel transport, taking into account the full, three-dimensional structure of the FWPs, including magnetic shadowing effects. The calculated ELM surface heat loads are used as input to the RACLETTE heat transfer code to estimate the FWP surface temperature rise. The results indicate that controlled ELMs in ITER during burning plasma operation (ΔWELM ≈ 0.6 M J) will not lead to melting or significant evaporation of the beryllium surfaces, even in the case of high ELM broadening and the minimum allowable distance between the primary and secondary separatrices. The ELM-averaged steady-state heat load also stays below the maximum power handling capability of the FWPs.
    Journal of Nuclear Materials 12/2014; 463. DOI:10.1016/j.jnucmat.2014.11.130 · 2.02 Impact Factor
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    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. DOI:10.1088/0029-5515/54/12/123006 · 3.24 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). DOI:10.1088/0029-5515/54/9/093012 · 3.24 Impact Factor
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    ABSTRACT: During the 2011 experimental campaign, one of the three ion cyclotron resonance heating (ICRH) antennas in the Tore Supra (TS) tokamak was equipped with a new type of Faraday screen (FS). The new design aimed at minimizing RF sheaths as well as increasing the heat exhaust capability of the actively cooled screen. It proved to be inefficient for attenuating the RF-sheaths on the screen itself on the contrary to the heat exhaust concept that allowed operation despite higher heat fluxes on the antenna. In parallel, a new approach has been proposed to model self-consistently RF sheaths: the SSWICH (Self-consistent Sheaths and Waves for IC Heating) code. Simulations results from SSWICH coupled with the TOPICA antenna code were able to reproduce the difference between the two FS designs and part of the spatial pattern of heat loads and floating potential. The poloidal pattern is a reliable result that mainly depends on the electrical design of the antenna while the radial pattern is on the contrary highly sensitive to loosely constrained parameters such as perpendicular conductivity that generates a DC current circulation from the the private region inside the antenna limiters to the free SOL outside these limiters. Moreover the cantilevered bars seem to be the element in the design of the new screen that enhanced RF sheaths.
    Physics of Plasmas 01/2014; 1580(1). DOI:10.1063/1.4864507 · 2.25 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. DOI:10.1088/0029-5515/54/1/013013 · 3.24 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-. DOI:10.1088/0029-5515/53/12/122001 · 3.24 Impact Factor
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    ABSTRACT: In parallel to the direct contribution to the procurement phase of ITER and Broader Approach, CEA has initiated research & development programmes, accompanied by experiments together with a significant modelling effort, aimed at ensuring robust operation, plasma performance, as well as mitigating the risks of the procurement phase. This overview reports the latest progress in both fusion science and technology including many areas, namely the mitigation of superconducting magnet quenches, disruption-generated runaway electrons, edge-localized modes (ELMs), the development of imaging surveillance, and heating and current drive systems for steady-state operation. The WEST (W Environment for Steady-state Tokamaks) project, turning Tore Supra into an actively cooled W-divertor platform open to the ITER partners and industries, is presented.
    Nuclear Fusion 10/2013; 53(10):104023. DOI:10.1088/0029-5515/53/10/104023 · 3.24 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: Comprehensive characterization (experiments and modelling) of a new Faraday screen concept tested on the Tore Supra antenna. A new model is developed for calculating the ICRH sheath rectification at the antenna vicinity. The model is applied to calculate the local heat flux on Tore Supra and ITER ICRH antennas. Full-wave modelling of ITER ICRH heating and current drive scenarios with the EVE code. With 20 MW of power, a current of +/- 400 kA could be driven on axis in the DT scenario. Comparison between DT and DT(He-3) scenario is given for heating and current drive efficiencies. First operation of CW test-bed facility, TITAN, designed for ITER ICRH components testing and could host up to a quarter of an ITER antenna. R&D of high permittivity materials to improve load of test facilities to better simulate ITER plasma antenna loading conditions.
    Nuclear Fusion 07/2013; 53(8):083012. DOI:10.1088/0029-5515/53/8/083012 · 3.24 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; DOI:10.1016/j.jnucmat.2013.01.105 · 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; DOI:10.1016/j.jnucmat.2013.01.103 · 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. DOI:10.1016/j.jnucmat.2013.01.055 · 2.02 Impact Factor
  • M Kočan · J P Gunn · T Lunt · O Meyer · J-Y Pascal
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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. DOI:10.1063/1.4812341 · 1.58 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. DOI:10.1016/j.jnucmat.2013.01.061 · 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. DOI:10.1016/j.jnucmat.2013.01.019 · 2.02 Impact Factor
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    ABSTRACT: Using a reciprocating Retarding Field Analyser (RFA), Scrape-Off Layer (SOL) modifications were investigated on ASDEX-Upgrade during heating with waves in the Ion Cyclotron Range of Frequencies (ICRF), suspected for enhanced impurity production in this all-metal machine. Two quantities involved in the sputtering were measured: the current I-slit on a saturated slit plate, proportional to the parallel ion flux and the mean parallel energy (t) of collected ions, averaged over many RF cycles. Combining multiple RFA reciprocations over a scan of q(95) provided 2D poloidal/radial resolution. In the outer SOL a localized RF-perturbed zone was evidenced on the RFA side magnetically connected to an active ICRF antenna. A flat 2D I-slit pattern surrounded by steep gradients was observed, correlatively with (t) exceeding 150eV. The centre of the zone is connected radially slightly behind the leading edge of antenna side limiters, with a radial extension up to +/- 2cm. The zone is broadest and (t) is largest near the bottom of the active antenna. This is interpreted as a zone of local plasma biasing via sheath rectification, creating density convection around it. The I-slit pattern is qualitatively consistent with simple considerations about ExB particle convection.
    AIP Conference Proceedings 06/2013; 1580. DOI:10.1063/1.4864537
  • 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. DOI:10.1088/0741-3335/55/4/045012 · 2.39 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. DOI:10.1088/0029-5515/53/3/033003 · 3.24 Impact Factor

Publication Stats

987 Citations
201.80 Total Impact Points


  • 2008–2015
    • Cea Leti
      Grenoble, Rhône-Alpes, France
    • The Police Academy of the Czech Republic in Prague
      Praha, Praha, Czech Republic
  • 2010–2013
    • Institute of Geophysics, China Earthquake Administration
      Peping, Beijing, China
  • 2012
    • Atomic Energy and Alternative Energies Commission
      Fontenay, Île-de-France, France