H. Reimerdes

École Polytechnique Fédérale de Lausanne, Lausanne, Vaud, Switzerland

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Publications (254)324.83 Total impact

  • G.P. Canal · T. Lunt · H. Reimerdes · B.P. Duval · B. Labit · W.A.J. Vijvers ·

    Nuclear Fusion 11/2015; 55(12):123023. DOI:10.1088/0029-5515/55/12/123023 · 3.06 Impact Factor
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    ABSTRACT: Since the 2012 IAEA-FEC Conference, FTU operations have been largely devoted to runaway electrons generation and control, to the exploitation of the 140 GHz electron cyclotron (EC) system and to liquid metal limiter elements. Experiments on runaway electrons have shown that the measured threshold electric field for their generation is larger than predicted by collisional theory and can be justified considering synchrotron radiation losses. A new runaway electrons control algorithm was developed and tested in presence of a runaway current plateau, allowing to minimize the interactions with plasma facing components and safely shut down the discharges. The experimental sessions with 140 GHz EC system have been mainly devoted to experiments on real-time control of magnetohydrodynamic (MHD) instabilities using the new EC launcher with fast steering capability. Experiments with central EC injection have shown the onset of 3/2 and 2/1 tearing modes, while EC assisted breakdown experiments have been focused on ITER start-up issues, exploring the polarization conversion at reflection from inner wall and the capability to assure plasma start-up even in presence of a large stray magnetic field. A new actively cooled lithium limiter has been installed and tested. The lithium limiter was inserted in the scrape-off layer, without any damage to the limiter surface. First elongated FTU plasmas with EC additional heating were obtained with the new cooled limiter. Density peaking and controlled MHD activity driven by neon injection were investigated at different plasma parameters. A full real-time algorithm for disruption prediction, based on MHD activity signals from Mirnov coils, was developed exploiting a large database of disruptions. Reciprocating Langmuir probes were used to measure the heat flux e-folding length in the scrape-off layer, with the plasma kept to lay on thea internal limiter to resemble the ITER start-up phase. New diagnostics were successfully installed and tested, as a diamond probe to detect Cherenkov radiation produced by fast electrons and a gamma camera for runaway electrons studies. Laser induced breakdown spectroscopy measurements were performed under vacuum and with toroidal magnetic field, so demonstrating their capability to provide useful information on the surface elemental composition and fuel retention in present and future tokamaks, such as ITER.
    Nuclear Fusion 10/2015; 55(10). DOI:10.1088/0029-5515/55/10/104005 · 3.06 Impact Factor
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    ABSTRACT: Experiments in the TCV tokamak show that high power central electron cyclotron heating (ECH) and current drive (ECCD) produce significant direct modification of the plasma rotation profile, as well as an effect on the equilibrium current density profile. In a regime of unsteady rotation, these effects contribute to the onset of neoclassical tearing instabilities, in the absence of triggers such as sawteeth, edge localised modes (ELMS) or relevant 'error' fields. In turn the growing tearing modes' breaking axisymmetry provides a nonlinear magnetic torque which converts the power absorption in a co-directed rotation with a flattening of the profile at the rational surfaces. The experimental results are presented and discussed in the context of theoretical models of neoclassical toroidal viscosity and ion inertial effects.
    Nuclear Fusion 09/2015; 55(9):093031. DOI:10.1088/0029-5515/55/9/093031 · 3.06 Impact Factor
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    ABSTRACT: Recent TCV experiments have examined the effect of the poloidal field strength in the vicinity of the x-point of diverted configurations on their ability to radiate a large fraction of the exhaust power. A larger region of low poloidal field is a key characteristic of the “snowflake” configuration, which has been proposed as an alternative divertor solution that decreases the power flux to the targets in a DEMO-size tokamak. In the investigated Ohmic discharges, increasing the plasma density and seeding neon both increased the radiated exhaust fraction up to 60–70%. In all cases, the highest radiation fraction was determined by the onset of MHD rather than a radiation instability. The experiments indicate that, while the conventional single-null configuration leads to more radiation (+10%) at higher densities, the snowflake configuration radiates more when seeding neon impurities (+15%). Extrapolation of these modest, but systematic, dependencies on the divertor geometry to reactor-relevant higher heating power and larger device size must be based on a physics model.
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    ABSTRACT: Edge intrinsic rotation was investigated in Ohmic L-mode discharges on the Tokamak à Configuration Variable, scanning the major radial position of the X point, R_{X}. Edge rotation decreased linearly with increasing R_{X}, vanishing or becoming countercurrent for an outboard X point, in agreement with theoretical expectations. The core rotation profile shifted fairly rigidly with the edge rotation, changing the central rotation speed by more than a factor of two. Core rotation reversals had little effect on the edge rotation velocity. Edge rotation was modestly more countercurrent in unfavorable than favorable ∇B shots.
    Physical Review Letters 06/2015; 114(24):245001. DOI:10.1103/PhysRevLett.114.245001 · 7.51 Impact Factor
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    ABSTRACT: Horizon 2020 is the largest EU Research and Innovation programme to date. The European fusion research programme for Horizon 2020 is outlined in the “Roadmap to the realization of fusion energy” and published in 2012 [1]. As part of it, the European Fusion Consortium (EUROfusion) has been established and will be responsible for implementing this roadmap through its members. The European fusion roadmap sets out a strategy for a collaboration to achieve the goal of generating fusion electricity by 2050. It is based on a goal-oriented approach with eight different missions including the development of heat-exhaust systems which must be capable of withstanding the large heat and particle fluxes of a fusion power plant (FPP). A summary of the main aims of the mission for a solution on heat-exhaust systems and the EUROfusion consortium strategy to set up an efficient Work Breakdown Structure and the collaborative efforts to address these challenges will be presented.
    Fusion Engineering and Design 05/2015; file:///C|/Users/turnyam.EFDA/Desktop/dx.doi.org/10.1016/j.fusengdes.2015.04.041. DOI:10.1016/j.fusengdes.2015.04.041 · 1.15 Impact Factor
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    ABSTRACT: Recent theoretical work predicts intrinsic toroidal rotation in the tokamak edge to depend strongly on the normalized major radial position of the X-point. With this motivation, we conducted a series of Ohmic L-mode shots on the Tokamak à Configuration Variable, moving the X-point from the inboard to the outboard edge of the last closed flux surface in both lower and upper single null configurations. The edge toroidal rotation evolved from strongly co-current for an inboard X-point to either vanishing or counter-current for an outboard X-point, in agreement with the theoretical expectations. The whole rotation profile shifted roughly rigidly with the edge rotation, resulting in variation of the peak core rotation by more than a factor of two. Core rotation reversals had little effect on the edge rotation. Edge rotation was slightly more counter-current for unfavorable than favorable ∇ B drift discharges.
    Physics of Plasmas 05/2015; 22(5):056118. DOI:10.1063/1.4921158 · 2.14 Impact Factor
  • J.-M. Moret · B. P. Duval · H. B. Le · S. Coda · F. Felici · H. Reimerdes ·
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    ABSTRACT: Equilibrium reconstruction consists in identifying, from experimental measurements, a distribution of the plasma current density that satisfies the pressure balance constraint. The LIUQE code adopts a computationally efficient method to solve this problem, based on an iterative solution of the Poisson equation coupled with a linear parametrisation of the plasma current density. This algorithm is unstable against vertical gross motion of the plasma column for elongated shapes and its application to highly shaped plasmas on TCV requires a particular treatment of this instability. TCV's continuous vacuum vessel has a low resistance designed to enhance passive stabilisation of the vertical position. The eddy currents in the vacuum vessel have a sizeable influence on the equilibrium reconstruction and must be taken into account. A real time version of LIUQE has been implemented on TCV's distributed digital control system with a cycle time shorter than 200 μs for a full spatial grid of 28 by 65, using all 133 experimental measurements and including the flux surface average of quantities necessary for the real time solution of 1.5 D transport equations. This performance was achieved through a thoughtful choice of numerical methods and code optimisation techniques at every step of the algorithm, and was coded in Matlab and Simulink for the off-line and real time version respectively.
    Fusion Engineering and Design 01/2015; 91. DOI:10.1016/j.fusengdes.2014.09.019 · 1.15 Impact Factor
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    R. Ambrosino · R. Albanese · S. Coda · M. Mattei · J.-M. Moret · H. Reimerdes ·
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    ABSTRACT: The design of a snowflake (SF) equilibrium requires a strong effort on the poloidal field (PF) currents in terms of MAturns and mechanical loads. This has limited the maximum plasma current in SF configurations on Tokamak à Configuration Variable (TCV) to values well below the intrinsic magnetohydrodynamic limits. In this paper the definition of optimized SF configurations in TCV and their experimental tests are illustrated. The PF current optimization procedure proposed in Albanese et al (2014 Plasma Phys. Control. Fusion 56 035008) is adapted and applied to a SF scenario in TCV where the PF currents were close to their operational limits with the aim of reducing the total MAturns in view of higher values of the plasma current. This procedure optimizes the PF currents while fulfilling the machine technological constraints for a given bound on the tolerable plasma shape changes. The method exploits the linearized relation between the plasma–wall gaps and the PF currents. In the investigated TCV scenario the optimization procedure allowed a 20% increase of the plasma current while keeping the plasma shape alignment with respect to the nominal shape within a tolerance of 1 cm. The predicted optimization potential was confirmed in a TCV experiment.
    Nuclear Fusion 11/2014; 54(12):123008. DOI:10.1088/0029-5515/54/12/123008 · 3.06 Impact Factor
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    ABSTRACT: The nonlinear response of a low-beta tokamak plasma to non-axisymmetric fields offers an alternative to direct measurement of the non-axisymmetric part of the vacuum magnetic fields, often termed 'error fields'. Possible approaches are discussed for determination of error fields and the required current in non-axisymmetric correction coils, with an emphasis on two relatively new methods: measurement of the torque balance on a saturated magnetic island, and measurement of the braking of plasma rotation in the absence of an island. The former is well suited to ohmically heated discharges, while the latter is more appropriate for discharges with a modest amount of neutral beam heating to drive rotation. Both can potentially provide continuous measurements during a discharge, subject to the limitation of a minimum averaging time. The applicability of these methods to ITER is discussed, and an estimate is made of their uncertainties in light of the specifications of ITER's diagnostic systems. The use of plasma response-based techniques in normal ITER operational scenarios may allow identification of the error field contributions by individual central solenoid coils, but identification of the individual contributions by the outer poloidal field coils or other sources is less likely to be feasible.
    Nuclear Fusion 04/2014; 54(7):073004. DOI:10.1088/0029-5515/54/7/073004 · 3.06 Impact Factor
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    ABSTRACT: First time experimental evidence is presented for a direct link between the decay of a n = 3 plasma response and the formation of a three-dimensional (3D) plasma boundary. We inspect a lower single-null L-mode plasma which first reacts at sufficiently high rotation with an ideal resonant screening response to an external toroidal mode number n = 3 resonant magnetic perturbation field. Decay of this response due to reduced bulk plasma rotation changes the plasma state considerably. Signatures such as density pump out and a spin up of the edge rotation-which are usually connected to formation of a stochastic boundary-are detected. Coincident, striation of the divertor single ionized carbon emission and a 3D emission structure in double ionized carbon at the separatrix is seen. The striated C II pattern follows in this stage the perturbed magnetic footprint modelled without a plasma response (vacuum approach). This provides for the first time substantial experimental evidence, that a 3D plasma boundary with direct impact on the divertor particle flux pattern is formed as soon as the internal plasma response decays. The resulting divertor structure follows the vacuum modelled magnetic field topology. However, the inward extension of the perturbed boundary layer can still not directly be determined from these measurements.
    Nuclear Fusion 03/2014; 54(1):012001. DOI:10.1088/0029-5515/54/1/012001 · 3.06 Impact Factor
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    ABSTRACT: Real-time control of multiple plasma actuators is a requirement in advanced tokamaks; for example for burn control, plasma current profile control and MHD stabilization - EC wave absorption is ideally suited especially for the latter. For example, on ITER 24 EC sources can be switched between 54 inputs at the torus. In the torus, 5 launchers direct the power to various locations across the plasma profile via 11 steerable mirrors. For optimal usage of the available power, the aiming and polarization of the beams must be adapted to the plasma configuration and the needs of the scenario. Since the EC system performs many competing tasks, present day systems should demonstrate the ability of an EC plant to deal with several targets in parallel and/or to switch smoothly between goals to attain overall satisfaction. Recently TCV has taken a first step towards such a demonstration. Several EC launchers are used simultaneously to regulate the sawtooth period and to preempt m/n = 3/2 NTMs, by controlling the power levels. In parallel, a second algorithm stabilizes any NTM that saturates [1]. These and real-time MHD control experiments on ELMs [2] are presented.
    20th Topical Conference on Radio Frequency Power in Plasmas; 02/2014
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    ABSTRACT: One of the approaches to solve the heat load problem in a divertor tokamak is the so called 'snowflake' (SF) configuration, a magnetic equilibrium with two nearby x-points and two additional divertor legs. Here we report on the first EMC3-Eirene simulations of plasma- and neutral particle transport in the scrape-off layer of a series of TCV SF equilibria with different values of σ, the distance between the x-points normalized to the minor radius of the plasma. The constant cross-field transport coefficients were chosen such that the power- and particle deposition profiles at the primary inner strike point (SP) match the Langmuir probe measurements for the σ = 0.1 case. At the secondary SP on the floor, however, a significantly larger power flux than that predicted by the simulation was measured by the probes, indicating an enhanced transport across the primary separatrix. As the ideal SF configuration (σ = 0) is approached, the density as well as the radiation maximum are predicted to move from the target plates upward to the x-point by the simulation.
    Plasma Physics and Controlled Fusion 02/2014; 56(3):035009. DOI:10.1088/0741-3335/56/3/035009 · 2.19 Impact Factor
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    ABSTRACT: The snowflake (SF) divertor is a plasma configuration that may enable tokamak operation at high performance and lower peak heat loads on the plasma-facing components than a standard single-null divertor. This paper reports on the results of experiments performed on the TCV tokamak in both the low- and high-confinement regimes, wherein the divertor configuration was continuously varied between a standard single-null and a 'SF-plus', which features auxiliary strike points (SPs) in the private flux region of the primary separatrix. The measured edge properties show that, in L-mode, the fraction of the exhaust power reaching the additional SPs is small. During edge-localized modes, up to similar to 20% of the exhausted energy is redistributed to the additional SPs even at an x-point separation of 0.6 times the plasma minor radius, thereby reducing the peak heat flux to the inner primary SP by a factor of 2-3. The observed behaviour is qualitatively consistent with a proposed model for enhanced cross-field transport through the SF's relatively large region of low poloidal field by instability-driven convection.
    Nuclear Fusion 01/2014; 54(2):023009. DOI:10.1088/0029-5515/54/2/023009 · 3.06 Impact Factor
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    ABSTRACT: TCV experiments demonstrate the basic power exhaust properties of the snowflake (SF) plus and SF minus divertor configurations by measuring the heat fluxes at each of their four divertor legs. The measurements indicate an enhanced transport into the private flux region and a reduction of peak heat fluxes compared to a similar single null configuration. There are indications that this enhanced transport cannot be explained by the modified field line geometry alone and likely requires an additional or enhanced cross-field transport channel. The measurements, however, do not show a broadening of the scrape-off layer (SOL) and, hence, no increased cross-field transport in the common flux region. The observations are consistent with the spatial limitation of several characteristic SF properties, such as a low poloidal magnetic field in the divertor region and a long connection length to the inner part of the SOL closest to the separatrix. Although this limitation is typical in a medium sized tokamak like TCV, it does not apply to significantly larger devices where the SF properties are enhanced across the entire expected extent of the SOL.
    Plasma Physics and Controlled Fusion 12/2013; 55(12):124027-124035. DOI:10.1088/0741-3335/55/12/124027 · 2.19 Impact Factor
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    ABSTRACT: Many tokamaks have observed that sawteeth of sufficient duration may trigger neoclassical tearing modes (NTMs) that lead to plasma performance degradation. In this paper, TCV's ability to accurately control the period of individual sawteeth is exploited, using localized electron cyclotron resonance heating and current drive (ECRH and ECCD), to trigger NTMs under controlled conditions, providing an excellent environment for the study of the seeding of NTMs by sawtooth crashes. The TCV experiments show evidence of a fast formation of seed islands with poloidal/toroidal mode numbers m/n = 3/2 and 2/1 within a few tens of microseconds following the sawtooth crash. Crashes of sawteeth with a longer period duration are observed to generate larger seed islands but also increase the plasma stability to conventional tearing modes. While these two effects compete, the NTM stability is found to decrease with increasing sawtooth period. The seed island size can be reduced and, thereby, the NTM stability improved, by increasing the value of the safety factor q(95). Alternatively, NTM stability can be increased by application of preemptive ECRH at the resonant surface of the NTM. Preemptive ECRH is found to enlarge the plasma operational domain by improving the conventional tearing stability and by reducing the coupling between the driving (m/n = 1/1 or 2/2) and the driven modes (m/n = 2/1 or 3/2), resulting in smaller sawtooth generated seed islands. The efficiency of preemptive ECRH increases when sufficient ECRH power is applied in a short time interval prior to the sawtooth crash.
    Nuclear Fusion 11/2013; 53(11):113026. DOI:10.1088/0029-5515/53/11/113026 · 3.06 Impact Factor
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    R.P. Wenninger · H. Reimerdes · O. Sauter · H. Zohm ·
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    ABSTRACT: Edge-localized modes (ELMs) are instabilities in the edge of tokamak plasmas in the high confinement regime (H-mode). Despite beneficial aspects of ELMs, in a future device the size of the energy loss per ELM must be controlled, in order to avoid intolerable divertor power flux densities. To proceed in understanding how the ELM size is determined and how ELM mitigation methods work it is necessary to characterize the non-linear evolution of ELMs. This publication presents a detailed analysis of the toroidal structure of dominant magnetic perturbations during type-I ELMs in TCV. These signatures of the instability can be observed most intensely in close temporal vicinity to the onset of enhanced D-alpha-radiation. In particular it is shown that dominant magnetic perturbations already have a rigid toroidal mode structure when they are detected with magnetic probes. This indicates that perturbations associated with this type of ELM at TCV cannot be observed in their linear phase. Furthermore it is demonstrated that the toroidal structure of dominant magnetic perturbations is most often dominated by the n = 1 component. This is in clear contrast to typical results of linear stability calculations, leading to the hypothesis that the dominant toroidal mode number from the linear to the non-linear phase has a transition from intermediate to low values. In general, the reported results show that non-linear coupling leads to a significant modification of the mode structure.
    Nuclear Fusion 09/2013; 53(11):113004. DOI:10.1088/0029-5515/53/11/113004 · 3.06 Impact Factor
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    ABSTRACT: Experimental studies of the novel snowflake divertor concept (D. Ryutov, Phys. Plasmas 14 (2007) 064502) performed in the NSTX and TCV tokamaks are reviewed in this paper. The snowflake divertor enables power sharing between divertor strike points, as well as the divertor plasma-wetted area, effective connection length and divertor volumetric power loss to increase beyond those in the standard divertor, potentially reducing heat flux and plasma temperature at the target. It also enables higher magnetic shear inside the separatrix, potentially affecting pedestal MHD stability. Experimental results from NSTX and TCV confirm the predicted properties of the snowflake divertor. In the NSTX, a large spherical tokamak with a compact divertor and lithium-coated graphite plasma-facing components (PFCs), the snowflake divertor operation led to reduced core and pedestal impurity concentration, as well as re-appearance of Type I ELMs that were suppressed in standard divertor H-mode discharges. In the divertor, an otherwise inaccessible partial detachment of the outer strike point with an up to 50% increase in divertor radiation and a peak divertor heat flux reduction from 3–7 MW/m2 to 0.5–1 MW/m2 was achieved. Impulsive heat fluxes due to Type-I ELMs were significantly dissipated in the high magnetic flux expansion region. In the TCV, a medium-size tokamak with graphite PFCs, several advantageous snowflake divertor features (cf. the standard divertor) have been demonstrated: an unchanged L–H power threshold, enhanced stability of the peeling–ballooning modes in the pedestal region (and generally an extended second stability region), as well as an H-mode pedestal regime with reduced (×2–3) Type I ELM frequency and slightly increased (20–30%) normalized ELM energy, resulting in a favorable average energy loss comparison to the standard divertor. In the divertor, ELM power partitioning between snowflake divertor strike points was demonstrated. The NSTX and TCV experiments are providing support for the snowflake divertor as a viable solution for the outstanding tokamak plasma–material interface issues.
    Journal of Nuclear Materials 07/2013; 438:S96–S101. DOI:10.1016/j.jnucmat.2013.01.015 · 1.87 Impact Factor
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    ABSTRACT: Tokamaks with weak to moderate reversed central shear in which the minimum inverse rotational transform (safety factor) qmin is in the neighbourhood of unity can trigger bifurcated magnetohydrodynamic equilibrium states, one of which is similar to a saturated ideal internal kink mode. Peaked prescribed pressure profiles reproduce the ‘snake’ structures observed in many tokamaks which has led to a novel explanation of the snake as a bifurcated equilibrium state. Snake equilibrium structures are computed in simulations of the tokamak `a configuration variable (TCV), DIII-D and mega amp spherical torus (MAST) tokamaks. The internal helical deformations only weakly modulate the plasma–vacuum interface which is more sensitive to ripple and resonant magnetic perturbations. On the other hand, the external perturbations do not alter the helical core deformation in a significant manner. The confinement of fast particles in MAST simulations deteriorate with the amplitude of the helical core distortion. These three-dimensional bifurcated solutions constitute a paradigm shift that motivates the applications of tools developed for stellarator research in tokamak physics investigations.
    Nuclear Fusion 07/2013; 53(7-7):073021. DOI:10.1088/0029-5515/53/7/073021 · 3.06 Impact Factor
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    ABSTRACT: The primary goal of hybrid scenarios in tokamaks is to enable high performance operation with large plasma currents whilst avoiding MHD instabilities. However, if a local minimum in the safety factor is allowed to approach unity, the energy required to overcome stabilizing magnetic field line bending is very small, and as a consequence, large MHD structures can be created, with typically dominant m = n = 1 helical component. If there is no exact q = 1 rational surface the essential character of these modes can be modelled assuming ideal nested magnetic flux surfaces. The methods used to characterize these structures include linear and non-linear ideal MHD stability calculations which evaluate the departure from an axisymmetric plasma state, and also equilibrium calculations using a 3D equilibrium code. While these approaches agree favourably for simulations of ITER relevant hybrid regimes in this paper, the relevance of the ideal MHD model itself is tested through empirical examination of helical states in MAST and TCV. While long lived modes in MAST do not have island structures, some of the continuous mode oscillations exhibited in high elongation experiments in TCV indicate that resistivity may play a role in further weakening the ability of the tokamak core to remain axisymmetric. The simulations and experiments consistently highlight the need to control the safety factor in hybrid scenarios planned for future fusion grade tokamaks such as ITER.
    Plasma Physics and Controlled Fusion 01/2013; 55(1):014005. DOI:10.1088/0741-3335/55/1/014005 · 2.19 Impact Factor

Publication Stats

3k Citations
324.83 Total Impact Points


  • 1998-2015
    • École Polytechnique Fédérale de Lausanne
      • Center for Research In Plasma Physics
      Lausanne, Vaud, Switzerland
    • Max Planck Institute for Plasma Physics
      • Max Planck Institute for Plasma Physics, Greifswald
      Arching, Bavaria, Germany
    • Academy of Sciences of the Czech Republic
      • Ústav fyziky plazmatu
      Praha, Hlavni mesto Praha, Czech Republic
  • 2002-2014
    • Columbia University
      • Department of Applied Physics and Applied Mathematics
      New York, New York, United States
  • 2011
    • Salt Lake City Community College
      Salt Lake City, Utah, United States
  • 2002-2008
    • General Atomics
      San Diego, California, United States
  • 2000
    • Kurchatov Institute
      Moskva, Moscow, Russia