A. C. C. Sips

European Commission, Bruxelles, Brussels Capital Region, Belgium

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Publications (199)258.65 Total impact

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    ABSTRACT: The former all-carbon wall on JET has been replaced with beryllium in the main torus and tungsten in the divertor to mimic the surface materials envisaged for ITER. Comparisons are presented between Type I H-mode characteristics in each design by examining respective scans over deuterium fuelling and impurity seeding, required to ameliorate exhaust loads both in JET at full capability and in ITER.
    06/2014;
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    ABSTRACT: The eXtreme Shape Controller (XSC) has been originally designed to control the plasma shape at JET during the flat-top phase, when the plasma current has a constant value. During the JET 2012 experimental campaigns, the XSC has been used to improve the shape control during the transient phases of plasma current ramp-up and ramp-down. In order to avoid the saturation of the actuators with these transient phases, a current limit avoidance system has been designed and implemented. This paper presents the experimental results achieved at JET during the 2012 campaigns using the XSC.
    Journal of Fusion Energy 04/2014; 33(2). · 1.00 Impact Factor
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    ABSTRACT: This paper will discuss simulations of the full ionization process (i.e. plasma burn-through), fundamental to creating high temperature plasma. By means of an applied electric field, the gas is partially ionized by the electron avalanche process. In order for the electron temperature to increase, the remaining neutrals need to be fully ionized in the plasma burn-through phase, as radiation is the main contribution to the electron power loss. The radiated power loss can be significantly affected by impurities resulting from interaction with the plasma facing components. The DYON code is a plasma burn-through simulator developed at Joint European Torus (JET) [1] [2]. The dynamic evolution of the plasma temperature and plasma densities including impurity content is calculated in a self-consistent way, using plasma wall interaction models. The recent installation of a beryllium wall at JET enabled validation of the plasma burn-through model in the presence of new, metallic plasma facing components. The simulation results of the plasma burn-through phase show consistent good agreement against experiments at JET, and explain differences observed during plasma initiation with the old carbon plasma facing components. In the International Thermonuclear Experimental Reactor (ITER), the allowable toroidal electric field is restricted to 0.35 [V/m], which is significantly lower compared to the typical value (~ 1 [V/m]) used in the present devices. The limitation on toroidal electric field also reduces the range of other operation parameters during plasma formation in ITER. Thus, predictive simulations of plasma burn-through in ITER using validated model is of crucial importance. This paper provides an overview of the DYON code and the validation, together with new predictive simulations for ITER using the DYON code.
    Plasma Physics and Controlled Fusion 03/2014; 55(12). · 2.37 Impact Factor
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    Nuclear Fusion 01/2014; 54:013015. · 2.73 Impact Factor
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    ABSTRACT: This paper presents the electromagnetic modeling of the plasma current breakdown phase of the JET tokamak. The first part of this paper models the presence of the JET iron core up-down asymmetry and the effects of the eddy currents in the reconstruction of the magnetic topology needed for the plasma start. The second part describes the approach used to evaluate the ionized particle connection length inside the vacuum chamber at breakdown. The results obtained are validated using JET experimental measurements.
    IEEE Transactions on Magnetics 01/2014; 50(2):937-940. · 1.42 Impact Factor
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    ABSTRACT: This paper reports the impact on confinement and power load of the high-shape 2.5MA ELMy H-mode scenario at JET of a change from an all carbon plasma facing components to an all metal wall. In preparation to this change, systematic studies of power load reduction and impact on confinement as a result of fuelling in combination with nitrogen seeding were carried out in JET-C and are compared to their counterpart in JET with a metallic wall. An unexpected and significant change is reported on the decrease of the pedestal confinement but is partially recovered with the injection of nitrogen.
    10/2013;
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    ABSTRACT: This paper presents the DYON simulations of the plasma burn-through phase at Joint European Torus (JET) with the ITER-like wall. The main purpose of the study is to validate the simulations with the ITER-like wall, made of beryllium. Without impurities, the burn-through process of a pure deuterium plasma is described using DYON simulations, and the criterion for deuterium burn-through is derived analytically. The plasma burn-through with impurities are simulated using wall-sputtering models in the DYON code, which are modified for the ITER-like wall. The wall-sputtering models and the validation against JET data are presented. The impact of the assumed plasma parameters in DYON simulations are discussed by means of parameter scans. As a result, the operation space of prefill gas pressure and toroidal electric field for plasma burn-through in JET is compared to the Townsend avalanche criterion.
    Nuclear Fusion 09/2013; 53(8). · 2.73 Impact Factor
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    ABSTRACT: Plasma Surface Interaction(PSI) effects on plasma burn-through are compared for the carbon wall and the ITER-Like Wall(ILW) at JET. For the carbon wall, the radiation barrier and C2+ influx have a significant linear correlation whereas the radiation barrier in the ILW does not have such a linear correlation with Be 1+ influx. The JET data are explained by the simulation results of the DYON code. The radiation barrier in the carbon wall JET is dominated by the carbon radiation, but the radiation barrier in the ILW is mainly from the deuterium radiation rather than the beryllium radiation.
    Journal of Nuclear Materials 06/2013; · 2.02 Impact Factor
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    ABSTRACT: The Current Limit Avoidance System (CLA) has been recently deployed at the JET tokamak to avoid current saturations in the poloidal field (PF) coils when the eXtreme Shape Controller is used to control the plasma shape. In order to cope with the current saturation limits, the CLA exploits the redundancy of the PF coils system to automatically obtain almost the same plasma shape using a different combination of currents in the PF coils. In the presence of disturbances it tries to avoid the current saturations by relaxing the constraints on the plasma shape control. The CLA system has been successfully implemented on the JET tokamak and fully commissioned in 2011. This paper presents the first experimental results achieved in 2011–2012 during the restart and the ITER-like wall campaigns at JET.
    Fusion Engineering and Design. 06/2013; 88(5):400–407.
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    ABSTRACT: To consolidate International Thermonuclear Experimental Reactor (ITER) design choices and prepare for its operation, Joint European Torus (JET) has implemented ITER's plasma facing materials, namely, Be for the main wall and W in the divertor. In addition, protection systems, diagnostics, and the vertical stability control were upgraded and the heating capability of the neutral beams was increased to over 30 MW. First results confirm the expected benefits and the limitations of all metal plasma facing components (PFCs) but also yield understanding of operational issues directly relating to ITER. H-retention is lower by at least a factor of 10 in all operational scenarios compared to that with C PFCs. The lower C content (≈ factor 10) has led to much lower radiation during the plasma burn-through phase eliminating breakdown failures. Similarly, the intrinsic radiation observed during disruptions is very low, leading to high power loads and to a slow current quench. Massive gas injection using a D2/Ar mixture restores levels of radiation and vessel forces similar to those of mitigated disruptions with the C wall. Dedicated L-H transition experiments indicate a 30% power threshold reduction, a distinct minimum density, and a pronounced shape dependence. The L-mode density limit was found to be up to 30% higher than for C allowing stable detached divertor operation over a larger density range. Stable H-modes as well as the hybrid scenario could be re-established only when using gas puff levels of a few 1021 es−1. On average, the confinement is lower with the new PFCs, but nevertheless, H factors up to 1 (H-Mode) and 1.3 (at βN ≈ 3, hybrids) have been achieved with W concentrations well below the maximum acceptable level.
    Physics of Plasmas 05/2013; 20(5). · 2.38 Impact Factor
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    ABSTRACT: The recent installation of a full metal, ITER-like, first wall provided the opportunity to study the impact of the plasma-facing materials on plasma initiation or breakdown. This study for the first time presents a full experimental characterisation of tokamak breakdown at JET, using all discharges since 2008, covering both operations with a main chamber carbon and a beryllium ITER-like main chamber wall. It was found that the avalanche phase was unaffected by the change in wall material. However, changes in out-gassing by the wall and lower carbon levels resulted in better controlled density and significantly lower radiation during the burn-through phase with the ITER-like wall. Breakdown failures, that usually developed with a carbon wall during the burn-through phase (especially after disruptions) were absent with the ITER-like wall. These observations match with the results obtained from a new model of plasma burn-through that includes plasma-surface interactions (Kim et al 2012 Nucl. Fusion 52 103016). This shows that chemical sputtering of carbon is the determining factor for the impurity content, and hence also radiation, during the burn-through phase for operations with a carbon wall. As seen experimentally, with a beryllium main wall, the plasma surface effects predicted by the model do not raise the radiation levels much above those expected for pure deuterium plasmas. With the ITER-like wall, operation with higher pre-fill pressures, and thus higher breakdown densities, was possible, which helped maintaining the density after breakdown.
    Nuclear Fusion 05/2013; 53(5):3003-. · 2.73 Impact Factor
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    ABSTRACT: In this paper, new models for a plasma burn-through simulation using the DYON code are introduced in detail, and the quantitative validation of the simulation results against JET data is presented for the first time. In order to calculate the particle confinement time, a dynamic effective connection length model including an eddy current effect is used assuming ambipolar transonic transport and the Bohm diffusion model for parallel and perpendicular particle losses, respectively. Plasma-surface interaction effects are treated with an impurity sputtering yield and an exponential saturation model of the deuterium recycling coefficient. The rate and power coefficients in the Atomic Data and Analysis Structure (ADAS) package are adopted to solve energy and particle balance. The neutral screening effects are taken into account according to particle species, and the sophisticated energy and particle balances are presented. The new burn-through simulation shows good agreement against carbon-wall JET data. This indicates that the burn-through simulation can be applied to investigate the key aspect of physics in plasma burn-through and to perform a predictive simulation for ITER start-up.
    Nuclear Fusion 10/2012; 52(10):3016-. · 2.73 Impact Factor
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    ABSTRACT: In ASDEX Upgrade the compatibility of improved H-modes with an all-W wall has been demonstrated. Under boronized conditions light impurities and the radiated power fraction in the divertor were reduced, requiring N seeding to cool the divertor plasma. The impurity seeding does not only protect the divertor tiles but also considerably improves the performance of improved H-mode discharges by up to 25%. The energy confinement increases to H98-factors up to 1.3 and thereby exceeds the best values in the carbon-dominated AUG at the same density and collisionality. This improvement is due to higher edge temperatures rather than to peaking of the electron density profile. Higher temperatures are reached at the pedestal top leading, via profile stiffness, to an increase in the total plasma pressure. There is no change to in the plasma core. The dilution at the plasma edge by nitrogen seems to play an important role since it allows higher ion temperatures at the same edge ion pressure as in the unseeded case. The dilution of the core plasma remains moderate.
    Nuclear Fusion 09/2011; 51(11):113003. · 2.73 Impact Factor
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    ABSTRACT: Recent progress on ITER steady-state (SS) scenario modelling by the ITPA-IOS group is reviewed. Code-to-code benchmarks as the IOS group's common activities for the two SS scenarios (weak shear scenario and internal transport barrier scenario) are discussed in terms of transport, kinetic profiles, and heating and current drive (CD) sources using various transport codes. Weak magnetic shear scenarios integrate the plasma core and edge by combining a theory-based transport model (GLF23) with scaled experimental boundary profiles. The edge profiles (at normalized radius ρ = 0.8–1.0) are adopted from an edge-localized mode-averaged analysis of a DIII-D ITER demonstration discharge. A fully noninductive SS scenario is achieved with fusion gain Q = 4.3, noninductive fraction fNI = 100%, bootstrap current fraction fBS = 63% and normalized beta βN = 2.7 at plasma current Ip = 8 MA and toroidal field BT = 5.3 T using ITER day-1 heating and CD capability. Substantial uncertainties come from outside the radius of setting the boundary conditions (ρ = 0.8). The present simulation assumed that βN (ρ) at the top of the pedestal (ρ = 0.91) is about 25% above the peeling–ballooning threshold. ITER will have a challenge to achieve the boundary, considering different operating conditions (Te/Ti ≈ 1 and density peaking). Overall, the experimentally scaled edge is an optimistic side of the prediction. A number of SS scenarios with different heating and CD mixes in a wide range of conditions were explored by exploiting the weak-shear steady-state solution procedure with the GLF23 transport model and the scaled experimental edge. The results are also presented in the operation space for DT neutron power versus stationary burn pulse duration with assumed poloidal flux availability at the beginning of stationary burn, indicating that the long pulse operation goal (3000 s) at Ip = 9 MA is possible. Source calculations in these simulations have been revised for electron cyclotron current drive including parallel momentum conservation effects and for neutral beam current drive with finite orbit and magnetic pitch effects.
    Nuclear Fusion 08/2011; 51(10):103006. · 2.73 Impact Factor
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    ABSTRACT: Electron cyclotron resonance heating (ECRH)-assisted plasma breakdown is foreseen with full and half magnetic field in ITER. As reported earlier, the corresponding O1- and X2-schemes have been successfully used to assist pre-ionization and breakdown in present-day devices. This contribution reports on common experiments studying the effect of toroidal inclination of the ECR beam, which is ≥20° in ITER. All devices could demonstrate successful breakdown assistance for this case also, although in some experiments the necessary power was almost a factor of 2 higher compared with perpendicular launch. Differences between the devices with regard to the required power and vertical field are discussed and analysed. In contrast to most of these experiments, ITER will build up loop voltage prior to the formation of the field null due to the strong shielding by the vessel. Possible consequences of this difference are discussed.
    Nuclear Fusion 07/2011; 51(8):083031. · 2.73 Impact Factor
  • Plasma Physics and Controlled Fusion 07/2011; · 2.37 Impact Factor
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    ABSTRACT: Sawtooth induced heat and density pulse measurements reported in the literature for the JET and TEXT experiments are discussed. In JET the heat pulse travels ten times faster than the density pulse, but in TEXT both pulses travel at the same speed. The measurements are analysed using coupled transport equations for energy and particles. It is shown that the different behaviour of the density pulse in the two experiments can be attributed to differences in the off-diagonal elements of the transport matrix. If the perturbed fluxes of heat and particles are expressed as linear combinations of the thermodynamic forces ∇p and ∇T (rather than ∇n and ∇T), the corresponding transport matrices are remarkably similar. However, minor differences in this transport matrix between JET and TEXT account for the qualitative difference in the density pulses.
    Nuclear Fusion 01/2011; 31(8):1545. · 2.73 Impact Factor
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    ABSTRACT: The perturbations of electron density and temperature profiles in a tokamak following a sawtooth collapse are considered. An analytic model for the interpretation of such perturbations is presented. It is shown that the perturbation can be decomposed into two contributions, which are eigenmodes of the linearized coupled diffusion equations for particles and energy. The approximations made in the analytic treatment are checked using computer simulations. Measurements of heat and density pulses in JET are used to illustrate the power of the new approach. It is shown that with the coupled equations an improved description of the heat and density pulses is obtained. The analysis yields the four diffusion coefficients in the linearized transport matrix. The non-zero off-diagonal elements explain certain salient features of the measurements, notably a marked decrease of the local density which occurs during the maximum of the temperature pulse.
    Nuclear Fusion 01/2011; 31(7):1261. · 2.73 Impact Factor
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    ABSTRACT: As the ITER project moves forward in the procurement phase of major components, the baseline design is subject to a series of modifications due to the finalization of the design and interaction with industry.Some modifications have a potential impact on the performance and operation of the machine. A substantial activity is being carried out in EU in support to the design, aimed at assessing such impact.The effect of the 2010 design of the central solenoid on the operational space was assessed. A detailed analysis of the plasma breakdown and start-up phases was performed to assess the impact of withdrawing the booster converters from the baseline design and of the modifications to the power supply voltage ratings. In order to reach a clear conclusion, a review of the start-up strategy was also carried out starting from an analysis of the avalanche and breakdown conditions in ITER. The controllability of the null quality and position was verified in presence of uncertainties and measurement errors. The advantages of higher voltage rating were also quantified.General aim of the paper is to present an overview of the analyses performed and of the issues identified providing an assessment of the impact of some of the proposed changes on ITER operation.
    Fusion Engineering and Design 01/2011; 86:1103-1106. · 0.84 Impact Factor
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    ABSTRACT: The current ramp-up phase of ITER demonstration discharges, performed at JET, is analysed and the capability of the empirical L-mode Bohm–gyroBohm and Coppi–Tang transport models as well as the theory-based GLF23 model to predict the temperature evolution in these discharges is examined. The analysed database includes ohmic (OH) plasmas with various current ramp rates and plasma densities and the L-mode plasmas with the ion cyclotron radio frequency (ICRF) and neutral beam injection (NBI) heating performed at various ICRF resonance positions and NBI heating powers. The emphasis of this analysis is a data consistency test, which is particularly important here because some parameters, useful for the transport model validation, are not measured in OH and ICRF heated plasmas (e.g. ion temperature, effective charge). The sensitivity of the predictive accuracy of the transport models to the unmeasured data is estimated. It is found that the Bohm–gyroBohm model satisfactorily predicts the temperature evolution in discharges with central heating (the rms deviation between the simulated and measured temperature is within 15%), but underestimates the thermal electron transport in the OH and off-axis ICRF heated discharges. The Coppi–Tang model strongly underestimates the thermal transport in all discharges considered. A re-normalization of these empirical models for improving their predictive capability is proposed. The GLF23 model, strongly dependent on the ion temperature gradient and tested only for NBI heated discharges with measured ion temperatures, predicts accurately the temperature in the low power NBI heated discharge (rms < 10%) while the discrepancy with the data increases at high power. Based on the analysis of the JET discharges, the modelling of the current ramp-up phase for the H-mode ITER scenario is performed with particular emphasis on the sensitivity of the sawtooth-free duration of this phase to transport model.
    Plasma Physics and Controlled Fusion 09/2010; 52(10):105011. · 2.37 Impact Factor

Publication Stats

1k Citations
258.65 Total Impact Points

Institutions

  • 2010–2014
    • European Commission
      Bruxelles, Brussels Capital Region, Belgium
  • 1999–2010
    • Max Planck Institute for Plasma Physics
      • Max Planck Institute for Plasma Physics, Greifswald
      Arching, Bavaria, Germany
  • 1998
    • General Atomics
      San Diego, California, United States
  • 1997
    • University of California, San Diego
      • Department of Mechanical and Aerospace Engineering (MAE)
      San Diego, CA, United States