A. C. C. Sips

European Commission, Bruxelles, Brussels Capital, Belgium

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Publications (206)349.85 Total impact

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    ABSTRACT: New experiments in 2013–2014 have investigated the physics responsible for the decrease in H-mode pedestal confinement observed in the initial phase of JET-ILW operation (2012 Experimental Campaigns). The effects of plasma triangularity, global beta and neutrals on pedestal confinement and stability have been investigated systematically. The stability of JET-ILW pedestals is analysed in the framework of the peeling–ballooning model and the model assumptions of the pedestal predictive code EPED. Low D neutrals content in the plasma, achieved either by low D2 gas injection rates or by divertor configurations with optimum pumping, and high beta are necessary conditions for good pedestal (and core) performance. In such conditions the pedestal stability is consistent with the peeling–ballooning paradigm. Moderate to high D2 gas rates, required for W control and stable H-mode operation with the ILW, lead to increased D neutrals content in the plasma and additional physics in the pedestal models may be required to explain the onset of the ELM instability. The changes in H-mode performance associated with the change in JET wall composition from C to Be/W point to D neutrals and low-Z impurities playing a role in pedestal stability, elements which are not currently included in pedestal models. These aspects need to be addressed in order to progress towards full predictive capability of the pedestal height.
    Nuclear Fusion 09/2015; 55(11). DOI:10.1088/0029-5515/55/11/113031 · 3.06 Impact Factor
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    ABSTRACT: In preparation of ITER operation, large machines have replaced their wall and divertor material to W (ASDEX Upgrade) or a combination of Be for the wall and W for the divertor (JET). Operation in these machines has shown that the influx of W can have a significant impact on the discharge evolution, which has made modelling of this impact for ITER an urgent task. This paper reports on such modelling efforts. Maximum tolerable W concentrations have been determined for various scenarios, both for the current ramp-up and flat-top phase. Results of two independent methods are presented, based on the codes ZIMPUR plus ASTRA and CRONOS, respectively. Both methods have been tested and benchmarked against ITER-like Ip RU experiments at JET. It is found that W significantly disturbs the discharge evolution when the W concentration approaches ~10−4; this critical level varies somewhat between scenarios.
    Nuclear Fusion 06/2015; 55(6):063031. DOI:10.1088/0029-5515/55/6/063031 · 3.06 Impact Factor
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    Dataset: Talk

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    Dataset: talk EX2-6

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    A. C. C. Sips · G. Giruzzi · S. Ide · C. Kessel · T. C. Luce · J. A. Snipes · J. K. Stober ·
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    ABSTRACT: The development of operating scenarios is one of the key issues in the research for ITER which aims to achieve a fusion gain (Q) of ̃10, while producing 500 MW of fusion power for ≥300 s. The ITER Research plan proposes a success oriented schedule starting in hydrogen and helium, to be followed by a nuclear operation phase with a rapid development towards Q ̃ 10 in deuterium/tritium. The Integrated Operation Scenarios Topical Group of the International Tokamak Physics Activity initiates joint activities among worldwide institutions and experiments to prepare ITER operation. Plasma formation studies report robust plasma breakdown in devices with metal walls over a wide range of conditions, while other experiments use an inclined EC launch angle at plasma formation to mimic the conditions in ITER. Simulations of the plasma burn-through predict that at least 4 MW of Electron Cyclotron heating (EC) assist would be required in ITER. For H-modes at q95 ̃ 3, many experiments have demonstrated operation with scaled parameters for the ITER baseline scenario at ne/nGW ̃ 0.85. Most experiments, however, obtain stable discharges at H98(y,2) ̃ 1.0 only for βN = 2.0-2.2. For the rampup in ITER, early X-point formation is recommended, allowing auxiliary heating to reduce the flux consumption. A range of plasma inductance (li(3)) can be obtained from 0.65 to 1.0, with the lowest values obtained in H-mode operation. For the rampdown, the plasma should stay diverted maintaining H-mode together with a reduction of the elongation from 1.85 to 1.4. Simulations show that the proposed rampup and rampdown schemes developed since 2007 are compatible with the present ITER design for the poloidal field coils. At 13-15 MA and densities down to ne/nGW ̃ 0.5, long pulse operation (〉1000 s) in ITER is possible at Q ̃ 5, useful to provide neutron fluence for Test Blanket Module assessments. ITER scenario preparation in hydrogen and helium requires high input power (〉50 MW). H-mode operation in helium may be possible at input powers above 35 MW at a toroidal field of 2.65 T, for studying H-modes and ELM mitigation. In hydrogen, H-mode operation is expected to be marginal, even at 2.65 T with 60 MW of input power. Simulation code benchmark studies using hybrid and steady state scenario parameters have proved to be a very challenging and lengthy task of testing suites of codes, consisting of tens of sophisticated modules. Nevertheless, the general basis of the modelling appears sound, with substantial consistency among codes developed by different groups. For a hybrid scenario at 12 MA, the code simulations give a range for Q = 6.5-8.3, using 30 MW neutral beam injection and 20 MW ICRH. For non-inductive operation at 7-9 MA, the simulation results show more variation. At high edge pedestal pressure (Tped ̃ 7 keV), the codes predict Q = 3.3-3.8 using 33 MW NB, 20 MW EC, and 20 MW ion cyclotron to demonstrate the feasibility of steady-state operation with the day-1 heating systems in ITER. Simulations using a lower edge pedestal temperature (̃3 keV) but improved core confinement obtain Q = 5-6.5, when ECCD is concentrated at mid-radius and ̃20 MW off-axis current drive (ECCD or LHCD) is added. Several issues remain to be studied, including plasmas with dominant electron heating, mitigation of transient heat loads integrated in scenario demonstrations and (burn) control simulations in ITER scenarios.
    Physics of Plasmas 01/2015; 22(2). DOI:10.1063/1.4904015 · 2.14 Impact Factor
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    ABSTRACT: Since the ITER-like wall in JET (JET-ILW) came into operation, dedicated ITER-like plasma current (I p) ramp-up (RU) and ramp-down (RD) experiments have been performed and matched to similar discharges with the carbon wall (JET-C). The experiments show that access to H-mode early in the I p RU phase and maintaining H-mode in the I p RD as long as possible are instrumental to achieve low internal plasma inductance (l i) and to minimize flux consumption. In JET-ILW, at a given current rise rate similar variations in l i (0.7–0.9) are obtained as in JET-C. In most discharges no strong W accumulation is observed. However, in some low density cases during the early phase of the I p &${\rm RU}(n_{\rm e}/n_{\rm e}^{\rm Gw} \sim 0.2)$ ; strong core radiation due to W influx led to hollow electron temperature (T e) profiles. In JET-ILW Z eff is significantly lower than in JET-C. W significantly disturbs the discharge evolution when the W concentration approaches 10−4; this threshold is confirmed by predictive transport modelling using the CRONOS code. I p RD experiments in JET-ILW confirm the result of JET-C that sustained H-mode and elongation reduction are both instrumental in controlling l i.
    Nuclear Fusion 01/2015; 55(1). DOI:10.1088/0029-5515/55/1/013009 · 3.06 Impact Factor
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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.
    Nuclear Fusion 06/2014; 54(7). DOI:10.1088/0029-5515/54/7/073016 · 3.06 Impact Factor
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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). DOI:10.1007/s10894-013-9652-7 · 0.99 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). DOI:10.1088/0741-3335/55/12/124032 · 2.19 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 02/2014; 50(2):937-940. DOI:10.1109/TMAG.2013.2282351 · 1.39 Impact Factor
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    ABSTRACT: In the recent JET experimental campaigns with the new ITER-like wall (JET-ILW), major progress has been achieved in the characterization and operation of the H-mode regime in metallic environments: (i) plasma breakdown has been achieved at the first attempt and X-point L-mode operation recovered in a few days of operation; (ii) stationary and stable type-I ELMy H-modes with βN ~ 1.4 have been achieved in low and high triangularity ITER-like shape plasmas and are showing that their operational domain at H = 1 is significantly reduced with the JET-ILW mainly because of the need to inject a large amount of gas (above 1022 D s−1) to control core radiation; (iii) in contrast, the hybrid H-mode scenario has reached an H factor of 1.2–1.3 at βN of 3 for 2–3 s; and, (iv) in comparison to carbon equivalent discharges, total radiation is similar but the edge radiation is lower and Zeff of the order of 1.3–1.4. Strong core radiation peaking is observed in H-mode discharges at a low gas fuelling rate (i.e. below 0.5 × 1022 D s−1) and low ELM frequency (typically less than 10 Hz), even when the tungsten influx from the diverter is constant. High-Z impurity transport from the plasma edge to the core appears to be the dominant factor to explain these observations. This paper reviews the major physics and operational achievements and challenges that an ITER-like wall configuration has to face to produce stable plasma scenarios with maximized performance.
    Nuclear Fusion 01/2014; 54(1):013011. DOI:10.1088/0029-5515/54/1/013011 · 3.06 Impact Factor
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    ABSTRACT: Since its inception in 2002, the International Tokamak Physics Activity topical group on Integrated Operational Scenarios (IOS) has coordinated experimental and modelling activity on the development of advanced inductive scenarios for applications in the ITER tokamak. The physics basis and the prospects for applications in ITER have been advanced significantly during that time, especially with respect to experimental results. The principal findings of this research activity are as follows. Inductive scenarios capable of higher normalized pressure (beta(N) >= 2.4) than the ITER baseline scenario (beta(N) = 1.8) with normalized confinement at or above the standard H-mode scaling are well established under stationary conditions on the four largest diverted tokamaks (AUG, DIII-D, JET, JT-60U), demonstrated in a database of more than 500 plasmas from these tokamaks analysed here. The parameter range where high performance is achieved is broad in q(95) and density normalized to the empirical density limit. MHD modes can play a key role in reaching stationary high performance, but also define the limits to achieved stability and confinement. Projection of performance in ITER from existing experiments uses empirical scalings and theory-based modelling. The status of the experimental validation of both approaches is summarized here. The database shows significant variation in the energy confinement normalized to standard H-mode confinement scalings, indicating the possible influence of additional physics variables absent from the scalings. Tests using the available information on rotation and the ratio of the electron and ion temperatures indicate neither of these variables in isolation can explain the variation in normalized confinement observed. Trends in the normalized confinement with the two dimensionless parameters that vary most from present-day experiments to ITER, gyroradius and collision frequency, are significant. Regression analysis on the multi-tokamak database has been performed, but it appears that the database is not conditioned sufficiently well to yield a new scaling for this type of plasma. Coordinated experiments on size scaling using the dimensionless parameter scaling approach find a weaker scaling with normalized gyroradius than the standard H-mode scaling. Preliminary studies on scaling with collision frequency show a favourable scaling stronger than the standard H-mode scaling. Coordinated modelling activity has resulted in successful benchmarking of modelling codes in the ITER regime. Validation of transport models using these codes on present-day experiments is in progress, but no single model has been shown to capture the variations seen in the experiments. However, projection to ITER using these models is in general agreement with the favourable projections found with the empirical scalings.
    Nuclear Fusion 01/2014; 54(1):013015. DOI:10.1088/0029-5515/54/1/013015 · 3.06 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.
    Nuclear Fusion 10/2013; 53(11). DOI:10.1088/0029-5515/53/11/113025 · 3.06 Impact Factor
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    Hyun-Tae Kim · A. C. C. Sips ·
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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). DOI:10.1088/0029-5515/53/8/083024 · 3.06 Impact Factor
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    ABSTRACT: The size and capability of JET to reach high plasma current and field enables a study of the plasma behaviour at ion Larmor radius and collisionality values approaching those of ITER. In this paper such study is presented. The achievement of stationary type I ELMy H-modes at high current proved to be quite challenging. As the plasma current was increased, it became more difficult to achieve stationary conditions. Nevertheless, it was possible to achieve stable operation at high plasma current (up to 4.5 MA) and low q(95) (2.65-3) at JET. One of the main reasons to revisit the high plasma current experiments done in 1997 is the higher power available and the improvement of the pedestal diagnostics. Indeed, compared with previous results, higher stored energy was achieved but confinement was still degraded. The causes of this confinement degradation are discussed in the paper.
    Nuclear Fusion 07/2013; 53(7):073020. DOI:10.1088/0029-5515/53/7/073020 · 3.06 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; 438. DOI:10.1016/j.jnucmat.2013.01.056 · 1.87 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. DOI:10.1016/j.fusengdes.2013.04.018 · 1.15 Impact Factor
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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). DOI:10.1063/1.4804411 · 2.14 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-. DOI:10.1088/0029-5515/53/5/053003 · 3.06 Impact Factor
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    ABSTRACT: In the all-tungsten ASDEX Upgrade tokamak nitrogen seeding is a reliable method for radiative cooling at the plasma edge, regularly applied in high power scenarios. Interestingly, in the presence of nitrogen seeding the energy confinement is observed to improve significantly, compared with similar unseeded discharges, with an increase by 10–25% in plasma stored energy and HIPB98(y,2). In this paper, we document the improvement and we analyse the transport properties in the core plasma, by comparing similar discharges with and without nitrogen seeding. The increase in the suprathermal energy content is assessed as well. The impurity nitrogen is shown not to penetrate significantly into the core plasma. Non-linear gyro-kinetic simulations predict that the improvement observed in the pedestal confinement is transferred to the core via profile stiffness, in agreement with the experimental evidence, whereas the direct contribution from core confinement is small.
    Plasma Physics and Controlled Fusion 01/2013; 55(1):015010. DOI:10.1088/0741-3335/55/1/015010 · 2.19 Impact Factor

Publication Stats

3k Citations
349.85 Total Impact Points


  • 2010-2015
    • European Commission
      Bruxelles, Brussels Capital, Belgium
  • 2012-2014
    • Imperial College London
      Londinium, England, United Kingdom
  • 2013
    • Queen's University Belfast
      Béal Feirste, Northern Ireland, United Kingdom
  • 1999-2010
    • Max Planck Institute for Plasma Physics
      • Max Planck Institute for Plasma Physics, Greifswald
      Arching, Bavaria, Germany
  • 2005
    • University College Cork
      • Department of Physics
      Corcaigh, Munster, Ireland
  • 1998
    • General Atomics
      San Diego, California, United States
  • 1994-1997
    • University of Toronto
      • Institute for Aerospace Studies
      Toronto, Ontario, Canada