Neoclassical and turbulent heavy impurity transport in tokamak core plasmas are determined by main ion temperature, density and toroidal rotation profiles. Thus, in order to understand and prevent experimental behaviour of W accumulation, flux-driven integrated modelling of main ion heat and particle transport over multiple confinement times is a vital prerequisite. For the first time, the quasilinear gyrokinetic code QuaLiKiz is applied for successful predictions of core kinetic profiles in an ASDEX Upgrade H-mode discharge in the turbulence dominated region within the integrated modelling suite JETTO. Neoclassical contributions are calculated by NCLASS; auxiliary heat and particle deposition profiles due to NBI and ECRH are prescribed from previous analysis with TRANSP. Turbulent and neoclassical contributions are insufficient in explaining main ion heat and particle transport inside the q = 1 surface, necessitating the prescription of further transport coefficients to mimic the impact of MHD activity on central transport. The ion to electron temperature ratio at the simulation boundary at p tor=0.85 stabilizes ion scale modes while destabilizing ETG modes when significantly exceeding unity. Careful analysis of experimental measurements using Gaussian process regression techniques is carried out to explore reasonable uncertainties. In following trace W impurity transport simulations performed with additionally NEO, neoclassical transport under consideration of poloidal asymmetries alone is found to be insufficient to establish hollow central W density profiles. Reproduction of these conditions measured experimentally is found possible only when assuming the direct impact of a saturated (m, n) = (1, 1) MHD mode on heavy impurity transport.

VL - 59 IS - 1 U1 -FP

U2 -IMT

U5 - 6d42c99c474c747f17f4b44236cecb27 ER - TY - JOUR T1 - Dependence on plasma shape and plasma fueling for small edge-localized mode regimes in TCV and ASDEX Upgrade JF - Nuclear Fusion Y1 - 2019 A1 - Labit, B. A1 - Eich, T. A1 - Harrer, G. A1 - Wolfrum, E. A1 - Bernert, M. A1 - Dunne, M. G. A1 - Frassinetti, L. A1 - Hogeweij, G. M. D. A1 - Perek, A. A1 - Vanovac, B. A1 - Hennequin, P. A1 - Maurizio, R. A1 - Merle, A. A1 - EUROfusion MST1 Team A1 - Meyer, H. A1 - Saarelma, S. A1 - Sheikh, H. AB - Within the EUROfusion MST1 work package, a series of experiments has been conducted on AUG and TCV devices to disentangle the role of plasma fueling and plasma shape for the onset of small ELM regimes. On both devices, small ELM regimes with high confinement are achieved if and only if two conditions are fulfilled at the same time. Firstly, the plasma density at the separatrix must be large enough (n e,sep / nG about 0.3), leading to a pressure profile flattening at the separatrix, which stabilizes type-I ELMs. Secondly, the magnetic configuration has to be close to a double null (DN), leading to a reduction of the magnetic shear in the extreme vicinity of the separatrix. As a consequence, its stabilizing effect on ballooning modes is weakened. VL - 59 IS - 8 U1 - FP U2 - PEPD U5 - aa2909729a2061f7cf725b5ee78cd3d1 ER - TY - JOUR T1 - Physics research on the TCV tokamak facility: from conventional to alternative scenarios and beyond JF - Nuclear Fusion Y1 - 2019 A1 - Coda, S. A1 - Agostini, M. A1 - Albanese, R. A1 - Alberti, S. A1 - Alessi, E. A1 - Allan, S. A1 - Hogeweij, G. M. D. A1 - Perek, A. A1 - Ravensbergen, T. A1 - Vijvers, W. A. J. A1 - Allcock, J. A1 - Ambrosino, R. A1 - Anand, H. A1 - Andrebe, Y. A1 - EUROfusion MST1 Team A1 - et al. AB - The research program of the TCV tokamak ranges from conventional to advanced-tokamak scenarios and alternative divertor configurations, to exploratory plasmas driven by theoretical insight, exploiting the device's unique shaping capabilities. Disruption avoidance by real-time locked mode prevention or unlocking with electron-cyclotron resonance heating (ECRH) was thoroughly documented, using magnetic and radiation triggers. Runaway generation with high-Z noble-gas injection and runaway dissipation by subsequent Ne or Ar injection were studied for model validation. The new 1 MW neutral beam injector has expanded the parameter range, now encompassing ELMy H-modes in an ITER-like shape and nearly non-inductive H-mode discharges sustained by electron cyclotron and neutral beam current drive. In the H-mode, the pedestal pressure increases modestly with nitrogen seeding while fueling moves the density pedestal outwards, but the plasma stored energy is largely uncorrelated to either seeding or fueling. High fueling at high triangularity is key to accessing the attractive small edge-localized mode (type-II) regime. Turbulence is reduced in the core at negative triangularity, consistent with increased confinement and in accord with global gyrokinetic simulations. The geodesic acoustic mode, possibly coupled with avalanche events, has been linked with particle flow to the wall in diverted plasmas. Detachment, scrape-off layer transport, and turbulence were studied in L- and H-modes in both standard and alternative configurations (snowflake, super-X, and beyond). The detachment process is caused by power 'starvation' reducing the ionization source, with volume recombination playing only a minor role. Partial detachment in the H-mode is obtained with impurity seeding and has shown little dependence on flux expansion in standard single-null geometry. In the attached L-mode phase, increasing the outer connection length reduces the in–out heat-flow asymmetry. A doublet plasma, featuring an internal X-point, was achieved successfully, and a transport barrier was observed in the mantle just outside the internal separatrix. In the near future variable-configuration baffles and possibly divertor pumping will be introduced to investigate the effect of divertor closure on exhaust and performance, and 3.5 MW ECRH and 1 MW neutral beam injection heating will be added. VL - 59 IS - 11 U1 - FP U2 - PEPD U5 - d50a5905a84255c86ad6d49b1cfa1569 ER - TY - JOUR T1 - Overview of the JET preparation for deuterium–tritium operation with the ITER like-wall JF - Nuclear Fusion Y1 - 2019 A1 - Joffrin, E. A1 - Abduallev, S. A1 - Abhangi, M A1 - Abreu, P. A1 - Afanasev, V. A1 - Citrin, J. A1 - Ho, A. A1 - Hogeweij, G. M. D. A1 - Marin, M. A1 - G. van Rooij A1 - Shumack, A. E. A1 - Jaulmes, F. A1 - Felici, F. A1 - den Harder, N. A1 - Tsalas, M. A1 - Afzal, M. A1 - Aggarwal, K. M. A1 - Ahlgren, T. A1 - Aho-Mantila, L. A1 - Aiba, N. A1 - et al. AB - For the past several years, the JET scientific programme (Pamela et al 2007 Fusion Eng. Des. 82 590) has been engaged in a multi-campaign effort, including experiments in D, H and T, leading up to 2020 and the first experiments with 50%/50% D–T mixtures since 1997 and the first ever D–T plasmas with the ITER mix of plasma-facing component materials. For this purpose, a concerted physics and technology programme was launched with a view to prepare the D–T campaign (DTE2). This paper addresses the key elements developed by the JET programme directly contributing to the D–T preparation. This intense preparation includes the review of the physics basis for the D–T operational scenarios, including the fusion power predictions through first principle and integrated modelling, and the impact of isotopes in the operation and physics of D–T plasmas (thermal and particle transport, high confinement mode (H-mode) access, Be and W erosion, fuel recovery, etc). This effort also requires improving several aspects of plasma operation for DTE2, such as real time control schemes, heat load control, disruption avoidance and a mitigation system (including the installation of a new shattered pellet injector), novel ion cyclotron resonance heating schemes (such as the three-ions scheme), new diagnostics (neutron camera and spectrometer, active Alfvèn eigenmode antennas, neutral gauges, radiation hard imaging systems...) and the calibration of the JET neutron diagnostics at 14 MeV for accurate fusion power measurement. The active preparation of JET for the 2020 D–T campaign provides an incomparable source of information and a basis for the future D–T operation of ITER, and it is also foreseen that a large number of key physics issues will be addressed in support of burning plasmas. VL - 59 IS - 11 U1 - FP U2 - IMT U5 - 307a9a4ea8f67c2769e405c252ba0db4 ER - TY - JOUR T1 - A systematic approach to optimize excitations for perturbative transport experiments JF - Physics of Plasmas Y1 - 2018 A1 - van Berkel, M. A1 - De Cock, A. A1 - Ravensbergen, T. A1 - Hogeweij, G. M. D. A1 - Zwart, H. J. A1 - Vandersteen, G. AB -In this paper, techniques for optimal input design are used to optimize the waveforms of perturbative experiments in modern fusion devices. The main focus of this paper is to find the modulation frequency for which the accuracy of the estimated diffusion coefficient is maximal. Mathematically, this problem can be formulated as an optimization problem in which the Fisher information matrix is maximized. First, this optimization problem is solved for a simplified diffusion model, while assuming a slab geometry and a semi-infinite domain. Later, the optimization is repeated under more general conditions such as a cylindrical geometry, finite domain, and simultaneous estimation of multiple transport coefficients. Based on the results of these optimizations, guidelines are offered to select the modulation frequency and to determine the optimality of the corresponding experiment. © 2018 EURATOM

VL - 25 IS - 8 U1 -FP

U2 -IMT

U5 - 83270781877438b0f4a781f307204968 ER - TY - JOUR T1 - Separation of transport in slow and fast time-scales using modulated heat pulse experiments (hysteresis in flux explained) JF - Nuclear Fusion Y1 - 2018 A1 - van Berkel, M. A1 - Vandersteen, G. A1 - Zwart, H. J. A1 - Hogeweij, G. M. D. A1 - Citrin, J. A1 - Westerhof, E. A1 - Peumans, D. A1 - M. R. de Baar AB - Old and recent experiments show that there is a direct response to the heating power of transport observed in modulated ECH experiments both in tokamaks and stellarators. This is most apparent for modulated experiments in the Large Helical Device (LHD) and in Wendelstein 7 advanced stellarator (W7-AS). In this paper we show that: 1) This power dependence can be reproduced by linear models and as such hysteresis (in flux) has no relationship to hysteresis as defined in the literature; 2) Observations of "hysteresis" (in flux) and a direct response to power can be perfectly reproduced by introducing an error in the estimated deposition profile as long as the errors redistribute the heat over a large radius; 3) Non-local models depending directly on the heating power can also explain the experimentally observed Lissajous curves (hysteresis); 4) How non-locality and deposition errors can be recognized in experiments and how they affect estimates of transport coefficients; 5) That non-linear-non-local transport models offer a path in discerning deposition errors from non-local fast transport components otherwise experimentally indistinguishable. To show all this, transport needs to be analyzed by separating the transport in a slow (diffusive) time-scale and a fast (heating/non-local) time-scale, which can only be done in the presence of perturbations. (DOI dataset, OA: 10.4121/uuid:5fcf4247-da0e-4119-adcd-fc90b85b7f03) VL - 58 IS - 10 U1 - FP U2 - IMT U5 - 29d19fbcbf73c33399d56c66632d0a33 ER - TY - JOUR T1 - Technical note on the linearity and power dependence of the diffusion coefficient in W7-AS JF - Plasma Physics and Controlled Fusion Y1 - 2017 A1 - van Berkel, M. A1 - Zwart, H. J. A1 - Hogeweij, G. M. D. A1 - M. R. de Baar AB -Transient electron temperature measurements of a step power experiment at W7-AS are reassessed by direct comparison of the up- and downward responses of the electron temperature. The analysis shows that the response at some distance to the center behaves linearly and the model predicted responses based on a power-dependent diffusion coefficient that vary from the measured step responses.

VL - 59 IS - 6 U1 -FP

U2 -IMT

U5 - 7ded523da87fbe72fd183fd1d4cceac0 ER - TY - JOUR T1 - Overview of the JET results in support to ITER JF - Nuclear Fusion Y1 - 2017 A1 - X. Litaudon A1 - Abduallev, S. A1 - Abhangi, M. A1 - Citrin, J. A1 - den Harder, N. A1 - Hogeweij, G. M. D. A1 - Jaulmes, F. A1 - Shumack, A. A1 - Tsalas, M. A1 - G. J. van Rooij A1 - et al. AB - The 2014–2016 JET results are reviewed in the light of their significance for optimising the ITER research plan for the active and non-active operation. More than 60 h of plasma operation with ITER first wall materials successfully took place since its installation in 2011. New multi-machine scaling of the type I-ELM divertor energy flux density to ITER is supported by first principle modelling. ITER relevant disruption experiments and first principle modelling are reported with a set of three disruption mitigation valves mimicking the ITER setup. Insights of the L–H power threshold in Deuterium and Hydrogen are given, stressing the importance of the magnetic configurations and the recent measurements of fine-scale structures in the edge radial electric. Dimensionless scans of the core and pedestal confinement provide new information to elucidate the importance of the first wall material on the fusion performance. H-mode plasmas at ITER triangularity (H = 1 at β N ~ 1.8 and n/n GW ~ 0.6) have been sustained at 2 MA during 5 s. The ITER neutronics codes have been validated on high performance experiments. Prospects for the coming D–T campaign and 14 MeV neutron calibration strategy are reviewed. VL - 57 IS - 10 U1 - FP U2 - IMT U5 - 38b74a22f1f4822d531a5a047a2ecc1c ER - TY - JOUR T1 - New evidence and impact of electron transport non-linearities based on new perturbative inter-modulation analysis JF - Nuclear Fusion Y1 - 2017 A1 - van Berkel, M. A1 - Kobayashi, T. A1 - Igami, H. A1 - Vandersteen, G. A1 - Hogeweij, G. M. D. A1 - Tanaka, K. A1 - Tamura, N. A1 - Zwart, H. J. A1 - Kubo, S. A1 - Ito, S. A1 - Tsuchiya, H. A1 - M. R. de Baar A1 - LHD Experiment Group AB -A new methodology to analyze non-linear components in perturbative transport experiments is introduced. The methodology has been experimentally validated in the Large Helical Device for the electron heat transport channel. Electron cyclotron resonance heating with different modulation frequencies by two gyrotrons has been used to directly quantify the amplitude of the non-linear component at the inter-modulation frequencies. The measurements show significant quadratic non-linear contributions and also the absence of cubic and higher order components. The non-linear component is analyzed using the Volterra series, which is the non-linear generalization of transfer functions. This allows us to study the radial distribution of the non-linearity of the plasma and to reconstruct linear profiles where the measurements were not distorted by non-linearities. The reconstructed linear profiles are significantly different from the measured profiles, demonstrating the significant impact that non-linearity can have.

VL - 57 IS - 12 U1 -FP

U2 -IMT

U5 - 253a5b36e4c90b5327e0cb8eb01a914e ER - TY - JOUR T1 - Control of the tokamak safety factor profile with time-varying constraints using MPC JF - Nuclear Fusion Y1 - 2015 A1 - Maljaars, E. A1 - Felici, F. A1 - M.R. de Baar A1 - van Dongen, J. A1 - Hogeweij, G. M. D. A1 - P. J. M. Geelen A1 - Steinbuch, M. AB -A controller is designed for the tokamak safety factor profile that takes real-time-varying operational and physics limits into account. This so-called model predictive controller (MPC) employs a prediction model in order to compute optimal control inputs that satisfy the given limits. The use of linearized models around a reference trajectory results in a quadratic programming problem that can easily be solved online. The performance of the controller is analysed in a set of ITER L-mode scenarios simulated with the non-linear plasma transport code RAPTOR. It is shown that the controller can reduce the tracking error due to an overestimation or underestimation of the modelled transport, while making a trade-off between residual error and amount of controller action. It is also shown that the controller can account for a sudden decrease in the available actuator power, while providing warnings ahead of time about expected violations of operational and physics limits. This controller can be extended and implemented in existing tokamaks in the near future.

VL - 55 IS - 2 U1 -FP

U2 -TP

U5 - 7d44399eb46a1150d2f9df7406831f56 ER - TY - JOUR T1 - Electromagnetic stabilization of tokamak microturbulence in a high- β regime JF - Plasma Physics and Controlled Fusion Y1 - 2015 A1 - Citrin, J. A1 - J. Garcia A1 - T. Görler A1 - Jenko, F. A1 - Mantica, P. A1 - Told, D. A1 - Bourdelle, C. A1 - Hatch, D. R. A1 - Hogeweij, G. M. D. A1 - Johnson, T. A1 - Pueschel, M. J. A1 - Schneider, M. AB -The impact of electromagnetic stabilization and flow shear stabilization on ITG turbulence is investigated. Analysis of a low- β JET L-mode discharge illustrates the relation between ITG stabilization and proximity to the electromagnetic instability threshold. This threshold is reduced by suprathermal pressure gradients, highlighting the effectiveness of fast ions in ITG stabilization. Extensive linear and nonlinear gyrokinetic simulations are then carried out for the high- β JET hybrid discharge 75225, at two separate locations at inner and outer radii. It is found that at the inner radius, nonlinear electromagnetic stabilization is dominant and is critical for achieving simulated heat fluxes in agreement with the experiment. The enhancement of this effect by suprathermal pressure also remains significant. It is also found that flow shear stabilization is not effective at the inner radii. However, at outer radii the situation is reversed. Electromagnetic stabilization is negligible while the flow shear stabilization is significant. These results constitute the high- β generalization of comparable observations found at low- β at JET. This is encouraging for the extrapolation of electromagnetic ITG stabilization to future devices. An estimation of the impact of this effect on the ITER hybrid scenario leads to a 20% fusion power improvement.

VL - 57 IS - 1 U1 -FP

U2 -CPP-HT

U5 - e0af60be4a46d18811b729f6beb4c203 ER - TY - JOUR T1 - Integrated core–SOL–divertor modelling for ITER including impurity: effect of tungsten on fusion performance in H-mode and hybrid scenario JF - Nuclear Fusion Y1 - 2015 A1 - Zagorski, R. A1 - Voitsekhovitch, I. A1 - Ivanova-Stanik, I. A1 - Kochl, F. A1 - da Silva Aresta Belo, P. A1 - Fable, E. A1 - J. Garcia A1 - Garzotti, L. A1 - Hobirk, J. A1 - Hogeweij, G. M. D. A1 - Joffrin, E. A1 - X. Litaudon A1 - Polevoi, A. R. A1 - Telesca, G. A1 - JET Contributors AB -The compatibility of two operational constraints—operation above the L–H power threshold and at low power to divertor—is examined for ITER long pulse H-mode and hybrid scenarios in integrated core–scrape off layer (SOL)–divertor modelling including impurities (intrinsic Be, He, W and seeded Ne). The core thermal, particle and momentum transport is simulated with the GLF23 transport model tested in the self-consistent simulations of temperatures, density and toroidal rotation velocity in JET hybrid discharges and extrapolated to ITER. The beneficial effect of toroidal rotation velocity on fusion gain is shown. The sensitivity studies with respect to operational (separatrix and pedestal density, Ne gas puff) and unknown physics (W convective velocity and perpendicular diffusion in SOL as well as W prompt re-deposition) parameters are performed to determine their influence on the operational window and fusion gain.

VL - 55 IS - 5 U1 -FP

U2 -CPP-HT

U5 - 88a082ba90add46139f1b78733b8692b ER - TY - JOUR T1 - ITER-like current ramps in JET with ILW: experiments, modelling and consequences for ITER JF - Nuclear Fusion Y1 - 2015 A1 - Hogeweij, G. M. D. A1 - Calabro, G. A1 - Sips, A.C.C. A1 - Maggi, C. F. A1 - G. M. De Tommasi A1 - Joffrin, E. A1 - Loarte, A. A1 - Maviglia, F. A1 - Mlynar, J. A1 - Rimini, F. G. A1 - Putterich, T. A1 - JET-EFDA Contributors AB -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

VL - 55 SN - 0029-5515 UR - http://www.iop.org/Jet/article?EFDP13051&EFDP13060 IS - 1 U1 -FP

U2 -CPP-HT

U5 - 8c52ed39444843fd7e56f57b21710499 ER - TY - JOUR T1 - Impact of W on scenario simulations for ITER JF - Nuclear Fusion Y1 - 2015 A1 - Hogeweij, G. M. D. A1 - Leonov, V. A1 - Schweinzer, J. A1 - Sips, A.C.C. A1 - Angioni, C. A1 - Calabro, G. A1 - Dux, R. A1 - Kallenbach, A. A1 - Lerche, E. A1 - Maggi, C. A1 - Putterich, T. A1 - ITPA Integrated Operating Scenarios topical group A1 - ASDEX Upgrade Team A1 - JET Contributors AB - 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 I p 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. VL - 55 IS - 6 U1 - FP U2 - CPP-HT U5 - d8553e78fee23a346493abe41c17a811 ER - TY - JOUR T1 - Ion temperature profile stiffness: non-linear gyrokinetic simulations and comparison with experiment JF - Nuclear Fusion Y1 - 2014 A1 - Citrin, J. A1 - Jenko, F. A1 - Mantica, P. A1 - Told, D. A1 - Bourdelle, C. A1 - Dumont, R. A1 - J. Garcia A1 - Haverkort, J. W. A1 - Hogeweij, G. M. D. A1 - Johnson, T. A1 - Pueschel, M. J. AB -Recent experimental observations at JET show evidence of reduced ion temperature profile stiffness. An extensive set of nonlinear gyrokinetic simulations are performed based on the experimental discharges, investigating the physical mechanism behind the observations. The impact on the ion heat flux of various parameters that differ within the data-set are explored. These parameters include the safety factor, magnetic shear, toroidal flow shear, effect of rotation on the magnetohydrodynamic equilibrium, R/L-n, beta(e), Z(eff), T-e/T-i, and the fast-particle content. While previously hypothesized to be an important factor in the stiffness reduction, the combined effect of toroidal flow shear and low magnetic shear is not predicted by the simulations to lead to a significant reduction in ion heat flux, due both to an insufficient magnitude of flow shear and significant parallel velocity gradient destabilization. It is however found that nonlinear electromagnetic effects due to both thermal and fast-particle pressure gradients, even at low beta(e), can significantly reduce the ion heat flux, and is a key factor in explaining the experimental observations. A total of four discharges are examined, at both inner and outer radii. For all cases studied, the simulated and experimental ion heat flux values agree within reasonable variations of input parameters around the experimental uncertainties.

VL - 54 SN - 0029-5515; 1741-4326 IS - 2 U1 -FP

U2 -CPP-HT

U5 - ba87938e30199199a6a17bf4846326c5 ER - TY - JOUR T1 - Comparison of bifurcation dynamics of turbulent transport models for the L-H transition JF - Physics of Plasmas Y1 - 2014 A1 - Weymiens, W. A1 - Paquay, S. A1 - de Blank, H. J. A1 - Hogeweij, G. M. D. AB -In more than three decades, a large amount of models and mechanisms have been proposed to describe a very beneficial feature of magnetically confined fusion plasmas: the L-H transition. Bifurcation theory can be used to compare these different models based on their dynamical transition structure. In this paper, we employ bifurcation theory to distinguish two fundamentally different descriptions of the interaction between turbulence levels and sheared flows. The analytic bifurcation analysis characterises the parameter space structure of the transition dynamics. Herewith, in these models three dynamically different types of transitions are characterised, sharp transitions, oscillatory transitions, and smooth transitions. One of the two models has a very robust transition structure and is therefore likely to be more accurate for such a robust phenomenon as the L-H transition. The other model needs more fine-tuning to get non-oscillatory transitions. These conclusions from the analytic bifurcation analysis are confirmed by dedicated numerical simulations, with the newly developed code Bifurcator

VL - 21 U1 -FP

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U5 - 8f4ef29f800c4a5ed648a8709306a54f ER - TY - JOUR T1 - Estimation of the thermal diffusion coefficient in fusion plasmas taking frequency measurement uncertainties into account JF - Plasma Physics and Controlled Fusion Y1 - 2014 A1 - van Berkel, M. A1 - Zwart, H. J. A1 - Hogeweij, G. M. D. A1 - Vandersteen, G. A1 - van den Brand, H. A1 - M.R. de Baar A1 - ASDEX Upgrade Team AB -In this paper, the estimation of the thermal diffusivity from perturbative experiments in fusion plasmas is discussed. The measurements used to estimate the thermal diffusivity suffer from stochastic noise. Accurate estimation of the thermal diffusivity should take this into account. It will be shown that formulas found in the literature often result in a thermal diffusivity that has a bias (a difference between the estimated value and the actual value that remains even if more measurements are added) or have an unnecessarily large uncertainty. This will be shown by modeling a plasma using only diffusion as heat transport mechanism and measurement noise based on ASDEX Upgrade measurements. The Fourier coefficients of a temperature perturbation will exhibit noise from the circular complex normal distribution (CCND). Based on Fourier coefficients distributed according to a CCND, it is shown that the resulting probability density function of the thermal diffusivity is an inverse non-central chi-squared distribution. The thermal diffusivity that is found by sampling this distribution will always be biased, and averaging of multiple estimated diffusivities will not necessarily improve the estimation. Confidence bounds are constructed to illustrate the uncertainty in the diffusivity using several formulas that are equivalent in the noiseless case. Finally, a different method of averaging, that reduces the uncertainty significantly, is suggested. The methodology is also extended to the case where damping is included, and it is explained how to include the cylindrical geometry.

VL - 56 IS - 10 U1 -FP

U2 -TP

U5 - f3be4e9703331bf6d0a635c164885fdc ER - TY - JOUR T1 - Explicit approximations to estimate the perturbative diffusivity in the presence of convectivity and damping. I. Semi-infinite slab approximations JF - Physics of Plasmas Y1 - 2014 A1 - van Berkel, M. A1 - Zwart, H. J. A1 - Tamura, N. A1 - Hogeweij, G. M. D. A1 - Inagaki, S. A1 - M.R. de Baar A1 - Ida, K. KW - approximation theory KW - convection KW - damping KW - diffusion KW - plasma simulation KW - plasma temperature KW - plasma waves AB -In this paper, a number of new approximations are introduced to estimate the perturbative diffusivity (χ), convectivity (V), and damping (τ) in cylindrical geometry. For this purpose, the harmonic components of heat waves induced by localized deposition of modulated power are used. The approximations are based on semi-infinite slab approximations of the heat equation. The main result is the approximation of χ under the influence of V and τ based on the phase of two harmonics making the estimate less sensitive to calibration errors. To understand why the slab approximations can estimate χ well in cylindrical geometry, the relationships between heat transport models in slab and cylindrical geometry are studied. In addition, the relationship between amplitude and phase with respect to their derivatives, used to estimate χ, is discussed. The results are presented in terms of the relative error for the different derived approximations for different values of frequency, transport coefficients, and dimensionless radius. The approximations show a significant region in which χ, V, and τ can be estimated well, but also regions in which the error is large. Also, it is shown that some compensation is necessary to estimate V and τ in a cylindrical geometry. On the other hand, errors resulting from the simplified assumptions are also discussed showing that estimating realistic values for V and τ based on infinite domains will be difficult in practice. This paper is the first part (Part I) of a series of three papers. In Part II and Part III, cylindrical approximations based directly on semi-infinite cylindrical domain (outward propagating heat pulses) and inward propagating heat pulses in a cylindrical domain, respectively, will be treated.

VL - 21 IS - 11 U1 -FP

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U5 - 09b280def3a08a562651dc157d2df688 ER - TY - JOUR T1 - Explicit approximations to estimate the perturbative diffusivity in the presence of convectivity and damping. II. Semi-infinite cylindrical approximations JF - Physics of Plasmas Y1 - 2014 A1 - van Berkel, M. A1 - Hogeweij, G. M. D. A1 - Tamura, N. A1 - Zwart, H. J. A1 - Inagaki, S. A1 - M.R. de Baar A1 - Ida, K. KW - approximation theory KW - convection KW - damping KW - diffusion KW - heat transfer KW - plasma transport processes AB -In this paper, a number of new explicit approximations are introduced to estimate the perturbative diffusivity (χ), convectivity (V), and damping (τ) in a cylindrical geometry. For this purpose, the harmonic components of heat waves induced by localized deposition of modulated power are used. The approximations are based upon the heat equation in a semi-infinite cylindrical domain. The approximations are based upon continued fractions, asymptotic expansions, and multiple harmonics. The relative error for the different derived approximations is presented for different values of frequency, transport coefficients, and dimensionless radius. Moreover, it is shown how combinations of different explicit formulas can yield good approximations over a wide parameter space for different cases, such as no convection and damping, only damping, and both convection and damping. This paper is the second part (Part II) of a series of three papers. In Part I, the semi-infinite slab approximations have been treated. In Part III, cylindrical approximations are treated for heat waves traveling towards the center of the plasma.

VL - 21 IS - 11 U1 -FP

U2 -CPP-HT

U5 - b8ba059ce7e27cc86b8a51d9285747a9 ER - TY - JOUR T1 - Explicit approximations to estimate the perturbative diffusivity in the presence of convectivity and damping. III. Cylindrical approximations for heat waves traveling inwards JF - Physics of Plasmas Y1 - 2014 A1 - van Berkel, M. A1 - Tamura, N. A1 - Hogeweij, G. M. D. A1 - Zwart, H. J. A1 - Inagaki, S. A1 - M.R. de Baar A1 - Ida, K. KW - approximation theory KW - Bessel functions KW - boundary-value problems KW - convection KW - damping KW - plasma transport processes AB -In this paper, a number of new explicit approximations are introduced to estimate the perturbative diffusivity (χ), convectivity (V), and damping (τ) in cylindrical geometry. For this purpose, the harmonic components of heat waves induced by localized deposition of modulated power are used. The approximations are based on the heat equation in cylindrical geometry using the symmetry (Neumann) boundary condition at the plasma center. This means that the approximations derived here should be used only to estimate transport coefficients between the plasma center and the off-axis perturbative source. If the effect of cylindrical geometry is small, it is also possible to use semi-infinite domain approximations presented in Part I and Part II of this series. A number of new approximations are derived in this part, Part III, based upon continued fractions of the modified Bessel function of the first kind and the confluent hypergeometric function of the first kind. These approximations together with the approximations based on semi-infinite domains are compared for heat waves traveling towards the center. The relative error for the different derived approximations is presented for different values of the frequency, transport coefficients, and dimensionless radius. Moreover, it is shown how combinations of different explicit formulas can be used to estimate the transport coefficients over a large parameter range for cases without convection and damping, cases with damping only, and cases with convection and damping. The relative error between the approximation and its underlying model is below 2% for the case, where only diffusivity and damping are considered. If also convectivity is considered, the diffusivity can be estimated well in a large region, but there is also a large region in which no suitable approximation is found. This paper is the third part (Part III) of a series of three papers. In Part I, the semi-infinite slab approximations have been treated. In Part II, cylindrical approximations are treated for heat waves traveling towards the plasma edge assuming a semi-infinite domain.

VL - 21 IS - 11 U1 -FP

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U5 - 78c3ab5e5ee32cfed4ba7cc0a4c84c2d ER - TY - JOUR T1 - Modelling of JET hybrid scenarios with GLF23 transport model: E × B shear stabilization of anomalous transport JF - Nuclear Fusion Y1 - 2014 A1 - Voitsekhovitch, I. A1 - da Silva Aresta Belo, P. A1 - Citrin, J. A1 - Fable, E. A1 - Ferreira, J. A1 - J. Garcia A1 - Garzotti, L. A1 - Hobirk, J. A1 - Hogeweij, G. M. D. A1 - Joffrin, E. A1 - Kochl, F. A1 - X. Litaudon A1 - Moradi, S. A1 - Nabais, F. A1 - JET-EFDA Contributors A1 - EU-ITM ITER Scenario Modelling group KW - E x B shear stabilization KW - hybrid scenario KW - transport modelling AB -The E × B shear stabilization of anomalous transport in JET hybrid discharges is studied via self-consistent predictive modelling of electron and ion temperature, ion density and toroidal rotation velocity performed with the GLF23 model. The E × B shear stabilization factor (parameter α E in the GLF23 model) is adjusted to predict accurately the four simulated quantities under different experimental conditions, and the uncertainty in α E determined by 15% deviation between simulated and measured quantities is estimated. A correlation of α E with toroidal rotation velocity and E × B shearing rate is found in the low density plasmas, suggesting that the turbulence quench rule may be more complicated than assumed in the GLF23 model with constant α E . For the selected discharges the best predictive accuracy is obtained by using weak/no E × B shear stabilization (i.e. α E ≈ 0) at low toroidal angular frequency (Ω < 60 krad s −1 ), even in the scenarios with the current overshoot, and α E = 0.9 at high frequency (Ω > 100 krad s −1 ). Interestingly, a weak E × B shear stabilization of anomalous transport is found in the medium density strongly rotating discharge. An importance of linear β e stabilization in this discharge is estimated and compared to the low density discharge with equally high β e . The toroidal rotation velocity is well predicted here by assuming that the momentum diffusion coefficient is a fraction of thermal ion diffusivity. Taking into account the α E and Prandtl number with their uncertainties determined in the modelling of JET hybrid discharges, the performance of ITER hybrid scenario with optimized heat mix (33 MW of NBI and 20 MW of ECCD) is estimated showing the importance of toroidal rotation for achieving Q > 5.

VL - 54 UR - http://www.iop.org/Jet/article?EFDP13041&EFDP13048 IS - 9 U1 -FP

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U5 - da2cf17455ede8d649edf343a88bdd87 ER - TY - JOUR T1 - Numerical optimization of actuator trajectories for ITER hybrid scenario profile evolution JF - Plasma Physics and Controlled Fusion Y1 - 2014 A1 - van Dongen, J. A1 - Felici, F. A1 - Hogeweij, G. M. D. A1 - Geelen, P. A1 - Maljaars, E. AB - Optimal actuator trajectories for an ITER hybrid scenario ramp-up are computed using a numerical optimization method. For both L-mode and H-mode scenarios, the time trajectory of plasma current, EC heating and current drive distribution is determined that minimizes a chosen cost function, while satisfying constraints. The cost function is formulated to reflect two desired properties of the plasma q profile at the end of the ramp-up. The first objective is to maximize the ITG turbulence threshold by maximizing the volume-averaged s / q ratio. The second objective is to achieve a stationary q profile by having a flat loop voltage profile. Actuator and physics-derived constraints are included, imposing limits on plasma current, ramp rates, internal inductance and q profile. This numerical method uses the fast control-oriented plasma profile evolution code RAPTOR, which is successfully benchmarked against more complete CRONOS simulations for L-mode and H-mode mode ITER hybrid scenarios. It is shown that the optimized trajectories computed using RAPTOR also result in an improved ramp-up scenario for CRONOS simulations using the same input trajectories. Furthermore, the optimal trajectories are shown to vary depending on the precise timing of the L–H transition. VL - 56 SN - 0741-3335 IS - 12 U1 - FP U2 - CPP-HT U5 - 66153de32ffd1ed712c0055e71e44bf6 ER - TY - JOUR T1 - The influence of an ITER-like wall on disruptions at JET JF - Physics of Plasmas Y1 - 2014 A1 - de Vries, P. C. A1 - Baruzzo, M. A1 - Hogeweij, G. M. D. A1 - Jachmich, S. A1 - Joffrin, E. A1 - Lomas, P. J. A1 - Matthews, G. F. A1 - Murari, A. A1 - Nunes, I. A1 - Putterich, T. A1 - Reux, C. A1 - Vega, J. A1 - JET-EFDA Contributors AB - In order to preserve the integrity of large tokamaks such as ITER, the number of disruptions has to be limited. JET has operated previously with a low frequency of disruptions (i.e., disruption rate) of 3.4% [P. C. de Vries et al., Nucl. Fusion 51, 053018 (2011)]. The start of operations with the new full-metal ITER-like wall at JET showed a marked rise in the disruption rate to 10%. A full survey was carried out to identify the root causes, the chain-of-events and classifying each disruption, similar to a previous analysis for carbon-wall operations. It showed the improvements made to avoid various disruption classes, but also indicated those disruption types responsible for the enhanced disruption rate. The latter can be mainly attributed to disruptions due to too high core radiation but also due to density control issues and error field locked modes. Detailed technical and physics understanding of disruption causes is essential for devising optimized strategies to avoid or mitigate these events. VL - 21 UR - http://scitation.aip.org/content/aip/journal/pop/21/5/10.1063/1.4872017 U1 - FP U2 - PDG U5 - 928816f24cf7568765c40767a08e9e4c ER - TY - JOUR T1 - Nonlinear Stabilization of Tokamak Microturbulence by Fast Ions JF - Physical Review Letters Y1 - 2013 A1 - Citrin, J. A1 - Jenko, F. A1 - Mantica, P. A1 - Told, D. A1 - Bourdelle, C. A1 - J. Garcia A1 - Haverkort, J. W. A1 - Hogeweij, G. M. D. A1 - Johnson, T. A1 - Pueschel, M. J. AB -Nonlinear electromagnetic stabilization by suprathermal pressure gradients found in specific regimes is shown to be a key factor in reducing tokamak microturbulence, augmenting significantly the thermal pressure electromagnetic stabilization. Based on nonlinear gyrokinetic simulations investigating a set of ion heat transport experiments on the JET tokamak, described by Mantica et al. [ Phys. Rev. Lett. 107 135004 (2011)], this result explains the experimentally observed ion heat flux and stiffness reduction. These findings are expected to improve the extrapolation of advanced tokamak scenarios to reactor relevant regimes.

PB - American Physical Society VL - 111 U1 -FP

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U5 - b79552ae74a4f49813ab40d1fa18d6b3 ER - TY - Generic T1 - Maximum Likelihood Estimation of diffusion and convection in tokamaks using infinite domains T2 - 2013 IEEE Multi-Conference on Systems and Control (MSC 2013) Y1 - 2013 A1 - van Berkel, M. A1 - Vandersteen, G. A1 - Zwart, H. A1 - Hogeweij, G. M. D. A1 - de M. Baar JF - 2013 IEEE Multi-Conference on Systems and Control (MSC 2013) PB - IEEE Conference Publications CY - Hyderabad, India N1 - 2013/08/30 U1 -FP

U2 -TP

U5 - 4d7f862dabf020d1474c73cfbb342d64 ER - TY - JOUR T1 - Bifurcation theory of a one-dimensional transport model for the L-H transition JF - Physics of Plasmas Y1 - 2013 A1 - Weymiens, W. A1 - de Blank, H. J. A1 - Hogeweij, G. M. D. KW - bifurcation KW - plasma density KW - plasma instability KW - plasma magnetohydrodynamics KW - plasma theory KW - plasma toroidal confinement KW - plasma transport processes KW - plasma turbulence KW - Tokamak devices PB - AIP VL - 20 U1 -FP

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U5 - a0badd3eefe13d34bf4057e7d08797f3 ER - TY - JOUR T1 - Global and local gyrokinetic simulations of high-performance discharges in view of ITER JF - Nuclear Fusion Y1 - 2013 A1 - Jenko, F. A1 - Told, D. A1 - T. Görler A1 - Citrin, J. A1 - Bañón Navarro, A. A1 - Bourdelle, C. A1 - Brunner, S. A1 - Conway, G. A1 - T. Dannert A1 - Doerk, H. A1 - Hatch, D. R. A1 - Haverkort, J. W. A1 - Hobirk, J. A1 - Hogeweij, G. M. D. A1 - Mantica, P. A1 - Pueschel, M. J. A1 - Sauter, O. A1 - Villard, L. A1 - Wolfrum, E. A1 - ASDEX Upgrade Team AB -One of the key challenges for plasma theory and simulation in view of ITER is to enhance the understanding and predictive capability concerning high-performance discharges. This involves, in particular, questions about high- β operation, ion temperature profile stiffness, and the physics of transport barriers. The goal of this contribution is to shed light on these issues by means of physically comprehensive ab initio simulations with the global gyrokinetic code GENE, applied to discharges in TCV, ASDEX Upgrade, and JET—with direct relevance to ITER.

VL - 53 U1 -FP

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U5 - b0a2b4afe7068d0bbd3b37e5f2c39afe ER - TY - JOUR T1 - Modelling of hybrid scenario: from present-day experiments towards ITER JF - Nuclear Fusion Y1 - 2013 A1 - X. Litaudon A1 - Voitsekhovitch, I. A1 - Artaud, J. F. A1 - Belo, P. A1 - Bizarro, J. P. S. A1 - Casper, T. A1 - Citrin, J. A1 - Fable, E. A1 - Ferreira, J. A1 - J. Garcia A1 - Garzotti, L. A1 - Giruzzi, G. A1 - Hobirk, J. A1 - Hogeweij, G. M. D. A1 - Imbeaux, F. A1 - Joffrin, E. A1 - Koechl, F. A1 - Liu, F. A1 - Lonnroth, J. A1 - Moreau, D. A1 - Parail, V. A1 - Schneider, M. A1 - Snyder, P. B. A1 - ASDEX Upgrade Team A1 - JET-EFDA Contributors A1 - EU-ITM ITER Scenario Modelling group AB -The ‘hybrid’ scenario is an attractive operating scenario for ITER since it combines long plasma duration with the reliability of the reference H-mode regime. We review the recent European modelling effort carried out within the Integrated Scenario Modelling group which aims at (i) understanding the underlying physics of the hybrid regime in ASDEX-Upgrade and JET and (ii) extrapolating them towards ITER. JET and ASDEX-Upgrade hybrid scenarios performed under different experimental conditions have been simulated in an interpretative and predictive way in order to address the current profile dynamics and its link with core confinement, the relative importance of magnetic shear, s , and E × B flow shear on the core turbulence, pedestal stability and H–L transition. The correlation of the improved confinement with an increased s / q at outer radii observed in JET and ASDEX-Upgrade discharges is consistent with the predictions based on the GLF23 model applied in the simulations of the ion and electron kinetic profiles. Projections to ITER hybrid scenarios have been carried out focusing on optimization of the heating/current drive schemes to reach and ultimately control the desired plasma equilibrium using ITER actuators. Firstly, access condition to the hybrid-like q -profiles during the current ramp–up phase has been investigated. Secondly, from the interpreted role of the s / q ratio, ITER hybrid scenario flat-top performance has been optimized through tailoring the q -profile shape and pedestal conditions. EPED predictions of pedestal pressure and width have been used as constraints in the interpretative modelling while the core heat transport is predicted by GLF23. Finally, model-based approach for real-time control of advanced tokamak scenarios has been applied to ITER hybrid regime for simultaneous magnetic and kinetic profile control.

VL - 53 U1 -FP

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U5 - 1e865efa4611fccf9bfb441bffcf216a ER - TY - JOUR T1 - Optimizing the current ramp-up phase for the hybrid ITER scenario JF - Nuclear Fusion Y1 - 2013 A1 - Hogeweij, G. M. D. A1 - Artaud, J. F. A1 - Casper, T. A. A1 - Citrin, J. A1 - Imbeaux, F. A1 - Kochl, F. A1 - X. Litaudon A1 - Voitsekhovitch, I. A1 - ITM-TF ITER Scenario Modelling Group AB -The current ramp-up phase for the ITER hybrid scenario is analysed with the CRONOS integrated modelling suite. The simulations presented in this paper show that the heating systems available at ITER allow, within the operational limits, the attainment of a hybrid q profile at the end of the current ramp-up. A reference ramp-up scenario is reached by a combination of NBI, ECCD (UPL) and LHCD. A heating scheme with only NBI and ECCD can also reach the target q profile; however, LHCD can play a crucial role in reducing the flux consumption during the ramp-up phase. The optimum heating scheme depends on the chosen transport model, and on assumptions of parameters like n e peaking, edge T e,i and Z eff . The sensitivity of the current diffusion on parameters that are not easily controlled, shows that development of real-time control is important to reach the target q profile. A first step in that direction has been indicated in this paper. Minimizing resistive flux consumption and optimizing the q profile turn out to be conflicting requirements. A trade-off between these two requirements has to be made. In this paper it is shown that fast current ramp with L-mode current overshoot is at the one extreme, i.e. the optimum q profile at the cost of increased resistive flux consumption, whereas early H-mode transition is at the other extreme.

VL - 53 U1 -FP

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U5 - 88a9a6c3198d406bdfbbe7a0eea1a68d ER - TY - JOUR T1 - Predictive analysis of q-profile influence on transport in JET and ASDEX Upgrade hybrid scenarios JF - Plasma Physics and Controlled Fusion Y1 - 2012 A1 - Citrin, J. A1 - Hobirk, J. A1 - Schneider, M. A1 - Artaud, J. F. A1 - Bourdelle, C. A1 - Crombe, K. A1 - Hogeweij, G. M. D. A1 - Imbeaux, F. A1 - Joffrin, E. A1 - Koechl, F. A1 - Stober, J. KW - BARRIERS KW - CONFINEMENT KW - DIII-D TOKAMAK KW - HIGH-PERFORMANCE DISCHARGES KW - IMPROVED H-MODE KW - ITER KW - OPERATION KW - SIMULATIONS KW - STABILITY KW - TORE-SUPRA AB -Hybrid scenarios in present machines are often characterized by improved confinement compared with the IPB98(y, 2) empirical scaling law expectations. This work concentrates on isolating the impact of increased s/q at outer radii (where s is the magnetic shear) on core confinement in low-triangularity JET and ASDEX Upgrade (AUG) experiments. This is carried out by predictive heat and particle transport modelling using the integrated modelling code CRONOS coupled to the GLF23 turbulent transport model. For both machines, discharge pairs were analysed displaying similar pedestal confinement yet significant differences in core confinement. From these comparisons, it is found that s/q shaping at outer radii may be responsible for up to similar to 50% of the relative core confinement improvement observed in these specific discharges. This relative improvement is independent of the degree of rotational shear turbulence suppression assumed in the GLF23 model. However, employing the full GLF23 rotational shear model leads to an overprediction of the ion temperatures in all discharges analysed. Additional mechanisms for core confinement improvement are discussed and estimated. Further linear threshold analysis with QuaLiKiz is carried out on both pairs of discharges. This work aims to validate recent predictions of the ITER hybrid scenario also employing CRONOS/GLF23, where a high level of confinement and resultant fusion power sensitivity to the s/q profile was found.

VL - 54 SN - 0741-3335 N1 - ISI Document Delivery No.: 947PATimes Cited: 0Cited Reference Count: 56 U1 -FP

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U5 - bda1f42e30092cecb77d13ab81a2b6af ER - TY - JOUR T1 - Quasilinear transport modelling at low magnetic shear JF - Physics of Plasmas Y1 - 2012 A1 - Citrin, J. A1 - Bourdelle, C. A1 - Cottier, P. A1 - Escande, D. F. A1 - Gurcan, O. D. A1 - Hatch, D. R. A1 - Hogeweij, G. M. D. A1 - Jenko, F. A1 - Pueschel, M. J. KW - GRADIENT-DRIVEN MODE KW - PARTICLE KW - STABILITY KW - TOKAMAKS KW - TORE-SUPRA KW - TURBULENCE AB -Accurate and computationally inexpensive transport models are vital for routine and robust predictions of tokamak turbulent transport. To this end, the QuaLiKiz [Bourdelle et al., Phys. Plasmas 14, 112501 (2007)] quasilinear gyrokinetic transport model has been recently developed. QuaLiKiz flux predictions have been validated by non-linear simulations over a wide range in parameter space. However, a discrepancy is found at low magnetic shear, where the quasilinear fluxes are significantly larger than the non-linear predictions. This discrepancy is found to stem from two distinct sources: the turbulence correlation length in the mixing length rule and an increase in the ratio between the quasilinear and non-linear transport weights, correlated with increased non-linear frequency broadening. Significantly closer agreement between the quasilinear and non-linear predictions is achieved through the development of an improved mixing length rule, whose assumptions are validated by non-linear simulations. (C) 2012 American Institute of Physics.

VL - 19 SN - 1070-664X N1 - ISI Document Delivery No.: 966JYTimes Cited: 0Cited Reference Count: 38 U1 -FP

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U5 - 79b0621193d64d70ed8b09d2bf00f7ca ER - TY - JOUR T1 - Bifurcation theory for the L-H transition in magnetically confined fusion plasmas JF - Physics of Plasmas Y1 - 2012 A1 - Weymiens, W. A1 - de Blank, H. J. A1 - Hogeweij, G. M. D. A1 - de Valenca, J. C. KW - FIELD KW - MODE KW - SHEAR KW - TOKAMAK KW - TRANSPORT KW - TURBULENCE KW - ZONAL FLOWS AB - The mathematical field of bifurcation theory is extended to be applicable to 1-dimensionally resolved systems of nonlinear partial differential equations, aimed at the determination of a certain specific bifurcation. This extension is needed to be able to properly analyze the bifurcations of the radial transport in magnetically confined fusion plasmas. This is of special interest when describing the transition from the low-energy-confinement state to the high-energy-confinement state of the radial transport in fusion plasmas (i.e., the L-H transition), because the nonlinear dynamical behavior during the transition corresponds to the dynamical behavior of a system containing such a specific bifurcation. This bifurcation determines how the three types (sharp, smooth, and oscillating) of observed L-H transitions are organized as function of all the parameters contained in the model. (C) 2012 American Institute of Physics. [http://dx.doi.org/10.1063/1.4739227] VL - 19 SN - 1070-664X IS - 7 U1 - FP U2 - CPP-HT U5 - cb4cd468aed34ad24099a3e0b0eee7c1 ER - TY - JOUR T1 - Degraded confinement and turbulence in tokamak experiments JF - Fusion Science and Technology Y1 - 2012 A1 - Hogeweij, G. M. D. KW - CONFINEMENT KW - DIII-D TOKAMAK KW - ENHANCED KW - IMPROVED MODE KW - JET KW - MAGNETIC SHEAR KW - PERFORMANCE KW - PLASMAS KW - REVERSED SHEAR DISCHARGES KW - TRANSITION KW - TRANSPORT AB - After a review on the state of tokamak transport theory, the methodology to derive experimental results will be described. Examples of confinement in ohmic plasmas and the deterioration with additional healing will be given. Some examples of improved confinement; modes will be discussed. VL - 61 SN - 1536-1055 UR - http://www.ans.org/pubs/journals/fst/a_13501 IS - 2T U1 - FP U5 - a2094639b661a13d41fcfdf3ed719dbd ER - TY - JOUR T1 - Transport studies using perturbative experiments JF - Fusion Science and Technology Y1 - 2012 A1 - Hogeweij, G. M. D. KW - COEFFICIENTS KW - ELECTRON-TRANSPORT KW - HEAT-PULSE-PROPAGATION KW - JET KW - MODULATION KW - NONLOCAL TRANSPORT KW - PARTICLE-TRANSPORT KW - PLASMAS KW - TEMPERATURE-GRADIENT KW - TOKAMAK AB - By inducing a small electron temperature perturbation in a plasma in steady state one can in principle determine the conductive and convective components of the electron heat flux, and the associated thermal diffusivity and convection velocity. The same can be done for other plasma, parameters, like density or ion temperature. In this paper experimental and analysis techniques are briefly reviewed. The fundamental question whether the fluxes are linear functions of the gradients or not is discussed. Experimental results are summarized, including so-called 'non-local' phenomena. VL - 61 SN - 1536-1055 UR - http://www.ans.org/pubs/journals/fst/a_13502 IS - 2T U1 - Fusion Physics U5 - 4ac25605b8e660d4a9c75d14c0f4be3e ER - TY - JOUR T1 - Optimizing the Current Ramp-Up Phase for Hybrid ITER Scenario JF - Plasma and Fusion Research Y1 - 2012 A1 - Hogeweij, G. M. D. A1 - Artaud, J. A1 - Casper, T. A1 - Citrin, J. A1 - Imbeaux, F. A1 - Kochl, F. A1 - X. Litaudon A1 - Voitsekhovitch, I. AB - This paper reports on a systematic effort to optimize the current ramp-up phase for the ITER hybrid scenario, and to assess the sensitivity of the results to the assumptions made. VL - 7 U1 - FP U2 - CPP-HT U4 - non, only conference fee (oral given) U5 - 430d712fff8a33fa43a15b1b9eef2d27 ER - TY - JOUR T1 - Core transport properties in JT-60U and JET identity plasmas JF - Nuclear Fusion Y1 - 2011 A1 - X. Litaudon A1 - Sakamoto, Y. A1 - de Vries, P. C. A1 - Salmi, A. A1 - Tala, T. A1 - Angioni, C. A1 - Benkadda, S. A1 - Beurskens, M. N. A. A1 - Bourdelle, C. A1 - Brix, M. A1 - Crombe, K. A1 - Fujita, T. A1 - Futatani, S. A1 - Garbet, X. A1 - Giroud, C. A1 - Hawkes, N. C. A1 - Hayashi, N. A1 - Hoang, G. T. A1 - Hogeweij, G. M. D. A1 - Matsunaga, G. A1 - Nakano, T. A1 - Oyama, N. A1 - Parail, V. A1 - Shinohara, K. A1 - Suzuki, T. A1 - Takechi, M. A1 - Takenaga, H. A1 - Takizuka, T. A1 - Urano, H. A1 - Voitsekhovitch, I. A1 - Yoshida, M. KW - BARRIERS KW - DENSITY PEAKING KW - NEOCLASSICAL TRANSPORT KW - TCV KW - TOKAMAK PLASMAS AB -The paper compares the transport properties of a set of dimensionless identity experiments performed between JET and JT-60U in the advanced tokamak regime with internal transport barrier, ITB. These International Tokamak Physics Activity, ITPA, joint experiments were carried out with the same plasma shape, toroidal magnetic field ripple and dimensionless profiles as close as possible during the ITB triggering phase in terms of safety factor, normalized Larmor radius, normalized collision frequency, thermal beta, ratio of ion to electron temperatures. Similarities in the ITB triggering mechanisms and sustainment were observed when a good match was achieved of the most relevant normalized profiles except the toroidal Mach number. Similar thermal ion transport levels in the two devices have been measured in either monotonic or non-monotonic q-profiles. In contrast, differences between JET and JT-60U were observed on the electron thermal and particle confinement in reversed magnetic shear configurations. It was found that the larger shear reversal in the very centre (inside normalized radius of 0.2) of JT-60U plasmas allowed the sustainment of stronger electron density ITBs compared with JET. As a consequence of peaked density profile, the core bootstrap current density is more than five times higher in JT-60U compared with JET. Thanks to the bootstrap effect and the slightly broader neutral beam deposition, reversed magnetic shear configurations are self-sustained in JT-60U scenarios. Analyses of similarities and differences between the two devices address key questions on the validity of the usual assumptions made in ITER steady scenario modelling, e. g. a flat density profile in the core with thermal transport barrier? Such assumptions have consequences on the prediction of fusion performance, bootstrap current and on the sustainment of the scenario.

VL - 51 SN - 0029-5515 IS - 7 N1 - ISI Document Delivery No.: 781LITimes Cited: 0Cited Reference Count: 31 U1 -FP

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U5 - 84b4980479d5b738ec4d26f9fd404727 ER - TY - JOUR T1 - Current ramps in tokamaks: from present experiments to ITER scenarios JF - Nuclear Fusion Y1 - 2011 A1 - Imbeaux, F. A1 - Citrin, J. A1 - Hobirk, J. A1 - Hogeweij, G. M. D. A1 - Kochl, F. A1 - Leonov, V. M. A1 - Miyamoto, S. A1 - Nakamura, Y. A1 - Parail, V. A1 - Pereverzev, G. A1 - Polevoi, A. A1 - Voitsekhovitch, I. A1 - Basiuk, V. A1 - Budny, R. A1 - Casper, T. A1 - Fereira, J. A1 - Fukuyama, A. A1 - J. Garcia A1 - Gribov, Y. V. A1 - Hayashi, N. A1 - Honda, M. A1 - Hutchinson, I. H. A1 - Jackson, G. A1 - Kavin, A. A. A1 - Kessel, C. E. A1 - Khayrutdinov, R. R. A1 - Labate, C. A1 - X. Litaudon A1 - Lomas, P. J. A1 - Lonnroth, J. A1 - Luce, T. A1 - Lukash, V. E. A1 - Mattei, M. A1 - Mikkelsen, D. A1 - Nunes, I. A1 - Peysson, Y. A1 - Politzer, P. A1 - Schneider, M. A1 - Sips, G. A1 - Tardini, G. A1 - Wolfe, S. M. A1 - Zhogolev, V. E. KW - DATABASE KW - DISCHARGES KW - HYBRID KW - MODEL KW - SIMULATION KW - TRANSPORT AB -In order to prepare adequate current ramp-up and ramp-down scenarios for ITER, present experiments from various tokamaks have been analysed by means of integrated modelling in view of determining relevant heat transport models for these operation phases. A set of empirical heat transport models for L-mode (namely, the Bohm-gyroBohm model and scaling based models with a specific fixed radial shape and energy confinement time factors of H(96-L) = 0.6 or H(IPB98) = 0.4) has been validated on a multi-machine experimental dataset for predicting the l(i) dynamics within +/- 0.15 accuracy during current ramp-up and ramp-down phases. Simulations using the Coppi-Tang or GLF23 models (applied up to the LCFS) overestimate or underestimate the internal inductance beyond this accuracy (more than +/- 0.2 discrepancy in some cases). The most accurate heat transport models are then applied to projections to ITER current ramp-up, focusing on the baseline inductive scenario (main heating plateau current of I(p) = 15 MA). These projections include a sensitivity study to various assumptions of the simulation. While the heat transport model is at the heart of such simulations (because of the intrinsic dependence of the plasma resistivity on electron temperature, among other parameters), more comprehensive simulations are required to test all operational aspects of the current ramp-up and ramp-down phases of ITER scenarios. Recent examples of such simulations, involving coupled core transport codes, free-boundary equilibrium solvers and a poloidal field (PF) systems controller are also described, focusing on ITER current ramp-down.

VL - 51 SN - 0029-5515 IS - 8 N1 - ISI Document Delivery No.: 818DDTimes Cited: 1Cited Reference Count: 18 U1 -FP

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U5 - 0d6030560b4587c9666dfb14d9e62d73 ER - TY - JOUR T1 - Impact of heating and current drive mix on the ITER hybrid scenario JF - Nuclear Fusion Y1 - 2010 A1 - Citrin, J. A1 - Artaud, J. F. A1 - J. Garcia A1 - Hogeweij, G. M. D. A1 - Imbeaux, F. KW - ASDEX UPGRADE KW - CONFINEMENT KW - CURRENT-DENSITY PROFILE KW - DIII-D KW - DISCHARGES KW - IMPROVED H-MODE KW - PHYSICS BASIS KW - TEMPERATURE KW - TOKAMAKS KW - TRANSPORT AB - Hybrid scenario performance in ITER is studied with the CRONOS integrated modelling suite, using the GLF23 anomalous transport model for heat transport prediction. GLF23 predicted core confinement is optimized through tailoring the q-profile shape by a careful choice of current drive actuators, affecting the transport due to the predicted dependence of the turbulence level on the absolute q-profile values and magnetic shear. A range of various heating and current drive choices are examined, as are different assumptions on the pedestal height. The optimum q-profile shape is predicted to be one that maximizes the ratio of s/q throughout the bulk of the plasma volume. Optimizing the confinement allows a minimization of the plasma density required in order to achieve a defined target fusion power of 350 MW. A lower density then allows a lower total current (I-p) at the same Greenwald fraction (f(G)), thus aiding in maintaining q > 1 as desired in a hybrid scenario, and in minimizing the flux consumption. The best performance is achieved with a combination of NBI and ECCD (e.g. 33/37 MW NBI/ECCD for a scenario with a pedestal height of 4 keV). The q-profile shape and plasma confinement properties are shown to be highly sensitive to the positioning of the ECCD deposition. Comparisons with the lower performing cases where some or all of the ECCD power is replaced with LHCD or ICRH are shown (e. g. 33/20/17 MW NBI/ECCD/LHCD or NBI/ECCD/ICRH). The inclusion of LHCD reduces confinement due to deleterious shaping of the q-profile, and the inclusion of ICRH, particularly in a stiff model, does not lead to significantly increased fusion power and furthermore does not contribute to the non-inductive current fraction. For the optimum NBI/ECCD current drive mix, the predictions show that a satisfactory ITER hybrid scenario (P-fus similar to 350 MW, Q >= 5, q(min) close to 1) may be achieved with T-ped >= 4 keV. In addition, predicted performance sensitivity analysis was carried out for several assumed parameters, such as Z(eff) and density peaking. VL - 50 SN - 0029-5515 UR -