N. C. Hawkes

Culham Centre for Fusion Energy, Abingdon-on-Thames, England, United Kingdom

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Publications (267)427.62 Total impact

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    ABSTRACT: Local electron and ion heating characteristics during merging reconnection startup on the MAST spherical tokamak have been revealed for the first time using a 130 channel YAG-TS system and a new 32 chord ion Doppler tomography diagnostic. 2D local profile measurement of $T_e$, $n_e$ and $T_i$ detect highly localized electron heating at the X point and bulk ion heating downstream. For the push merging experiment under high guide field condition, thick layer of closed flux surface formed by reconnected field sustains the heating profile for more than electron and ion energy relaxation time $\tau^E_{ei}\sim4-10$ms, both heating profiles finally form triple peak structure at the X point and downstream. Toroidal guide field mostly contributes the formation of peaked electron heating profile at the X point. The localized heating increases with higher guide field, while bulk downstream ion heating is unaffected by the change in the guide field under MAST conditions ($B_t>3B_{rec}$).
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    ABSTRACT: The replacement of the JET carbon wall (C-wall) by a Be/W ITER-like wall (ILW) has affected the plasma energy confinement. To investigate this, experiments have been performed with both the C-wall and ILW to vary the heating power over a wide range for plasmas with different shapes.
    Nuclear Fusion 01/2015; 55(5). DOI:10.1088/0029-5515/55/5/053031 · 3.24 Impact Factor
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    ABSTRACT: The phenomenon of redistribution of neutral beam fast-ions due to MHD activity in plasma has been observed on many tokamaks and more recently has been a focus of research on MAST (Turnyanskiy M. et al, 2011 Nucl. Fusion 53 053016). n=1 fishbone modes are observed to cause a large decrease in the neutron emission rate corresponding to a significant perturbation of the fast-ion population in the plasma. Theoretical work on fishbone modes states that the fast-ion distribution itself acts as the source of free energy driving the modes that cause the redistribution. Therefore a series of experiments have been carried out on MAST to investigate a range of plasma density levels at two neutral beam power levels to determine the region within this parameter space in which MHD activity and consequent fast-ion redistribution is suppressed. Analysis of these experiments shows complete suppression of MHD activity at high density with increasing activity and fast-ion redistribution at lower densities and higher NB power accompanied by strong evidence for localisation of the redistribution effect to a specific region in the plasma core. The results also indicate correlations between the form of the modelled fast-ion distribution function, the amplitude and growth rate of the fishbone modes, and the magnitude of the redistribution effect. The same analysis has been carried out on models of MAST-Upgrade baseline plasma scenarios to determine whether significant fast-ion redistribution is likely to occur in that device. A simple change to the neutral-beam injector geometry is proposed which is shown to have a significant mitigating effect in terms of the fishbone mode drive and is therefore expected to allow effective plasma heating and current drive over a wider range of plasma conditions in MAST-Upgrade.
    Nuclear Fusion 01/2015; 55:013021. DOI:10.1088/0029-5515/55/1/013021 · 3.24 Impact Factor
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    ABSTRACT: In a Tokamak the configuration of the magnetic fields remains the key element to improve performance and to maximise the scientific exploitation of the device. On the other hand, the quality of the reconstructed fields depends crucially on the measurements available. Traditionally in the least square minimisation phase of the algorithms, used to obtain the magnetic field topology, all the diagnostics are given the same weights, a part from a corrective factor taking into account the error bars. This assumption unduly penalises complex diagnostics, such as polarimetry, which have a limited number of highly significant measurements. A completely new method to choose the weights, to be given to the internal measurements of the magnetic fields for improved equilibrium reconstructions, is presented in this paper. The approach is based on various statistical indicators applied to the residuals, the difference between the actual measurements and their estimates from the reconstructed equilibrium. The potential of the method is exemplified using the measurements of the Faraday rotation derived from JET polarimeter. The results indicate quite clearly that the weights have to be determined carefully, since the inappropriate choice can have significant repercussions on the quality of the magnetic reconstruction both in the edge and in the core. These results confirm the limitations of the assumption that all the diagnostics have to be given the same weight, irrespective of the number of measurements they provide and the region of the plasma they probe.
    Review of Scientific Instruments 12/2014; 85(12):123507. DOI:10.1063/1.4904450 · 1.58 Impact Factor
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    ABSTRACT: Charge exchange spectroscopy has long been a key diagnostic tool for fusion plasmas and is well developed in devices with Carbon Plasma-Facing Components. Operation with the ITER-like wall at JET has resulted in changes to the spectrum in the region of the Carbon charge exchange line at 529.06 nm and demonstrates the need to revise the core charge exchange analysis for this line. An investigation has been made of this spectral region in different plasma conditions and the revised description of the spectral lines to be included in the analysis is presented.
    Review of Scientific Instruments 10/2014; 85(11). DOI:10.1063/1.4890118 · 1.58 Impact Factor
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    ABSTRACT: The behaviour of tungsten in the core of hybrid scenario plasmas in JET with the ITER-like wall is analysed and modelled with a combination of neoclassical and gyrokinetic codes. In these discharges, good confinement conditions can be maintained only for the first 2–3 s of the high power phase. Later W accumulation is regularly observed, often accompanied by the onset of magneto-hydrodynamical activity, in particular neoclassical tearing modes (NTMs), both of which have detrimental effects on the global energy confinement. The dynamics of the accumulation process is examined, taking into consideration the concurrent evolution of the background plasma profiles, and the possible onset of NTMs. Two time slices of a representative discharge, before and during the accumulation process, are analysed with two independent methods, in order to reconstruct the W density distribution over the poloidal cross-section. The same time slices are modelled, computing both neoclassical and turbulent transport components and consistently including the impact of centrifugal effects, which can be significant in these plasmas, and strongly enhance W neoclassical transport. The modelling closely reproduces the observations and identifies inward neoclassical convection due to the density peaking of the bulk plasma in the central region as the main cause of the accumulation. The change in W neoclassical convection is directly produced by the transient behaviour of the main plasma density profile, which is hollow in the central region in the initial part of the high power phase of the discharge, but which develops a significant density peaking very close to the magnetic axis in the later phase. The analysis of a large set of discharges provides clear indications that this effect is generic in this scenario. The unfavourable impact of the onset of NTMs on the W behaviour, observed in several discharges, is suggested to be a consequence of a detrimental combination of the effects of neoclassical transport and of the appearance of an island.
    Nuclear Fusion 08/2014; 54(8). DOI:10.1088/0029-5515/54/8/083028 · 3.24 Impact Factor
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    ABSTRACT: A sixteen channel millimeter-wave diagnostic system, covering the frequency range 30-75 GHz, has been installed on MAST [B. Lloyd et al., Nucl. Fusion 43, 1665 (2003)] and has been successfully used for both Doppler backscattering (DBS) and conventional (normal-incidence) fluctuation reflectometry. DBS has become a well-established and versatile diagnostic technique for the measurement of intermediate- k ($k_{\bot} \rho_i \sim 1$, and higher) density fluctuations and flows in magnetically confined fusion experiments. The $180^{\circ}$ backscattering for DBS requires three dimensional wave-vector matching between the launched beam and the plasma fluctuations inducing the scattering, which are expected to be highly elongated along the magnetic field. The large pitch angle in MAST means that DBS implementation depends strongly on the capability to accurately launch the probing beam at a toroidal and poloidal angle that is matched to the magnetic field at the scattering location. We report on the scattering considerations and ray tracing calculations used to optimize the design, a description of the implementation including 2D beam steering, and present initial data demonstrating measurement capabilities and comparing to optimization calculations. Initial results confirm the applicability of the design and implementation approaches, showing the dependence of scattering alignment on toroidal launch angle and demonstrating DBS is sensitive to the local magnetic field pitch angle. We also present comparisons of DBS plasma velocity measurements with charge exchange recombination and beam emission spectroscopy measurements, which show good agreement in most cases. The 2D steering is shown to enable high-k measurements with DBS, at $k_{\bot}>20\ \mathrm{cm}^{-1}$ ($k_{\bot} \rho_i>10$) for launch frequencies less than 75 GHz.
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    ABSTRACT: A comparison of the L-H power threshold (Pthr) in JET with all carbon, JET-C, and beryllium/tungsten wall (the ITER-like choice), JET-ILW, has been carried out in experiments with slow input power ramps and matched plasma shapes, divertor configuration and IP/BT pairs. The low density dependence of the L-H power threshold, namely an increase below a minimum density ne,min, which was first observed in JET with the MkII-GB divertor and C wall and subsequently not observed with the current MkII-HD geometry, is observed again with JET-ILW. At plasma densities above ne,min, Pthr is reduced by ̃30%, and by ̃40% when the radiation from the bulk plasma is subtracted (Psep), with JET-ILW compared to JET-C. At the L-H transition the electron temperature at the edge, where the pedestal later develops, is also lower with JET-ILW, for a given edge density. With JET-ILW the minimum density is found to increase roughly linearly with magnetic field, n_{e,min} \sim B_{T}^{4/5} , while the power threshold at the minimum density scales as P_{sep,\min} \sim B_{T}^{5/2} . The H-mode power threshold in JET-ILW is found to be sensitive both to variations in main plasma shape (Psep decreases with increasing lower triangularity and increases with upper triangularity) and in divertor configuration. When the data are recast in terms of Psep and Zeff or subdivertor neutral pressure a linear correlation is found, pointing to a possible role of Zeff and/or subdivertor neutral pressure in the L-H transition physics. Depending on the chosen divertor configuration, Pthr can be up to a factor of two lower than the ITPA scaling law for densities above ne,min. A shallow edge radial electric field well is observed at the L-H transition. The edge impurity ion poloidal velocity remains low, close to its L-mode values, 5 km s-1 ± 2-3 km s-1, at the L-H transition and throughout the H-mode phase, with no measureable increase within the experimental uncertainties. The edge toroidal rotation profile does not contribute to the depth of the negative Er well and thus may not be correlated with the formation of the edge transport barrier in JET.
    Nuclear Fusion 01/2014; 54(2). DOI:10.1088/0029-5515/54/2/023007 · 3.24 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; 54:013011. DOI:10.1088/0029-5515/54/1/013011 · 3.24 Impact Factor
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    ABSTRACT: One of the main approaches to thermonuclear fusion relies on confining high temperature plasmas with properly shaped magnetic fields. The determination of the magnetic topology is, therefore, essential for controlling the experiments and for achieving the required performance. In Tokamaks, the reconstruction of the fields is typically formulated as a free boundary equilibrium problem, described by the Grad-Shafranov equation in toroidal geometry and axisymmetric configurations. Unfortunately, this results in mathematically very ill posed problems and, therefore, the quality of the equilibrium reconstructions depends sensitively on the measurements used as inputs and on the imposed constraints. In this paper, it is shown how the different diagnostics (Magnetics Measurements, Polarimetry and Motional Stark Effect), together with the edge current density and plasma pressure constraints, can have a significant impact on the quality of the equilibrium on JET. Results show that both the Polarimetry and Motional Stark Effect internal diagnostics are crucial in order to obtain reasonable safety factor profiles. The impact of the edge current density constraint is significant when the plasma is in the H-mode of confinement. In this plasma scenario the strike point positions and the plasma last closed flux surface can change even by centimetres, depending on the edge constraints, with a significant impact on the remapping of the equilibrium-dependent diagnostics and of pedestal physics studies. On the other hand and quite counter intuitively, the pressure constraint can severely affect the quality of the magnetic reconstructions in the core. These trends have been verified with several JET discharges and consistent results have been found. An interpretation of these results, as interplay between degrees of freedom and available measurements, is provided. The systematic analysis described in the paper emphasizes the importance of having sufficient diagnostic inputs and of pro- erly validating the results of the codes with independent measurements.
    Review of Scientific Instruments 10/2013; 84(10):103508-103508-11. DOI:10.1063/1.4824200 · 1.58 Impact Factor
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    ABSTRACT: New diagnostic, modelling and plant capability on the Mega Ampère Spherical Tokamak (MAST) have delivered important results in key areas for ITER/DEMO and the upcoming MAST Upgrade, a step towards future ST devices on the path to fusion currently under procurement. Micro-stability analysis of the pedestal highlights the potential roles of micro-tearing modes and kinetic ballooning modes for the pedestal formation. Mitigation of edge localized modes (ELM) using resonant magnetic perturbation has been demonstrated for toroidal mode numbers n = 3, 4, 6 with an ELM frequency increase by up to a factor of 9, compatible with pellet fuelling. The peak heat flux of mitigated and natural ELMs follows the same linear trend with ELM energy loss and the first ELM-resolved Ti measurements in the divertor region are shown. Measurements of flow shear and turbulence dynamics during L–H transitions show filaments erupting from the plasma edge whilst the full flow shear is still present. Off-axis neutral beam injection helps to strongly reduce the redistribution of fast-ions due to fishbone modes when compared to on-axis injection. Low-k ion-scale turbulence has been measured in L-mode and compared to global gyro-kinetic simulations. A statistical analysis of principal turbulence time scales shows them to be of comparable magnitude and reasonably correlated with turbulence decorrelation time. Te inside the island of a neoclassical tearing mode allow the analysis of the island evolution without assuming specific models for the heat flux. Other results include the discrepancy of the current profile evolution during the current ramp-up with solutions of the poloidal field diffusion equation, studies of the anomalous Doppler resonance compressional Alfvén eigenmodes, disruption mitigation studies and modelling of the new divertor design for MAST Upgrade. The novel 3D electron Bernstein synthetic imaging shows promising first data sensitive to the edge current profile and flows.
    Nuclear Fusion 10/2013; 53(10):104008. DOI:10.1088/0029-5515/53/10/104008 · 3.24 Impact Factor
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    ABSTRACT: JET has been recently refurbished with an ITER-like Be first wall and W divertor (ILW), to study plasma wall interaction processes and integrated scenario development for ITER. With the change of the divertor material, and the related presence of radiating W ions in the plasma, several changes in the pre-existing MHD phenomenology have been observed. The experimental signature of the new MHD behaviour will be characterized in this work using high bandwidth pick up coils, fast ECE signals and fast Soft X-ray signals
    40th EPS Conference on Plasma Physics, EPS 2013, Espoo, Finland, 1st—5th July 2013; 07/2013
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    ABSTRACT: Detailed experimental studies of ion heat transport are carried out in JET to explore the Te/Ti dependence of ion heat transport in ITER relevant range of parameters (Te/Ti ≥ 1) using low rotation plasmas with dominant ion cyclotron resonance heating to avoid the coupling of the effects of Te/Ti and rotation which affected previous experiments. This experimental setup has led to an accurate determination of the ion temperature gradient (ITG) threshold at varying Te/Ti, offering unique opportunities for validation of the well-established theory of ITG driven modes. A rather mild decrease in threshold with increasing Te/Ti in the interval of ITER interest was found. The new experimental result has found good agreement with theoretical predictions based on quasi-linear fluid and linear gyrokinetic models.
    Plasma Physics and Controlled Fusion 04/2013; 55(5):055003. DOI:10.1088/0741-3335/55/5/055003 · 2.39 Impact Factor
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    ABSTRACT: JET has been recently refurbished with an ITER-like Be first wall and W divertor, to study plasma wall interaction processes for ITER. In this work the new behaviour of the MHD instabilities will be characterized in the hybrid scenario, which with the C-wall in JET achieved high energy confinement, combined with good MHD stability to NTMs and ideal kinks. The same scenario developed for the ILW has produced good confinement, but interactions are observed between MHD phenomena and impurities coming from the wall. The q=1 MHD activity with the JET C-wall showed a negligible effect on plasma confinement, except NTM triggering. In some ILW hybrid pulses at the start of the heating phase a q=1 fishbone occurs, as with the C-wall, but it is often replaced by a continuous q=1 mode, with a significant reduction of confinement. ECE measurements also highlight a change from pure kink fluctuations to islands centered on q=1. NTMs have also been observed in these plasmas. Their appearance is coincident with a flattening of electron temperature profile within the island (the effect with the C-wall), but it is also correlated with enhanced radiation from the plasma core and a slow decrease of central electron temperature.
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    ABSTRACT: The configuration of magnetic fields is an essential ingredient of tokamak physics. In modern day devices, the magnetic topology is normally derived from equilibrium codes, which solve the Grad–Shafranov equation with constraints imposed by the available measurements. On JET, the main code used for this purpose is EFIT and the more commonly used diagnostics are external pick-up coils. Both the code and the measurements present worse performance during edge localized modes (ELMs). To quantify this aspect, various statistical indicators, based on the values of the residuals and their probability distribution, are defined and calculated. They all show that the quality of EFIT reconstructions is clearly better in the absence of ELMs. To investigate the possible causes of the detrimental effects of ELMs on the reconstruction, the pick-up coils are characterized individually and both the spatial distribution and time behaviour of their residuals are analysed in detail. The coils with a faster time response are the ones reproduced less well by EFIT. The constraints of current and pressure at the separatrix are also varied but the effects of such modifications do not result in decisive improvements in the quality of the reconstructions. The interpretation of this experimental evidence is not absolutely compelling but strongly indicative of deficiencies in the physics model on which the JET reconstruction code is based.
    Plasma Physics and Controlled Fusion 10/2012; 54(10). DOI:10.1088/0741-3335/54/10/105005 · 2.39 Impact Factor
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    Plasma Physics and Controlled Fusion 09/2012; DOI:10.1088/0741-3335/54/9/095001 · 2.39 Impact Factor
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    ABSTRACT: Experiments on JET with a carbon-fibre composite wall have explored the reduction of steady-state power load in an ELMy H-mode scenario at high Greenwald fraction ∼0.8, constant power and close to the L to H transition. This paper reports a systematic study of power load reduction due to the effect of fuelling in combination with seeding over a wide range of pedestal density ((4–8) × 1019 m−3) with detailed documentation of divertor, pedestal and main plasma conditions, as well as a comparative study of two extrinsic impurity nitrogen and neon. It also reports the impact of steady-state power load reduction on the overall plasma behaviour, as well as possible control parameters to increase fuel purity. Conditions from attached to fully detached divertor were obtained during this study. These experiments provide reference plasmas for comparison with a future JET Be first wall and an all W divertor where the power load reduction is mandatory for operation.
    Nuclear Fusion 06/2012; 52(6). DOI:10.1088/0029-5515/52/6/063022 · 3.24 Impact Factor
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    ABSTRACT: Detailed experimental studies of ion heat transport have been carried out in JET exploiting the upgrade of active charge exchange spectroscopy and the availability of multi-frequency ion cyclotron resonance heating with 3He minority. The determination of ion temperature gradient (ITG) threshold and ion stiffness offers unique opportunities for validation of the well-established theory of ITG driven modes. Ion stiffness is observed to decrease strongly in the presence of toroidal rotation when the magnetic shear is sufficiently low. This effect is dominant with respect to the well-known ωE×B threshold up-shift and plays a major role in enhancing core confinement in hybrid regimes and ion internal transport barriers. The effects of Te/Ti and s/q on ion threshold are found rather weak in the domain explored. Quasi-linear fluid/gyro-fluid and linear/non-linear gyro-kinetic simulations have been carried out. Whilst threshold predictions show good match with experimental observations, some significant discrepancies are found on the stiffness behaviour.
    Plasma Physics and Controlled Fusion 11/2011; 53(12):124033. DOI:10.1088/0741-3335/53/12/124033 · 2.39 Impact Factor
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    ABSTRACT: The ITER core charge exchange recombination spectroscopy (core CXRS) diagnostic system is designed to provide experimental access to various measurement quantities in the ITER core plasma such as ion densities, temperatures and velocities. The implementation of the approved CXRS diagnostic principle on ITER faces significant challenges: First, a comparatively low CXRS signal intensity is expected, together with a high noise level due to bremsstrahlung, while the requested measurement accuracy and stability for the core CXRS system go far beyond the level commonly achieved in present-day fusion experiments. Second, the lifetime of the first mirror surface is limited due to either erosion by fast particle bombardment or deposition of impurities. Finally, the hostile technical environment on ITER imposes challenging boundary conditions for the diagnostic integration and operation, including high neutron loads, electro-magnetic loads, seismic events and a limited access for maintenance. A brief overview on the R&D and design activities for the core CXRS system is presented here, while the details are described in parallel papers.
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    ABSTRACT: The paper presents component concepts developed for the ITER core charge exchange recombination spectroscopy (cCXRS). They are based on the cCXRS layout of 2009, named as reference design option. It includes an outer shell, carrying a blanket shield module and a shielding cassette. The cassette, attached to the outer shell, carries secondary mirrors and a retractable tube. The tube holds a first mirror (M1) and a shutter.Development priority is given to the M1, M1 holder, shutter, calibration system and retractable tube. These units are the most critical from reliability point of view. The developed design approaches could be likely used as generic or prototype solutions for a cCXRS diagnostic port plug in future.Before prototyping critical components as the M1 holder and shutter, the final cCXRS conceptual design has to meet forthcoming changes in the ITER diagnostic upper port plug (UPP) and blanket system. The integrated first wall—diagnostic shield modules, first wall recession, blanket shaping and generic UPP layout have considerable impact on the layout of the cCXRS and its components. The paper presents a preliminary solution for integration of the customized reference cCXRS into expected ITER blanket and UPP layouts.
    Fusion Engineering and Design 10/2011; 86(9):2055-2059. DOI:10.1016/j.fusengdes.2011.01.086 · 1.15 Impact Factor

Publication Stats

3k Citations
427.62 Total Impact Points


  • 2004–2011
    • Culham Centre for Fusion Energy
      Abingdon-on-Thames, England, United Kingdom
  • 2009
    • Ghent University
      • Department of Applied Physics
      Gand, Flanders, Belgium
  • 2005
    • École Polytechnique Fédérale de Lausanne
      • Center for Research In Plasma Physics
      Lausanne, Vaud, Switzerland
  • 2003
    • Massachusetts Institute of Technology
      • Plasma Science and Fusion Center (PSFC)
      Cambridge, Massachusetts, United States
  • 2002
    • Princeton University
      • Princeton Plasma Physics Laboratory
      Princeton, New Jersey, United States
  • 1994–1995
    • University of Toronto
      Toronto, Ontario, Canada
  • 1987–1989
    • United Kingdom Atomic Energy Authority
      Abingdon-on-Thames, England, United Kingdom