G. Counsell

Fusion for Energy, Barcino, Catalonia, Spain

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Publications (139)178.47 Total impact

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    ABSTRACT: In a tokamak, plasma-wall interactions can result in production of dust. During operation, the tritium present in the Vacuum Vessel (VV) can then be trapped in the in-vessel materials but also in dust. The vacuum vessel represents the first confinement barrier to this radioactive material. In the event of a postulated accident involving ingress of steam into the VV, hydrogen could in principle be produced by chemical reaction with hot metal and dust. If the ingress of air into the VV is also postulated, reaction of air with hydrogen and/or dust cannot be completely excluded and could lead to a possible explosion which could challenge the VV tightness. In order to prevent such accidents and their radiological consequences, limitations on the accumulation of dust and tritium in the VV and on the air ingress are imposed. Correlatively, ITER has defined a strategy for the control of in-vessel dust and tritium inventories based on both measurement and removal techniques. In this context, this paper reports on the status of tasks under F4E responsibility aiming at developing some of the measurement systems and necessary R&D for the validation of the ITER strategy.
    2011 IEEE/NPSS 24th Symposium on Fusion Engineering; 06/2011
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    ABSTRACT: This paper reports on the current status of integration of ITER microwave diagnostics, such as ECE, reflectometry systems and Collective Thomson scattering, and gives an outlook on the upcoming technical and design activity. Some open issues are addressed and discussed.
    02/2011;
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    ABSTRACT: In a Tokamak vacuum vessel, plasma–wall interactions can result in the production of radioactive dust and H isotopes (including tritium) can be trapped both in in-vessel material and in dust. The vacuum vessel represents the most important confinement barrier to this radioactive material. In the event of an accident involving ingress of steam to the vacuum vessel, hydrogen could be produced by chemical reactions with hot metal and dust. Hydrogen isotopes could also be desorbed from in-vessel components, e.g. cryopumps. In events where an ingress of air to the vacuum vessel occurs, reaction of the air with hydrogen and/or dust therefore cannot be completely excluded. Due to the radiological risks highlighted by the safety evaluation studies for ITER in normal conditions (e.g. in-vessel maintenance chronic release) and accidental ones (e.g. challenge of vacuum vessel tightness in the event of a hydrogen/dust explosion with air), limitations on the accumulation of dust and tritium in the vacuum vessel are imposed as well as controls over the maximum extent of the quantity of accidental air ingress. ITER IO has defined a strategy for the control of in-vessel dust and tritium inventories below the safety limits based primarily on the measurement and removal of dust and tritium. In this context, this paper will report on the efforts under F4E responsibility to develop a number of the new ITER baseline systems. In particular this paper, after a review of safety constraints and ITER strategy, provides the status of: (1) tasks being launched on diagnostics for in-vessel dust inventory measurement, (2) experiments to enrich the data about the effectiveness of desorption of tritium from Be at 350 °C (divertor baking aiming to release significant amount of tritium trapped in Be co-deposit), (3) on-going R&D programme (experimental and numerical simulation) at FZK, CEA and ENEA on in-vacuum vessel H2 dust explosion.
    Fusion Engineering and Design 01/2011; 86:2753-2757. · 0.84 Impact Factor
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    ABSTRACT: ITER will explore a plasma parameter envelope currently not available in tokamaks. This will require a set of diagnostics that can follow this envelope. To implement these diagnostics in a reliable and robust way requires development of current techniques in many areas to make them applicable to ITER: they need to be operable in the ITER environment and satisfy the physics and engineering requirements. In some cases, the exploitation of new techniques will be required. While much work has been carried out in this area, significant further work remains to bring the system to implementation.
    01/2011;
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    ABSTRACT: The International Thermonuclear Experimental Reactor will have wide angle viewing systems and a divertor thermography diagnostic, which shall provide infrared coverage of the divertor and large parts of the first wall surfaces with spatial and temporal resolution adequate for operational purposes and higher resolved details of the divertor and other areas for physics investigations. We propose specifications for each system such that they jointly respond to the requirements. Risk analysis driven priorities for future work concern mirror degradation, interfaces with other diagnostics, radiation damage to refractive optics, reflections, and the development of calibration and measurement methods for varying optical and thermal target properties.
    Review of Scientific Instruments 11/2010; · 1.60 Impact Factor
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    ABSTRACT: Active (beam-based) spectroscopic measurements are intended to provide a number of crucial parameters for the ITER device being built in Cadarache, France. These measurements include the determination of impurity ion temperatures, absolute densities, and velocity profiles, as well as the determination of the plasma current density profile. Because ITER will be the first experiment to study long timescale (∼1 h) fusion burn plasmas, of particular interest is the ability to study the profile of the thermalized helium ash resulting from the slowing down and confinement of the fusion alphas. These measurements will utilize both the 1 MeV heating neutral beams and a dedicated 100 keV hydrogen diagnostic neutral beam. A number of separate instruments are being designed and built by several of the ITER partners to meet the different spectroscopic measurement needs and to provide the maximum physics information. In this paper, we describe the planned measurements, the intended diagnostic ensemble, and we will discuss specific physics and engineering challenges for these measurements in ITER.
    The Review of scientific instruments 10/2010; 81(10):10D725. · 1.52 Impact Factor
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    ABSTRACT: Dust samples collected in the MAST tokamak have been characterized. Mass measurements have been correlated to regions of collection. Electron microscopy has revealed the presence of large quantities of carbon nanoparticles produced in gas phase as well as the presence of rolled-up carbon thin layers whatever the collection region. Shape, structure and chemical composition have been established by means of complementary diagnostics such as electron microscopy, electron diffraction, energy dispersed X-ray spectroscopy, micro-Raman spectroscopy and infrared absorption spectroscopy.
    Journal of Nuclear Materials 01/2010; 401(1):130-137. · 1.21 Impact Factor
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    ABSTRACT: Dust will be present in ITER and will be an issue in terms of safety. The ITER strategy for dust management is based among others on the definition and respect of dust inventory limits. To ensure that the safety limits are fulfilled, dust diagnostics and removal techniques need to be developed considering the Tokamak environment constraints (magnetic field, radiation, vacuum and temperature). This paper presents possible dust inventory build-up monitoring strategies for different periods of the machine operation (during/between pulses and during short or long maintenance periods). The strategies rely on the use of a set of complementary techniques for dust diagnostics and removal.
    Journal of Nuclear Materials 04/2009; 386-388:882. · 1.21 Impact Factor
  • E Delchambre, G Counsell, A Kirk
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    ABSTRACT: The non-uniformity of the target temperature due to micrometric hot spots (Hermann et al 2004 Phys. Scr. T 111 98) is an explanation for the experimental fact that near-infrared measurements yield higher temperature values than mid-infrared measurements (Hildebrandt et al 2003 InfraMation 2003 Proc. (Las Vegas, USA, October 2003), Delchambre et al 2005 J. Nucl. Mater. 337–339 1069). The issue of micrometric hot spot disturbance in the surface temperature (Tsurf) measurement and heat load calculation is addressed in this paper. The theoretical investigation at 3, 5 and 12 µm and experiments in the range 3.5–5 µm indicate that the surface state can play an important role in the non-uniform heating surface and consequently in the overestimation of the bulk temperature. The contribution of the hot spots to temperature measurements and flux calculations has been simulated at different wavelengths. Calculations show that (1) the overestimation of the bulk temperature decreases with the wavelength and (2) the overestimation depends on the temperature difference, ΔT, between the bulk and the micrometric hot spots. In addition, experiments have been carried out in order to compare the flux calculations at different wavelengths on different graphite (polished, dusty). The results obtained are very sensitive to the surface state pointing out the difficulties in improving the heat flux calculation model, since the surface state can change during the plasma discharges. This paper shows that the problem of non-homogenous surface temperature can be significantly diminished on working at longer wavelengths.
    Plasma Physics and Controlled Fusion 03/2009; 51(5):055012. · 2.37 Impact Factor
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    ABSTRACT: A new B2SOLPS5.2 transport code has been developed and implemented for the simulation of H-mode shots. A new equation system is proposed, which is equivalent to the system which was used in B2SOLPS5.0 previously. The main idea is to replace the major part of the large radial ∇B driven convective fluxes by poloidal fluxes with the same divergence both in the particle balance and in the energy balance equations. This is of special importance for the H-mode where the diffusion coefficient is strongly reduced inside the barrier and large radial convective flows are strongly undesirable from the numerical point of view. The H-mode shots of ASDEX-Upgrade and MAST have been simulated with the new version with reasonable time steps and convergence. It is demonstrated that the radial electric field inside the edge transport barrier and in the pedestal region is close to the neoclassical electric field as in previous simulations of Ohmic shots. The toroidal rotation is co-current directed as in L-mode but is significantly larger in absolute value. It is shown that the shear of the poloidal drift at the inner side of the barrier is close to the value of the shear before the transition, while inside the barrier the value of the shear is significantly bigger. This fact determines self-consistently the width of the edge transport barrier. It is demonstrated that to match the experimental density and temperature radial profiles the drop in the diffusion coefficient within the barrier needs to be significantly larger than the drop in the electron heat conductivity coefficient.For the H-mode the pedestal region usually corresponds to the collisionless regime, so several corrections were introduced into the transport coefficients to extend the applicability of the code to the plateau and banana regimes in the inner regions of the simulation domain.
    Nuclear Fusion 01/2009; 49(2):025007. · 2.73 Impact Factor
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    ABSTRACT: Plasma wall interaction (PWI) is important for the material choice in ITER and for the plasma scenarios compatible with material constraints. In this paper, different aspects of the PWI are assessed in their importance for the initial wall materials choice: CFC for the strike point tiles, W in the divertor and baffle and Be on the first wall. Further material options are addressed for comparison, such as W divertor/Be first wall and all-W or all-C. One main parameter in this evaluation is the particle flux to the main vessel wall. One detailed plasma scenario exists for a Q = 10 ITER discharge [G. Federici et al., J. Nucl. Mater. 290–293 (2001) 260] which was taken as the basis of further erosion and tritium retention evaluations. As the assessment of steady state wall fluxes from a scaling of present fusion devices indicates that global wall fluxes may be a factor of 4 ± 3 higher, this margin has been adopted as uncertainty of the scaling. With these wall and divertor fluxes, important PWI processes such as erosion and tritium accumulation have been evaluated: It was found that the steady state erosion is no problem for the lifetime of plasma-facing divertor components. Be wall erosion may pose a problem in case of a concentration of the wall fluxes to small wall areas. ELM erosion may drastically limit the PFC lifetime if ELMs are not mitigated to energies below 0.5 MJ. Dust generation is still a process which requires more attention. Conversion from gross or net erosion to dust and the assessment of dust on hot surfaces need to be investigated. For low-Z materials the build-up of the tritium inventory is dominated by co-deposition with eroded wall atoms. For W, where erosion and tritium co-deposition are small, the implantation, diffusion and bulk trapping constitute the dominant retention processes. First extrapolations with models based on laboratory data show small contributions to the inventory. For later ITER phases and the extrapolation to DEMO additional tritium trapping sites due to neutron-irradiation damage need to be taken into account. Finally, the expected values for erosion and tritium retention are compared to the ITER administrative limits for the lifetime, dust and tritium inventory.
    Journal of Nuclear Materials. 01/2009;
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    ABSTRACT: Several improvements to the MAST plant and diagnostics have facilitated new studies advancing the physics basis for ITER and DEMO, as well as for future spherical tokamaks (STs). Using the increased heating capabilities P-NBI <= 3.8 MW H-mode at I-P = 1.2 MA was accessed showing that the energy confinement on MAST scales more weakly with I-P and more strongly with B-t than in the ITER IPB98(y, 2) scaling. Measurements of the fuel retention of shallow pellets extrapolate to an ITER particle throughput of 70% of its original designed total throughput capacity. The anomalous momentum diffusion, chi(phi), is linked to the ion diffusion, chi(i), with a Prandtl number close to P-phi approximate to chi(phi)/chi(i) approximate to 1, although chi(i) approaches neoclassical values. New high spatial resolution measurements of the edge radial electric field, E-r, show that the position of steepest gradients in electron pressure and E-r (i.e. shearing rate) are coincident, but their magnitudes are not linked. The T-e pedestal width on MAST scales with root beta(ped)(pol) rather than rho(pol). The edge localized mode (ELM) frequency for type-IV ELMs, new in MAST, was almost doubled using n = 2 resonant magnetic perturbations from a set of four external coils (n = 1, 2). A new internal 12 coil set (n <= 3) has been commissioned. The filaments in the inter-ELM and L-mode phase are different from ELM filaments, and the characteristics in L-mode agree well with turbulence calculations. A variety of fast particle driven instabilities were studied from 10 kHz saturated fishbone like activity up to 3.8 MHz compressional Alfven eigenmodes. Fast particle instabilities also affect the off-axis NBI current drive, leading to fast ion diffusion of the order of 0.5 m(2) s(-1) and a reduction in the driven current fraction from 40% to 30%. EBW current drive start-up is demonstrated for the first time in a ST generating plasma currents up to 55 kA. Many of these studies contributed to the physics basis of a planned upgrade to MAST.
    Nuclear Fusion. 01/2009; 49(10):104017.
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    ABSTRACT: A new version of the B2SOLPS5.0 transport code, which is free from numerical problems in the barrier region, has been used to simulate H-mode shots from ASDEX-Upgrade and MAST. The radial electric field inside the edge transport barrier and in the pedestal region is close to the neoclassical prediction. The shear of poloidal E→×B→ drift at the inner side of the barrier is close to the value before the transition, while inside the barrier it is significantly larger. It is demonstrated that to match the experimental density and temperature radial profiles the drop in the diffusion coefficient within the barrier should be significantly larger than the drop in the electron heat conductivity.
    Journal of Nuclear Materials. 01/2009; 390:408-411.
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    ABSTRACT: The tokamak boundary plasma is inherently 2D/3D, which impedes the detailed validation of transport models due to the limited spatial coverage of most diagnostics. A potential method for determining the 2D profiles of ne and Te in the plasma volume is to resolve spatially the emission ratios of atomic helium, principally the lines at 667, 706 and 728 nm. Unfortunately, there are several challenges associated with this approach: the traditional use of line-of-sight spectrometer data; crowding by other impurity lines; low signal levels; reliance on accurate atomic physics; and He I meta-stables. Several of these issues are explored in the present study, which utilises tomographic reconstruction of filtered CCD camera images to measure He emission throughout the divertor. Plasma gradients are resolved and compared with results from the OSM–EIRENE code. The near-target ne and Te values agree with Langmuir probe measurements to within a factor ∼2.
    Journal of Nuclear Materials 01/2009; 390:1078-1080. · 1.21 Impact Factor
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    ABSTRACT: Electrical currents in the scrape-off layer (SOL) of MAST are modelled using an interpretive Onion-Skin Model (OSM) constrained with experimental data from MAST diagnostics. The model was extended to include the effects of the magnetic mirror force, which has a strong influence on the particle and momentum balance in spherical tokamaks, such as MAST [1]. These modifications serve to more accurately model the parallel electric fields present in the MAST SOL, which can alter plasma dynamics via the E×B drift. Simulations show that the electrical current at the divertor targets is predominantly thermoelectric, whereas Pfirsch–Schlüter currents have a greater contribution to the total current in the bulk of the SOL plasma.
    Journal of Nuclear Materials - J NUCL MATER. 01/2009; 390:392-394.
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    ABSTRACT: Experimental and simulation results on filamentary structures observed in the Mega-Amp Spherical Tokamak (MAST) are presented and discussed. Fast camera data have been used to determine the mode number, toroidal and radial sizes and velocities of the filaments observed in L-mode, inter-edge localized mode (ELM) periods and ELMs which are summarized. Automated methods are applied to the analysis of L-mode image data in order to measure dependence with plasma parameters. This indicates that the mode number of L-mode edge turbulence increases with density and decreases with q95, while filament width has the opposite dependence.Simulations of L-mode discharges using the 3D, 2-fluid BOUT code produce similar sizes and radial velocities to the observations, and indicate that the source of these filaments is within a region ~2 cm from the plasma edge in a spontaneously formed E × B shear layer. Ion temperature fluctuations in these filaments are found to be approximately double the magnitude of electron temperature fluctuations, probably due to fast parallel electron heat transport.
    Plasma Physics and Controlled Fusion 11/2008; 50(12):124012. · 2.37 Impact Factor
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    ABSTRACT: A comparison is presented of measured and simulated parallel flows in the low field side scrape-off layer of MAST Ohmic shots. Simulations with the B2SOLPS5.0 code reproduce the experimentally observed co-current rotation direction. The absolute values of the simulated Mach number are smaller than those of the measured ones; the difference might reach a factor of two. It is demonstrated that in both the simulations and the experiment the parallel velocity increases with temperature and decreases with poloidal magnetic field.
    Plasma Physics and Controlled Fusion 10/2008; 50(11):115010. · 2.37 Impact Factor
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    ABSTRACT: A comparison of the spatial and temporal evolution of the filamentary structures observed during type I ELMs is presented from a variety of diagnostics and machines. There is evidence that these filaments can be detected inside the LCFS prior to ELMs. The filaments do not have a circular cross section instead they are elongated in the perpendicular (poloidal) direction and this size appears to increase linearly with the minor radius of the machine. The filaments start off rotating toroidally/poloidally with velocities close to that of the pedestal. This velocity then decreases as the filaments propagate radially. By comparing the results from all measurements and from comparison with models it is most likely that the filaments have at least their initial radial velocity when they are far out into the SOL and before they have interacted with the nearest limiter surface. There is a general consensus that the dominant loss mechanism in the separated filaments is through parallel transport and that the transport to the wall is through the radial propagation of these filaments. Measurements of the filament energy content show that each filament contains up to 2.5% of the energy released by the ELM at the time it separates from the LCFS, assuming Ti = Te. The parallel flux e-folding length measured on DIII-D, AUG and MAST has a weaker scaling with normalised ELM size than appears to be necessary to explain the deficit in the ELM energy arriving in the divertor on JET, assuming a purely exponential decay of the filament energy with time.
    Journal of Physics Conference Series 08/2008; 123(1):012011.
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    ABSTRACT: A study of the evolution of the filaments observed during Type I ELMs on ASDEX Upgrade and MAST is presented. The filaments start off rotating toroidally/poloidally with velocities close to that of the pedestal. This velocity then decreases as the filaments propagate radially. On both devices the ion saturation current e-folding lengths of the filaments show a weak, if any, dependence on the size of the ELM (δWELM/Wped). On MAST the measured radial velocities of the filaments also show at most a weak dependence on δWELM/Wped.
    Journal of Physics Conference Series 08/2008; 123(1):012012.
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    ABSTRACT: Interactions between the plasma and the vessel walls constitute a major engineering problem for next step fusion devices, such as ITER, determining the choice of the plasma-facing materials. A prominent issue in this choice is the tritium inventory build-up in the vessel, which must be limited for safety reasons. The initial material selection, i.e. beryllium (Be) on the main vessel walls, tungsten (W) on the divertor upper baffle and dome, and carbon fibre composite around the strike points on the divertor plates, results both from the attempt to reduce the tritium inventory and to optimize the lifetime of the plasma-facing components.In the framework of the EU Task Force on Plasma–Wall Interaction (PWI TF), the many physics aspects governing the tritium inventory are brought together. Together with supporting information from international experts represented by the ITPA SOL/DIV section, this paper describes the present status of knowledge of the in-vessel tritium inventory build-up. Firstly, the main results from present fusion devices in this field are briefly reviewed. Then, the processes involved are discussed: implantation, trapping and diffusion in plasma-facing materials are considered as well as surface erosion and co-deposition of tritium with eroded material. The intermixing of the different materials and its influence on hydrogen retention and co-deposition is a major source of uncertainty on present estimates and is also addressed.Based on the previous considerations, estimates for the tritium inventory build-up are given for the initial choice of ITER materials, as well as for alternative options. Present estimates indicate a build-up of the tritium inventory to the administrative limit within a few hundred nominal full power D : T discharges, co-deposition with carbon being the dominant process. Therefore, tritium removal methods are also an active area of research within the EU PWI TF, and are discussed. An integrated operational scheme to slow the rate of tritium accumulation is presented, which includes plasma operation as well as conditioning procedures.
    Plasma Physics and Controlled Fusion 08/2008; 50(10):103001. · 2.37 Impact Factor

Publication Stats

918 Citations
178.47 Total Impact Points

Institutions

  • 2009–2010
    • Fusion for Energy
      Barcino, Catalonia, Spain
  • 2001–2008
    • Culham Centre for Fusion Energy
      Abingdon-on-Thames, England, United Kingdom
  • 2007
    • St. Petersburg State Polytechnical University
      Sankt-Peterburg, St.-Petersburg, Russia
  • 2003
    • Lawrence Livermore National Laboratory
      • Physics Division
      Livermore, California, United States
  • 1999
    • Massachusetts Institute of Technology
      • Plasma Science and Fusion Center (PSFC)
      Cambridge, Massachusetts, United States