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  • 51. Beurskens, M N A
    et al.
    Osborne, T H
    Schneider, P A
    Wolfrum, E
    Frassinetti, Lorenzo
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Groebner, R
    Lomas, P
    Nunes, I
    Saarelma, S
    Scannell, R
    Snyder, P B
    Zarzoso, D
    Balboa, I
    Bray, B
    Brix, M
    Flanagan, J
    Giroud, C
    Giovannozzi, E
    Kempenaars, M
    Loarte, A
    de la Luna, E
    Maddison, G
    Maggi, C F
    McDonald, D
    Pasqualotto, R
    Saibene, G
    Sartori, R
    Solano, E
    Walsh, M
    Zabeo, L
    Team, D I I I-D
    Team, ASDEX Upgrade
    Contributors, J E T-E F D A
    H-mode pedestal scaling in DIII-D, ASDEX Upgrade, and JET2011In: Physics of Plasmas, ISSN 1070-664X, E-ISSN 1089-7674, Vol. 18, no 5Article in journal (Refereed)
    Abstract [en]

    Multidevice pedestal scaling experiments in the DIII-D, ASDEX Upgrade (AUG), and JET tokamaks are presented in order to test two plasma physics pedestal width models. The first model proposes a scaling of the pedestal width Delta/a proportional to rho*(1/2) to rho* based on the radial extent of the pedestal being set by the point where the linear turbulence growth rate exceeds the E x B velocity. In the multidevice experiment where rho* at the pedestal top was varied by a factor of four while other dimensionless parameters where kept fixed, it has been observed that the temperature pedestal width in real space coordinates scales with machine size, and that therefore the gyroradius scaling suggested by the model is not supported by the experiments. The density pedestal width is not invariant with rho* which after comparison with a simple neutral fuelling model may be attributed to variations in the neutral fuelling patterns. The second model, EPED1, is based on kinetic ballooning modes setting the limit of the radial extent of the pedestal region and leads to Delta(psi) proportional to beta p(1/2). All three devices show a scaling of the pedestal width in normalised poloidal flux as Delta(psi) proportional to beta p(1/2), as described by the kinetic ballooning model; however, on JET and AUG, this could not be distinguished from an interpretation where the pedestal is fixed in real space. Pedestal data from all three devices have been compared with the predictive pedestal model EPED1 and the model produces pedestal height values that match the experimental data well.

  • 52. Beurskens, M. N. A.
    et al.
    Schweinzer, J.
    Angioni, C.
    Burckhart, A.
    Challis, C. D.
    Chapman, I.
    Fischer, R.
    Flanagan, J.
    Frassinetti, Lorenzo
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Giroud, C.
    Hobirk, J.
    Joffrin, E.
    Kallenbach, A.
    Kempenaars, M.
    Leyland, M.
    Lomas, P.
    Maddison, G.
    Maslov, M.
    McDermott, R.
    Neu, R.
    Nunes, I.
    Osborne, T.
    Ryter, F.
    Saarelma, S.
    Schneider, P. A.
    Snyder, P.
    Tardini, G.
    Viezzer, E.
    Wolfrum, E.
    The effect of a metal wall on confinement in JET and ASDEX Upgrade2013In: Plasma Physics and Controlled Fusion, ISSN 0741-3335, E-ISSN 1361-6587, Vol. 55, no 12, p. 124043-Article in journal (Refereed)
    Abstract [en]

    In both JET and ASDEX Upgrade (AUG) the plasma energy confinement has been affected by the presence of a metal wall by the requirement of increased gas fuelling to avoid tungsten pollution of the plasma. In JET with a beryllium/tungsten wall the high triangularity baseline H-mode scenario (i.e. similar to the ITER reference scenario) has been the strongest affected and the benefit of high shaping to give good normalized confinement of H-98 similar to 1 at high Greenwald density fraction of f(GW) similar to 0.8 has not been recovered to date. In AUG with a full tungsten wall, a good normalized confinement H-98 similar to 1 could be achieved in the high triangularity baseline plasmas, albeit at elevated normalized pressure beta(N) > 2. The confinement lost with respect to the carbon devices can be largely recovered by the seeding of nitrogen in both JET and AUG. This suggests that the absence of carbon in JET and AUG with a metal wall may have affected the achievable confinement. Three mechanisms have been tested that could explain the effect of carbon or nitrogen (and the absence thereof) on the plasma confinement. First it has been seen in experiments and by means of nonlinear gyrokinetic simulations (with the GENE code), that nitrogen seeding does not significantly change the core temperature profile peaking and does not affect the critical ion temperature gradient. Secondly, the dilution of the edge ion density by the injection of nitrogen is not sufficient to explain the plasma temperature and pressure rise. For this latter mechanism to explain the confinement improvement with nitrogen seeding, strongly hollow Z(eff) profiles would be required which is not supported by experimental observations. The confinement improvement with nitrogen seeding cannot be explained with these two mechanisms. Thirdly, detailed pedestal structure analysis in JET high triangularity baseline plasmas have shown that the fuelling of either deuterium or nitrogen widens the pressure pedestal. However, in JET-ILW this only leads to a confinement benefit in the case of nitrogen seeding where, as the pedestal widens, the obtained pedestal pressure gradient is conserved. In the case of deuterium fuelling in JET-ILW the pressure gradient is strongly degraded in the fuelling scan leading to no net confinement gain due to the pedestal widening. The pedestal code EPED correctly predicts the pedestal pressure of the unseeded plasmas in JET-ILW within +/- 5%, however it does not capture the complex variation of pedestal width and gradient with fuelling and impurity seeding. Also it does not predict the observed increase of pedestal pressure by nitrogen seeding in JET-ILW. Ideal peeling ballooning MHD stability analysis shows that the widening of the pedestal leads to a down shift of the marginal stability boundary by only 10-20%. However, the variations in the pressure gradient observed in the JET-ILW fuelling experiment is much larger and spans a factor of more than two. As a result the experimental points move from deeply unstable to deeply stable on the stability diagram in a deuterium fuelling scan. In AUG-W nitrogen seeded plasmas, a widening of the pedestal has also been observed, consistent with the JET observations. The absence of carbon can thus affect the pedestal structure, and mainly the achieved pedestal gradient, which can be recovered by seeding nitrogen. The underlying physics mechanism is still under investigation and requires further understanding of the role of impurities on the pedestal stability and pedestal structure formation.

  • 53. Beurskens, M.N.A.
    et al.
    Schweinzer, J
    Angioni, C
    Bourdelle, C
    Challis, C
    Frassinetti, Lorenzo
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Giroud, C
    Hobirk, J
    Joffrin, E
    Kallenbach, A
    Maddison, G.
    Neu, R.
    Osborne, T.
    Ryter, F.
    Saarelma, S.
    Schneider, P.
    Snyder, P.
    Wolfrum, E.
    The Effect of a Metal Wall on Confinement in JET and ASDEX-Upgrade2013In: 40th European Physical Society Conference on Plasma Physics: Espoo, Finland, 1st - 5th July 2013, European Physical Society , 2013Conference paper (Refereed)
  • 54. Bilato, R.
    et al.
    Bertelli, N.
    Brambilla, M.
    Dumont, R.
    Jaeger, E. F.
    Johnson, Thomas
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Lerche, E.
    Sauter, O.
    Van Eester, D.
    Villard, L.
    Status of the benchmark activity of ICRF full-wave codes within EUROfusion WPCD and beyond2015In: RADIOFREQUENCY POWER IN PLASMAS, American Institute of Physics (AIP), 2015, article id UNSP 060001Conference paper (Refereed)
    Abstract [en]

    As follow-up of the benchmark activity of ICRF full-wave codes within the EUROfusion Code Development for Integrated Modelling project (WPCD), a simple-to-complex approach has been devised for verification of the European ICRF codes, imported in the European-Integrated Modelling infrastructure, which represents a unique environment for input-data sharing and result analysis. This benchmark activity has been recently extended to non-European codes, in particular the ICRF full-wave AORSA code. Here we discussed the results of this benchmark.

  • 55. Bilato, R.
    et al.
    Coster, D.
    Dumont, R.
    Johnson, Tomas
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Klingshirn, H. -J
    Lerche, E.
    Sauter, O.
    Brambilla, M.
    Figini, L.
    Van Eester, D.
    Villard, L.
    Farina, D.
    ICRF-code benchmark activity in the framework of the European task-force on integrated Tokamak Modelling2014In: AIP Conference Proceedings, 2014, Vol. 1580, p. 291-294Conference paper (Refereed)
    Abstract [en]

    The grand aim of the Integrated Tokamak Modelling (ITM) task-force is to provide a flexible, modular and reliable plasma simulator in view of planning and analyzing ITER discharges. Since radio-frequency (rf) heating in the ion cyclotron range of frequencies (ICRF) is foreseen as one of the main additional heating systems in ITER, physics modules that simulate ICRF wave propagation and absorption are necessary for the ITM project. Here, we report on the status of the benchmark activity of ICRF codes, already imported in ITM environment platform. We consider various scenarios for ITER, limiting the comparisons to wave propagation and absorption in Maxwellian plasmas.

  • 56. Bogomolov, A. V.
    et al.
    Classen, I. G. J.
    Donne, A. J. H.
    Meyer, H.
    Dunne, M.
    Schneider, P. A.
    Wolfrum, E.
    Vanovac, B.
    Fischer, R.
    Frassinetti, Lorenzo
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Luhmann, N.C., Jr.
    The effect of nitrogen seeding on ELM filaments2015In: 42nd European Physical Society Conference on Plasma Physics, EPS 2015, European Physical Society (EPS) , 2015Conference paper (Refereed)
  • 57. Bohm, P.
    et al.
    Aftanas, M.
    Bilkova, P.
    Stefanikova, E.
    Mikulin, O.
    Melich, R.
    Janky, F.
    Havlicek, J.
    Sestak, D.
    Weinzettl, V.
    Stockel, J.
    Hron, M.
    Panek, R.
    Scannell, R.
    Frassinetti, Lorenzo
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Fassina, A.
    Naylor, G.
    Walsh, M. J.
    Edge Thomson scattering diagnostic on COMPASS tokamak: Installation, calibration, operation, improvements2014In: Review of Scientific Instruments, ISSN 0034-6748, E-ISSN 1089-7623, Vol. 85, no 11, p. 11E431-Article in journal (Refereed)
    Abstract [en]

    The core Thomson scattering diagnostic (TS) on the COMPASS tokamak was put in operation and reported earlier. Implementation of edge TS, with spatial resolution along the laser beam up to similar to 1/100 of the tokamak minor radius, is presented now. The procedure for spatial calibration and alignment of both core and edge systems is described. Several further upgrades of the TS system, like a triggering unit and piezo motor driven vacuum window shutter, are introduced as well. The edge TS system, together with the core TS, is now in routine operation and provides electron temperature and density profiles.

  • 58. Bohm, P
    et al.
    Bilkova, P
    Aftanas, M
    Stefanikova, E
    Mikulin, O
    Melich, R
    Janky, J
    Havlicek, H
    Sestak, D
    Weinzettl, V
    Stockel, A
    Hron, R
    Panek, R
    Scannell, R
    Frassinetti, Lorenzo
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Fassina, A
    Naylor, N
    Walsh, J
    Thomson Scattering on COMPASS tokamak: Plasma Edge Profile2013In: 16th International Symposium on Laser Aided Plasma Diagnostics Conference: Madison, Wisconsin, USA, 22-26 September, 2013, 2013Conference paper (Other academic)
  • 59. Borodkina, I.
    et al.
    Borodin, D.
    Brezinsek, S.
    Tsvetkov, I. V.
    Kurnaev, V. A.
    Guillemaut, C.
    Maslov, M.
    Frassinetti, Lorenzo
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Intra-ELM tungsten sputtering in JET ITER-like wall: analytical studies of Be impurity and ELM type influence2017In: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T170, article id 014065Article in journal (Refereed)
    Abstract [en]

    The W source strength in JET H-mode discharges depends on the W sputtering in the inter and the intra-ELM phase due to impinging hydrogenic ions (D or H) and impurities (mainly Be). The analytical approach for interpretation of the Langmuir probe measurements is applied to model the ELM ion parallel transport and the W sputtering flux in intra-ELM and inter-ELM conditions in JET-ITER-like wall (ILW) hydrogen and deuterium plasmas. The impact of the Be ion charge and the Be concentration in the impinging ion flux on the W sputtering was estimated. Be2+ concentrations of 0.5% and 1% in the impinging ion flux increases the W sputtering fluence per ELM by 20%-30% and 35%-55% correspondingly with respect to pure deuterium plasma; the charge state of Be ions has no substantial effect on W sputtering in the intra-ELM phase. The analysis of JET ELMy H-mode discharges in hydrogen with different types of ELMs is presented. The W sputtering source under inter-and intra-ELM conditions is estimated using the analytical approach and validated by optical emission spectroscopy in these discharges. The intra-ELM W sputtering flux increases 2-4 times in comparison to the inter-ELM flux.

  • 60. Boswell, C. J.
    et al.
    Berk, H. L.
    Borba, D.
    Johnson, Tomas
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Nave, M. F. F.
    Pinches, S. D.
    Sharapov, S. E.
    Observation and explanation of the JET n = 0 chirping mode2005Conference paper (Refereed)
  • 61. Boswell, C. J.
    et al.
    Berk, H. L.
    Borba, D. N.
    Johnson, Thomas J.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Pinches, S. D.
    Sharapov, S. E.
    Observation and explanation of the JET n=0 chirping mode2006In: Physics Letters A, ISSN 0375-9601, E-ISSN 1873-2429, Vol. 358, no 2, p. 154-158Article in journal (Refereed)
    Abstract [en]

    Persistent rapid up and down frequency chirping modes with a toroidal mode number of zero (n = 0) have been observed in the JET tokamak when energetic ions, with a mean energy similar to 500 keV, were created by high field side ion cyclotron resonance frequency heating. This heating method enables the formation of an energetically inverted ion distribution function that allows ions to spontaneously excite the observed instability, identified as a global geodesic acoustic mode. The interpretation is that phase space structures form and interact with the fluid zonal flow to produce the pronounced frequency chirping.

  • 62. Bowman, C.
    et al.
    Dickinson, D.
    Horvath, L.
    Lunniss, A. E.
    Wilson, H. R.
    Cziegler, I.
    Frassinetti, Lorenzo
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Space and Plasma Physics.
    Gibson, K.
    Kirk, A.
    Lipschultz, B.
    Maggi, C. F.
    Roach, C. M.
    Saarelma, S.
    Snyder, P. B.
    Thornton, A.
    Wynn, A.
    Pedestal evolution physics in low triangularity JET tokamak discharges with ITER-like wall2018In: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 58, no 1, article id 016021Article in journal (Refereed)
    Abstract [en]

    The pressure gradient of the high confinement pedestal region at the edge of tokamak plasmas rapidly collapses during plasma eruptions called edge localised modes (ELMs), and then re-builds over a longer time scale before the next ELM. The physics that controls the evolution of the JET pedestal between ELMs is analysed for 1.4 MA, 1.7 T, low triangularity, delta = 0.2, discharges with the ITER-like wall, finding that the pressure gradient typically tracks the ideal magneto-hydrodynamic ballooning limit, consistent with a role for the kinetic ballooning mode. Furthermore, the pedestal width is often influenced by the region of plasma that has second stability access to the ballooning mode, which can explain its sometimes complex evolution between ELMs. A local gyrokinetic analysis of a second stable flux surface reveals stability to kinetic ballooning modes; global effects are expected to provide a destabilising mechanism and need to be retained in such second stable situations. As well as an electronscale electron temperature gradient mode, ion scale instabilities associated with this flux surface include an electro-magnetic trapped electron branch and two electrostatic branches propagating in the ion direction, one with high radial wavenumber. In these second stability situations, the ELM is triggered by a peeling-ballooning mode; otherwise the pedestal is somewhat below the peeling-ballooning mode marginal stability boundary at ELM onset. In this latter situation, there is evidence that higher frequency ELMs are paced by an oscillation in the plasma, causing a crash in the pedestal before the peeling-ballooning boundary is reached. A model is proposed in which the oscillation is associated with hot plasma filaments that are pushed out towards the plasma edge by a ballooning mode, draining their free energy into the cooler plasma there, and then relaxing back to repeat the process. The results suggest that avoiding the oscillation and maximising the region of plasma that has second stability access will lead to the highest pedestal heights and, therefore, best confinement-a key result for optimising the fusion performance of JET and future tokamaks, such as ITER.

  • 63.
    Brenning, Nils
    et al.
    KTH, School of Electrical Engineering (EES), Space and Plasma Physics.
    Axnäs, Ingvar
    KTH, School of Electrical Engineering (EES), Space and Plasma Physics.
    Koepke, Mark
    KTH.
    Raadu, Michael A.
    KTH, School of Electrical Engineering (EES), Space and Plasma Physics.
    Tennfors, Einar
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Radiation from an electron beam in magnetized plasma: excitation of a whistler mode wave packet by interacting, higher-frequency, electrostatic-wave eigenmodes2017In: Plasma Physics and Controlled Fusion, ISSN 0741-3335, E-ISSN 1361-6587, Vol. 59, no 12, article id 124006Article in journal (Refereed)
    Abstract [en]

    Infrequent, bursty, electromagnetic, whistler-mode wave packets, excited spontaneously in the laboratory by an electron beam from a hot cathode, appear transiently, each with a time duration tau around similar to 1 mu s. The wave packets have a center frequency f(W) that is broadly distributed in the range 7 MHz < f(W) < 40 MHz. They are excited in a region with separate electrostatic (es) plasma oscillations at values of f(hf), 200 MHz < f(hf) < 500 MHz, that are hypothesized to match eigenmode frequencies of an axially localized hf es field in a well-defined region attached to the cathode. Features of these es-eigenmodes that are studied include: the mode competition at times of transitions from one dominating es-eigenmode to another, the amplitude and spectral distribution of simultaneously occurring es-eigenmodes that do not lead to a transition, and the correlation of these features with the excitation of whistler mode waves. It is concluded that transient coupling of es-eigenmode pairs at f(hf) such that vertical bar f(1, hf) - f(2, hf)vertical bar = f(W) < f(ge) can explain both the transient lifetime and the frequency spectra of the whistler-mode wave packets (f(W)) as observed in lab. The generalization of the results to bursty whistler-mode excitation in space from electron beams, created on the high potential side of double layers, is discussed.

  • 64. Brezinsek, S.
    et al.
    Fundamenski, W.
    Eich, T.
    Coad, J. P.
    Giroud, C.
    Huber, A.
    Jachmich, S.
    Joffrin, E.
    Krieger, K.
    McCormick, K.
    Lehnen, M.
    Loarer, T.
    de la Luna, E.
    Maddison, G.
    Matthews, G. F.
    Mertens, Ph.
    Nunes, I.
    Philipps, V.
    Riccardo, V.
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Stamp, M. F.
    Tsalas, M.
    Overview of experimental preparation for the ITER-Like Wall at JET2011In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 415, no 1, p. S936-S942Article in journal (Refereed)
    Abstract [en]

    Experiments in JET with carbon-based plasma-facing components have been carried out in preparation of the ITER-Like Wall with beryllium main chamber and full tungsten divertor. The preparatory work was twofold: (i) development of techniques, which ensure safe operation with the new wall and (ii) provision of reference plasmas, which allow a comparison of operation with carbon and metallic wall. (i) Compatibility with the W divertor with respect to energy loads could be achieved in N-2 seeded plasmas at high densities and low temperatures, finally approaching partial detachment, with only moderate confinement reduction of 10%. Strike-point sweeping increases the operational space further by re-distributing the load over several components. (ii) Be and C migration to the divertor has been documented with spectroscopy and QMBs under different plasma conditions providing a database which will allow a comparison of the material transport to remote areas with metallic walls. Fuel retention rates of 1.0-2.0 x 10(21) D s(-1) were obtained as references in accompanied gas balance studies.

  • 65. Brezinsek, S.
    et al.
    Jachmich, S.
    Stamp, M. F.
    Meigs, A. G.
    Coenen, J. W.
    Krieger, K.
    Giroud, C.
    Groth, M.
    Philipps, V.
    Grünhagen, S.
    Smith, R.
    Van Rooij, G. J.
    Ivanova, Darya
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Matthews, G. F.
    Residual carbon content in the initial ITER-Like Wall experiments at JET2013In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 438, no Suppl., p. S303-S308Article in journal (Refereed)
    Abstract [en]

    The residual carbon content and carbon edge flux in JET have been assessed by three independent diagnostic techniques after start of plasma operation with the ITER-Like Wall (ILW) with beryllium first wall and tungsten divertor: (i) in-situ measurements with optical spectroscopy on low ionisation stages of carbon, (ii) charge-exchange recombination spectroscopy, and (iii) residual gas composition analysis in dedicated global gas balance experiments. Direct comparison experiments in L-mode discharges were carried out between references from the previously installed material configuration with plasma-facing components made of carbon-fibre composite (JET-CFC) and the JET-ILW. The temporal evolution of the C divertor flux since installation of the ILW has been studied in the ohmic phase of dedicated monitoring discharges which have been executed regularly throughout the experimental exploitation so far (60000 plasma seconds). The C flux behaviour in the divertor can be divided in three phases: initial fast drop, moderate reduction phase, and a long lasting phase with almost constant C flux. The Be flux in both divertor legs mirrors the behaviour of C. All experiments and diagnostic techniques demonstrate a strong reduction in C fluxes and C content of more than one order of magnitude with respect to JET-CFC which is in line with the reduction in long-term fuel retention due to co-deposition. There is no evidence of an increase in residual carbon in time, thus no indication that a damage of the thin tungsten coatings on CFC substrate in the divertor occurred.

  • 66. Brezinsek, S.
    et al.
    Petersson, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Ratynskaia, Svetlana
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Ström, Petter
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Tolias, Panagiotis
    KTH, School of Electrical Engineering (EES), Space and Plasma Physics.
    Weckmann, Armin
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Zaplotnik, R.
    et al.,
    Plasma-wall interaction studies within the EUROfusion consortium: Progress on plasma-facing components development and qualification2017In: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 57, no 11, article id 116041Article in journal (Refereed)
    Abstract [en]

    The provision of a particle and power exhaust solution which is compatible with first-wall components and edge-plasma conditions is a key area of present-day fusion research and mandatory for a successful operation of ITER and DEMO. The work package plasma-facing components (WP PFC) within the European fusion programme complements with laboratory experiments, i.e. in linear plasma devices, electron and ion beam loading facilities, the studies performed in toroidally confined magnetic devices, such as JET, ASDEX Upgrade, WEST etc. The connection of both groups is done via common physics and engineering studies, including the qualification and specification of plasma-facing components, and by modelling codes that simulate edge-plasma conditions and the plasma-material interaction as well as the study of fundamental processes. WP PFC addresses these critical points in order to ensure reliable and efficient use of conventional, solid PFCs in ITER (Be and W) and DEMO (W and steel) with respect to heat-load capabilities (transient and steady-state heat and particle loads), lifetime estimates (erosion, material mixing and surface morphology), and safety aspects (fuel retention, fuel removal, material migration and dust formation) particularly for quasi-steady-state conditions. Alternative scenarios and concepts (liquid Sn or Li as PFCs) for DEMO are developed and tested in the event that the conventional solution turns out to not be functional. Here, we present an overview of the activities with an emphasis on a few key results: (i) the observed synergistic effects in particle and heat loading of ITER-grade W with the available set of exposition devices on material properties such as roughness, ductility and microstructure; (ii) the progress in understanding of fuel retention, diffusion and outgassing in different W-based materials, including the impact of damage and impurities like N; and (iii), the preferential sputtering of Fe in EUROFER steel providing an in situ W surface and a potential first-wall solution for DEMO.

  • 67. Brezinsek, S.
    et al.
    Widdowson, A.
    Mayer, M.
    Philipps, V.
    Baron-Wiechec, P.
    Coenen, J. W.
    Heinola, K.
    Huber, A.
    Likonen, J.
    Petersson, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Stamp, M. F.
    Borodin, D.
    Coad, J. P.
    Carrasco, Alvaro Garcia
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Kirschner, A.
    Krat, S.
    Krieger, K.
    Lipschultz, B.
    Linsmeier, Ch.
    Matthews, G. F.
    Schmid, K.
    Beryllium migration in JET ITER-like wall plasmas2015In: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 55, no 6, article id 063021Article in journal (Refereed)
    Abstract [en]

    JET is used as a test bed for ITER, to investigate beryllium migration which connects the lifetime of first-wall components under erosion with tokamak safety, in relation to long-term fuel retention. The (i) limiter and the (ii) divertor configurations have been studied in JET-ILW (JET with a Be first wall and W divertor), and compared with those for the former JET-C (JET with carbon-based plasma-facing components (PFCs)). (i) For the limiter configuration, the Be gross erosion at the contact point was determined in situ by spectroscopy as between 4% (E-in = 35 eV) and more than 100%, caused by Be self-sputtering (E-in = 200 eV). Chemically assisted physical sputtering via BeD release has been identified to contribute to the effective Be sputtering yield, i.e. at E-in = 75 eV, erosion was enhanced by about 1/3 with respect to the bare physical sputtering case. An effective gross yield of 10% is on average representative for limiter plasma conditions, whereas a factor of 2 difference between the gross erosion and net erosion, determined by post-mortem analysis, was found. The primary impurity source in the limiter configuration in JET-ILW is only 25% higher (in weight) than that for the JET-C case. The main fraction of eroded Be stays within the main chamber. (ii) For the divertor configuration, neutral Be and BeD from physically and chemically assisted physical sputtering by charge exchange neutrals and residual ion flux at the recessed wall enter the plasma, ionize and are transported by scrape-off layer flows towards the inner divertor where significant net deposition takes place. The amount of Be eroded at the first wall (21 g) and the Be amount deposited in the inner divertor (28 g) are in fair agreement, though the balancing is as yet incomplete due to the limited analysis of PFCs. The primary impurity source in the JET-ILW is a factor of 5.3 less in comparison with that for JET-C, resulting in lower divertor material deposition, by more than one order of magnitude. Within the divertor, Be performs far fewer re-erosion and transport steps than C due to an energetic threshold for Be sputtering, and inhibits as a result of this the transport to the divertor floor and the pump duct entrance. The target plates in the JET-ILW inner divertor represent at the strike line a permanent net erosion zone, in contrast to the net deposition zone in JET-C with thick carbon deposits on the CFC (carbon-fibre composite) plates. The Be migration identified is consistent with the observed low long-term fuel retention and dust production with the JET-ILW.

  • 68. Brezinsek, Sebastijan
    et al.
    Wirtz, Marius
    Dorrow-Gesprach, Daniel
    Loewenhoff, Thorsten
    Rubel, Marek
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    16th International Conference on Plasma-Facing Materials and Components for Fusion Applications2017In: Physica Scripta, ISSN 0031-8949, E-ISSN 1402-4896, Vol. T170, article id 010201Article in journal (Refereed)
  • 69.
    Brunsell, P R
    et al.
    KTH, Superseded Departments, Alfvén Laboratory.
    Bergsåker, H
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Brzozowski, J H
    KTH, Superseded Departments, Alfvén Laboratory.
    Cecconello, M
    KTH, Superseded Departments, Alfvén Laboratory.
    Drake, J R
    KTH, Superseded Departments, Alfvén Laboratory.
    Malmberg, J-A
    KTH, Superseded Departments, Alfvén Laboratory.
    Scheffel, J
    KTH, Superseded Departments, Alfvén Laboratory.
    Schnack, D D
    Mode dynamics and confinement in the reversed-field pinch2000In: IAEA-CN-77: Fusion Energy 2000, 2000, p. Paper EXP3/14-Conference paper (Refereed)
  • 70.
    Brunsell, Per
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Frassinetti, Lorenzo
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Volpe, Francesco
    Columbia University, New York, NY, USA.
    Olofsson, Erik
    Columbia University, New York, NY, USA.
    Fridström, Rickard
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Setiadi, Agung Chris
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Resistive Wall Mode Studies utilizing External Magnetic Perturbations2014In: Proceeding of the 25th IAEA Fusion Energy Conference, 2014, article id Paper EX/P4-20Conference paper (Other academic)
  • 71.
    Brunsell, Per
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Frassinetti, Lorenzo
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Volpe, Francesco
    Columbia University, New York, NY, USA.
    Olofsson, Erik
    Columbia University, New York, NY, USA.
    Fridström, Rickard
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Setiadi, Agung Chris
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Resistive Wall Mode Studies utilizing External Magnetic PerturbationsManuscript (preprint) (Other academic)
  • 72.
    Brunsell, Per
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Olofsson, K. Erik J.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Frassinetti, Lorenzo
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Menmuir, Sheena
    KTH, School of Engineering Sciences (SCI), Physics.
    Cecconello, Marco
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Yadikin, D.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Kuldkepp, Mattias
    KTH, School of Engineering Sciences (SCI), Physics.
    Drake, James Robert
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Rachlew, Elisabethth
    KTH, School of Engineering Sciences (SCI), Physics.
    Resistive wall mode feedback control experiments in EXTRAP T2R2007In: 34th EPS Conference on Plasma Physics 2007, EPS 2007 - Europhysics Conference Abstracts, 2007, p. 544-547Conference paper (Refereed)
    Abstract [en]

    Experiments in EXTRAP T2R on RWM stabilization using intelligent shell feedback with a P-controller showed that mode suppression improves with increasing gain up to the system stability limit. A PD-controller gives faster response and allows operation with higher gain. The PI-controller is useful for suppression of modes driven by external resonant field error. Best mode suppression was in the present study achieved with a PID-controller.

  • 73. Brunsell, Per R.
    et al.
    Olofsson, K. E. J.
    Frassinetti, L.
    Drake, James R.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Resistive wall mode feedback control in EXTRAP T2R with improved steady-state error and transient response2007In: Physics of Plasmas, ISSN 1070-664X, E-ISSN 1089-7674, Vol. 14, no 10Article in journal (Refereed)
  • 74.
    Brunsell, Per
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Yadikin, Dmitriy
    KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Cecconello, Marco
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Drake, James Robert
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Menmuir, Sheena
    KTH, School of Engineering Sciences (SCI), Physics.
    Rachlew, Elisabeth
    KTH, School of Engineering Sciences (SCI), Physics, Atomic and Molecular Physics.
    Zanca, P.
    Active control of multiple resistive wall modes2005In: Plasma Physics and Controlled Fusion, ISSN 0741-3335, E-ISSN 1361-6587, Vol. 47, no 12 B, p. B25-B36Article in journal (Refereed)
    Abstract [en]

     A two-dimensional array of saddle coils at M-c poloidal and N-c toroidal positions is used on the EXTRAP T2R reversed-field pinch (Brunsell P R et al 2001 Plasma Phys. Control. Fusion 43 1457) to study active control of resistive wall modes (RWMs). Spontaneous growth of several RWMs with poloidal mode number m = 1 and different toroidal mode number n is observed experimentally, in agreement with linear MHD modelling. The measured plasma response to a controlled coil field and the plasma response computed using the linear circular cylinder MHD model are in quantitive agreement. Feedback control introduces a linear coupling of modes with toroidal mode numbers n, n' that fulfil the condition vertical bar n - n'vertical bar = N-c. Pairs of coupled unstable RWMs are present in feedback experiments with an array of Mc x Nc = 4 x 16 coils. Using intelligent shell feedback, the coupled modes are generally not controlled even though the field is suppressed at the active coils. A better suppression of coupled modes may be achieved in the case of rotating modes by using the mode control feedback scheme with individually set complex gains. In feedback with a larger array of Mc x Nc = 4 x 32 coils, the coupling effect largely disappears, and with this array, the main internal RWMs n = -11, -10, +5, +6 are all simultaneously suppressed throughout the discharge (7-8 wall times). With feedback there is a two-fold extension of the pulse length, compared to discharges without feedback.

  • 75. Bucalossi, J.
    et al.
    Neu, R.
    Joffrin, E.
    Lomas, P.
    Nunes, I.
    Rimini, F.
    Beurskens, M.
    Frassinetti, Lorenzo
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Baruzzo, M.
    Bourdelle, C.
    Challis, C.
    Coenen, J.
    De Vries, P.
    Dux, R.
    Giroud, C.
    Giruzzi, G.
    Maddison, G.
    Mayoral, M.
    Characterization of the ELMy H-mode regime with the ITER-like wall in JET2012In: 39th EPS Conference on Plasma Physics 2012, EPS 2012 and the 16th International Congress on Plasma Physics: Volume 1, 2012, 2012, p. 45-48Conference paper (Refereed)
  • 76. Budny, R. V.
    et al.
    Indireshkumar, K.
    McCune, D.
    Mayoral, M. -L
    Ongena, J.
    Van Eester, D.
    Conboy, J.
    Voitsekhovitch, I.
    Johnson, Thomas Joe
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Sartori, R.
    Progress testing TRANSP-TORIC simulations of ICRH in JET2009In: 36th EPS Conference on Plasma Physics 2009, EPS 2009 - Europhysics Conference Abstracts, 2009, p. 1455-1458Conference paper (Refereed)
  • 77.
    Bykov, Igor
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Experimental studies of materials migration in magnetic confinement fusion devices: Novel methods for measurement of macro particle migration, transport of atomic impurities and characterization of exposed surfaces2014Doctoral thesis, comprehensive summary (Other academic)
    Abstract [en]

    During several decades of research and development in the field of Magnetically Confined Fusion (MCF) the preferred selection of materials for Plasma Facing Components (PFC) has changed repeatedly. Without doubt, endurance of the first wall will decide research availability and lifespan of the first International Thermonuclear Research Reactor (ITER). Materials erosion, redeposition and mixing in the reactor are the critical processes responsible for modification of materials properties under plasma impact. This thesis presents several diagnostic techniques and their applications for studies of materials transport in fusion devices. The measurements were made at the EXTRAP T2R Reversed Field Pinch operated in Alfvén laboratory at KTH (Sweden), the TEXTOR tokamak, recently shut down at Forschungszentrum Jülich (Germany) and in the JET tokamak at CCFE (UK). The main outcomes of the work are:

    • Development and application of a method for non-destructive capture and characterization of fast dust particles moving in the edge plasma of fusion devices, as well as particles generated upon laser-assisted cleaning of plasma exposed surfaces. 
    • Advancement of conventional broad beam and micro ion beam techniques to include measurement of tritium in the surfaces exposed in future D-T experiments. 
    • Adaption of the micro ion beam method for precision mapping of non uniform elements concentrations on irregular surfaces. 
    • Implementation of an isotopic marker to study the large scale materials migration in a tokamak and development of a method for fast non destructive sampling of the marker on surfaces of PFCs.
  • 78.
    Bykov, Igor
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Bergsaker, Henric
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Possnert, G.
    Zhou, Y.
    Heinola, K.
    Pettersson, J.
    Conroy, S.
    Likonen, J.
    Petersson, Per
    Widdowson, A.
    Studies of Be migration in the JET tokamak using AMS with Be-10 marker2016In: Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, ISSN 0168-583X, E-ISSN 1872-9584, Vol. 371, p. 370-375Article in journal (Refereed)
    Abstract [en]

    The JET tokamak is operated with beryllium limiter tiles in the main chamber and tungsten coated carbon fiber composite tiles and solid W tiles in the divertor. One important issue is how wall materials are migrating during plasma operation. To study beryllium redistribution in the main chamber and in the divertor, a Be-10 enriched limiter tile was installed prior to plasma operations in 2011-2012. Methods to take surface samples have been developed, an abrasive method for bulk Be tiles in the main chamber, which permits reuse of the tiles, and leaching with hot HCl to remove all Be deposited at W coated surfaces in the divertor. Quantitative analysis of the total amount of Be in cm(2) sized samples was made with inductively coupled plasma atomic emission spectroscopy (ICP-AES). The Be-10/Be-9 ratio in the samples was measured with accelerator mass spectrometry (AMS). The experimental setup and methods are described in detail, including sample preparation, measures to eliminate contributions in AMS from the B-10 isobar, possible activation due to plasma generated neutrons and effects of diffusive isotope mixing. For the first time marker concentrations are measured in the divertor deposits. They are in the range 0.4-1.2% of the source concentration, with moderate poloidal variation. (C) 2015 Elsevier B.V. All rights reserved.

  • 79.
    Bykov, Igor
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Bergsåker, Henric
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Ogata, Douglas
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Petersson, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Ratynskaia, Svetlana
    KTH, School of Electrical Engineering (EES), Space and Plasma Physics.
    Collection of mobile dust in the T2R reversed field pinch2012In: Nukleonika, ISSN 0029-5922, E-ISSN 1508-5791, Vol. 57, no 1, p. 55-60Article in journal (Refereed)
    Abstract [en]

    Intensive plasma-wall interactions in fusion devices result in the impurity production and the formation of films of redeposited material, debris and dust. In present day devices, with short pulses, the mobile dust does not pose any serious operational problems, but it is a matter of serious concern for ITER and for later power producing devices with a high duty cycle. We report results of a dust collection experiment carried out at the T2R reversed field pinch device and related heavy impurity flux measurements. Dust and impurities were collected on passive Si surface probes and on ultralow density silica aerogel collectors. The advantage of the latter method is the possibility of nondestructive capture of the micron- and submicron-sized dust particles. The toroidal and radial deposition fluxes of dust particles and impurities are estimated and discussed in the light of the dominant forces acting on the dust.

  • 80.
    Bykov, Igor
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Bergsåker, Henric
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Petersson, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Likonen, Jari
    Possnert, G.
    Widdowson, C.
    Combined ion micro probe and SEM analysis of strongly non uniform deposits in fusion devices2015In: Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, ISSN 0168-583X, E-ISSN 1872-9584, Vol. 342, p. 19-28Article in journal (Refereed)
    Abstract [en]

    Conventional ion beam analysis (IBA) of deposited layers from fusion devices may have insufficient accuracy due to strongly uneven appearance of the layers. Surface roughness and spatial variation of the matrix composition make interpretation of broad beam spectra complex and non obvious. We discuss complications of applied IBA arising for fusion-relevant surfaces and demonstrate how quantification can be improved by employing micro IBA methods. The analysis is bound to pre-defined regions on the sample surface and can be extended by employing beams of several types, scanning electron microscopy (SEM) and stereo SEM techniques.

  • 81.
    Bykov, Igor
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Bergsåker, Henric
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Petersson, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Likonen, Jari
    Possnert, Göran
    Quantitative plasma-fuel and impurity profiling in thick plasma-deposited layers by means of micro ion beam analysis and SIMS2014In: Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, ISSN 0168-583X, E-ISSN 1872-9584, Vol. 332, p. 280-285Article in journal (Refereed)
    Abstract [en]

    The operation of the Joint European Torus (JET) with full-carbon wall during the last decades has proven the importance of material re-deposition processes in remote areas of the tokamak. The thickness of the deposits in shadowed areas can reach 1 mm. The main constituent is carbon, with little inclusion of Inconel components. Atomic fractions Be/C and D/C can locally reach 1. Three methods were used to study thick deposits on JET divertor surfaces: (i) NRA analysis with a 15 mu m wide, 3 MeV He-3 ion microbeam on a polished cross section of the layer to determine the concentration distribution of D, Be and C and the distribution of Ni by particle induced X-ray emission; (ii) elastic proton scattering (EPS) from the top of the layers with a broad proton beam at 3.5 and 4.6 MeV. These methods were absolutely calibrated using thick elemental targets. (iii) Depth profiling of D, Be and Ni was done with secondary ion mass spectrometry (SIMS), sputtering the layers from the surface. The three methods are complementary. The thickest layers are accessible only by microbeam mapping of the cross sections, albeit with limited spatial resolution. The SIMS has the best depth resolution, but is difficult for absolute quantification and is limited in accessible depth. The probed depth with proton backscattering is limited to about 30 mu m. The combination of all three methods provided a coherent picture of the layer composition. It was possible to correlate the SIMS profiling results to quantitative data obtained by the microbeam method.

  • 82.
    Bykov, Igor
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Bergsåker, Henric
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Ratynskaia, Svetlana
    KTH, School of Electrical Engineering (EES), Space and Plasma Physics.
    Litnovsky, A.
    Petersson, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Possnert, G.
    Time resolved collection and characterization of dust particles moving in the TEXTOR scrape-off layer2013In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 438, no Suppl., p. S681-S685Article in journal (Refereed)
    Abstract [en]

    Moving dust has been collected in the SOL of TEXTOR in a time-resolved way with silica aerogel collectors [1-3]. The collectors were exposed to the toroidal particle flux in NBI heated discharges during the startup and flat top phase. Intrinsic dust was collected in several discharges. Other discharges were accompanied with injection of known amounts of pre-characterized dust (W, C flakes and C microspheres) from a position toroidally 120° away from the collector. Particle flux, composition and dust size distribution have been determined with SEM and EDX. Calibration allowed particle velocity estimates to be made. Upper limits for the deuterium content of individual dust grains have been determined by NRA.

  • 83.
    Bykov, Igor
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Bergsåker, Henrik
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Possnert, G.
    Heinola, K.
    Miettunen, J.
    Groth, M.
    Petersson, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Widdowson, A.
    Likonen, J.
    Materials migration in JET with ITER-like wall traced with a Be-10 isotopic marker2015In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 463, p. 773-776Article in journal (Refereed)
    Abstract [en]

    The current configuration of JET with ITER-like Wall (ILW) is the best available proxy for the ITER first wall. Beryllium redistribution in JET-ILW can be used for estimates of its migration in ITER. To trace it, a localized isotopic Be marker has been implemented. A bulk Be-9 tile has been enriched with Be-10 up to atomic concentrations of 1.7 x 10(-9) and installed at the inner midplane of JET before the campaign. During the 2012 shutdown over 100 surface samples were taken non destructively from surfaces of two toroidally opposite limiter beams. The absolute areal densities of the marker were inferred from Be-15 atomic concentration in each sample, measured with Accelerator Mass Spectrometry with sensitivity <10(-14). The results of marker mapping are compared with predictions made with the ASCOT orbit following code.

  • 84.
    Bykov, Igor
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Petersson, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Bergsåker, Henric
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Hallén, A.
    Possnert, G.
    Investigation of tritium analysis methods for ion microbeam application2012In: Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, ISSN 0168-583X, E-ISSN 1872-9584, Vol. 273, p. 250-253Article in journal (Refereed)
    Abstract [en]

    The trapping and retention of tritium in deposited layers on plasma-facing components is a critical issue for the international tokamak experimental reactor (ITER) and for future power producing tokamak fusion reactors. Cross sections of deposited layers at surfaces in the JET tokamak divertor are being investigated using ion microbeam analysis. To include tritium analysis with high spatial resolution, a number of plausible ion beam techniques have been investigated. Calibration samples with 150 nm tritiated titanium films were used. Absolute concentrations were determined with classical ERD using 2.5-3.5 MeV C-12(+). Cross sections for non-Rutherford ERD and for the T(C-12,p)C-14 and T(C-12,alpha)B-11 nuclear reactions were measured for different angles in the energy range 2.5-15 MeV. Background spectra were collected from pure carbon, beryllium and deuterium enriched samples and the sensitivity for microbeam NRA measurements of the tritium concentration in thick targets with predominantly Be-C-D matrix was estimated.

  • 85.
    Bykov, Igor
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Vignitchouk, Ladislas
    KTH, School of Electrical Engineering (EES), Space and Plasma Physics.
    Ratynskaia, Svetlana
    KTH, School of Electrical Engineering (EES), Space and Plasma Physics.
    Banon, Jean-Philippe
    KTH, School of Electrical Engineering (EES), Space and Plasma Physics.
    Tolias, Panagiotis
    KTH, School of Electrical Engineering (EES), Space and Plasma Physics.
    Bergsåker, Henric
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Frassinetti, Lorenzo
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Brunsell, Per R.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Transport asymmetry and release mechanisms of metal dust in the reversed-field pinch configuration2014In: Plasma Physics and Controlled Fusion, ISSN 0741-3335, E-ISSN 1361-6587, Vol. 56, no 3, p. 035014-Article in journal (Refereed)
    Abstract [en]

    Experimental data on dust resident in the EXTRAP T2R reversed-field pinch are reported. Mobile dust grains are captured in situ by silicon collectors, whereas immobile grains are sampled post mortem from the wall by adhesive tape. The simulation of collection asymmetries by the MIGRAINe dust dynamics code in combination with the experimental results is employed to deduce some characteristics of the mechanism of intrinsic dust release. All evidence suggests that re-mobilization is dominant with respect to dust production.

  • 86. Cavinato, M.
    et al.
    Gregoratto, D.
    Marchiori, G.
    Paccagnella, R.
    Brunsell, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Yadikin, Dmitriy
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Comparison of strategies and regulator design for active control of MHD modes2005In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 74, no 1-4, p. 549-553Article in journal (Refereed)
    Abstract [en]

    A system of evenly spaced poloidal arrays of saddle coils was recently installed on the reversed field pinch device EXTRAP T2R to perform experiments on the active control of MHD modes. The implementation of different control strategies, such as "intelligent shell" and "mode control", was made possible by a flexible digital control system. After giving some results on the performances of the innermost coil current control loop, two versions of "mode control" recently tested on the machine are presented. In the "wise shell" approach, equilibrium related modes are ruled out and a systematic increase of the pulse length is obtained. In a second, more model based, approach, a mode estimator/controller is designed aiming at a full state feedback by including modes, which are not directly measurable due to the limited number of available real-time signals.

  • 87.
    Cecconello, Marco
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Brunsell, Per R.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Yadikin, Dmitriy
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Drake, James Robert
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Rotation evolution of tearing modes during feedback stabilization of resistive wall modes in a reversed field pinch2005Conference paper (Refereed)
  • 88.
    Cecconello, Marco
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Frassinetti, Lorenzo
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Olofsson, Erik
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Khan, Muhammad Waqas Mehmood
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Brunsell, Per
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Resistive tearing modes dynamics with plasma control in a reversed field pinch2008In: 35th EPS Conference on Plasma Physics 2008, EPS 2008 - Europhysics Conference Abstracts: Volume 32, Issue 1, 2008, p. 429-432Conference paper (Refereed)
  • 89.
    Cecconello, Marco
    et al.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Kuldkepp, Mattias
    KTH, School of Engineering Sciences (SCI), Physics.
    Menmuir, Sheena
    KTH, School of Engineering Sciences (SCI), Physics.
    Hedqvist, Anders
    KTH, School of Engineering Sciences (SCI), Physics.
    Brunsell, Per R.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. KTH, School of Electrical Engineering (EES), Centres, Alfvén Laboratory Centre for Space and Fusion Plasma Physics.
    Current profile modifications with active feedback stabilization of resistive wall modes in a reversed field pinch2006In: Proceedings of the 33rd European Physical Society Conference on Plasma Physics, 2006, p. 1680-1683Conference paper (Refereed)
  • 90. Challis, C. D.
    et al.
    Garcia, J.
    Beurskens, M.
    Buratti, P.
    Delabie, E.
    Drewelow, P.
    Frassinetti, Lorenzo
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Giroud, C.
    Hawkes, N.
    Hobirk, J.
    Joffrin, E.
    Keeling, D.
    King, D. B.
    Maggi, C. F.
    Mailloux, J.
    Marchetto, C.
    McDonald, D.
    Nunes, I.
    Pucella, G.
    Saarelma, S.
    Simpson, J.
    Improved confinement in JET high β plasmas with an ITER-like wall2015In: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 55, no 5, article id 053031Article in journal (Refereed)
    Abstract [en]

    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. It was found that the power degradation of thermal energy confinement was weak with the ILW; much weaker than the IPB98(y,2) scaling and resulting in an increase in normalized confinement from H<inf>98</inf> ∼ 0.9 at β<inf>N</inf> ∼ 1.5 to H<inf>98</inf> ∼ 1.2-1.3 at β<inf>N</inf> ∼ 2.5 - 3.0 as the power was increased (where H<inf>98</inf> = τ<inf>E</inf>/τ<inf>IPB98(y,2)</inf> and β<inf>N</inf> = β<inf>T</inf>B<inf>T</inf>/aI<inf>P</inf> in % T/mMA). This reproduces the general trend in JET of higher normalized confinement in the so-called 'hybrid' domain, where normalized β is typically above 2.5, compared with 'baseline' ELMy H-mode plasmas with β<inf>N</inf> ∼ 1.5 - 2.0. This weak power degradation of confinement, which was also seen with the C-wall experiments at low triangularity, is due to both increased edge pedestal pressure and core pressure peaking at high power. By contrast, the high triangularity C-wall plasmas exhibited elevated H<inf>98</inf> over a wide power range with strong, IPB98(y,2)-like, power degradation. This strong power degradation of confinement appears to be linked to an increase in the source of neutral particles from the wall as the power increased, an effect that was not reproduced with the ILW. The reason for the loss of improved confinement domain at low power with the ILW is yet to be clarified, but contributing factors may include changes in the rate of gas injection, wall recycling, plasma composition and radiation. The results presented in this paper show that the choice of wall materials can strongly affect plasma performance, even changing confinement scalings that are relied upon for extrapolation to future devices.

  • 91. Chapman, I. T.
    et al.
    Graves, J. P.
    Johnson, Thomas
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Asunta, O.
    Bonoli, P.
    Choi, M.
    Jaeger, E. F.
    Jucker, M.
    Sauter, O.
    Sawtooth control in ITER using ion cyclotron resonance heating2011In: Plasma Physics and Controlled Fusion, ISSN 0741-3335, E-ISSN 1361-6587, Vol. 53, no 12, p. 124003-Article in journal (Refereed)
    Abstract [en]

    Numerical modelling of the effects of ion cyclotron resonance heating (ICRH) on the stability of the internal kink mode suggests that ICRH should be considered as an essential sawtooth control tool in ITER. Sawtooth control using ICRH is achieved by directly affecting the energy of the internal kink mode rather than through modification of the magnetic shear by driving localized currents. Consequently, ICRH can be seen as complementary to the planned electron cyclotron current drive actuator, and indeed will improve the efficacy of current drive schemes. Simulations of the ICRH distribution using independent RF codes give confidence in numerical predictions that the stabilizing influence of the fusion-born alphas can be negated by appropriately tailored minority (3)He ICRH heating in ITER. Finally, the effectiveness of all sawtooth actuators is shown to increase as the q = 1 surface moves towards the manetic axis, whilst the passive stabilization arising from the alpha and NBI particles decreases.

  • 92. Chapman, I. T.
    et al.
    Graves, J. P.
    Johnson, Thomas
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Jaeger, E. F.
    Jucker, M.
    Sauter, O.
    Sawtooth Control in ITER using Ion Cyclotron Resonance Heating2011In: Proceedings of the EPS Conference on Plasma Physics, 2011Conference paper (Other academic)
    Abstract [en]

    Numerical modelling of the effects of ion cyclotron resonance heating (ICRH) on the stability of the internal kink mode suggests that ICRH should be considered as an essential sawtooth control tool in ITER. Sawtooth control using ICRH is achieved by directly affecting the energy of the kink mode rather than through modification of the magnetic shear by driving localised currents. Consequently, ICRH can be seen as complementary to the planned electron cyclotron current drive actuator. Simulations of the ICRH distribution using independent RF codes give confidence in numerical predictions that the stabilising influence of the fusion-born alphas can be negated by appropriately tailored minority 3He ICRH heating in ITER.

  • 93. Chapman, I. T.
    et al.
    Graves, J. P.
    Lennholm, M.
    Faustin, J.
    Lerche, E.
    Johnson, Thomas
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. EUROfusion Consortium, England.
    Tholerus, Simon
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics. EUROfusion Consortium, England.
    The merits of ion cyclotron resonance heating schemes for sawtooth control in tokamak plasmas2015In: Journal of Plasma Physics, ISSN 0022-3778, E-ISSN 1469-7807, Vol. 81, no 06, article id 365810601Article in journal (Refereed)
    Abstract [en]

    JET experiments have compared the efficacy of low-and high field side ion cyclotron resonance heating (ICRH) as an actuator to deliberately minimise the sawtooth period. It is found that low-field side ICRH with low minority concentration is optimal for saw tooth control for two main reasons. Firstly, low-field side heating means that any toroidal phasing of the ICRH (-90 degrees, +90 degrees or dipole) has a destabilising effect on the sawteeth, meaning that dipole phasing can be employed, since tins is preferable due to less plasma wall interaction from Resonant Frequency (RI) sheaths. Secondly, the resonance position of the low field side ICRH does not have to be very accurately placed to achieve saw tooth control, relaxing the requirement for real-time control of the RF frequency. These empirical observations have been confirmed by hybrid kinetic-magnetohydrodynamic modelling, and suggest that the ICRH antenna design for ITER is well positioned to provide a control actuator capable of having a significant effect on the sawtooth behaviour.

  • 94. Chapman, I. T.
    et al.
    Graves, J. P.
    Sauter, O.
    Zucca, C.
    Asunta, O.
    Buttery, R. J.
    Coda, S.
    Goodman, T.
    Igochine, V.
    Johnson, Thomas
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Jucker, M.
    La Haye, R. J.
    Lennholm, M.
    Power requirements for electron cyclotron current drive and ion cyclotron resonance heating for sawtooth control in ITER2013In: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 53, no 6, p. 066001-Article in journal (Refereed)
    Abstract [en]

    13MW of electron cyclotron current drive (ECCD) power deposited inside the q = 1 surface is likely to reduce the sawtooth period in ITER baseline scenario below the level empirically predicted to trigger neoclassical tearing modes (NTMs). However, since the ECCD control scheme is solely predicated upon changing the local magnetic shear, it is prudent to plan to use a complementary scheme which directly decreases the potential energy of the kink mode in order to reduce the sawtooth period. In the event that the natural sawtooth period is longer than expected, due to enhanced a particle stabilization for instance, this ancillary sawtooth control can be provided from >10MW of ion cyclotron resonance heating (ICRH) power with a resonance just inside the q = 1 surface. Both ECCD and ICRH control schemes would benefit greatly from active feedback of the deposition with respect to the rational surface. If the q = 1 surface can be maintained closer to the magnetic axis, the efficacy of ECCD and ICRH schemes significantly increases, the negative effect on the fusion gain is reduced, and off-axis negative-ion neutral beam injection (NNBI) can also be considered for sawtooth control. Consequently, schemes to reduce the q = 1 radius are highly desirable, such as early heating to delay the current penetration and, of course, active sawtooth destabilization to mediate small frequent sawteeth and retain a small q = 1 radius. Finally, there remains a residual risk that the ECCD + ICRH control actuators cannot keep the sawtooth period below the threshold for triggering NTMs (since this is derived only from empirical scaling and the control modelling has numerous caveats). If this is the case, a secondary control scheme of sawtooth stabilization via ECCD + ICRH + NNBI, interspersed with deliberate triggering of a crash through auxiliary power reduction and simultaneous pre-emptive NTM control by off-axis ECCD has been considered, permitting long transient periods with high fusion gain. The power requirements for the necessary degree of sawtooth control using either destabilization or stabilization schemes are expected to be within the specification of anticipated ICRH and ECRH heating in ITER, provided the requisite power can be dedicated to sawtooth control.

  • 95. Chapman, I. T.
    et al.
    Liu, Y. Q.
    Asunta, O.
    Graves, J. P.
    Johnson, Thomas
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Jucker, M.
    Kinetic damping of resistive wall modes in ITER2012In: Physics of Plasmas, ISSN 1070-664X, E-ISSN 1089-7674, Vol. 19, no 5, p. 052502-Article in journal (Refereed)
    Abstract [en]

    Full drift kinetic modelling including finite orbit width effects has been used to assess the passive stabilisation of the resistive wall mode (RWM) that can be expected in the ITER advanced scenario. At realistic plasma rotation frequency, the thermal ions have a stabilising effect on the RWM, but the stability limit remains below the target plasma pressure to achieve Q = 5. However, the inclusion of damping arising from the fusion-born alpha particles, the NBI ions, and ICRH fast ions extends the RWM stability limit above the target beta for the advanced scenario. The fast ion damping arises primarily from finite orbit width effects and is not due to resonance between the particle frequencies and the instability.

  • 96. Chapman, I. T.
    et al.
    Pinches, S. D.
    Graves, J. P.
    Akers, R. J.
    Appel, L. C.
    Budny, R. V.
    Coda, S.
    Conway, N. J.
    de Bock, M.
    Eriksson, L-G
    Hastie, R. J.
    Hender, T. C.
    Huysmans, G. T. A.
    Johnson, Thomas J.
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Koslowski, H. R.
    Kraemer-Flecken, A.
    Lennholm, M.
    Liang, Y.
    Saarelma, S.
    Sharapov, S. E.
    Voitsekhovitch, I.
    The physics of sawtooth stabilization2007In: Plasma Physics and Controlled Fusion, ISSN 0741-3335, E-ISSN 1361-6587, Vol. 49, no 12B, p. B385-B394Article in journal (Refereed)
    Abstract [en]

    Long period sawteeth have been observed to result in low-beta triggering of neo-classical tearing modes, which can significantly degrade plasma confinement. Consequently, a detailed physical understanding of sawtooth behaviour is critical, especially for ITER where fusion-born a particles are likely to lead to very long sawtooth periods. Many techniques have been developed to control, and in particular to destabilize the sawteeth. The application of counter-current neutral beam injection (NBI) in JET has resulted in shorter sawtooth periods than in Ohmic plasmas. This result has been explained because, firstly, the counter-passing fast ions give a destabilizing contribution to the n=1 internal kink mode-which is accepted to be related to sawtooth oscillations-and secondly, the flow shear strongly influences the stabilizing trapped particles. A similar experimental result has been observed in counter-NBI heated plasmas in MAST. However, the strong toroidal flows in spherical tokamaks mean that the sawtooth behaviour is determined by the gyroscopic flow stabilization of the kink mode rather than kinetic effects. In NBI heated plasmas in smaller conventional aspect-ratio tokamaks, such as TEXTOR, the flow and kinetic effects compete to give different sawtooth behaviour. Other techniques applied to destabilize sawteeth are the application of electron cyclotron current drive (ECCD) or ion cyclotron resonance heating (ICRH). In JET, it has been observed that localized ICRH is able to destabilize sawteeth which were otherwise stabilized by a co-existing population of energetic trapped ions in the core. This is explained through the dual role of the ICRH in reducing the critical magnetic shear required to trigger a sawtooth crash, and the increase in the local magnetic shear which results from driving current near the q=1 rational surface. Sawtooth control in ITER could be provided by a combination of ECCD and co-passing off-axis negative-NBI fast ions.

  • 97. Cherigier-Kovacic, L.
    et al.
    Ström, Petter
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Lejeune, A.
    Doveil, F.
    Electric field induced Lyman-alpha emission of a hydrogen beam for electric field measurements2015In: Review of Scientific Instruments, ISSN 0034-6748, E-ISSN 1089-7623, Vol. 86, no 6, article id 063504Article in journal (Refereed)
    Abstract [en]

    Electric field induced Lyman-alpha emission is a new way of measuring weak electric fields in vacuum and in a plasma. It is based on the emission of Lyman-alpha radiation (121.6 nm) by a low-energy metastable H atom beam due to Stark-quenching of the 2s level induced by the field. In this paper, we describe the technique in detail. Test measurements have been performed in vacuum between two plates polarized at a controlled voltage. The intensity of emitted radiation, proportional to the square of the field modulus, has been recorded by a lock-in technique, which gives an excellent signal to noise ratio. These measurements provide an in situ calibration that can be used to obtain the absolute value of the electric field. A diagnostic of this type can help to address a long standing challenge in plasma physics, namely, the problem of measuring electric fields without disturbing the equilibrium of the system that is being studied.

  • 98. Chérigier-Kovacic, L.
    et al.
    Ström, Petter
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Doveil, F.
    Electric field induced lyman-α Emission (EFILE) diagnostic for electric field measurements2015In: Proceedings of Science, Proceedings of Science (PoS) , 2015Conference paper (Refereed)
    Abstract [en]

    When a metastable hydrogen test beam is exposed to a constant or oscillating electric field, Lyman-A (121.6 nm) emission occurs. This results from the Stark quenching of the metastable 2s level, induced by the field. The intensity of the radiation is proportional to the square of the electric field amplitude and it is recorded by a lock-in technique, which gives an excellent signal to noise ratio. This provides us with a very sensitive and non intrusive method to measure the electric field value, called EFILE (Electric Field Induced Lyman-A emission). Sensitivity is as good as 0.1 V/cm in the case of an oscillatory field resonant with the Lamb shift frequency ≈ 1 GHz. Hydrogen ions are produced in a magnetic multicusp source by a thermo-electronic discharge. The ions are extracted from the source, focused by a series of electrostatic lenses and accelerated to 500 eV. The beam interacts with cesium vapor which produces atoms in the metastable 2s1=2 state. In the diagnosed volume, the beam passes between a pair of plane electrodes separated by 5 cm. One of them is grounded, the other one is polarized to generate an electric field. The diagnosed volume can be kept under vacuum or exposed to an argon plasma. Lyman-A emission from the beam passing between the plates is measured as a function of the polarized plate voltage. A saturation of the signal is observed at large field amplitudes, which is explained through oscillatory and geometrical mechanisms. A function that takes this saturation into account is used as a calibration for the subsequent electric field profile measurements in the case of a constant voltage applied between the plates in vacuum. We find a good agreement between our results and a finite element method calculation of the profile.

  • 99. Citrin, J.
    et al.
    Garcia, J.
    Görier, T.
    Jenko, F.
    Mantica, P.
    Told, D.
    Bourdelle, C.
    Hatch, D. R.
    Hogeweij, G. M. D.
    Johnson, Thomas
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Pueschel, M. J.
    Schneider, M.
    Electromagnetic stabilization of tokamak microturbulence in a high-β regime2014In: 41st EPS Conference on Plasma Physics, EPS 2014, European Physical Society (EPS) , 2014Conference paper (Refereed)
    Abstract [en]

    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.

  • 100. Citrin, J.
    et al.
    Garcia, J.
    Görler, T.
    Jenko, F.
    Mantica, P.
    Told, D.
    Bourdelle, C.
    Hatch, D. R.
    Hogeweij, G. M. D.
    Johnson, Thomas
    KTH, School of Electrical Engineering (EES), Fusion Plasma Physics.
    Pueschel, M. J.
    Schneider, M.
    Electromagnetic stabilization of tokamak microturbulence in a high-beta regime2015In: Plasma Physics and Controlled Fusion, ISSN 0741-3335, E-ISSN 1361-6587, Vol. 57, no 1, p. 014032-Article in journal (Refereed)
    Abstract [en]

    The impact of electromagnetic stabilization and flow shear stabilization on ITG turbulence is investigated. Analysis of a low-beta 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-beta 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-beta generalization of comparable observations found at low-beta 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.

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