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  • 1. Aiba, N.
    et al.
    Giroud, C.
    Honda, M.
    Delabie, E.
    Saarelma, S.
    Frassinetti, L
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Lupelli, I.
    Casson, F. J.
    Pamela, S.
    Urano, H.
    Maggi, C. F.
    Numerical analysis of ELM stability with rotation and ion diamagnetic drift effects in JET2017Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 57, nr 12, artikel-id 126001Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Stability to the type-I edge localized mode (ELM) in JET plasmas was investigated numerically by analyzing the stability to a peeling-ballooning mode with the effects of plasma rotation and ion diamagnetic drift. The numerical analysis was performed by solving the extended Frieman-Rotenberg equation with the MINERVA-DI code. To take into account these effects in the stability analysis self-consistently, the procedure of JET equilibrium reconstruction was updated to include the profiles of ion temperature and toroidal rotation, which are determined based on the measurement data in experiments. With the new procedure and MINERVA-DI, it was identified that the stability analysis including the rotation effect can explain the ELM trigger condition in JET with ITER like wall (JET-ILW), though the stability in JET with carbon wall (JET-C) is hardly affected by rotation. The key difference is that the rotation shear in JET-ILW plasmas analyzed in this study is larger than that in JET-C ones, the shear which enhances the dynamic pressure destabilizing a peeling-ballooning mode. In addition, the increase of the toroidal mode number of the unstable MHD mode determining the ELM trigger condition is also important when the plasma density is high in JET-ILW. Though such modes with high toroidal mode number are strongly stabilized by the ion diamagnetic drift effect, it was found that plasma rotation can sometimes overcome this stabilizing effect and destabilizes the peeling-ballooning modes in JET-ILW.

  • 2. Baron-Wiechec, A.
    et al.
    Fortuna-Zalesna, E.
    Grzonka, J.
    Rubel, Marek
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Widdowson, A.
    Ayres, C.
    Coad, J. P.
    Hardie, C.
    Heinola, K.
    Matthews, G. F.
    First dust study in JET with the ITER-like wall: sampling, analysis and classification2015Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 55, nr 11, artikel-id 113033Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Results of the first dust survey in JET with the ITER-Like Wall (JET-ILW) are presented. The sampling was performed using adhesive stickers from the divertor tiles where the greatest material deposition was detected after the first JET-ILW campaign in 2011-2012. The emphasis was especially on sampling and analysis of metal particles (Be and W) with the aim to determine the composition, size, surface topography and internal dust structure using a large set of methods: high-resolution scanning and transmission electron microscopy, focused ion beam, electron diffraction and also wavelength and energy dispersive x-ray spectroscopy. The most important was the identification of beryllium dust both in the form of flakes and droplets with dimensions in the micrometer range. Tungsten, molybdenum, inconel constituents were identified along with many impurity species. The particles are categorised and the origin of the various constituents discussed.

  • 3.
    Ben Yaala, M.
    et al.
    Univ Basel, Dept Phys, Klingelbergstr 82, CH-4056 Basel, Switzerland..
    Moser, L.
    Univ Basel, Dept Phys, Klingelbergstr 82, CH-4056 Basel, Switzerland..
    Steiner, R.
    Univ Basel, Dept Phys, Klingelbergstr 82, CH-4056 Basel, Switzerland..
    Butoi, B.
    Natl Inst Laser Plasma & Radiat Phys, 409 Atomistilor St, Magurele 077125, Romania..
    Dinca, P.
    Natl Inst Laser Plasma & Radiat Phys, 409 Atomistilor St, Magurele 077125, Romania..
    Petersson, Per
    KTH, Skolan för elektroteknik och datavetenskap (EECS), Fusionsplasmafysik.
    Marot, L.
    Univ Basel, Dept Phys, Klingelbergstr 82, CH-4056 Basel, Switzerland..
    Meyer, E.
    Univ Basel, Dept Phys, Klingelbergstr 82, CH-4056 Basel, Switzerland..
    Deuterium as a cleaning gas for ITER first mirrors: experimental study on beryllium deposits from laboratory and JET-ILW2019Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 59, nr 9, artikel-id 096027Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Cleaning techniques for metallic first mirrors are needed in more than 20 optical diagnostic systems from ITER to avoid reflectivity losses. Plasma sputtering is considered as one of the most promising techniques to remove deposits coming from the main wall (mainly beryllium and tungsten). Previous plasma cleaning studies were conducted on mirrors contaminated with beryllium and tungsten where argon and/or helium were employed as process gas, demonstrating removal of contamination and recovery of optical properties. Still, both abovementioned process gases have a non-negligible sputtering yield on mirrors. In this work, we explored the possibility to use a sputter gas having a small impact on mirrors while being efficient on Be deposits, e.g. deuterium. Two sputtering regimes were studied, on laboratory deposits as well as on mirrors exposed in .TET-ILW, namely physical sputtering (220eV ion energy) and chemically assisted physical sputtering (60 eV ion energy) using capacitively coupled plasma with radio frequency. The removal of Be and mixed Be/W contaminants, as well as the recovery of reflectivity, was achieved when deuterium was employed at 220eV while cleaning at 60eV was only fully efficient on laboratory beryllium deposits. On mirrors exposed in JET-ILW, the situation is more complex due to the presence of tungsten in the contaminant film, leading to the formation of a tungsten enriched surface that is not easily sputtered, especially at 60eV.

  • 4.
    Bergkvist, Tommy
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Hellsten, Torbjörn
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Holmström, Kerstin
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Non-linear dynamics of Alfvén eigenmodes excited by thermonulcear alpha particles in the presence of ion cyclotron resonance heating2007Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 47, nr 9, s. 1131-1141Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Alfvén eigenmodes (AEs) excited by thermonuclear α-particles can degrade the heating efficiency by spatial redistribution of the resonant α-particles. Changes of the orbit invariants in phase space by collisions and interactions with other waves, such as magnetosonic waves during ion cyclotron resonance heating (ICRH), lead to changes in the phase between the α-particles and AEs, causing a decorrelation of the interactions and stronger redistribution of the α-particles. Cyclotron interactions increase the decorrelation of the AE interactions with the high-energy ions and hence a stronger radial redistribution of the high-energy α-particles by the AEs. Renewal of the distribution function by thermonuclear reactions and losses of α-particles to the wall lead to a continuous drive of the AEs and a radial redistribution of the α-particles. The condition for excitation of AEs is shown to depend on the heating scenario where heating at the low field side creates a significant population of high-energy non-standard orbits which drive the modes. The redistribution results in a reduction in the averaged α-particle energy and a degradation of the heating efficiency. The effect on the distribution function in the presence of several unstable modes is not additive and the particle redistribution is found to saturate with an increasing number of modes.

  • 5.
    Bergkvist, Tommy
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Hellsten, Torbjörn
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Johnson, T.
    Laxåback, Martin
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Non-linear study of fast particle excitation of global Alfvén eigenmodes during ICRH2005Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 45, s. 485-493Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    High-power ion–cyclotron resonance heating (ICRH) can produce centrally peaked fast ion distributions with wide non-standard drift orbits exciting Alfvén eigenmodes (AEs). The dynamics of the AE excitation depends not only on the anisotropy and the peaking of the fast ion distribution but also on the decorrelation of the AE interactions and the renewal of the fast ions resonant with the AE by ion–cyclotron interactions. A method of self-consistently including the evolution of the distribution function of fast ions during excitation of AEs and ICRH has been developed and implemented in the SELFO code. Numerical simulations of the AE dynamics and ICRH give a variation of the AE amplitude consistent with the experimentally observed splitting of the mode frequency. The experimentally observed fast damping of the mode as the ICRH is switched off is also evident in the simulations.

  • 6.
    Bergsåker, Henric
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Possnert, G.
    Bykov, Igor
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Heinola, K.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Miettunen, J.
    Widdowson, A.
    Riccardo, V.
    Nunes, I.
    Stamp, M.
    Brezinsek, S.
    Groth, M.
    Kurki-Suonio, T.
    Likonen, J.
    Coad, J. P.
    Borodin, D.
    Kirschner, A.
    Schmid, K.
    Krieger, K.
    First results from the Be-10 marker experiment in JET with ITER-like wall2014Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 54, nr 8, s. 082004-Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    When the ITER-like wall was installed in JET, one of the 218 Be inner wall guard limiter tiles had been enriched with Be-10 as a bulk isotopic marker. During the shutdown in 2012-2013, a set of tiles were sampled nondestructively to collect material for accelerator mass spectroscopy measurements of Be-10 concentration. The letter shows how the marker experiment was set up, presents first results and compares them to preliminary predictions of marker redistribution, made with the ASCOT numerical code. Finally an outline is shown of what experimental data are likely to become available later and the possibilities for comparison with modelling using the WallDYN, ERO and ASCOT codes are discussed.

  • 7. Berk, H. L.
    et al.
    Boswell, C. J.
    Borba, D.
    Figueiredo, A. C. A.
    Johnson, Thomas J.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Nave, M. F. F.
    Pinches, S. D.
    Sharapov, S. E.
    Explanation of the JET n=0 chirping mode2006Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 46, nr 10, s. S888-S897Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Persistent rapid up and down frequency chirping modes with a toroidal mode number of zero (n = 0) are observed in the JET tokamak when energetic ions, in the range of several hundred keV, are created by high field side ion cyclotron resonance frequency heating. Fokker-Planck calculations demonstrate that the heating method enables the formation of an energetically inverted ion distribution which supplies the free energy for the ions to excite a mode related to the geodesic acoustic mode. The large frequency shifts of this mode are attributed to the formation of phase space structures whose frequencies, which are locked to an ion orbit bounce resonance frequency, are forced to continually shift so that energetic particle energy can be released to counterbalance the energy dissipation present in the background plasma.

  • 8. Beurskens, M. N. A.
    et al.
    Arnoux, G.
    Brezinsek, A. S.
    Rachlew, Elisabeth
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Atom- och molekylfysik.
    Saarelma, S.
    Solano, E.
    et al,
    Pedestal and ELM response to impurity seeding in JET advanced scenario plasmas2008Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 48, nr 9Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Advanced scenario plasmas must often be run at low densities and high power, leading to hot edge temperatures and consequent power handling issues at plasma - surface interaction zones. Experiments at JET are addressing this issue by exploring the use of extrinsic impurity seeding and D-2 puffing to reduce heat fluxes. The experiments presented in this paper continue the line of advanced tokamak ( AT) scenario studies at high triangularity in JET by concentrating on the characterization of the edge pedestal and the ELM behaviour with deuterium and/or light impurity fuelling (neon, nitrogen). Both injection of extrinsic impurities and D2 puffing are shown to have a significant impact on the edge pedestal in typical JET AT conditions. The ELM energy loss, Delta W-ELM/W-dia, can be reduced to below 3% and the maximum ELM penetration depth can be limited to r/a > 0.7, thus enhancing the possibility for sustainable internal transport barriers at large plasma radius. These conditions can be achieved in two separate domains, either at a radiated power fraction (F-rad) of 30% or at a fraction of > 50%. At the lower Frad the ELMs are type I and a high pedestal pressure is maintained, but the occasional large ELM may still occur. At F-rad > 50% the pedestal pressure is degraded by 30-50%, but the ELMs are degraded to type III. The intermediate regime at F-rad similar to 40% is unattractive for ITB scenarios because large type I ELMs occur intermittently during the predominantly type III ELM phases (compound type I/III). F-rad = 30% can be obtained with D-2 fuelling alone, whereas neon or nitrogen seeding is needed to achieve F-rad > 50%. Only a limited number of tests have been carried out with nitrogen seeding, with the preliminary conclusion that the plasma edge behaviour is similar to that with neon seeding once the radiated fraction is matched.

  • 9. Beurskens, M. N. A.
    et al.
    Dunne, M. G.
    Frassinetti, Lorenzo
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Bernert, M.
    Cavedon, M.
    Fischer, R.
    Järvinen, A.
    Kallenbach, A.
    Laggner, F. M.
    McDermott, R. M.
    Potzel, S.
    Schweinzer, J.
    Tardini, G.
    Viezzer, E.
    Wolfrum, E.
    The role of carbon and nitrogen on the H-mode confinement in ASDEX Upgrade with a metal wall2016Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 56, nr 5, artikel-id 056014Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Carbon (CD4) and nitrogen (N2) have been seeded in ASDEX Upgrade (AUG) with a tungsten wall and have both led to a 20-30% confinement improvement. The reference plasma is a standard target plasma with I p /B T = 1 MA/2.5 T, total input power P tot ∼ 12 MW and normalized pressure of β N ∼ 1.8. Carbon and nitrogen are almost perfectly exchangeable for the core, pedestal and divertor plasma in this experiment where impurity concentrations of C and N of 2% are achieved and Z eff only mildly increases from ∼1.3 to ∼1.7. As the radiation potentials of C and N are similar and peak well below 100 eV, both impurities act as divertor radiators and radiate well outside the pedestal region. The outer divertor is purposely kept in an attached state when C and N are seeded to avoid confinement degradation by detachment. As reported in earlier publications for nitrogen, carbon is also seen to reduce the high field side high density (the so-called HFSHD) in the scrape off layer above the inner divertor strike point by about 50%. This is accompanied by a confinement improvement for both low (δ ∼ 0.25) and high (δ ∼ 0.4) triangularity configurations for both seeding gases, due to an increase of pedestal temperature and stiff core temperature profiles. The electron density profiles show no apparent change due to the seeding. As an orthogonal effect, increasing the triangularity leads to an additionally increased pedestal density, independent of the impurity seeding. This experiment further closes the gap in understanding the confinement differences observed in carbon and metal wall devices; the absence of carbon can be substituted by nitrogen which leads to a similar confinement benefit. So far, no definite physics explanation for the confinement enhancement has been obtained, but the experimental observations in this paper provide input for further model development.

  • 10. Beurskens, M. N. A.
    et al.
    Frassinetti, Lorenzo
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Challis, C.
    Giroud, C.
    Saarelma, S.
    Alper, B.
    Angioni, C.
    Bilkova, P.
    Bourdelle, C.
    Brezinsek, S.
    Buratti, P.
    Calabro, G.
    Eich, T.
    Flanagan, J.
    Giovannozzi, E.
    Groth, M.
    Hobirk, J.
    Joffrin, E.
    Leyland, M. J.
    Lomas, P.
    de la Luna, E.
    Kempenaars, M.
    Maddison, G.
    Maggi, C.
    Mantica, P.
    Maslov, M.
    Matthews, G.
    Mayoral, M-L
    Neu, R.
    Nunes, I.
    Osborne, T.
    Rimini, F.
    Scannell, R.
    Solano, E. R.
    Snyder, P. B.
    Voitsekhovitch, I.
    de Vries, Peter
    Global and pedestal confinement in JET with a Be/W metallic wall2014Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 54, nr 4, s. 043001-Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Type I ELMy H-mode operation in JET with the ITER-like Be/W wall (JET-ILW) generally occurs at lower pedestal pressures compared to those with the full carbon wall (JET-C). The pedestal density is similar but the pedestal temperature where type I ELMs occur is reduced and below to the so-called critical type I-type III transition temperature reported in JET-C experiments. Furthermore, the confinement factor H-98(y,H- 2) in type I ELMy H-mode baseline plasmas is generally lower in JET-ILWcompared to JET-C at low power fractions Ploss/P-thr,(08)< 2 (where P-loss is (P-in-dW/dt), and P-thr,(08) the L-H power threshold from Martin et al 2008 (J. Phys. Conf. Ser. 123 012033)). Higher power fractions have thus far not been achieved in the baseline plasmas. At Ploss/P-thr,P- 08 > 2, the confinement in JET-ILW hybrid plasmas is similar to that in JET-C. A reduction in pedestal pressure is the main reason for the reduced confinement in JET-ILW baseline ELMy H-mode plasmas where typically H-98((y, 2)) = 0.8 is obtained, compared to H-98((y, 2)) = 1.0 in JET-C. In JET-ILW hybrid plasmas a similarly reduced pedestal pressure is compensated by an increased peaking of the core pressure profile resulting in H-98((y, 2)) <= 1.25. The pedestal stability has significantly changed in high triangularity baseline plasmas where the confinement loss is also most apparent. Applying the same stability analysis for JET-C and JET-ILW, the measured pedestal in JET-ILW is stable with respect to the calculated peeling-ballooning stability limit and the ELM collapse time has increased to 2ms from typically 200 mu s in JET-C. This indicates that changes in the pedestal stability may have contributed to the reduced pedestal confinement in JET-ILW plasmas. A comparison of EPED1 pedestal pressure prediction with JET-ILW experimental data in over 500 JET-C and JET-ILW baseline and hybrid plasmas shows a good agreement with 0.8 < (measured p(ped))/(predicted p(ped), EPED) < 1.2, but that the role of triangularity is generally weaker in the JET-ILW experimental data than in the model predictions.

  • 11. Beurskens, M. N. A.
    et al.
    Frassinetti, Lorenzo
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Challis, C.
    Osborne, T.
    Snyder, P. B.
    Alper, B.
    Angioni, C.
    Bourdelle, C.
    Buratti, P.
    Crisanti, F.
    Giovannozzi, E.
    Giroud, C.
    Groebner, R.
    Hobirk, J.
    Jenkins, I.
    Joffrin, E.
    Leyland, M. J.
    Lomas, P.
    Mantica, P.
    McDonald, D.
    Nunes, I.
    Rimini, F.
    Saarelma, S.
    Voitsekhovitch, I.
    De Vries, P.
    Zarzoso, D.
    Comparison of hybrid and baseline ELMy H-mode confinement in JET with the carbon wall2013Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 53, nr 1, s. 013001-Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The confinement in JET baseline type I ELMy H-mode plasmas is compared to that in so-called hybrid H-modes in a database study of 112 plasmas in JET with the carbon fibre composite (CFC) wall. The baseline plasmas typically have βN ∼ 1.5-2, H98 ∼ 1, whereas the hybrid plasmas have βN ∼ 2.5-3, H98 &lt; 1.5. The database study contains both low- (δ ∼ 0.2-0.25) and high-triangularity (δ ∼ 0.4) hybrid and baseline H-mode plasmas from the last JET operational campaigns in the CFC wall from the period 2008-2009. Based on a detailed confinement study of the global as well as the pedestal and core confinement, there is no evidence that the hybrid and baseline plasmas form separate confinement groups; it emerges that the transition between the two scenarios is of a gradual kind rather than demonstrating a bifurcation in the confinement. The elevated confinement enhancement factor H98 in the hybrid plasmas may possibly be explained by the density dependence in the τ98 scaling as n0.41 and the fact that the hybrid plasmas operate at low plasma density compared to the baseline ELMy H-mode plasmas. A separate regression on the confinement data in this study shows a reduction in the density dependence as n0.09±0.08. Furthermore, inclusion of the plasma toroidal rotation in the confinement regression provides a scaling with the toroidal Alfvén Mach number as and again a reduced density dependence as n0.15±0.08. The differences in pedestal confinement can be explained on the basis of linear MHD stability through a coupling of the total and pedestal poloidal pressure and the pedestal performance can be improved through plasma shaping as well as high β operation. This has been confirmed in a comparison with the EPED1 predictive pedestal code which shows a good agreement between the predicted and measured pedestal pressure within 20-30% for a wide range of βN ∼ 1.5-3.5. The core profiles show a strong degree of pressure profile consistency. No beneficial effect of core density peaking on confinement could be identified for the majority of the plasmas presented here as the density peaking is compensated by a temperature de-peaking resulting in no or only a weak variation in the pressure peaking. The core confinement could only be optimized in case the ions and electrons are decoupled, in which case the ion temperature profile peaking can be enhanced, which benefits confinement. In this study, the latter has only been achieved in the low-triangularity hybrid plasmas, and can be attributed to low-density operation. Plasma rotation has been found to reduce core profile stiffness, and can explain an increase in profile peaking at small radius ρtor = 0.3.

  • 12.
    Blanken, T. C.
    et al.
    Eindhoven Univ Technol, Control Syst Technol Grp, Dept Mech Engn, POB 513, NL-5600 MB Eindhoven, Netherlands.;Eindhoven Univ Technol, POB 513, NL-5600 MB Eindhoven, Netherlands..
    Frassinetti, Lorenzo
    KTH, Skolan för elektroteknik och datavetenskap (EECS), Fusionsplasmafysik.
    Fridström, Richard
    KTH, Skolan för elektroteknik och datavetenskap (EECS), Fusionsplasmafysik.
    Garcia-Carrasco, Alvaro
    KTH, Skolan för elektroteknik och datavetenskap (EECS), Fusionsplasmafysik.
    Hellsten, Torbjörn
    KTH, Skolan för elektroteknik och datavetenskap (EECS), Fusionsplasmafysik.
    Jonsson, T.
    KTH, Skolan för elektroteknik och datavetenskap (EECS), Fusionsplasmafysik.
    Petersson, Per
    KTH, Skolan för elektroteknik och datavetenskap (EECS), Fusionsplasmafysik.
    Ratynskaia, Svetlana
    KTH, Skolan för elektroteknik och datavetenskap (EECS), Fusionsplasmafysik.
    Tolias, Panagiotis
    KTH, Skolan för elektroteknik och datavetenskap (EECS), Fusionsplasmafysik.
    Vallejos, Pablo
    KTH, Skolan för elektroteknik och datavetenskap (EECS), Fusionsplasmafysik.
    Vignitchouk, Ladislas
    KTH, Skolan för elektroteknik och datavetenskap (EECS), Fusionsplasmafysik.
    Dori, V
    Univ Split, Fac Elect Engn Mech Engn & Naval Architecture, R Boskovica 32, Split 21000, Croatia..
    Real-time plasma state monitoring and supervisory control on TCV2019Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 59, nr 2, artikel-id 026017Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    In ITER and DEMO, various control objectives related to plasma control must be simultaneously achieved by the plasma control system (PCS), in both normal operation as well as off-normal conditions. The PCS must act on off-normal events and deviations from the target scenario, since certain sequences (chains) of events can precede disruptions. It is important that these decisions are made while maintaining a coherent prioritization between the real-time control tasks to ensure high-performance operation. In this paper, a generic architecture for task-based integrated plasma control is proposed. The architecture is characterized by the separation of state estimation, event detection, decisions and task execution among different algorithms, with standardized signal interfaces. Central to the architecture are a plasma state monitor and supervisory controller. In the plasma state monitor, discrete events in the continuous-valued plasma state arc modeled using finite state machines. This provides a high-level representation of the plasma state. The supervisory controller coordinates the execution of multiple plasma control tasks by assigning task priorities, based on the finite states of the plasma and the pulse schedule. These algorithms were implemented on the TCV digital control system and integrated with actuator resource management and existing state estimation algorithms and controllers. The plasma state monitor on TCV can track a multitude of plasma events, related to plasma current, rotating and locked neoclassical tearing modes, and position displacements. In TCV experiments on simultaneous control of plasma pressure, safety factor profile and NTMs using electron cyclotron heating (ECI I) and current drive (ECCD), the supervisory controller assigns priorities to the relevant control tasks. The tasks are then executed by feedback controllers and actuator allocation management. This work forms a significant step forward in the ongoing integration of control capabilities in experiments on TCV, in support of tokamak reactor operation.

  • 13. Bonanomi, N.
    et al.
    Mantica, P.
    Di Siena, A.
    Delabie, E.
    Giroud, C.
    Johnson, Thomas
    KTH.
    Lerche, E.
    Menmuir, S.
    Tsalas, M.
    Van Eester, D.
    Turbulent transport stabilization by ICRH minority fast ions in low rotating JET ILW L-mode plasmas2018Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 58, nr 5, artikel-id 056025Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The first experimental demonstration that fast ion induced stabilization of thermal turbulent transport takes place also at low values of plasma toroidal rotation has been obtained in JET ILW (ITER-like wall) L-mode plasmas with high (He-3)-D ICRH (ion cyclotron resonance heating) power. A reduction of the gyro-Bohm normalized ion heat flux and higher values of the normalized ion temperature gradient have been observed at high ICRH power and low NBI (neutral beam injection) power and plasma rotation. Gyrokinetic simulations indicate that ITG (ion temperature gradient) turbulence stabilization induced by the presence of high-energetic He-3 ions is the key mechanism in order to explain the experimental observations. Two main mechanisms have been identified to be responsible for the turbulence stabilization: a linear electrostatic wave-fast particle resonance mechanism and a nonlinear electromagnetic mechanism. The dependence of the stabilization on the He-3 distribution function has also been studied.

  • 14. Bowman, C.
    et al.
    Dickinson, D.
    Horvath, L.
    Lunniss, A. E.
    Wilson, H. R.
    Cziegler, I.
    Frassinetti, Lorenzo
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. KTH, Skolan för elektro- och systemteknik (EES), Rymd- och plasmafysik.
    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 wall2018Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 58, nr 1, artikel-id 016021Artikel i tidskrift (Refereegranskat)
    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.

  • 15.
    Brezinsek, S.
    et al.
    Forschungszentrum Julich, Inst Energie & Klimaforsch Plasmaphys, TEC, D-52425 Julich, Germany..
    Kirschner, A.
    Forschungszentrum Julich, Inst Energie & Klimaforsch Plasmaphys, TEC, D-52425 Julich, Germany..
    Mayer, M.
    Max Planck Inst Plasma Phys, D-85748 Garching, Germany..
    Baron-Wiechec, A.
    CCFE Fus Assoc, Culham Sci Ctr, Abingdon OX14 3DB, Oxon, England..
    Borodkina, I
    Czech Acad Sci, Inst Plasma Phys, Prague 18200, Czech Republic..
    Borodin, D.
    Forschungszentrum Julich, Inst Energie & Klimaforsch Plasmaphys, TEC, D-52425 Julich, Germany..
    Coffey, I
    CCFE Fus Assoc, Culham Sci Ctr, Abingdon OX14 3DB, Oxon, England..
    Coenen, J.
    Forschungszentrum Julich, Inst Energie & Klimaforsch Plasmaphys, TEC, D-52425 Julich, Germany..
    den Harder, N.
    Max Planck Inst Plasma Phys, D-85748 Garching, Germany..
    Eksaeva, A.
    Forschungszentrum Julich, Inst Energie & Klimaforsch Plasmaphys, TEC, D-52425 Julich, Germany..
    Guillemaut, C.
    Univ Lisbon, Inst Super Tecn, Inst Plasmas & Fusao Nucl, Lisbon, Portugal..
    Heinola, K.
    IAEA, POB 100, A-1400 Vienna, Austria.;Univ Helsinki, Dept Phys, POB 64, FIN-00014 Helsinki, Finland..
    Huber, A.
    Forschungszentrum Julich, Inst Energie & Klimaforsch Plasmaphys, TEC, D-52425 Julich, Germany..
    Huber, V
    Forschungszentrum Julich, Inst Energie & Klimaforsch Plasmaphys, TEC, D-52425 Julich, Germany..
    Imrisek, M.
    Czech Acad Sci, Inst Plasma Phys, Prague 18200, Czech Republic..
    Jachmich, S.
    Ecole Royale Mil, LPP, Koninkllijke Mil Sch, B-1000 Brussels, Belgium..
    Pawelec, E.
    Opole Univ, Inst Phys, Oleska 48, PL-45052 Opole, Poland..
    $$$Rubel, M.
    Royal Inst Technol KTH, SE-10044 Stockholm, Sweden..
    Krat, S.
    Max Planck Inst Plasma Phys, D-85748 Garching, Germany..
    Sergienko, G.
    Forschungszentrum Julich, Inst Energie & Klimaforsch Plasmaphys, TEC, D-52425 Julich, Germany..
    Matthews, G. F.
    CCFE Fus Assoc, Culham Sci Ctr, Abingdon OX14 3DB, Oxon, England..
    Meigs, A. G.
    CCFE Fus Assoc, Culham Sci Ctr, Abingdon OX14 3DB, Oxon, England..
    Wiesen, S.
    Forschungszentrum Julich, Inst Energie & Klimaforsch Plasmaphys, TEC, D-52425 Julich, Germany..
    Widdowson, A.
    CCFE Fus Assoc, Culham Sci Ctr, Abingdon OX14 3DB, Oxon, England..
    Erosion, screening, and migration of tungsten in the JET divertor2019Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 59, nr 9, artikel-id 096035Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The erosion of tungsten (W), induced by the bombardment of plasma and impurity particles, determines the lifetime of plasma-facing components as well as impacting on plasma performance by the influx of W into the confined region. The screening of W by the divertor and the transport of W in the plasma determines largely the W content in the plasma core, but the W source strength itself has a vital impact on this process. The JET tokamak experiment provides access to a large set of W erosion-determining parameters and permits a detailed description of the W source in the divertor closest to the ITER one: (i) effective sputtering yields and fluxes as function of impact energy of intrinsic (Be, C) and extrinsic (Ne, N) impurities as well as hydrogenic isotopes (H, D) are determined and predictions for the tritium (T) isotope are made. This includes the quantification of intra- and inter-edge localised mode (ELM) contributions to the total W source in H-mode plasmas which vary owing to the complex flux compositions and energy distributions in the corresponding phases. The sputtering threshold behaviour and the spectroscopic composition analysis provides an insight in the dominating species and plasma phases causing W erosion. (ii) The interplay between the net and gross W erosion source is discussed considering (prompt) re-deposition, thus, the immediate return of W ions back to the surface due to their large Larmor radius, and surface roughness, thus, the difference between smooth bulk-W and rough W-coating components used in the JET divertor. Both effects impact on the balance equation of local W erosion and deposition. (iii) Post-mortem analysis reveals the net erosion/deposition pattern and the W migration paths over long periods of plasma operation identifying the net W transport to remote areas. This W transport is related to the divertor plasma regime, e.g. attached operation with high impact energies of impinging particles or detached operation, as well as to the applied magnetic configuration in the divertor, e.g. close divertor with good geometrical screening of W or open divertor configuration with poor screening. JET equipped with the ITER-like wall (ILW) provided unique access to the net W erosion rate within a series of 151 subsequent H-mode discharges (magnetic field: B-t = 2.0 T, plasma current: I-p = 2.0 MA, auxiliary power P-aux = 12 MW) in one magnetic configuration accumulating 900 s of plasma with particle fluences in the range of 5-6 x 10(26) D(+ )m(-2) in the semi-detached inner and attached outer divertor leg. The comparison of W spectroscopy in the intra-ELM and inter-ELM phases with post-mortem analysis of W marker tiles provides a set of gross and net W erosion values at the outer target plate. ERO code simulations are applied to both divertor leg conditions and reproduce the erosion/deposition pattern as well as confirm the high experimentally observed prompt W re-deposition factors of more than 95% in the intra- and inter-ELM phase of the unseeded deuterium H-mode plasma. Conclusions to the expected divertor conditions in ITER as well as to the JET operation in the DT plasma mixture are drawn on basis of this unique benchmark experiment.

  • 16. Brezinsek, S.
    et al.
    Petersson, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Ratynskaia, Svetlana
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Ström, Petter
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Tolias, Panagiotis
    KTH, Skolan för elektro- och systemteknik (EES), Rymd- och plasmafysik.
    Weckmann, Armin
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Zaplotnik, R.
    et al.,
    Plasma-wall interaction studies within the EUROfusion consortium: Progress on plasma-facing components development and qualification2017Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 57, nr 11, artikel-id 116041Artikel i tidskrift (Refereegranskat)
    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.

  • 17. 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, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Rubel, Marek
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Stamp, M. F.
    Borodin, D.
    Coad, J. P.
    Carrasco, Alvaro Garcia
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Kirschner, A.
    Krat, S.
    Krieger, K.
    Lipschultz, B.
    Linsmeier, Ch.
    Matthews, G. F.
    Schmid, K.
    Beryllium migration in JET ITER-like wall plasmas2015Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 55, nr 6, artikel-id 063021Artikel i tidskrift (Refereegranskat)
    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.

  • 18.
    Brunsell, Per R.
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Kuldkepp, Mattias
    KTH, Skolan för teknikvetenskap (SCI), Fysik.
    Menmuir, Sheena
    KTH, Skolan för teknikvetenskap (SCI), Fysik.
    Cecconello, Marco
    KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Hedqvist, Anders
    KTH, Skolan för teknikvetenskap (SCI), Fysik.
    Yadikin, Dimitry
    KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Drake, James Robert
    KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Rachlew, Elisabeth
    KTH, Skolan för teknikvetenskap (SCI), Fysik.
    Reversed field pinch operation with intelligent shell feedback control in EXTRAP T2R2006Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 46, nr 11, s. 904-913Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Discharges in the thin shell reversed field pinch (RFP) device EXTRAP T2R without active feedback control are characterized by growth of non-resonant m = 1 unstable resistive wall modes (RWMs) in agreement with linear MHD theory. Resonant m = 1 tearing modes (TMs) exhibit initially fast rotation and the associated perturbed radial fields at the shell are small, but eventually TMs wall-lock and give rise to a growing radial field. The increase in the radial field at the wall due to growing RWMs and wall-locked TMs is correlated with an increase in the toroidal loop voltage, which leads to discharge termination after 3-4 wall times. An active magnetic feedback control system has been installed in EXTRAP T2R. A two-dimensional array of 128 active saddle coils (pair-connected into 64 independent m = 1 coils) is used with intelligent shell feedback control to suppress the m = 1 radial field at the shell. With feedback control, active stabilization of the full toroidal spectrum of 16 unstable m = 1 non-resonant RWMs is achieved, and TM wall locking is avoided. A three-fold extension of the pulse length, up to the power supply limit, is observed. Intelligent shell feedback control is able to maintain the plasma equilibrium for 10 wall times, with plasma confinement parameters sustained at values comparable to those obtained in thick shell devices of similar size.

  • 19. Castaldo, C.
    et al.
    Ratynskaia, Svetlana V.
    KTH, Skolan för elektro- och systemteknik (EES), Rymd- och plasmafysik.
    Pericoli, V.
    de Angelis, U.
    Rypdal, K.
    Pieroni, L.
    Giovannozzi, E.
    Maddaluno, G.
    Marmolino, C.
    Rufoloni, A.
    Tuccillo, A.
    Kretschmer, M.
    Morfill, G. E.
    Diagnostics of fast dust particles in tokamak edge plasmas2007Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 47, nr 7, s. L5-L9Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The use of electrostatic probes as a diagnostic tool of the dust particles in the tokamak edge plasmas is investigated. Probe measurements of electrostatic fluctuations in the scrape-off layer of the Frascati Tokamak Upgrade revealed that some features of the signals can be explained only by a local non-propagating phenomenon. These signal features are shown to be both in qualitative and quantitative agreement with ionization, and consequent extra charge collected by the probes, due to the impact of micrometre-sized dust at a velocity of the order of 10 km s(- 1). Electron microscope analysis of the probe surface yielded direct support for such an interpretation.

  • 20.
    Causa, F.
    et al.
    CNR, Ist Fis Plasma Piero Caldirola, Via R Cozzi 53, I-20125 Milan, Italy.;CNR, Ist Fis Plasma, Via R Cozzi 53, I-20125 Milan, Italy..
    Ratynskaia, Svetlana
    KTH, Skolan för elektroteknik och datavetenskap (EECS), Rymd- och plasmafysik.
    Tolias, Panagiotis
    KTH, Skolan för elektroteknik och datavetenskap (EECS), Rymd- och plasmafysik.
    Zito, P.
    ENEA, Fus & Nucl Safety Dept, CR Frascati, Via E Fermi 45, I-00044 Rome, Italy..
    Analysis of runaway electron expulsion during tokamak instabilities detected by a single-channel Cherenkov probe in FTU2019Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 59, nr 4, artikel-id 046013Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The expulsion of runaway electrons (REs) during different types of tokamak instabilities is analysed by means of a Cherenkov probe inserted into the scrape-off layer of the FTU tokamak. One such type of instability, the well-known tearing mode, is involved in disruptive plasma termination events, during which the risk of RE avalanche multiplication is highest. The second type, known as anomalous Doppler instability, influences RE dynamics by enhancing pitch angle scattering. Three scenarios are analysed here, characterised by different RE generation rates and mechanisms. The main conclusions are drawn from correlations between the Cherenkov probe and other diagnostics. In particular, the Cherenkov probe permits the detection of fast electron expulsion with a high level of detail, presenting peaks with 100% signal contrast during tearing mode growth and rotation, and sub-peak structures reflecting the interplay between the magnetic island formed by the tearing mode, RE diffusion during island rotation and the geometry of obstacles in the vessel. Correlations between the Cherenkov signal, hard x-ray emission and electron cyclotron emission reveal the impulsive development of the anomalous Doppler instability with instability rise time in the microsecond scale resolved by the high time-resolution of the Cherenkov probe.

  • 21. Challis, C. D.
    et al.
    Garcia, J.
    Beurskens, M.
    Buratti, P.
    Delabie, E.
    Drewelow, P.
    Frassinetti, Lorenzo
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    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 wall2015Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 55, nr 5, artikel-id 053031Artikel i tidskrift (Refereegranskat)
    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.

  • 22. 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, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Jucker, M.
    La Haye, R. J.
    Lennholm, M.
    Power requirements for electron cyclotron current drive and ion cyclotron resonance heating for sawtooth control in ITER2013Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 53, nr 6, s. 066001-Artikel i tidskrift (Refereegranskat)
    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.

  • 23. Citrin, J.
    et al.
    Jenko, F.
    Mantica, P.
    Told, D.
    Bourdelle, C.
    Dumont, R.
    Garcia, J.
    Haverkort, J. W.
    Hogeweij, G. M. D.
    Johnson, Thomas
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Pueschel, M. J.
    Ion temperature profile stiffness: non-linear gyrokinetic simulations and comparison with experiment2014Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 54, nr 2, s. 023008-Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Recent experimental observations at JET show evidence of reduced ion temperature profile stiffness. An extensive set of nonlinear gyrokinetic simulations are performed based on the experimental discharges, investigating the physical mechanism behind the observations. The impact on the ion heat flux of various parameters that differ within the data-set are explored. These parameters include the safety factor, magnetic shear, toroidal flow shear, effect of rotation on the magnetohydrodynamic equilibrium, R/L-n, beta(e), Z(eff), T-e/T-i, and the fast-particle content. While previously hypothesized to be an important factor in the stiffness reduction, the combined effect of toroidal flow shear and low magnetic shear is not predicted by the simulations to lead to a significant reduction in ion heat flux, due both to an insufficient magnitude of flow shear and significant parallel velocity gradient destabilization. It is however found that nonlinear electromagnetic effects due to both thermal and fast-particle pressure gradients, even at low beta(e), can significantly reduce the ion heat flux, and is a key factor in explaining the experimental observations. A total of four discharges are examined, at both inner and outer radii. For all cases studied, the simulated and experimental ion heat flux values agree within reasonable variations of input parameters around the experimental uncertainties.

  • 24. Coad, J. P.
    et al.
    Likonen, J.
    Rubel, Marek J.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Vainonen-Ahlgren, E.
    Hole, D. E.
    Sajavaara, T.
    Renvall, T.
    Matthews, G. F.
    Overview of material re-deposition and fuel retention studies at JET with the Gas Box divertor2006Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 46, nr 2, s. 350-366Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    in the period 1998-2001 the JET tokamak was operated with the MkII Gas Box divertor. On two occasions during that period a number of limiter and divertor tiles were retrieved from the torus and then examined ex situ with surface sensitive techniques. Erosion and deposition patterns were determined in order to assess the material erosion, material migration and fuel inventory on plasma facing components. Tracer techniques, e.g. injection of C-13 labelled methane and tiles coated with a low-Z and high-Z marker layer, were used to enhance the volume of information on the material transport. The results show significant asymmetry in the distribution of fuel and plasma impurity species between the inner (net deposition area) and the outer (net erosion) divertor channels. No significant formation of highly hydrogenated carbon films has been found in the Gas Box structure. The important processes for material migration, and the influence of operation scenarios on the morphology of the deposits are discussed. Comparison is also made with results obtained following previous divertor campaigns.

  • 25. Coda, S.
    et al.
    Ahn, J.
    Albanese, R.
    Alberti, S.
    Alessi, E.
    Allan, S.
    Anand, H.
    Anastassiou, G.
    Andrèbe, Y.
    Angioni, C.
    Ariola, M.
    Bernert, M.
    Beurskens, M.
    Bin, W.
    Blanchard, P.
    Blanken, T. C.
    Boedo, J. A.
    Bolzonella, T.
    Bouquey, F.
    Braunmüller, F. H.
    Bufferand, H.
    Buratti, P.
    Calabró, G.
    Camenen, Y.
    Carnevale, D.
    Carpanese, F.
    Causa, F.
    Cesario, R.
    Chapman, I. T.
    Chellai, O.
    Choi, D.
    Cianfarani, C.
    Ciraolo, G.
    Citrin, J.
    Costea, S.
    Crisanti, F.
    Cruz, N.
    Czarnecka, A.
    Decker, J.
    De Masi, G.
    De Tommasi, G.
    Douai, D.
    Dunne, M.
    Duval, B. P.
    Eich, T.
    Elmore, S.
    Esposito, B.
    Faitsch, M.
    Fasoli, A.
    Fedorczak, N.
    Felici, F.
    Février, O.
    Ficker, O.
    Fietz, S.
    Fontana, M.
    Frassinetti, Lorenzo
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Furno, I.
    Galeani, S.
    Gallo, A.
    Galperti, C.
    Garavaglia, S.
    Garrido, I.
    Geiger, B.
    Giovannozzi, E.
    Gobbin, M.
    Goodman, T. P.
    Gorini, G.
    Gospodarczyk, M.
    Granucci, G.
    Graves, J. P.
    Guirlet, R.
    Hakola, A.
    Ham, C.
    Harrison, J.
    Hawke, J.
    Hennequin, P.
    Hnat, B.
    Hogeweij, D.
    Hogge, J. -P
    Honoré, C.
    Hopf, C.
    Horáček, J.
    Huang, Z.
    Igochine, V.
    Innocente, P.
    Ionita Schrittwieser, C.
    Isliker, H.
    Jacquier, R.
    Jardin, A.
    Kamleitner, J.
    Karpushov, A.
    Keeling, D. L.
    Kirneva, N.
    Kong, M.
    Koubiti, M.
    Kovacic, J.
    Krämer-Flecken, A.
    Krawczyk, N.
    Kudlacek, O.
    Labit, B.
    Lazzaro, E.
    Le, H. B.
    Lipschultz, B.
    Llobet, X.
    Lomanowski, B.
    Loschiavo, V. P.
    Lunt, T.
    Maget, P.
    Maljaars, E.
    Malygin, A.
    Maraschek, M.
    Marini, C.
    Martin, P.
    Martin, Y.
    Mastrostefano, S.
    Maurizio, R.
    Mavridis, M.
    Mazon, D.
    McAdams, R.
    McDermott, R.
    Merle, A.
    Meyer, H.
    Militello, F.
    Miron, I. G.
    Molina Cabrera, P. A.
    Moret, J. -M
    Moro, A.
    Moulton, D.
    Naulin, V.
    Nespoli, F.
    Nielsen, A. H.
    Nocente, M.
    Nouailletas, R.
    Nowak, S.
    Odstrčil, T.
    Papp, G.
    Papřok, R.
    Pau, A.
    Pautasso, G.
    Pericoli Ridolfini, V.
    Piovesan, P.
    Piron, C.
    Pisokas, T.
    Porte, L.
    Preynas, M.
    Ramogida, G.
    Rapson, C.
    Juul Rasmussen, J.
    Reich, M.
    Reimerdes, H.
    Reux, C.
    Ricci, P.
    Rittich, D.
    Riva, F.
    Robinson, T.
    Saarelma, S.
    Saint-Laurent, F.
    Sauter, O.
    Scannell, R.
    Schlatter, C.
    Schneider, B.
    Schneider, P.
    Schrittwieser, R.
    Sciortino, F.
    Sertoli, M.
    Sheikh, U.
    Sieglin, B.
    Silva, M.
    Sinha, J.
    Sozzi, C.
    Spolaore, M.
    Stange, T.
    Stoltzfus-Dueck, T.
    Tamain, P.
    Teplukhina, A.
    Testa, D.
    Theiler, C.
    Thornton, A.
    Tophøj, L.
    Tran, M. Q.
    Tsironis, C.
    Tsui, C.
    Uccello, A.
    Vartanian, S.
    Verdoolaege, G.
    Verhaegh, K.
    Vermare, L.
    Vianello, N.
    Vijvers, W. A. J.
    Vlahos, L.
    Vu, N. M. T.
    Walkden, N.
    Wauters, T.
    Weisen, H.
    Wischmeier, M.
    Zestanakis, P.
    Zuin, M.
    Overview of the TCV tokamak program: Scientific progress and facility upgrades2017Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 57, nr 10, artikel-id 102011Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The TCV tokamak is augmenting its unique historical capabilities (strong shaping, strong electron heating) with ion heating, additional electron heating compatible with high densities, and variable divertor geometry, in a multifaceted upgrade program designed to broaden its operational range without sacrificing its fundamental flexibility. The TCV program is rooted in a three-pronged approach aimed at ITER support, explorations towards DEMO, and fundamental research. A 1 MW, tangential neutral beam injector (NBI) was recently installed and promptly extended the TCV parameter range, with record ion temperatures and toroidal rotation velocities and measurable neutral-beam current drive. ITER-relevant scenario development has received particular attention, with strategies aimed at maximizing performance through optimized discharge trajectories to avoid MHD instabilities, such as peeling-ballooning and neoclassical tearing modes. Experiments on exhaust physics have focused particularly on detachment, a necessary step to a DEMO reactor, in a comprehensive set of conventional and advanced divertor concepts. The specific theoretical prediction of an enhanced radiation region between the two X-points in the low-field-side snowflake-minus configuration was experimentally confirmed. Fundamental investigations of the power decay length in the scrape-off layer (SOL) are progressing rapidly, again in widely varying configurations and in both D and He plasmas; in particular, the double decay length in L-mode limited plasmas was found to be replaced by a single length at high SOL resistivity. Experiments on disruption mitigation by massive gas injection and electron-cyclotron resonance heating (ECRH) have begun in earnest, in parallel with studies of runaway electron generation and control, in both stable and disruptive conditions; a quiescent runaway beam carrying the entire electrical current appears to develop in some cases. Developments in plasma control have benefited from progress in individual controller design and have evolved steadily towards controller integration, mostly within an environment supervised by a tokamak profile control simulator. TCV has demonstrated effective wall conditioning with ECRH in He in support of the preparations for JT-60SA operation.

  • 26.
    Coda, S.
    et al.
    Ecole Polytech Fed Lausanne, Swiss Plasma Ctr, CH-1015 Lausanne, Switzerland..
    Frassinetti, Lorenzo
    KTH, Skolan för elektroteknik och datavetenskap (EECS), Fusionsplasmafysik.
    Zuin, M.
    Consorzio RFX, Corso Stati Uniti 4, I-35127 Padua, Italy..
    et al.,
    Physics research on the TCV tokamak facility: from conventional to alternative scenarios and beyond2019Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 59, nr 11, artikel-id 112023Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The research program of the TCV tokamak ranges from conventional to advanced-tokamak scenarios and alternative divertor configurations, to exploratory plasmas driven by theoretical insight, exploiting the device's unique shaping capabilities. Disruption avoidance by real-time locked mode prevention or unlocking with electron-cyclotron resonance heating (ECRH) was thoroughly documented, using magnetic and radiation triggers. Runaway generation with high-Z noble-gas injection and runaway dissipation by subsequent Ne or Ar injection were studied for model validation. The new 1 MW neutral beam injector has expanded the parameter range, now encompassing ELMy H-modes in an ITER-like shape and nearly noninductive II-mode discharges sustained by electron cyclotron and neutral beam current drive. In the H-mode, the pedestal pressure increases modestly with nitrogen seeding while fueling moves the density pedestal outwards, but the plasma stored energy is largely uncorrelated to either seeding or fueling. High fueling at high triangularity is key to accessing the attractive small edge-localized mode (type-II) regime. Turbulence is reduced in the core at negative triangularity, consistent with increased confinement and in accord with global gyrokinetic simulations. The geodesic acoustic mode, possibly coupled with avalanche events, has been linked with particle flow to the wall in diverted plasmas. Detachment, scrape-off layer transport, and turbulence were studied in L- and H-modes in both standard and alternative configurations (snowflake, super-X, and beyond). The detachment process is caused by power `starvation' reducing the ionization source, with volume recombination playing only a minor role. Partial detachment in the H-mode is obtained with impurity seeding and has shown little dependence on flux expansion in standard single-null geometry. In the attached 1,-mode phase, increasing the outer connection length reduces the in-out heat-flow asymmetry. A doublet plasma, featuring an internal X-point, was achieved successfully, and a transport barrier was observed in the mantle just outside the internal separatrix. In the near future variableconfiguration baffles and possibly divertor ptunping will be introduced to investigate the effect of divertor closure on exhaust and performance, and 3.5 MW ECR and 1 MW neutral beam injection heating will be added.

  • 27. Coenen, J. W.
    et al.
    Arnoux, G.
    Bazylev, B.
    Matthews, G. F.
    Autricque, A.
    Balboa, I.
    Clever, M.
    Dejarnac, R.
    Coffey, I.
    Corre, Y.
    Devaux, S.
    Frassinetti, Lorenzo
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Gauthier, E.
    Horacek, J.
    Jachmich, S.
    Komm, M.
    Knaup, M.
    Krieger, K.
    Marsen, S.
    Meigs, A.
    Mertens, Ph.
    Pitts, R. A.
    Puetterich, T.
    Rack, M.
    Stamp, M.
    Sergienko, G.
    Tamain, P.
    Thompson, V.
    ELM-induced transient tungsten melting in the JET divertor2015Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 55, nr 2, artikel-id 023010Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The original goals of the JET ITER-like wall included the study of the impact of an all W divertor on plasma operation (Coenen et al 2013 Nucl. Fusion 53 073043) and fuel retention (Brezinsek et al 2013 Nucl. Fusion 53 083023). ITER has recently decided to install a full-tungsten (W) divertor from the start of operations. One of the key inputs required in support of this decision was the study of the possibility of W melting and melt splashing during transients. Damage of this type can lead to modifications of surface topology which could lead to higher disruption frequency or compromise subsequent plasma operation. Although every effort will be made to avoid leading edges, ITER plasma stored energies are sufficient that transients can drive shallow melting on the top surfaces of components. JET is able to produce ELMs large enough to allow access to transient melting in a regime of relevance to ITER. Transient W melt experiments were performed in JET using a dedicated divertor module and a sequence of I-P = 3.0 MA/B-T = 2.9 T H-mode pulses with an input power of P-IN = 23 MW, a stored energy of similar to 6 MJ and regular type I ELMs at Delta W-ELM = 0.3 MJ and f(ELM) similar to 30 Hz. By moving the outer strike point onto a dedicated leading edge in the W divertor the base temperature was raised within similar to 1 s to a level allowing transient, ELM-driven melting during the subsequent 0.5 s. Such ELMs (delta W similar to 300 kJ per ELM) are comparable to mitigated ELMs expected in ITER (Pitts et al 2011 J. Nucl. Mater. 415 (Suppl.) S957-64). Although significant material losses in terms of ejections into the plasma were not observed, there is indirect evidence that some small droplets (similar to 80 mu m) were released. Almost 1 mm (similar to 6 mm(3)) of W was moved by similar to 150 ELMs within 7 subsequent discharges. The impact on the main plasma parameters was minor and no disruptions occurred. The W-melt gradually moved along the leading edge towards the high-field side, driven by j x B forces. The evaporation rate determined from spectroscopy is 100 times less than expected from steady state melting and is thus consistent only with transient melting during the individual ELMs. Analysis of IR data and spectroscopy together with modelling using the MEMOS code Bazylev et al 2009 J. Nucl. Mater. 390-391 810-13 point to transient melting as the main process. 3D MEMOS simulations on the consequences of multiple ELMs on damage of tungsten castellated armour have been performed. These experiments provide the first experimental evidence for the absence of significant melt splashing at transient events resembling mitigated ELMs on ITER and establish a key experimental benchmark for the MEMOS code.

  • 28. Coenen, J. W.
    et al.
    Sertoli, M.
    Brezinsek, S.
    Coffey, I.
    Dux, R.
    Giroud, C.
    Groth, M.
    Huber, A.
    Ivanova, Darya
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Krieger, K.
    Lawson, K.
    Marsen, S.
    Meigs, A.
    Neu, R.
    Puetterich, T.
    van Rooij, G. J.
    Stamp, M. F.
    Long-term evolution of the impurity composition and impurity events with the ITER-like wall at JET2013Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 53, nr 7, s. 073043-Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    This paper covers aspects of long-term evolution of intrinsic impurities in the JET tokamak with respect to the newly installed ITER-like wall (ILW). At first the changes related to the change over from the JET-C to the JET-ILW with beryllium (Be) as the main wall material and tungsten (W) in the divertor are discussed. The evolution of impurity fluxes in the newly installed W divertor with respect to studying material migration is described. In addition, a statistical analysis of transient impurity events causing significant plasma contamination and radiation losses is shown. The main findings comprise a drop in carbon content (x20) (see also Brezinsek et al (2013 J. Nucl. Mater. 438 S303)), low oxygen content (x10) due to the Be first wall (Douai et al 2013 J. Nucl. Mater. 438 S1172-6) as well as the evolution of the material mix in the divertor. Initially, a short period of repetitive ohmic plasmas was carried out to study material migration (Krieger et al 2013 J. Nucl. Mater. 438 S262). After the initial 1600 plasma seconds the material surface composition is, however, still evolving. With operational time, the levels of recycled C are increasing slightly by 20% while the Be levels in the deposition-dominated inner divertor are dropping, hinting at changes in the surface layer material mix made of Be, C and W. A steady number of transient impurity events, consisting of W and constituents of inconel, is observed despite the increase in variation in machine operation and changes in magnetic configuration as well as the auxiliary power increase.

  • 29.
    Dahlin, Jon-Erik
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Scheffel, Jan
    KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Numerical studies of confinement scalings for the dynamo-free reversed-field pinch2007Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 47, nr 1, s. 9-16Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    In the reversed-field pinch (RFP), tearing modes associated with the dynamo are responsible for reduced energy- and particle confinement. In this study, it is observed that by implementing current profile control (CPC) in the RFP, a dynamo-free state can be achieved. The effect of CPC in the RFP is examined by the use of numerical simulations, and scaling laws are presented for confinement parameters. The model is nonlinear MHD in 3D including finite resistivity and pressure. A linear regression analysis is performed on simulation data from a series of computer runs for a set of initial parameter values. Scaling laws are determined for radial magnetic field, energy confinement time, poloidal beta and temperature. Confinement is improved substantially as compared with the conventional RFP - the temperature reaches reactor relevant levels by ohmic heating alone. It is observed that the configuration spontaneously develops into a quasi single helicity state. The CPC scheme is designed to eliminate the fluctuating electric dynamo field Ef ≤ -〈v × B〉, using feedback of an externally imposed electric field. The focus of this study is on obtaining principal theoretical optimization of confinement in the RFP by implementing CPC and to formulate scaling laws for confinement parameters, thus investigating the reactor viability of the concept.

  • 30.
    Dahlin, Jon-Erik
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Scheffel, Jan
    KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Ultra-high beta in numerical simulations of a tearing-mode reduced reversed-field pinch2007Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 47, nr 9, s. 1184-1188Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    In the advanced reversed-field pinch (RFP), current profile control (CPC) enables energy confinement time and poloidal beta to increase substantially as compared with the conventional RFP due to reduced magnetic field stochasticity. Numerical simulations using the three-dimensional non-linear resistive MHD-code DEBSP are performed showing that the poloidal beta is not limited to the m ≤ 0 stability criterion βθ < 1/2. Instead, as tearing modes are diminished, it may approach unity. The beta criterion is theoretically analysed and a new, more general, criterion is derived. Analytic estimates of the resistive tearing and g-mode growth rates are derived for m ≤ 0, and it is shown that both tearing and g-mode growth rates decrease significantly as CPC is employed. Furthermore, quasi-steady state operation with increased confinement due to active control of the current profile is numerically demonstrated for the advanced RFP for a scenario with βθ < 1/2.

  • 31.
    De Angeli, M.
    et al.
    CNR, Ist Sci & Tecnol Plasmi, Milan, Italy..
    Lazzaro, E.
    CNR, Ist Sci & Tecnol Plasmi, Milan, Italy..
    Tolias, Panagiotis
    KTH, Skolan för elektroteknik och datavetenskap (EECS), Rymd- och plasmafysik.
    Ratynskaia, Svetlana V.
    KTH, Skolan för elektroteknik och datavetenskap (EECS), Rymd- och plasmafysik.
    Vignitchouk, Ladislas
    KTH, Skolan för elektroteknik och datavetenskap (EECS), Rymd- och plasmafysik.
    Castaldo, C.
    ENEA, CR Frascati, I-00044 Rome, Italy..
    Apicella, M. L.
    ENEA, CR Frascati, I-00044 Rome, Italy..
    Gervasini, G.
    CNR, Ist Sci & Tecnol Plasmi, Milan, Italy..
    Giacomi, G.
    ENEA, CR Frascati, I-00044 Rome, Italy..
    Giovannozzi, E.
    ENEA, CR Frascati, I-00044 Rome, Italy..
    Granucci, G.
    CNR, Ist Sci & Tecnol Plasmi, Milan, Italy..
    Iafrati, M.
    ENEA, CR Frascati, I-00044 Rome, Italy..
    Iraji, D.
    Amirkabir Univ Technol, Energy Engn & Phys Dept, Tehran, Iran..
    Maddaluno, G.
    ENEA, CR Frascati, I-00044 Rome, Italy..
    Riva, G.
    CNR, Ist Chim Mat Condensata & Tecnol Energia, Via R Cozzi 53, I-20125 Milan, Italy..
    Uccello, A.
    CNR, Ist Sci & Tecnol Plasmi, Milan, Italy..
    Pre-plasma remobilization of ferromagnetic dust in FTU and possible interference with tokamak operations2019Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 59, nr 10, artikel-id 106033Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Experimental evidence of the pre-plasma remobilization of ferromagnetic dust in FTU is presented. Thomson scattering data and IR camera observations document the occurrence of intrinsic dust remobilization prior to discharge start-up and allow for a rough calculation of the average mobilized dust density. Exposures of calibrated extrinsic non-magnetic and ferromagnetic dust to sole magnetic field discharges reveal that the magnetic moment force is the main mobilizing force, as confirmed by theoretical estimates. Pre-plasma remobilization probabilities are computed for varying dust sizes. The impact of prematurely remobilized dust on the breakdown and burn-through start-up phases is investigated together with the discharge termination induced once the plasma plateau is established.

  • 32. de Vries, P. C.
    et al.
    Salmi, A.
    Parail, V.
    Giroud, C.
    Andrew, Y.
    Biewer, T. M.
    Crombe, K.
    Jenkins, I.
    Johnson, Thomas J.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Kiptily, V.
    Loarte, A.
    Lonnroth, J.
    Meigs, A.
    Oyama, N.
    Sartori, R.
    Saibene, G.
    Urano, H.
    Zastrow, K. D.
    Effect of toroidal field ripple on plasma rotation in JET2008Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 48, nr 3Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Dedicated experiments on TF ripple effects on the performance of tokamak plasmas have been carried out at JET. The TF ripple was found to have a profound effect on the plasma rotation. The central Mach number, M, defined as the ratio of the rotation velocity and the thermal velocity, was found to drop as a function of TF ripple amplitude (3) from an average value of M = 0.40-0.55 for operations at the standard JET ripple of 6 = 0.08% to M = 0.25-0.40 for 6 = 0.5% and M = 0.1-0.3 for delta = 1%. TF ripple effects should be considered when estimating the plasma rotation in ITER. With standard co-current injection of neutral beam injection (NBI), plasmas were found to rotate in the co-current direction. However, for higher TF ripple amplitudes (delta similar to 1%) an area of counter rotation developed at the edge of the plasma, while the core kept its co-rotation. The edge counter rotation was found to depend, besides on the TF ripple amplitude, on the edge temperature. The observed reduction of toroidal plasma rotation with increasing TF ripple could partly be explained by TF ripple induced losses of energetic ions, injected by NBI. However, the calculated torque due to these losses was insufficient to explain the observed counter rotation and its scaling with edge parameters. It is suggested that additional TF ripple induced losses of thermal ions contribute to this effect.

  • 33. Dejarnac, R.
    et al.
    Podolnik, A.
    Komm, M.
    Arnoux, G.
    Coenen, J. W.
    Devaux, S.
    Frassinetti, Lorenzo
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Gunn, J. P.
    Matthews, G. F.
    Pitts, R. A.
    Numerical evaluation of heat flux and surface temperature on a misaligned JET divertor W lamella during ELMs2014Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 54, nr 12, s. 123011-Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    A series of experiments has been performed on JET to investigate the dynamics of transient melting due to edge localized modes (ELMs). The experiment employs a deliberately misaligned lamella in one module of the JET bulk tungsten outer divertor, allowing the combination of stationary power flux and ELMs to transiently melt the misaligned edge. During the design of the experiment a number of calculations were performed using 2D particle-in-cell simulations and a heat transfer code to investigate the influence on the deposited power flux of finite Larmor radius effects associated with the energetic ELM ions. This has been performed using parameter scans inside a range of pedestal temperatures and densities to scope different experimentally expected ELM energies. On the one hand, we observe optimistic results, with smoothing of the heat flux due to the Larmor gyration on the protruding side of the lamella which sees the direct parallel flux-the deposited power tends to be lower than the nominal value expected from geometric magnetic field line impact over a distance smaller than 2 Larmor radii, a finding which is always valid during ELMs for such a geometry. On the other hand, the fraction of the flux not reaching the directly wetted side is transferred and spread to the top surface of the lamella. The hottest point of the lamella (corner side/top) does not always benefit from the gain from the Larmor smoothing effect because of an enhanced power deposition from the second contribution.

  • 34.
    Drake, James Robert
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Brunsell, Per
    KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Yadikin, Dimitry
    KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Cecconello, Marco
    KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Malmberg, Jenny
    KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Liu, Y.
    Experimental and theoretical studies of active control of resistive wall mode growth in the EXTRAP T2R reversed-field pinch2005Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 45, nr 7, s. 557-564Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Active feedback control of resistive wall modes (RWMs) has been demonstrated in the EXTRAP T2R reversed-field pinch experiment. The control system includes a sensor consisting of an array of magnetic coils (measuring mode harmonics) and an actuator consisting of a saddle coil array (producing control harmonics). Closed-loop (feedback) experiments using a digital controller based on a real time Fourier transform of sensor data have been studied for cases where the feedback gain was constant and real for all harmonics (corresponding to an intelligent-shell) and cases where the feedback gain could be set for selected harmonics, with both real and complex values (targeted harmonics). The growth of the dominant RWMs can be reduced by feedback for both the intelligent-shell and targeted-harmonic control systems. Because the number of toroidal positions of the saddle coils in the array is half the number of the sensors, it is predicted and observed experimentally that the control harmonic spectrum has sidebands. Individual unstable harmonics can be controlled with real gains. However if there are two unstable mode harmonics coupled by the sideband effect, control is much less effective with real gains. According to the theory, complex gains give better results for (slowly) rotating RWMs, and experiments support this prediction. In addition, open loop experiments have been used to observe the effects of resonant field errors applied to unstable, marginally stable and robustly stable modes. The observed effects of field errors are consistent with the thin-wall model, where mode growth is proportional to the resonant field error amplitude and the wall penetration time for that mode harmonic.

  • 35.
    Dumont, R. J.
    et al.
    CEA, IRFM, F-13108 St Paul Les Durance, France..
    Mailloux, J.
    Culham Sci Ctr, CCFE, Abingdon OX14 3DB, Oxon, England..
    Aslanyan, V
    MIT, PSFC, 175 Albany St, Cambridge, MA 02039 USA..
    Baruzzo, M.
    Consorzio RFX, Corso Stati Uniti 4, I-35127 Padua, Italy..
    Challis, C. D.
    Culham Sci Ctr, CCFE, Abingdon OX14 3DB, Oxon, England..
    Coffey, I
    Queens Univ, Dept Pure & Appl Phys, Belfast BT7 1NN, Antrim, North Ireland..
    Czarnecka, A.
    Inst Plasma Phys & Laser Microfus, Hery St 23, PL-00908 Warsaw, Poland..
    Delabie, E.
    Oak Ridge Natl Lab, Oak Ridge, TN USA..
    Eriksson, J.
    Uppsala Univ, Dept Phys & Astron, SE-75119 Uppsala, Sweden..
    Faustin, J.
    Ecole Polytech Fed Lausanne, SPC, CH-1015 Lausanne, Switzerland..
    Ferreira, J.
    Univ Lisbon, IST, Inst Plasmas & Fusao Nucl, Lisbon, Portugal..
    Fitzgerald, M.
    Culham Sci Ctr, CCFE, Abingdon OX14 3DB, Oxon, England..
    Garcia, J.
    CEA, IRFM, F-13108 St Paul Les Durance, France..
    Giacomelli, L.
    Univ Milano Bicocca, Piazza Sci 3, I-20126 Milan, Italy..
    Giroud, C.
    Culham Sci Ctr, CCFE, Abingdon OX14 3DB, Oxon, England..
    Hawkes, N.
    Culham Sci Ctr, CCFE, Abingdon OX14 3DB, Oxon, England..
    Jacquet, Ph
    Culham Sci Ctr, CCFE, Abingdon OX14 3DB, Oxon, England..
    Joffrin, E.
    CEA, IRFM, F-13108 St Paul Les Durance, France..
    Johnson, Thomas
    KTH, Skolan för elektroteknik och datavetenskap (EECS), Fusionsplasmafysik.
    Keeling, D.
    Culham Sci Ctr, CCFE, Abingdon OX14 3DB, Oxon, England..
    King, D.
    Culham Sci Ctr, CCFE, Abingdon OX14 3DB, Oxon, England..
    Kiptily, V
    Culham Sci Ctr, CCFE, Abingdon OX14 3DB, Oxon, England..
    Lomanowski, B.
    Aalto Univ, POB 14100, FIN-00076 Aalto, Finland..
    Lerche, E.
    Ass EUROFUS Belgian State, LPP ERM KMS, TEC Partner, Brussels, Belgium..
    Mantsinen, M.
    Barcelona Supercomp Ctr, Barcelona, Spain.;ICREA, Barcelona, Spain..
    Meneses, L.
    Univ Lisbon, IST, Inst Plasmas & Fusao Nucl, Lisbon, Portugal..
    Menmuir, S.
    Culham Sci Ctr, CCFE, Abingdon OX14 3DB, Oxon, England..
    McClements, K.
    Culham Sci Ctr, CCFE, Abingdon OX14 3DB, Oxon, England..
    Moradi, S.
    Culham Sci Ctr, CCFE, Abingdon OX14 3DB, Oxon, England..
    Nabais, F.
    Univ Lisbon, IST, Inst Plasmas & Fusao Nucl, Lisbon, Portugal..
    Nocente, M.
    Univ Milano Bicocca, Piazza Sci 3, I-20126 Milan, Italy..
    Patel, A.
    Culham Sci Ctr, CCFE, Abingdon OX14 3DB, Oxon, England..
    Patten, H.
    Ecole Polytech Fed Lausanne, SPC, CH-1015 Lausanne, Switzerland..
    Puglia, P.
    Ecole Polytech Fed Lausanne, SPC, CH-1015 Lausanne, Switzerland..
    Scannell, R.
    Culham Sci Ctr, CCFE, Abingdon OX14 3DB, Oxon, England..
    Sharapov, S.
    Culham Sci Ctr, CCFE, Abingdon OX14 3DB, Oxon, England..
    Solano, E. R.
    CIEMAT, Lab Nacl Fus, Madrid, Spain..
    Tsalas, M.
    FOM Inst DIFFER, NL-3430 BE Nieuwegein, Netherlands.;ITER Org, Route Vinon Sur Verdon, F-13067 St Paul Les Durance, France..
    Vallejos, Pablo
    KTH, Skolan för elektroteknik och datavetenskap (EECS), Fusionsplasmafysik.
    Weisen, H.
    Ecole Polytech Fed Lausanne, SPC, CH-1015 Lausanne, Switzerland..
    Scenario development for the observation of alpha-driven instabilities in JET DT plasmas2018Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 58, nr 8, artikel-id 082005Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    In DT plasmas, toroidal Alfven eigenmodes (TAEs) can be made unstable by the alpha particles resulting from fusion reactions, and may induce a significant redistribution of fast ions. Recent experiments have been conducted in JET deuterium plasmas in order to prepare scenarios aimed at observing alpha-driven TAEs in a future JET DT campaign. Discharges at low density, large core temperatures associated with the presence of internal transport barriers and characterised by good energetic ion confinement have been performed. ICRH has been used in the hydrogen minority heating regime to probe the TAE stability. The consequent presence of MeV ions has resulted in the observation of TAEs in many instances. The impact of several key parameters on TAE stability could therefore be studied experimentally. Modeling taking into account NBI and ICRH fast ions shows good agreement with the measured neutron rates, and has allowed predictions for DT plasmas to be performed.

  • 36. Eriksson, J.
    et al.
    Nocente, M.
    Binda, F.
    Cazzaniga, C.
    Conroy, S.
    Ericsson, G.
    Giacomelli, L.
    Gorini, G.
    Hellesen, C.
    Hellsten, Torbjörn
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Hjalmarsson, A.
    Jacobsen, A. S.
    Johnson, Thomas
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Kiptily, V.
    Koskela, T.
    Mantsinen, M.
    Salewski, M.
    Schneider, M.
    Sharapov, S.
    Skiba, M.
    Tardocchi, M.
    Weiszflog, M.
    Dual sightline measurements of MeV range deuterons with neutron and gamma-ray spectroscopy at JET2015Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 55, nr 12, artikel-id 123026Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Observations made in a JET experiment aimed at accelerating deuterons to the MeV range by third harmonic radio-frequency (RF) heating coupled into a deuterium beam are reported. Measurements are based on a set of advanced neutron and gamma-ray spectrometers that, for the first time, observe the plasma simultaneously along vertical and oblique lines of sight. Parameters of the fast ion energy distribution, such as the high energy cut-off of the deuteron distribution function and the RF coupling constant, are determined from data within a uniform analysis framework for neutron and gamma-ray spectroscopy based on a one-dimensional model and by a consistency check among the individual measurement techniques. A systematic difference is seen between the two lines of sight and is interpreted to originate from the sensitivity of the oblique detectors to the pitch-angle structure of the distribution around the resonance, which is not correctly portrayed within the adopted one dimensional model. A framework to calculate neutron and gamma-ray emission from a spatially resolved, two-dimensional deuteron distribution specified by energy/pitch is thus developed and used for a first comparison with predictions from ab initio models of RF heating at multiple harmonics. The results presented in this paper are of relevance for the development of advanced diagnostic techniques for MeV range ions in high performance fusion plasmas, with applications to the experimental validation of RF heating codes and, more generally, to studies of the energy distribution of ions in the MeV range in high performance deuterium and deuterium-tritium plasmas.

  • 37. Eriksson, L. G.
    et al.
    Johnson, Thomas J.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Mayoral, M. L.
    Coda, S.
    Sauter, O.
    Buttery, R. J.
    McDonald, D.
    Hellsten, Torbjörn A. K.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Mantsinen, M. J.
    Mueck, A.
    Noterdaeme, J. M.
    Santala, M.
    Westerhor, E.
    de Vries, P.
    On ion cyclotron current drive for sawtooth control2006Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 46, nr 10, s. S951-S964Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Experiments using ion cyclotron current drive (ICCD) to control sawteeth are presented. In particular, discharges demonstrating shortening of fast ion induced long sawteeth reported in (Eriksson et al 2004 Phys. Rev. Lett. 92 235004) by ICCD have been analysed in detail. Numerical simulations of the ICCD driven currents are shown to be consistent with the experimental observations. They support the hypothesis that an increase in the magnetic shear, due to the driven current, at the surface where the safety factor is unity was the critical factor for the shortening of the sawteeth. In view of the potential utility of ICCD, the mechanisms for the current drive have been further investigated experimentally. This includes the influence of the averaged energy of the resonating ions carrying the current and the spectrum of the launched waves. The results of these experiments are discussed in the light of theoretical considerations.

  • 38. Falchetto, G. L.
    et al.
    Coster, D.
    Coelho, R.
    Scott, B. D.
    Figini, L.
    Kalupin, D.
    Nardon, E.
    Nowak, S.
    Alves, L. L.
    Artaud, J. F.
    Basiuk, V.
    Bizarro, Jao P. S.
    Boulbe, C.
    Dinklage, A.
    Farina, D.
    Faugeras, B.
    Ferreira, J.
    Figueiredo, A.
    Huynh, Ph
    Imbeaux, F.
    Ivanova-Stanik, I.
    Johnson, Thomas
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Klingshirn, H-J
    Konz, C.
    Kus, A.
    Marushchenko, N. B.
    Pereverzev, G.
    Owsiak, M.
    Poli, E.
    Peysson, Y.
    Reimer, R.
    Signoret, J.
    Sauter, O.
    Stankiewicz, R.
    Strand, P.
    Voitsekhovitch, I.
    Westerhof, E.
    Zok, T.
    Zwingmann, W.
    The European Integrated Tokamak Modelling (ITM) effort: achievements and first physics results2014Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 54, nr 4, s. 043018-Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    A selection of achievements and first physics results are presented of the European Integrated Tokamak Modelling Task Force (EFDA ITM-TF) simulation framework, which aims to provide a standardized platform and an integrated modelling suite of validated numerical codes for the simulation and prediction of a complete plasma discharge of an arbitrary tokamak. The framework developed by the ITM-TF, based on a generic data structure including both simulated and experimental data, allows for the development of sophisticated integrated simulations (workflows) for physics application.The equilibrium reconstruction and linear magnetohydrodynamic (MHD) stability simulation chain was applied, in particular, to the analysis of the edgeMHDstability of ASDEX Upgrade type-I ELMy H-mode discharges and ITER hybrid scenario, demonstrating the stabilizing effect of an increased Shafranov shift on edge modes. Interpretive simulations of a JET hybrid discharge were performed with two electromagnetic turbulence codes within ITM infrastructure showing the signature of trapped-electron assisted ITG turbulence. A successful benchmark among five EC beam/ray-tracing codes was performed in the ITM framework for an ITER inductive scenario for different launching conditions from the equatorial and upper launcher, showing good agreement of the computed absorbed power and driven current. Selected achievements and scientific workflow applications targeting key modelling topics and physics problems are also presented, showing the current status of the ITM-TF modelling suite.

  • 39. Fasoli, A.
    et al.
    Gormenzano, C.
    Berk, H. L.
    Breizman, B.
    Briguglio, S.
    Darrow, D. S.
    Gorelenkov, N.
    Heidbrink, W. W.
    Jaun, Andre
    KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Konovalov, S. V.
    Nazikian, R.
    Noterdaeme, J. M.
    Sharapov, S.
    Shinohara, K.
    Testa, D.
    Tobita, K.
    Todo, Y.
    Vlad, G.
    Zonca, F.
    Chapter 5: Physics of energetic ions2007Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 47, nr 6, s. S264-S284Artikel, forskningsöversikt (Refereegranskat)
    Abstract [en]

    This chapter reviews the progress accomplished since the redaction of the first ITER Physics Basis (1999 Nucl. Fusion 39 2137-664) in the field of energetic ion physics and its possible impact on burning plasma regimes. New schemes to create energetic ions simulating the fusion-produced alphas are introduced, accessing experimental conditions of direct relevance for burning plasmas, in terms of the Alfvenic Mach number and of the normalised pressure gradient of the energetic ions, though orbit characteristics and size cannot always match those of ITER. Based on the experimental and theoretical knowledge of the effects of the toroidal magnetic field ripple on direct fast ion losses, ferritic inserts in ITER are expected to provide a significant reduction of ripple alpha losses in reversed shear configurations. The nonlinear fast ion interaction with kink and tearing modes is qualitatively understood, but quantitative predictions are missing, particularly for the stabilisation of sawteeth by fast particles that can trigger neoclassical tearing modes. A large database on the linear stability properties of the modes interacting with energetic ions, such as the Alfven eigenmode has been constructed. Comparisons between theoretical predictions and experimental measurements of mode structures and drive/damping rates approach a satisfactory degree of consistency, though systematic measurements and theory comparisons of damping and drive of intermediate and high mode numbers, the most relevant for ITER, still need to be performed. The nonlinear behaviour of Alfven eigenmodes close to marginal stability is well characterized theoretically and experimentally, which gives the opportunity to extract some information on the particle phase space distribution from the measured instability spectral features. Much less data exists for strongly unstable scenarios, characterised by nonlinear dynamical processes leading to energetic ion redistribution and losses, and identified in nonlinear numerical simulations of Alfven eigenmodes and energetic particle modes. Comparisons with theoretical and numerical analyses are needed to assess the potential implications of these regimes on burning plasma scenarios, including in the presence of a large number of modes simultaneously driven unstable by the fast ions.

  • 40.
    Frassinetti, Lorenzo
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Alfier, A.
    Pasqualotto, R.
    Bonomo, F.
    Innocente, P.
    Heat diffusivity model and temperature simulations in RFX-mod2008Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 48, nr 4, s. 045007-Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The core transport properties of reversed field pinch (RFP) plasmas in the standard regime are generally associated with a high level of magnetic chaos. Indeed, in the RFX-mod RFP device, the core temperature profile is often very flat, indicating that the heat diffusivity is very high. In contrast, the temperature edge profile has a steep gradient, indicating that the edge is characterized by low heat transport. These simple experimental evidences are the basis of a heat diffusivity model that is used as an input to a numerical code for plasma temperature simulation. The simulated temperature reproduces with good accuracy both the experimental T, time evolution and its radial profiles in different plasma scenarios, showing that the model is useful for estimating the plasma heat diffusivity. This work suggests that the heat transport properties in the RFP plasma core are dominated by magnetic chaos in standard discharges and suggests a simple way to estimate electron heat diffusivity from density, input power and magnetic fluctuation measurements.

  • 41.
    Frassinetti, Lorenzo
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Beurskens, M. N. A.
    Saarelma, S.
    Boom, J. E.
    Delabie, E.
    Flanagan, J.
    Kempenaars, M.
    Giroud, C.
    Lomas, P.
    Meneses, L.
    Maggi, C. S.
    Menmuir, S.
    Nunes, I.
    Rimini, F.
    Stefanikova, E.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Urano, H.
    Verdoolaege, G.
    Global and pedestal confinement and pedestal structure in dimensionless collisionality scans of low-triangularity H-mode plasmas in JET-ILW2017Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 57, nr 1, artikel-id 061012Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    A dimensionless collisionality scan in low-triangularity plasmas in the Joint European Torus with the ITER-like wall (JET-ILW) has been performed. The increase of the normalized energy confinement (defined as the ratio between thermal energy confinement and Bohm confinement time) with decreasing collisionality is observed. Moreover, at low collisionality, a confinement factor H-98, comparable to JET-C, is achieved. At high collisionality, the low normalized confinement is related to a degraded pedestal stability and a reduction in the density-profile peaking. The increase of normalized energy confinement is due to both an increase in the pedestal and in the core regions. The improvement in the pedestal is related to the increase of the stability. The improvement in the core is driven by (i) the core temperature increase via the temperature-profile stiffness and by (ii) the density-peaking increase driven by the low collisionality. Pedestal stability analysis performed with the ELITE (edge-localized instabilities in tokamak equilibria) code has a reasonable qualitative agreement with the experimental results. An improvement of the pedestal stability with decreasing collisionality is observed. The improvement is ascribed to the reduction of the pedestal width, the increase of the bootstrap current and the reduction of the relative shift between the positions of the pedestal density and pedestal temperature. The EPED1 model predictions for the pedestal pressure height are qualitatively well correlated with the experimental results. Quantitatively, EPED1 overestimates the experimental pressure by 15-35%. In terms of the pedestal width, a correct agreement (within 10-15%) between the EPED1 and the experimental width is found at low collisionality. The experimental pedestal width increases with collisionality. Nonetheless, an extrapolation to low-collisionality values suggests that the width predictions from the KBM constraint are reasonable for ITER.

  • 42.
    Frassinetti, Lorenzo
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Brunsell, Per R.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Cecconello, Marco
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Drake, James R.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Heat transport modelling in EXTRAP T2R2009Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 49, nr 2Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    A model to estimate the heat transport in the EXTRAP T2R reversed field pinch (RFP) is described. The model, based on experimental and theoretical results, divides the RFP electron heat diffusivity chi(e) into three regions, one in the plasma core, where chi(e) is assumed to be determined by the tearing modes, one located around the reversal radius, where chi(e) is assumed not dependent on the magnetic fluctuations and one in the extreme edge, where high chi(e) is assumed. The absolute values of the core and of the reversal chi(e) are determined by simulating the electron temperature and the soft x-ray and by comparing the simulated signals with the experimental ones. The model is used to estimate the heat diffusivity and the energy confinement time during the flat top of standard plasmas, of deep F plasmas and of plasmas obtained with the intelligent shell.

  • 43.
    Frassinetti, Lorenzo
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Brunsell, Per R.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Drake, James R.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Experiments and modelling of active quasi-single helicity regime generation in a reversed field pinch2009Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 49, nr 7Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The interaction of a static resonant magnetic perturbation (RMP) with a tearing mode (TM) is becoming a relevant topic in fusion plasma physics. RMPs can be generated by active coils and then used to affect the properties of TMs and of the corresponding magnetic islands. This paper shows how the feedback system of the EXTRAP T2R reversed field pinch (RFP) can produce a RMP that affects a rotating TM and stimulate the transition to the so-called quasi-single helicity (QSH) regime, a RFP plasma state characterized by a magnetic island surrounded by low magnetic chaos. The application of the RMP can increase the QSH probability up to 10% and enlarge the size of the corresponding island. Part of the experimental results are supported by a theoretical study that models the effect of the active coils on the magnetic island.

  • 44.
    Frassinetti, Lorenzo
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Dodt, D.
    Beurskens, M. N. A.
    Sirinelli, A.
    Boom, J. E.
    Eich, T.
    Flanagan, J.
    Giroud, C.
    Jachmich, M. S.
    Kempenaars, M.
    Lomas, P.
    Maddison, G.
    Maggi, C.
    Neu, R.
    Nunes, I.
    von Thun, C. Perez
    Sieglin, B.
    Stamp, M.
    Effect of nitrogen seeding on the energy losses and on the time scales of the electron temperature and density collapse of type-I ELMs in JET with the ITER-like wall2015Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 55, nr 2, artikel-id 023007Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The baseline type-I ELMy H-mode scenario has been re-established in JET with the new tungsten MKII-HD divertor and beryllium on the main wall (hereafter called the ITER-like wall, JET-ILW). The first JET-ILW results show that the confinement is degraded by 20-30% in the baseline scenarios compared to the previous carbon wall JET (JET-C) plasmas. The degradation is mainly driven by the reduction in the pedestal temperature. Stored energies and pedestal temperature comparable to the JET-C have been obtained to date in JET-ILW baseline plasmas only in the high triangularity shape using N-2 seeding. This work compares the energy losses during ELMs and the corresponding time scales of the temperature and density collapse in JET-ILWbaseline plasmas with and without N-2 seeding with similar JET-C baseline plasmas. ELMs in the JET-ILW differ from those with the carbon wall both in terms of time scales and energy losses. The ELM time scale, defined as the time to reach the minimum pedestal temperature soon after the ELM collapse, is similar to 2ms in the JET-ILW and lower than 1 ms in the JET-C. The energy losses are in the range Delta W-ELM/W-ped approximate to 7-12% in the JET-ILWand Delta W-ELM/W-ped approximate to 10-20% in JET-C, and fit relatively well with earlier multi-machine empirical scalings of Delta W-ELM/W-ped with collisionality. The time scale of the ELM collapse seems to be related to the pedestal collisionality. Most of the non-seeded JET-ILW ELMs are followed by a further energy drop characterized by a slower time scale similar to 8-10 ms (hereafter called slow transport events), that can lead to losses in the range Delta W-slow/W-ped approximate to 15-22%, slightly larger than the losses in JET-C. The N-2 seeding in JET-ILW significantly affects the ELMs. The JET-ILW plasmas with N-2 seeding are characterized by ELM energy losses and time scales similar to the JET-C and by the absence of the slow transport events.

  • 45.
    Frassinetti, Lorenzo
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Dunne, M. G.
    Beurskens, M.
    Wolfrum, E.
    Bogomolov, A.
    Carralero, D.
    Cavedon, M.
    Fischer, R.
    Laggner, F. M.
    McDermott, R. M.
    Meyer, H.
    Tardini, G.
    Viezzer, E.
    ELM behavior in ASDEX Upgrade with and without nitrogen seeding2017Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 57, nr 2, artikel-id 022004Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The Type I ELM behavior in ASDEX Upgrade with full W plasma facing components is studied in terms of time scales and energy losses for a large set of shots characterized by similar operational parameters but different nitrogen seeding rate and input power. ELMs with no nitrogen can have two typical behaviors, that can be classified depending on their duration, the long and the short ELMs. The work shows that both short and long ELMs have a similar first phase, but the long ELMs are characterized by a second phase with further energy losses. The second phase disappears when nitrogen is seeded with a flux rate above 10(22) (e s(-1)). The phenomenon is compatible with a threshold effect. The presence of the second phase is related to a high divertor/scrape-off layer (SOL) temperature and/or to a low pedestal temperature. The ELM energy losses of the two phases are regulated by different mechanisms. The energy losses of the first phase increase with nitrogen which, in turn, produce the increase of the pedestal temperature. So the energy losses of the first phase are regulated by the pedestal top parameters and the increase with nitrogen is due to the decreasing pedestal collisionality. The energy losses of the second phase are related to the divertor/ SOL conditions. The long ELMs energy losses increase with increasing divertor temperature and with the number of the expelled filaments. In terms of the power lost by the plasma, the nitrogen seeding increases the power losses of the short ELMs. The long ELMs have a first phase with power losses comparable to the short ELMs losses. Assuming no major difference in the wetted area, these results suggest that (i) the nitrogen might increase the divertor heat fluxes during the short ELMs and that (ii) the long ELMs, despite the longer time scale, are not beneficial in terms of divertor heat loads.

  • 46.
    Frassinetti, Lorenzo
    et al.
    Consorzio RFX.
    Gobbin, M
    Marrelli, L
    Piovesan, P
    Franz, P
    Martin, P
    Chapman, BE
    Perturbative transport studies in the reversed-field pinch2005Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 45, nr 11, s. 1342-1349Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    In this paper we present the results of transient transport experiments in a reversed-field pinch device. Measurements have been made in the Madison Symmetric Torus experiment using a novel soft x-ray diagnostic. Spontaneous transient transport events are observed in enhanced confinement shots obtained using the pulsed parallel current drive technique, as a consequence of bursts of magnetic fluctuations triggered by an edge resonant m = 0 instability. The perturbed electron heat diffusivity, chi(e), is estimated through a numerical transient heat transport model, and the values thus obtained are compared with those measured in similar unperturbed enhanced confinement and standard plasmas using the power balance technique.

  • 47.
    Frassinetti, Lorenzo
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Menmuir, Sheena
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Olofsson, K. Erik J.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Brunsell, Per R.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Drake, James Robert
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Tearing mode velocity braking due to resonant magnetic perturbations2012Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 52, nr 10, s. 103014-Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The effect of resonant magnetic perturbations (RMPs) on the tearing mode (TM) velocity is studied in EXTRAP T2R. Experimental results show that the RMP produces TM braking until a new steady velocity or wall locking is reached. The braking is initially localized at the TM resonance and then spreads to the other TMs and to the rest of the plasma producing a global velocity reduction via the viscous torque. The process has been used to experimentally estimate the kinematic viscosity profile, in the range 2-40 m 2 s -1, and the electromagnetic torque produced by the RMP, which is strongly localized at the TM resonance. Experimental results are then compared with a theoretical model which gives a reasonable qualitative explanation of the entire process.

  • 48.
    Frassinetti, Lorenzo
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Olofsson, Erik
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Brunsell, Per
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Drake, James robert
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik. KTH, Skolan för elektro- och systemteknik (EES), Centra, Alfvénlaboratoriet.
    Implementation of advanced feedback control algorithms for controlled resonant magnetic perturbation physics studies on EXTRAP T2R2011Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 51, nr 6, s. 063018-Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The EXTRAP T2R feedback system (active coils, sensor coils and controller) is used to study and develop new tools for advanced control of the MHD instabilities in fusion plasmas. New feedback algorithms developed in EXTRAP T2R reversed-field pinch allow flexible and independent control of each magnetic harmonic. Methods developed in control theory and applied to EXTRAP T2R allow a closed-loop identification of the machine plant and of the resistive wall modes growth rates. The plant identification is the starting point for the development of output-tracking algorithms which enable the generation of external magnetic perturbations. These algorithms will then be used to study the effect of a resonant magnetic perturbation (RMP) on the tearing mode (TM) dynamics. It will be shown that the stationary RMP can induce oscillations in the amplitude and jumps in the phase of the rotating TM. It will be shown that the RMP strongly affects the magnetic island position.

  • 49.
    Frassinetti, Lorenzo
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Olofsson, Erik
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Brunsell, Per R.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Drake, James R.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Resonant magnetic perturbation effect on tearing mode dynamics2010Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 50, nr 3, s. 035005-Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The effect of a resonant magnetic perturbation (RMP) on the tearing mode (TM) dynamics is experimentally studied in the EXTRAP T2R device. EXTRAP T2R is equipped with a set of sensor coils and active coils connected by a digital controller allowing a feedback control of the magnetic instabilities. The recently upgraded feedback algorithm allows the suppression of all the error field harmonics but keeping a selected harmonic to the desired amplitude, therefore opening the possibility of a clear study of the RMP effect on the corresponding TM. The paper shows that the RMP produces two typical effects: (1) a weak oscillation in the TM amplitude and a modulation in the TM velocity or (2) a strong modulation in the TM amplitude and phase jumps. Moreover, the locking mechanism of a TM to a RMP is studied in detail. It is shown that before the locking, the TM dynamics is characterized by velocity modulation followed by phase jumps. Experimental results are reasonably explained by simulations obtained with a model.

  • 50.
    Frassinetti, Lorenzo
    et al.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Sun, Y.
    Fridström, Richard
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Menmuir, Sheena
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Olofsson, K. E. J.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Brunsell, Per R.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Khan, M. W. M.
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Liang, Y.
    Drake, James Robert
    KTH, Skolan för elektro- och systemteknik (EES), Fusionsplasmafysik.
    Braking due to non-resonant magnetic perturbations and comparison with neoclassical toroidal viscosity torque in EXTRAP T2R2015Ingår i: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 55, nr 11, artikel-id 112003Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The non-resonant magnetic perturbation (MP) braking is studied in the EXTRAP T2R reversed-field pinch (RFP) and the experimental braking torque is compared with the torque expected by the neoclassical toroidal viscosity (NTV) theory. The EXTRAP T2R active coils can apply magnetic perturbations with a single harmonic, either resonant or non-resonant. The non-resonant MP produces velocity braking with an experimental torque that affects a large part of the core region. The experimental torque is clearly related to the plasma displacement, consistent with a quadratic dependence as expected by the NTV theory. The work show a good qualitative agreement between the experimental torque in a RFP machine and NTV torque concerning both the torque density radial profile and the dependence on the non-resonant MP harmonic.

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