Impurity seeding in ASDEX upgrade tokamak modeled by COREDIV code

Energy Technology Data Exchange (ETDEWEB)

Galazka, K.; Ivanova-Stanik, I.; Czarnecka, A.; Zagoerski, R. [Institute of Plasma Physics and Laser Microfusion, Warsaw (Poland); Bernert, M.; Kallenbach, A. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Collaboration: ASDEX Upgrade Team

2016-08-15

The self-consistent COREDIV code is used to simulate discharges in a tokamak plasma, especially the influence of impurities during nitrogen and argon seeding on the key plasma parameters. The calculations are performed with and without taking into account the W prompt redeposition in the divertor area and are compared to the experimental results acquired on ASDEX Upgrade tokamak (shots 29254 and 29257). For both impurities the modeling shows a better agreement with the experiment in the case without prompt redeposition. It is attributed to higher average tungsten concentration, which on the other hand seriously exceeds the experimental value. By turning the prompt redeposition process on, the W concentration is lowered, what, in turn, results in underestimation of the radiative power losses. By analyzing the influence of the transport coefficients on the radiative power loss and average W concentration it is concluded that the way to compromise the opposing tendencies is to include the edge-localized mode flushing mechanism into the code, which dominates the experimental particle and energy balance. Also performing the calculations with both anomalous and neoclassical diffusion transport mechanisms included is suggested. (copyright 2016 The Authors. Contributions to Plasma Physics published by Wiley-VCH Verlag GmbH and Co. KGaA Weinheim. This)

Profile measurements of the electron temperature on the ASDEX Upgrade, COMPASS, and ISTTOK tokamak using Thomson scattering, triple, and ball-pen probes

Energy Technology Data Exchange (ETDEWEB)

Adamek, J., E-mail: adamek@ipp.cas.cz; Horacek, J.; Bilkova, P.; Böhm, P.; Aftanas, M.; Vondracek, P.; Stöckel, J.; Panek, R. [Institute of Plasma Physics, Prague (Czech Republic); Müller, H. W. [Max-Planck-Institute for Plasma Physics, Garching near Munich (Germany); Institute of Materials Chemistry & Research, University of Vienna, Vienna (Austria); Silva, C.; Fernandes, H.; Figueiredo, H. [Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade de Lisboa, 1049-001 Lisboa (Portugal); Schrittwieser, R.; Ionita, C.; Mehlmann, F.; Costea, S. [Institute for Ion Physics and Applied Physics, University of Innsbruck, Innsbruck (Austria); Kurzan, B. [Max-Planck-Institute for Plasma Physics, Garching near Munich (Germany)

2016-04-15

The ball-pen probe (BPP) technique is used successfully to make profile measurements of the electron temperature on the ASDEX Upgrade (Axially Symmetric Divertor Experiment), COMPASS (COMPact ASSembly), and ISTTOK (Instituto Superior Tecnico TOKamak) tokamak. The electron temperature is provided by a combination of the BPP potential (Φ{sub BPP}) and the floating potential (V{sub fl}) of the Langmuir probe (LP), which is compared with the Thomson scattering diagnostic on ASDEX Upgrade and COMPASS. Excellent agreement between the two diagnostics is obtained for circular and diverted plasmas and different heating mechanisms (Ohmic, NBI, ECRH) in deuterium discharges with the same formula T{sub e} = (Φ{sub BPP} − V{sub fl})/2.2. The comparative measurements of the electron temperature using BPP/LP and triple probe (TP) techniques on the ISTTOK tokamak show good agreement of averaged values only inside the separatrix. It was also found that the TP provides the electron temperature with significantly higher standard deviation than BPP/LP. However, the resulting values of both techniques are well in the phase with the maximum of cross-correlation function being 0.8.

Profile measurements of the electron temperature on the ASDEX Upgrade, COMPASS, and ISTTOK tokamak using Thomson scattering, triple, and ball-pen probes

International Nuclear Information System (INIS)

Adamek, J.; Horacek, J.; Bilkova, P.; Böhm, P.; Aftanas, M.; Vondracek, P.; Stöckel, J.; Panek, R.; Müller, H. W.; Silva, C.; Fernandes, H.; Figueiredo, H.; Schrittwieser, R.; Ionita, C.; Mehlmann, F.; Costea, S.; Kurzan, B.

2016-01-01

The ball-pen probe (BPP) technique is used successfully to make profile measurements of the electron temperature on the ASDEX Upgrade (Axially Symmetric Divertor Experiment), COMPASS (COMPact ASSembly), and ISTTOK (Instituto Superior Tecnico TOKamak) tokamak. The electron temperature is provided by a combination of the BPP potential (Φ_B_P_P) and the floating potential (V_f_l) of the Langmuir probe (LP), which is compared with the Thomson scattering diagnostic on ASDEX Upgrade and COMPASS. Excellent agreement between the two diagnostics is obtained for circular and diverted plasmas and different heating mechanisms (Ohmic, NBI, ECRH) in deuterium discharges with the same formula T_e = (Φ_B_P_P − V_f_l)/2.2. The comparative measurements of the electron temperature using BPP/LP and triple probe (TP) techniques on the ISTTOK tokamak show good agreement of averaged values only inside the separatrix. It was also found that the TP provides the electron temperature with significantly higher standard deviation than BPP/LP. However, the resulting values of both techniques are well in the phase with the maximum of cross-correlation function being 0.8.

The strongest magnetic barrier in the DIII-D tokamak and comparison with the ASDEX UG

Science.gov (United States)

Ali, Halima; Punjabi, Alkesh

2013-05-01

Magnetic perturbations in tokamaks lead to the formation of magnetic islands, chaotic field lines, and the destruction of flux surfaces. Controlling or reducing transport along chaotic field lines is a key challenge in magnetically confined fusion plasmas. A local control method was proposed by Chandre et al. [Nucl. Fusion 46, 33-45 (2006)] to build barriers to magnetic field line diffusion by addition of a small second-order control term localized in the phase space to the field line Hamiltonian. Formation and existence of such magnetic barriers in Ohmically heated tokamaks (OHT), ASDEX UG and piecewise analytic DIII-D [Luxon, J.L.; Davis, L.E., Fusion Technol. 8, 441 (1985)] plasma equilibria was predicted by the authors [Ali, H.; Punjabi, A., Plasma Phys. Control. Fusion 49, 1565-1582 (2007)]. Very recently, this prediction for the DIII-D has been corroborated [Volpe, F.A., et al., Nucl. Fusion 52, 054017 (2012)] by field-line tracing calculations, using experimentally constrained Equilibrium Fit (EFIT) [Lao, et al., Nucl. Fusion 25, 1611 (1985)] DIII-D equilibria perturbed to include the vacuum field from the internal coils utilized in the experiments. This second-order approach is applied to the DIII-D tokamak to build noble irrational magnetic barriers inside the chaos created by the locked resonant magnetic perturbations (RMPs) (m, n)=(3, 1)+(4, 1), with m and n the poloidal and toroidal mode numbers of the Fourier expansion of the magnetic perturbation with amplitude δ. A piecewise, analytic, accurate, axisymmetric generating function for the trajectories of magnetic field lines in the DIII-D is constructed in magnetic coordinates from the experimental EFIT Grad-Shafranov solver [Lao, L, et al., Fusion Sci. Technol. 48, 968 (2005)] for the shot 115,467 at 3000 ms in the DIII-D. A symplectic mathematical map is used to integrate field lines in the DIII-D. A numerical algorithm [Ali, H., et al., Radiat. Eff. Def. Solids Inc. Plasma Sc. Plasma Tech. 165, 83

Axisymmetric disruption dynamics including current profile changes in the ASDEX-Upgrade tokamak

International Nuclear Information System (INIS)

Nakamura, Y.; Pautasso, G.; Gruber, O.; Jardin, S.C.

2002-01-01

Axisymmetric MHD simulations have revealed a new driving mechanism that governs the vertical displacement event (VDE) dynamics in tokamak disruptions. A rapid flattening of the plasma current profile during the disruption plays a substantial role in dragging a single null-diverted plasma vertically towards the divertor. As a consequence, the occurrence of downward-going VDEs predominates over the upward-going ones in bottom-diverted discharges. This dragging effect, due to an abrupt change in the current profile, is absent in up-down symmetric limiter discharges. These simulation results are consistent with experiments in ASDEX-Upgrade. Together with the attractive force that arises from passive shell currents induced by the plasma current quench, the dragging effect explains many details of the VDE dynamics over the whole period of the disruptive termination. (author)

Overview of ASDEX Upgrade results

DEFF Research Database (Denmark)

Stroth, U.; Adamek, J.; Aho-Mantila, L.

2013-01-01

The medium size divertor tokamak ASDEX Upgrade (major and minor radii 1.65 m and 0.5 m, respectively, magnetic-field strength 2.5 T) possesses flexible shaping and versatile heating and current drive systems. Recently the technical capabilities were extended by increasing the electron cyclotron r...

Overview of ASDEX Upgrade results

NARCIS (Netherlands)

Stroth, U.; Adamek, J.; Aho-Mantila, L.; Akaslompolo, S.; Amdor, C.; Angioni, C.; Balden, M.; Bardin, S.; L. Barrera Orte,; Behler, K.; Belonohy, E.; Bergmann, A.; Bernert, M.; Bilato, R.; Birkenmeier, G.; Bobkov, V.; Boom, J.; Bottereau, C.; Bottino, A.; Braun, F.; Brezinsek, S.; Brochard, T.; M. Brüdgam,; Buhler, A.; Burckhart, A.; Casson, F. J.; Chankin, A.; Chapman, I.; Clairet, F.; Classen, I.G.J.; Coenen, J. W.; Conway, G. D.; Coster, D. P.; Curran, D.; da Silva, F.; P. de Marné,; D' Inca, R.; Douai, D.; Drube, R.; Dunne, M.; Dux, R.; Eich, T.; Eixenberger, H.; Endstrasser, N.; Engelhardt, K.; Esposito, B.; Fable, E.; Fischer, R.; H. Fünfgelder,; Fuchs, J. C.; K. Gál,; M. García Muñoz,; Geiger, B.; Giannone, L.; T. Görler,; da Graca, S.; Greuner, H.; Gruber, O.; Gude, A.; Guimarais, L.; S. Günter,; Haas, G.; Hakola, A. H.; Hangan, D.; Happel, T.; T. Härtl,; Hauff, T.; Heinemann, B.; Herrmann, A.; Hobirk, J.; H. Höhnle,; M. Hölzl,; Hopf, C.; Houben, A.; Igochine, V.; Ionita, C.; Janzer, A.; Jenko, F.; Kantor, M.; C.-P. Käsemann,; Kallenbach, A.; S. Kálvin,; Kantor, M.; Kappatou, A.; Kardaun, O.; Kasparek, W.; Kaufmann, M.; Kirk, A.; H.-J. Klingshirn,; Kocan, M.; Kocsis, G.; Konz, C.; Koslowski, R.; Krieger, K.; Kubic, M.; Kurki-Suonio, T.; Kurzan, B.; Lackner, K.; Lang, P. T.; Lauber, P.; Laux, M.; Lazaros, A.; Leipold, F.; Leuterer, F.; Lindig, S.; Lisgo, S.; Lohs, A.; Lunt, T.; Maier, H.; Makkonen, T.; Mank, K.; M.-E. Manso,; Maraschek, M.; Mayer, M.; McCarthy, P. J.; McDermott, R.; Mehlmann, F.; Meister, H.; Menchero, L.; Meo, F.; Merkel, P.; Merkel, R.; Mertens, V.; Merz, F.; Mlynek, A.; Monaco, F.; Müller, S.; H.W. Müller,; M. Münich,; Neu, G.; Neu, R.; Neuwirth, D.; Nocente, M.; Nold, B.; Noterdaeme, J. M.; Pautasso, G.; Pereverzev, G.; B. Plöckl,; Podoba, Y.; Pompon, F.; Poli, E.; Polozhiy, K.; Potzel, S.; Puschel, M. J.; Putterich, T.; Rathgeber, S. K.; Raupp, G.; Reich, M.; Reimold, F.; Ribeiro, T.; Riedl, R.; Rohde, V.; van Rooij, G. J.; Roth, J.; Rott, M.; Ryter, F.; Salewski, M.; Santos, J.; Sauter, P.; Scarabosio, A.; Schall, G.; Schmid, K.; Schneider, P. A.; Schneider, W.; Schrittwieser, R.; Schubert, M.; Schweinzer, J.; Scott, B.; Sempf, M.; Sertoli, M.; Siccinio, M.; Sieglin, B.; Sigalov, A.; Silva, A.; Sommer, F.; A. Stäbler,; Stober, J.; Streibl, B.; Strumberger, E.; Sugiyama, K.; Suttrop, W.; Tala, T.; Tardini, G.; Teschke, M.; Tichmann, C.; Told, D.; Treutterer, W.; Tsalas, M.; VanZeeland, M. A.; Varela, P.; Veres, G.; Vicente, J.; Vianello, N.; Vierle, T.; Viezzer, E.; Viola, B.; Vorpahl, C.; Wachowski, M.; Wagner, D.; Wauters, T.; Weller, A.; Wenninger, R.; Wieland, B.; Willensdorfer, M.; Wischmeier, M.; Wolfrum, E.; E. Würsching,; Yu, Q.; Zammuto, I.; Zasche, D.; Zehetbauer, T.; Zhang, Y.; Zilker, M.; Zohm, H.

2013-01-01

The medium size divertor tokamak ASDEX Upgrade (major and minor radii 1.65 m and 0.5 m, respectively, magnetic-field strength 2.5 T) possesses flexible shaping and versatile heating and current drive systems. Recently the technical capabilities were extended by increasing the electron cyclotron

Investigation of limiter recycling in the divertor tokamak ASDEX

International Nuclear Information System (INIS)

Wagner, F.

1981-08-01

A divertor experiment like the ASDEX tokamak is especially suited for studying ion recycling at a material limiter, because the plasma can alternatively be limited by a magnetic limiter (separatrix) or by a material limiter. The role of the material limiter in ion recycling is documented by observing the increase in charge exchange flux emitted at the limiter position, and the decrease in external gas input necessary to keep the plasma line density invariant, when the material limiter is moved to the plasma. Ion recycling occurs predominantly at the outside section of a ring limiter. The limiter material saturates shortly after the start of the discharge. About 60% of the total recycling occurs at the limiter, which is nearly 100% of the ion recycling. The remaining 40% of the total recycling is carried by charge exchange neutrals. Due to saturation, the recycling coefficient at the limiter is 1; the recycling coefficient of the charge exchange neutrals at the wall is approximately 0.5 giving rise to a total recycling coefficient of limiter discharges of 0.8-0.9. It is observed that the plasma resistivity increases when the material limiter is moved toward the separatrix. The increase in Zsub(eff) can tentatively be explained by proton sputtering. (orig.)

Optimized tokamak power exhaust with double radiative feedback in ASDEX Upgrade

Science.gov (United States)

Kallenbach, A.; Bernert, M.; Eich, T.; Fuchs, J. C.; Giannone, L.; Herrmann, A.; Schweinzer, J.; Treutterer, W.; the ASDEX Upgrade Team

2012-12-01

A double radiative feedback technique has been developed on the ASDEX Upgrade tokamak for optimization of power exhaust with a standard vertical target divertor. The main chamber radiation is measured in real time by a subset of three foil bolometer channels and controlled by argon injection in the outer midplane. The target heat flux is in addition controlled by nitrogen injection in the divertor private flux region using either a thermoelectric sensor or the scaled divertor radiation obtained by a bolometer channel in the outer divertor. No negative interference of the two radiation controllers has been observed so far. The combination of main chamber and divertor radiative cooling extends the operational space of a standard divertor configuration towards high values of P/R. Pheat/R = 14 MW m-1 has been achieved so far with nitrogen seeding alone as well as with combined N + Ar injection, with the time-averaged divertor peak heat flux below 5 MW m-2. Good plasma performance can be maintained under these conditions, namely H98(y,2) = 1 and βN = 3.

Overview of ASDEX Upgrade results

Czech Academy of Sciences Publication Activity Database

Stroth, U.; Adámek, Jiří; Aho-Mantila, L.; Äkäslompolo, S.; Amdor, C.; Angioni, C.; Balden, M.; Bardin, S.; Barrera Orte, L.; Behler, K.; Belonohy, E.; Bergmann, A.; Bernert, M.; Bilato, R.; Birkenmeier, G.; Bobkov, V.; Boom, J.; Bottereau, C.; Bottino, A.; Braun, F.; Brezinsek, S.; Brochard, T.; Brüdgam, M.; Buhler, A.; Burckhart, A.; Casson, F.J.; Chankin, A.; Chapman, I.; Clairet, F.; Classen, I.G.J.; Coenen, J.W.; Conway, G.D.; Coster, D.P.; Curran, D.; da Silva, F.; de Marné, P.; D’Inca, R.; Douai, D.; Drube, R.; Dunne, M.; Dux, R.; Eich, T.; Eixenberger, H.; Endstrasser, N.; Engelhardt, K.; Esposito, B.; Fable, E.; Fischer, R.; Fünfgelder, H.; Fuchs, J.C.; Gál, K.; García Munoz, M.; Geiger, B.; Giannone, L.; Görler, T.; da Graca, S.; Greuner, H.; Gruber, O.; Gude, A.; Guimarais, L.; Günter, S.; Haas, G.; Hakola, A.H.; Hangan, D.; Happel, T.; Härtl, T.; Hauff, T.; Heinemann, B.; Herrmann, A.; Hobirk, J.; Höhnle, H.; Hölzl, M.; Hopf, C.; Igochine, V.; Ionita, C.; Janzer, A.; Jenko, F.; Käsemann, C.-P.; Kallenbach, A.; Kálvin, S.; Kantor, M.; Kappatou, A.; Kardaun, O.; Kasparek, W.; Kaufmann, M.; Kirk, A.; Klingshirn, H.-J.; Kocan, M.; Kocsis, G.; Konz, C.; Koslowski, R.; Krieger, K.; Kubic, M.; Kurki-Suonio, T.; Kurzan, B.; Lackner, K.; Lang, P.T.; Lauber, P.; Laux, M.; Lazaros, A.; Leipold, F.; Leuterer, F.; Lindig, S.; Lisgo, S.; Lohs, A.; Lunt, T.; Maier, H.; Makkonen, T.; Mank, K.; Manso, M.-E.; Maraschek, M.; Mayer, M.; McCarthy, P.J.; McDermott, R.; Mehlmann, F.; Meister, H.; Menchero, L.; Meo, F.; Merkel, P.; Merkel, R.; Mertens, V.; Merz, F.; Mlynek, A.; Monaco, F.; Müller, S.; Müller, H.W.; Münich, M.; Neu, G.; Neu, R.; Neuwirth, D.; Nocente, M.; Nold, B.; Noterdaeme, J.-M.; Pautasso, G.; Pereverzev, G.; Plöckl, B.; Podoba, Y.; Pompon, F.; Poli, E.; Polozhiy, K.; Potzel, S.; Püschel, M.J.; Pütterich, T.; Rathgeber, S.K.; Raupp, G.; Reich, M.; Reimold, M.; Ribeiro, T.; Riedl, R.; Rohde, V.; v.Rooij, G.; Roth, J.; Rott, M.; Ryter, F.; Salewski, M.; Santos, J.; Sauter, P.; Scarabosio, A.; Schall, G.; Schmid, K.; Schneider, P.A.; Schneider, W.; Schrittwieser, R.; Schubert, M.; Schweinzer, J.; Scott, B.; Sempf, M.; Sertoli, M.; Siccinio, M.; Sieglin, B.; Sigalov, A.; Silva, A.; Sommer, F.; Stäbler, A.; Stober, J.; Streibl, B.; Strumberger, E.; Sugiyama, K.; Suttrop, W.; Tala, T.; Tardini, G.; Teschner, M.; Tichmann, C.; Told, D.; Treutterer, W.; Tsalas, M.; Van Zeeland, M.A.; Varela, P.; Véres, G.; Vicente, J.; Vianello, N.; Vierle, T.; Viezzer, E.; Viola, B.; Vorpahl, C.; Wachowski, M.; Wagner, D.; Wauters, T.; Weller, A.; Wenninger, R.; Wieland, B.; Willensdorfer, M.; Wischmeier, M.; Wolfrum, E.; Würsching, E.; Yu, Q.; Zammuto, I.; Zasche, D.; Zehetbauer, T.; Zhang, Y.; Zilker, M.; Zohm, H.

2013-01-01

Roč. 53, č. 10 (2013), s. 104003-104003 ISSN 0029-5515. [IAEA Fusion Energy Conference/24./. San Diego, 08.10.2012-13.10.2012] Institutional support: RVO:61389021 Keywords : tokamak * ASDEX * ITER * ICRH system Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.243, year: 2013 http://m.iopscience.iop.org/0029-5515/53/10/104003/pdf/0029-5515_53_10_104003.pdf

Fast-ion losses induced by ACs and TAEs in the ASDEX Upgrade tokamak

International Nuclear Information System (INIS)

GarcIa-Munoz, M.; Hicks, N.; Classen, I.G.J.; Bilato, R.; Bobkov, V.; Brambilla, M.; Bruedgam, M.; Fahrbach, H.-U.; Igochine, V.; Maraschek, M.; Sassenberg, K.; Van Voornveld, R.; Jaemsae, S.

2010-01-01

The phase-space of convective and diffusive fast-ion losses induced by shear Alfven eigenmodes has been characterized in the ASDEX Upgrade tokamak. Time-resolved energy and pitch-angle measurements of fast-ion losses correlated in frequency and phase with toroidal Alfven eigenmodes (TAEs) and Alfven cascades (ACs) have allowed to identify both loss mechanisms. While single ACs and TAEs eject resonant fast-ions in a convective process, the overlapping of AC and TAE spatial structures leads to a large fast-ion diffusion and loss. The threshold for diffusive fast-ion losses depends on the ion energy (gyroradius). Diffusive fast-ion losses with gyroradius ∼70 mm have been observed with a single TAE for local radial displacements of the magnetic field lines larger than ∼2 mm. Multiple frequency chirping ACs cause an enhancement of the diffusive losses. The ACs and TAEs radial structures have been reconstructed by means of cross-correlation techniques between the fast-ion loss detector and the electron cyclotron emission radiometer.

Measurement of turbulent electron temperature fluctuations on the ASDEX Upgrade tokamak using correlated electron cyclotron emission

Energy Technology Data Exchange (ETDEWEB)

Freethy, S. J., E-mail: simon.freethy@ipp.mpg.de [Max Planck Institute for Plasma Physics, 85748 Garching (Germany); Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Conway, G. D.; Happel, T.; Köhn, A. [Max Planck Institute for Plasma Physics, 85748 Garching (Germany); Classen, I.; Vanovac, B. [FOM Institute DIFFER, 5612 AJ Eindhoven (Netherlands); Creely, A. J.; White, A. E. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States)

2016-11-15

Turbulent temperature fluctuations are measured on the ASDEX Upgrade tokamak using pairs of closely spaced, narrow-band heterodyne radiometer channels and a standard correlation technique. The pre-detection spacing and bandwidth of the radiometer channel pairs is chosen such that they are physically separated less than a turbulent correlation length, but do not overlap. The radiometer has 4 fixed filter frequency channels and two tunable filter channels for added flexibility in the measurement position. Relative temperature fluctuation amplitudes are observed in a helium plasma to be δT/T = (0.76 ± 0.02)%, (0.67 ± 0.02)%, and (0.59 ± 0.03)% at normalised toroidal flux radius of ρ{sub tor} = 0.82, 0.75, and 0.68, respectively.

Edge fluctuations and global confinement with lower hybrid current drive in the ASDEX tokamak

Energy Technology Data Exchange (ETDEWEB)

Stoeckel, J; Soeldner, F X; Giannone, L.; Leuterer, F; Steuer, K H [Association Euratom-Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); ASDEX Team

1992-03-01

Electrostatic edge fluctuations were investigated by means of Langmuir probes on the ASDEX tokamak in lower hybrid current drive regimes, simultaneously with the global particle and energy balances. It was found that the edge fluctuations are reduced and the global particle/energy confinement improves when the LH power is below the initial ohmic power. The maximum reduction of the fluctuations and the best confinement occur when the total power input (OH + LH) is minimum. With a LH power higher than the initial OH value, the fluctuation level increases noticeably, while no improvement of the global confinement is observed. The increase of the edge fluctuations seems to be poloidally localized and caused by local power deposition in front of the grill antenna. Therefore, the relative positions of the probe and antenna structure have to be taken account for correct interpretation of the fluctuation data. (orig.).

Edge fluctuations and global confinement with lower hybrid current drive in the ASDEX tokamak

International Nuclear Information System (INIS)

Stoeckel, J.; Soeldner, F.X.; Giannone, L.; Leuterer, F.; Steuer, K.H.

1992-03-01

Electrostatic edge fluctuations were investigated by means of Langmuir probes on the ASDEX tokamak in lower hybrid current drive regimes, simultaneously with the global particle and energy balances. It was found that the edge fluctuations are reduced and the global particle/energy confinement improves when the LH power is below the initial ohmic power. The maximum reduction of the fluctuations and the best confinement occur when the total power input (OH + LH) is minimum. With a LH power higher than the initial OH value, the fluctuation level increases noticeably, while no improvement of the global confinement is observed. The increase of the edge fluctuations seems to be poloidally localized and caused by local power deposition in front of the grill antenna. Therefore, the relative positions of the probe and antenna structure have to be taken account for correct interpretation of the fluctuation data. (orig.)

Creation of a magnetic barrier at a noble q close to physical midpoint between two resonant surfaces in the ASDEX UG tokamak

Science.gov (United States)

Vazquez, Justin; Ali, Halima; Punjabi, Alkesh

2009-11-01

Ciraolo, Vittot and Chandre method of building invariant manifolds inside chaos in Hamiltonian systems [Ali H. and Punjabi A, Plasma Phys. Control. Fusion, 49, 1565--1582 (2007)] is used in the ASDEX UG tokamak. In this method, a second order perturbation is added to the perturbed Hamiltonian [op cit]. It creates an invariant torus inside the chaos, and reduces the plasma transport. The perturbation that is added to the equilibrium Hamiltonian is at least an order of magnitude smaller than the perturbation that causes chaos. This additional term has a finite, limited number of Fourier modes. Resonant magnetic perturbations (m,n) = (3,2)+(4,3) are added to the field line Hamiltonian for the ASDEX UG. An area-preserving map for the field line trajectories in the ASDEX UG is used. The common amplitude δ of these modes that gives complete chaos between the resonant surfaces ψ43 and ψ32 is determined. A magnetic barrier is built at a surface with noble q that is very nearly equals to the q at the physical midpoint between the two resonant surfaces. The maximum amplitude of magnetic perturbation for which this barrier can be sustained is determined. This work is supported by US Department of Energy grants DE-FG02-07ER54937, DE-FG02-01ER54624 and DE-FG02-04ER54793.

A data bank of disruptive discharges in ASDEX

International Nuclear Information System (INIS)

Ludescher, C.; Lackner, K.; Schneider, F.

1994-03-01

The compilation of data banks relating to plasma disruptions is important for the design of next-step devices and tokamak reactors, as a means of establishing safe operation regimes and assessing the residual risk from such events. ASDEX has an operational history of 33509 plasma shots covering an exceptionally wide range of machine conditions: Divertor/limiter configurations; Ohmic, NBI, ICRH and LH heating; carbonization, boronization wall-conditioning, gas-puff and pellet refuelling. We have compiled a data base of the Disruptive Operationl Regimes in ASDEX (DORA), which contains the relevant information for all ASDEX-discharges and is available on tape and readable by different data bank systems for further evaluation. We first describe the criteria applied to recognize and classify disruptions and the information about them stored in the file. In a second part we use the DORA file for some sample applications of physical or engineering interest. In an appendix we give the data and format information necessary to read the DORA file. (orig.)

Plasma control techniques of the ASDEX feedback system

International Nuclear Information System (INIS)

Schneider, F.

1981-01-01

In the ASDEX tokamak the shots are exactly preprogrammed and most of the disturbances are reproducible. So a computer can learn from one shot how to correct the next one. With this sort of disturbance feedforward one can also introduce a 'negative delay' in the program to compensate even fast and strong disturbances withous unwanted overswing or oscillations. The feedforward in conjunction with feedback control allows production of a magnetically limited plasma from the very beginning without any wall or limiter contact. This is a reason why in ASDEX the loop voltage on breakdown can be as low as 5 V/sup 2/. The plasma column can be controlled in the vacuum vessel even after disruptions have occurred

Density, potential and temperature fluctuations in Wendelstein 7-AS and ASDEX

International Nuclear Information System (INIS)

Balbin, R.; Hidalgo, C.; Carlson, A.; Endler, M.; Giannone, L.; Herre, G.; Niedermeyer, H.; Rudyj, A.; Theimer, G.

1992-01-01

Measurements of ion saturation current, floating potential and temperature fluctuations in Wendelstein 7-AS stellarator (W7-AS) and ASDEX tokamak have been carried out. A reciprocating Langmuir probe with an array of 19 graphite tips has been used to obtain the radial profiles of these fluctuations in W7-AS and ASDEX. In both devices, a reversal of the radial electric field and an associated velocity shear layer at the plasma boundary have been observed. At the radial position where the phase velocity the poloidal direction of the fluctuations goes to zero, the normalised ion saturation current fluctuation level of 0.2 is the same for edge plasma parameters of similar temperatures and densities. A spatial crosscorrelation between floating potential and ion saturation current fluctuations has been observed in both machines and this feature can be explained in terms of turbulent eddies. A comparison of fluctuations in a tokamak and stellarator therefore shows many features in common. (orig.)

Fusion reaction product diagnostics in ASDEX

International Nuclear Information System (INIS)

Bosch, H.S.

1987-01-01

A diagnostic method was developed to look for the charged fusion products from the D(D,p)T-reactions in the divertor tokamak ASDEX. With a semi-conductor detector it was possible to evaluate the ion temperature in thermal plasmas from the proton energy spectra as well as from the triton spectra. In lower-hybrid wave heated plasmas non-thermal (fast) ions were observed. These ions create fusion products with a characteristically different energy spectrum. (orig.)

Interpretation of the effects of electron cyclotron power absorption in pre-disruptive tokamak discharges in ASDEX Upgrade

Energy Technology Data Exchange (ETDEWEB)

Nowak, S.; Lazzaro, E.; Granucci, G. [Associazione Euratom-CNR sulla Fusione, IFP-CNR, Via R. Cozzi 53, 20125 Milano (Italy); Esposito, B. [Associazione Euratom-CNR sulla Fusione, CR Frascati, C.P. 65, 00044 Frascati (Italy); Maraschek, M.; Zohm, H. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Boltzmannstr.2, 85748 Garching bei Munchen (Germany); Sauter, O.; Brunetti, D. [CRPP, Association Euratom-Confederation Suisse, EPFL, 1015 Lausanne (Switzerland); Collaboration: ASDEX Upgrade Team

2012-09-15

Tokamak disruptions are events of fatal collapse of the magnetohydrodynamic (MHD) confinement configuration, which cause a rapid loss of the plasma thermal energy and the impulsive release of magnetic energy and heat on the tokamak first wall components. The physics of the disruptions is very complex and non-linear, strictly associated with the dynamics of magnetic tearing perturbations. The crucial problem of the response to the effects of localized heat deposition and current driven by external (rf) sources to avoid or quench the MHD tearing instabilities has been investigated both experimentally and theoretically on the ASDEX Upgrade tokamak. The analysis of the conditions under which a disruption can be prevented by injection of electron cyclotron (EC) rf power, or, alternatively, may be caused by it, shows that the local EC heating can be more significant than EC current drive in ensuring neoclassical tearing modes (NTMs) stability, due to two main reasons: first, the drop of temperature associated with the island thermal short circuit tends to reduce the neoclassical character of the instability and to limit the EC current drive generation; second, the different effects on the mode evolution of both the location of the power deposition relative to the island separatrix and the island shape deformation lead to less strict requirements of precise power deposition focussing. A contribution to the validation of theoretical models of the events associated with NTM is given and can be used to develop concepts for their control, relevant also for ITER-like scenarios.

The effect of off-axis neutral beam injection on sawtooth stability in ASDEX Upgrade and Mega-Ampere Spherical Tokamak

International Nuclear Information System (INIS)

Chapman, I. T.; de Bock, M. F.; Pinches, S. D.; Turnyanskiy, M. R.; Igochine, V. G.; Maraschek, M.; Tardini, G.

2009-01-01

Sawtooth behavior has been investigated in plasmas heated with off-axis neutral beam injection in ASDEX Upgrade [A. Herrmann and O. Gruber, Fusion Sci. Technol. 44, 569 (2003)] and the Mega-Ampere Spherical Tokamak (MAST) [A. Sykes et al., Nucl. Fusion 41, 1423 (2001)]. Provided that the fast ions are well confined, the sawtooth period is found to decrease as the neutral beam is injected further off-axis. Drift kinetic modeling of such discharges qualitatively shows that the passing fast ions born outside the q=1 rational surface can destabilize the n=1 internal kink mode, thought to be related to the sawtooth instability. This effect can be enhanced by optimizing the deposition of the off-axis beam energetic particle population with respect to the mode location.

Status of the new multi-frequency ECRH system for ASDEX Upgrade

DEFF Research Database (Denmark)

Wagner, D.; Grünwald, G.; Leuterer, F.

2008-01-01

Currently, a new multi-frequency ECRH system is under construction at the ASDEX Upgrade tokamak experiment. This system employs, for the first time in a fusion device, multi-frequency gyrotrons, step-tunable in the range 105-140 GHz. The first two gyrotrons, working at 105 and 140 GHz, were...

Strain and stress of the ASDEX multipole magnetic coils

International Nuclear Information System (INIS)

Jandl, O.; Pillsticker, M.

1978-01-01

A brief description of the technical concept of the multipole magnetic field coils for the ASDEX tokamak is given. The various loads of the coils are explained in quality. To compute displacement and stress of the coils FEM computer programs are used. The computing models applied to this problem are founded and the results and the conclusions are reported. (orig.) [de

Performance of the first ASDEX Upgrade neutral beam injector

International Nuclear Information System (INIS)

Staebler, A.; Vollmer, O.; Feist, J.H.; Speth, E.; Heinemann, B.; Melkus, W.; Obermayer, S.; Riedl, R.; Schaerich, W.; Wittenbecher, K.

1995-01-01

Plasmas of the ASDEX Upgrade tokamak have been heated with H 0 beams of up to 7 MW and D 0 beams of up to 10 MW. Beam modulation allows to inject at any power level between zero and full power. Measurements characterizing the NBI system performance, the power accountability, and the operational experience obtained so far are discussed. (orig.)

Profile measurements of the electron temperature on the ASDEX Upgrade, COMPASS, and ISTTOK tokamak using Thomson scattering, triple, and ball-pen probes

Czech Academy of Sciences Publication Activity Database

Adámek, Jiří; Müller, H.W.; Silva, C.; Schrittwieser, R.; Ionita, C.; Mehlmann, F.; Costea, S.; Horáček, Jan; Kurzan, B.; Bílková, Petra; Böhm, Petr; Aftanas, Milan; Vondráček, Petr; Stöckel, Jan; Pánek, Radomír; Fernandes, H.; Figueiredo, H.

2016-01-01

Roč. 87, č. 4 (2016), č. článku 043510. ISSN 0034-6748 R&D Projects: GA ČR(CZ) GA15-10723S; GA ČR(CZ) GAP205/12/2327; GA ČR(CZ) GA14-35260S; GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : ball- pen probe (BPP) * ASDEX Upgrade * Langmuir probe (LP) * ISTTOK (Instituto Superior Tecnico TOKamak) * COMPASS (COMPact ASSembly), Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: 2.11 Other engineering and technologies Impact factor: 1.515, year: 2016 http://scitation.aip.org/content/aip/journal/rsi/87/4/10.1063/1.4945797

Profile measurements of the electron temperature on the ASDEX Upgrade, COMPASS, and ISTTOK tokamak using Thomson scattering, triple, and ball-pen probes

Czech Academy of Sciences Publication Activity Database

Adámek, Jiří; Müller, H.W.; Silva, C.; Schrittwieser, R.; Ionita, C.; Mehlmann, F.; Costea, S.; Horáček, Jan; Kurzan, B.; Bílková, Petra; Böhm, Petr; Aftanas, Milan; Vondráček, Petr; Stöckel, Jan; Pánek, Radomír; Fernandes, H.; Figueiredo, H.

2016-01-01

Roč. 87, č. 4 (2016), č. článku 043510. ISSN 0034-6748 R&D Projects: GA ČR(CZ) GA15-10723S; GA ČR(CZ) GAP205/12/2327; GA ČR(CZ) GA14-35260S; GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : ball-pen probe (BPP) * ASDEX Upgrade * Langmuir probe (LP) * ISTTOK (Instituto Superior Tecnico TOKamak) * COMPASS (COMPact ASSembly), Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: 2.11 Other engineering and technologies Impact factor: 1.515, year: 2016 http://scitation.aip.org/content/aip/journal/rsi/87/4/10.1063/1.4945797

Low-noise cable for diagnostics, control and instrumentation of the ASDEX tokamak fusion experiment

International Nuclear Information System (INIS)

Gernhardt, J.

1988-11-01

ASDEX (Axially Symmetric Divertor EXperiment) is a large tokamak (R=1.65 m; a=0.4 m) with an air transformer. The relatively large stray field, Bz=10 mT=(100 G); for ρ=5 m, Bz=40 mT=(400 G); for ρ=3 m, BΦ=0.3 T=(3 kG); for ρ=3m, compared with that of an iron transformer, and the cable length 1≤30 m from the experiment to the control room, make mainly the magnetically induced and capacitively coupled noise signal in the cable relatively high. As a result of neutral injection (> 4 MW; 40 kV) and lower hybrid ion cyclotron and Alfven wave heating strong E-fields are produced and noise is coupled into the cables. These magnetic and electric field gradients during the plasma shot vary with time and location. This report tries to show how these noise signals can be reduced without reducing the broadcast frequency of the signal. The Electro Magnetic Compatibility and Interference (EMC, EMI) are discussed. The cost of diagnostic cable, connectors and cable ducts without mounting is approximately DM 700,000.--. (orig.)

Confinement and transport properties during current ramps in the ASDEX Upgrade tokamak

Science.gov (United States)

Fable, E.; Angioni, C.; Hobirk, J.; Pereverzev, G.; Fietz, S.; Hein, T.; ASDEX Upgrade Team

2011-04-01

A detailed analysis of experimental data from the ASDEX Upgrade tokamak is carried out to shed light on the properties of confinement and transport in the current ramp-up and ramp-down phases of the plasma discharge. The experimental database is used to identify the relevant ranges of parameters explored during the ramp-up and the ramp-down. The energy confinement time observed in the two ramps displays interesting evolution, in many cases attaining different values at the same current level between ramp-up and ramp-down. The possible reasons for this behaviour are investigated. Interpretative transport simulations are used as a tool to clarify the interplay between different parameters, which are coupled in a non-linear way. In addition, a theory-based transport model is used to understand the behaviour of confinement as observed in the experiment, evidencing the role of both turbulent and neoclassical transport. Linear gyrokinetic calculations are performed to identify the relevant turbulence regime, showing that a broad range of frequencies, in the trapped electron modes (TEMs) and in the ion temperature gradient modes (ITGs) regimes, is explored during both the ramp-up and ramp-down. In the same framework, a quasi-linear model is applied to calculate the value of the local logarithmic density gradient and compare it with the experimental value. Finally, first non-linear simulations of heat transport during the current ramps are presented.

Density, potential and temperature fluctuations in Wendelstein 7-AS and ASDEX

Energy Technology Data Exchange (ETDEWEB)

Balbin, R; Hidalgo, C [Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas (CIEMAT), Madrid (Spain); Carlson, A; Endler, M; Giannone, L.; Niedermeyer, H; Rudyj, A; Theimer, G [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

1993-12-31

Measurements of density, potential and temperature fluctuations in Wendelstein 7-AS stellarator (W7-AS) and ASDEX tokamak have been carried out. The properties of plasma fluctuations in a tokamak and stellarator can then be compared. A reciprocating Langmuir probe with an array of 19 graphite tips has been used to measure the radial profiles of fluctuations in the ion saturation current and floating potential in W7-AS and ASDEX. In both devices, a reversal in radial electric field and an associated velocity shear layer at the plasma boundary have been observed and in both cases the normalized ion saturation current fluctuation level decreases monotonically moving towards the plasma centre and through the shear layer. At the radial position where the phase velocity in the poloidal direction of the fluctuations goes to zero, the normalized ion saturation current fluctuation level of 0.25 are similar for edge plasma parameters of similar temperatures and densities. A spatial crosscorrelation between fluctuations in floating potential and ion saturation current has been observed in both machines. (author) 6 refs., 4 figs.

Creation of second order magnetic barrier inside chaos created by NTMs in the ASDEX UG

Science.gov (United States)

Ali, Halima; Punjabi, Alkesh

2012-10-01

Understanding and stabilization of neoclassical tearing modes (NTM) in tokamaks is an important problem. For low temperature plasmas, tearing modes are believed to be mainly driven by current density gradient. For collisionless plasmas, even when plasma is stable to classical tearing modes, helical reduction in bootstrap current in O-point of an island can destabilize NTMs when an initial island is seeded by other global MHD instabilities or when microturbulence triggers the transition from a linear to nonlinear instability. The onset of NTMs leads to the most serious beta limit in ASDEX UG tokamak [O. Gubner et al 2005 NF 39 1321]. The important NTMs in the ASDDEX UG are (m,n)=(3,2)+(4,3)+(1,1). Realistic parameterization of these NTMs and the safety factor in ASDEX UG are given in [O. Dumbrajs et al 2005 POP 12 1107004]. We use a symplectic map in magnetic coordinates for the ASDEX UG to integrate field lines in presence of the NTMs. We add a second order control term [H. Ali and A. Punjabi 2007 PPCF 49 1565] to this ASDEX UG field line Hamiltonian to create an invariant magnetic surface inside the chaos generated by the NTMs. The relative strength, robustness, and resilience of this barrier are studied to ascertain the most desirable noble barrier in the ASDEX UG with NTMs. We present preliminary results of this work, and discuss its implications with regard to magnetic transport barriers for increasing strength of magnetic perturbations. This work is supported by the grants DE-FG02-01ER54624 and DE-FG02-04ER54793.

Plasma shut-down with fast impurity puff on ASDEX Upgrade

International Nuclear Information System (INIS)

Pautasso, G.; Fuchs, C.J.; Gruber, O.; Maggi, C.F.; Maraschek, M.; Puetterich, T.; Rohde, V.; Wittmann, C.; Wolfrum, E.; Cierpka, P.; Beck, M.

2007-01-01

The massive injection of impurity gas into a plasma has been proved to reduce forces and localized thermal loads caused by disruptions in tokamaks. This mitigation system is routinely used on ASDEX Upgrade to shut down plasmas with a locked mode. The plasma response to impurity injection and the mechanism of reduction of the mechanical forces is discussed in the paper

High-speed lithium pellet injector commissioning in ASDEX Upgrade to investigate impact of Li in an all-metal wall tokamak

Energy Technology Data Exchange (ETDEWEB)

Arredondo Parra, Rodrigo; Lang, Peter Thomas; Ploeckl, Bernhard [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Cardella, Antonino [Technische Universitaet Muenchen, Garching (Germany); Fusion for Energy, Garching (Germany); Macian Juan, Rafael [Technische Universitaet Muenchen, Garching (Germany); Neu, Rudolf [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Technische Universitaet Muenchen, Garching (Germany)

2015-05-01

Encouraging results with respect to plasma performance have been observed in several tokamak devices (TFTR, NSTX, etc) when injecting Lithium. Recently, a pedestal broadening resulting in an enhanced energy content during transient ELM-free H-mode phases was achieved in DIII-D. Experiments are planned at ASDEX Upgrade, aiming to investigate the impact of Li in an all-metal wall tokamak and to enhance the pedestal operational space. For this purpose, a Lithium pellet injector has been developed, capable of injecting pellets with a particle content up to 1.64 . 10{sup 20} atoms (1.89 mg) at a foreseen maximum repetition rate of 3 Hz. Free flight launch from the torus outboard side without a guiding tube is envisaged. A transfer efficiency exceeding 90 % was achieved in the test bed. Pellets will be accelerated in a gas gun; hence special care must be taken to avoid deleterious effects by the propellant gas pulse, this being the main plasma gas, leading to speeds ranging from 500 (m)/(s) to 800 (m)/(s). Additionally, a large expansion volume equipped with a cryopump is added in to the flight path. The injector is expected to commence operation by May 2015.

Pellet imaging techniques on ASDEX

International Nuclear Information System (INIS)

Wurden, G.A.; Buechl, K.; Hofmann, J.; Lang, R.; Loch, R.; Rudyj, A.; Sandmann, W.

1990-01-01

As part of a USDOE/ASDEX collaboration, a detailed examination of pellet ablation in ASDEX with a variety of diagnostics has allowed a better understanding of a number of features of hydrogen ice pellet ablation in a plasma. In particular, fast gated photos with an intensified Xybion CCD video camera allow in-situ velocity measurements of the pellet as it penetrates the plasma. With time resolution of typically 100 nanoseconds and exposures every 50 microseconds, the evolution of each pellet in a multi-pellet ASDEX tokamak plasma discharge can be followed. When the pellet cloud track has striations, the light intensity profile through the cloud is hollow (dark near the pellet), whereas at the beginning or near the end of the pellet trajectory the track is typically smooth (without striations) and has a gaussian-peaked light emission profile. New, single pellet Stark broadened D α D β , and D γ spectra, obtained with a tangentially viewing scanning mirror/spectrometer with Reticon array readout, are consistent with cloud densities of 2 x 10 17 cm -3 or higher in the regions of strongest light emission. A spatially resolved array of D α detectors shows that the light variations during the pellet ablation are not caused solely by a modulation of the incoming energy flux as the pellet crosses rational q-surfaces, but instead are a result of a dynamic, non-stationary, ablation process. 20 refs., 4 figs

Spectroscopy as a major programme in ASDEX - a discussion study

International Nuclear Information System (INIS)

Fussmann, G.

1986-03-01

This report deals with the objectives and possibilities of a spectroscopy programme in ASDEX and provides some basic information on the relevant processes of atomic physics in tokamaks. The spectroscopic analogies found in observation of astrophysical objects are also briefly treated. In addition, the possibilities for conducting investigations in alternative high-Z ion sources are discussed. A first proposal for an appropriate programme is then formulated. (orig.)

Inter-ELM evolution of the edge current density profile on the ASDEX Upgrade tokamak

International Nuclear Information System (INIS)

Dunne, Michael G.

2014-01-01

The sudden decrease of plasma stored energy and subsequent power deposition on the first wall of a tokamak device due to edge localised modes (ELMs) is potentially detrimental to the success of a future fusion reactor. Understanding and control of ELMs is critical for the longevity of these devices and also to maximise their performance. The commonly accepted picture of ELMs posits a critical pressure gradient and current density in the plasma edge, above which coupled magnetohydrodynamic (MHD) peeling-ballooning modes are driven unstable. Much analysis has been presented in recent years on the spatial and temporal evolution of the edge pressure gradient. However, the edge current density has typically been overlooked due to the difficulties in measuring this quantity. In this thesis, a novel method of current density recovery is presented, using the equilibrium solver CLISTE to reconstruct a high resolution equilibrium utilising both external magnetic and internal edge kinetic data measured on the ASDEX Upgrade (AUG) tokamak. The evolution of the edge current density relative to an ELM crash is presented, showing that a resistive delay in the buildup of the current density is unlikely. An uncertainty analysis shows that the edge current density can be determined with an accuracy consistent with that of the kinetic data used. A comparison with neoclassical theory demonstrates excellent agreement between the current density determined by CLISTE and the calculated profiles. Three ELM mitigation regimes are investigated: Type-II ELMs, ELMs suppressed by external magnetic perturbations (MPs), and Nitrogen seeded ELMs. In the first two cases, the current density is found to decrease as mitigation onsets, indicating a more ballooning-like plasma behaviour. In the latter case, the flux surface averaged current density can decrease while the local current density increases, thus providing a mechanism to suppress both the peeling and ballooning modes.

Inter-ELM evolution of the edge current density profile on the ASDEX Upgrade tokamak

Energy Technology Data Exchange (ETDEWEB)

Dunne, Michael G.

2014-02-15

The sudden decrease of plasma stored energy and subsequent power deposition on the first wall of a tokamak device due to edge localised modes (ELMs) is potentially detrimental to the success of a future fusion reactor. Understanding and control of ELMs is critical for the longevity of these devices and also to maximise their performance. The commonly accepted picture of ELMs posits a critical pressure gradient and current density in the plasma edge, above which coupled magnetohydrodynamic (MHD) peeling-ballooning modes are driven unstable. Much analysis has been presented in recent years on the spatial and temporal evolution of the edge pressure gradient. However, the edge current density has typically been overlooked due to the difficulties in measuring this quantity. In this thesis, a novel method of current density recovery is presented, using the equilibrium solver CLISTE to reconstruct a high resolution equilibrium utilising both external magnetic and internal edge kinetic data measured on the ASDEX Upgrade (AUG) tokamak. The evolution of the edge current density relative to an ELM crash is presented, showing that a resistive delay in the buildup of the current density is unlikely. An uncertainty analysis shows that the edge current density can be determined with an accuracy consistent with that of the kinetic data used. A comparison with neoclassical theory demonstrates excellent agreement between the current density determined by CLISTE and the calculated profiles. Three ELM mitigation regimes are investigated: Type-II ELMs, ELMs suppressed by external magnetic perturbations (MPs), and Nitrogen seeded ELMs. In the first two cases, the current density is found to decrease as mitigation onsets, indicating a more ballooning-like plasma behaviour. In the latter case, the flux surface averaged current density can decrease while the local current density increases, thus providing a mechanism to suppress both the peeling and ballooning modes.

The energy principle applied to diverted tokamak configurations

International Nuclear Information System (INIS)

Atanasiu, C. V.; Guenter, S.; Lackner, K.; Moraru, A.; Zakharov, L. E.; Subbotin, A. A.

2008-01-01

Writing the expression of the potential energy in terms of the perturbation of the flux function, and performing an Euler minimisation, one obtains a system of ordinary differential equations in that perturbation. For a diverted configuration, the usual vanishing boundary conditions for the perturbed flux function at the magnetic axis and at infinity can no longer be used. In place of the vanishing boundary conditions at infinity, an approach to fix 'natural' boundary conditions for the system of differential equations for the perturbed flux function, just at the plasma boundary has been developed. As an example of application of our approaches, a particular equilibrium configuration of the ASDEX Upgrade tokamak has been considered and a detailed investigation of the dependence of the tearing stability parameter Δ' on plasma shape is given for a realistic tokamak equilibrium. The results shown are at least in qualitative agreement with experimental observations on ASDEX Upgrade and JET of a stabilizing influence of triangularity. The knowledge of Δ' for realistic tokamak plasmas is especially important for understanding of the plasma stability against NTMs. (authors)

Stability investigations of the ASDEX feedback system with filters for reducing thyristor noise

International Nuclear Information System (INIS)

Crisanti, F.; Schneider, F.

1983-06-01

A computer program for analysing the absolute and relative stabilities of any complex system by the root-locus method was developed. It is used to reanalyse the present horizontal position feed-back control in the ASDEX tokamak and to select the optimum parameters for this system with RCL filters for reducing thyristor noise. (orig.)

The MHD stability analysis of type I ELMS in ASDEX Upgrade Tokamak

International Nuclear Information System (INIS)

Saarelma, S.

2000-01-01

The ELMs or edge localized modes are plasma instabilities localized in the edge region of a tokamak plasma. They cause periodic expulsions of particles and energy. The ELMs play a significant role in the confinement of the plasma, helium exhaust and diverter erosion. These are crucial issues in tokamak operation and, thus, understanding the underlying physical mechanism behind the ELM phenomenon is very important. The ELMs are classified into three different types based on the plasma conditions, where they are observed, and, on the ELM frequency response to the heating power. In this thesis, type I ELMs which are the most intense and the most damaging to the diverters, are studied. A model for the ELMs presented by Connor et al. is tested in experimental ASDEX Upgrade plasmas. In the Connor model, the ELMs are explained as a result of two instabilities, ballooning and peeling modes. Also a phenomenon called the bootstrap current plays a significant role by being the destabilising trigger to the peeling modes. The method used to study the model is MHD or magnetohydrodynamics. The theory of the ideal MHD equilibrium and the linear stability analysis is described. Inclusion of the bootstrap current to the equilibrium construction is introduced. The equilibria are created using experimental data from plasma shots that display type I ELMs. The stability analysis indicates that the investigated ELM model is a feasible explanation for type I ELMs. The pressure gradient near the plasma edge was found to be close to the ballooning stability boundary as predicted by the model. The peeling mode stability analysis confirms the prediction of the model that as the bootstrap current increases, the plasma becomes unstable for peeling modes with low to intermediate toroidal mode numbers. The mode numbers agree with the experimental results. In the experiments with high triangularity, low ELM frequency and ELM-free periods were observed. This indicates better stability of the plasma

Fast-ion transport in the presence of magnetic reconnection induced by sawtooth oscillations in ASDEX Upgrade

NARCIS (Netherlands)

Geiger, B.; M. García-Muñoz,; Dux, R.; Ryter, F.; Tardini, G.; Orte, L. B.; Classen, I.G.J.; Fable, E.; Fischer, R.; Igochine, V.; McDermott, R. M.

2014-01-01

The transport of beam-generated fast ions has been investigated experimentally at the ASDEX Upgrade tokamak in the presence of sawtooth crashes. After sawtooth crashes, phase space resolved fast-ion D-alpha measurements show a significant reduction of the central fast-ion density-more than

Calculation of the electromagnetic forces on the ASDEX upgrade vacuum vessel on disruption of the plasma current

International Nuclear Information System (INIS)

Preis, H.

1986-01-01

This study investigates the magnetic field diffusion through the vacuum vessel of the ASDEX Upgrade tokamak that occurs on sudden disruption of the plasma current. Eddy currents are thereby produced in the vessel wall. Their time behaviour and distribution are determined. Furthermore, the vessel is permeated by various magnetic fields which, together with the eddy currents, exert magnetic forces in the vessel wall. These are also calculated. These numerical analyses are performed for two of the modes of operation envisaged for ASDEX Upgrade: the so-called limiter and single-null magnetic field configurations. (orig.)

Flux measurements with a sniffer probe near the wall in ASDEX

International Nuclear Information System (INIS)

Poschenrieder, W.; Venus, G.; Wang, Y.G.; Mueller, E.R.; Bartiromo, R.; Becker, G.; Bosch, H.S.; Brocken, H.; Eberhagen, A.; Fussmann, G.; Gehre, O.; Gernhardt, J.; Gierke, G. v.; Glock, E.; Gruber, O.; Haas, G.; Janeschitz, G.; Karger, F.; Kotze, P.B.; Keilhacker, M.; Klueber, O.; Kornherr, M.; Lackner, K.; Lenoci, M.; Lisitano, G.; Mayer, H.M.; McCormick, K.; Meisel, D.; Mertens, V.; Murmann, H.; Niedermeyer, H.; Rapp, H.; Roehr, H.; Ryter, F.; Schneider, F.; Siller, G.; Smeulders, P.; Soeldner, F.; Speth, E.; Steuer, K.H.; Vollmer, O.; Wagner, F.

1985-01-01

For a detailed assessment of particle recycling in a tokamak it is necessary to know quality and quantity of the particle fluxes directed to the elements of the wall. In a divertor machine like ASDEX we have to differentiate between at least four distinct elements: main chamber wall, protective limiters, collector plates, and divertor walls. Relevant data about the divertor region are obtained from pressure and flux measurements. (orig./GG)

Numerical simulations of fast ion loss measurements induced by magnetic islands in the ASDEX Upgrade tokamak

International Nuclear Information System (INIS)

Gobbin, M.; Marrelli, L.; Martin, P.; Fahrbach, H.U.; Garcia-Munoz, M.; Guenter, S.; White, R.B.

2009-01-01

A test particle approach, implemented with the Hamiltonian code ORBIT, is used to simulate measurements of fast ion losses induced by magnetic islands in the ASDEX Upgrade tokamak. In particular, the numerical simulations reproduce the toroidal localization of losses and the lost ions pitch angle and energy distribution experimentally measured with the fast ion losses detector (FILD) in the presence of a neoclassical tearing mode (NTM). The simulated NTM induced losses occurring on time scales longer than 100 μs are composed of mainly trapped or barely passing particles, consistently with the slow decay of the experimental signal from one FILD channel after the beam switch-off. The numerical simulations have been performed by taking into account the D-shaped plasma geometry, the collision mechanisms, the losses due to ripple effects and the rotation of the mode. The radial profile of the magnetic perturbation is adjusted in order to match ECE measurements. While statistical properties of FILD measurements are rather well reproduced, the simulated total amount of losses is found to be significantly affected by edge details of the magnetic perturbation as it determines the loss mechanism.

Disruption studies in ASDEX Upgrade

International Nuclear Information System (INIS)

Pautasso, G.

2002-01-01

Disruption generate large thermal and mechanical stresses on the tokamak components. For a future reactor disruptions have a significant impact on the design since all loading conditions must be analyzed in accordance with stricter design criteria (due to safety or difficult maintenance). Therefore the uncertainties affecting the predicted stresses must be reduced as much as possible with a more comprehensive set of measurements and analyses in this generation of experimental machines, and avoidance/ predictive methods must be developed further. The study of disruptions on ASDEX Upgrade is focused on these subjects, namely on: (1) understanding the physical mechanisms leading to this phenomenon and learning to avoid it or to predict its occurrence (with neural networks, for example) and to mitigate its effects; (2) analyzing the effects of disruptions on the machine to determine the functional dependence of the thermal and mechanical loads upon the discharge parameters. This allows to dimension or reinforce the machine components to withstand these loads and to extrapolate them to tokamaks still in the design phase; (3) learning to mitigate the consequence of disruptions. (author)

Nitrogen implantation in tungsten and migration in the fusion experiment ASDEX upgrade

International Nuclear Information System (INIS)

Meisl, Gerd Korbinian

2015-01-01

The implantation of nitrogen ions into tungsten was studied in laboratory experiments to understand the interaction of nitrogen containing fusion plasmas with tungsten walls. The resulting model of W-N interaction was tested by experiments in the tokamak ASDEX Upgrade. Using the measurements from these experiments as boundary condition, nitrogen transport and re-distribution in the plasma were modeled by self-consistent WallDYN-DIVIMP simulations.

Impact of lithium on the plasma performance in the all-metal-wall tokamak ASDEX upgrade

Energy Technology Data Exchange (ETDEWEB)

Lang, P.T.; Moreno Quicios, R.; Arredondo Parra, R.; Ploeckl, B.; McDermott, R.; Neu, R.; Wolfrum, E. [MPI fuer Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Maingi, R.; Mansfield, D.K.; Diallo, A. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Collaboration: ASDEX Upgrade Team

2016-07-01

Several tokamaks reported improvement in key plasma parameters concurrent with the presence of lithium in the plasma. At ASDEX Upgrade explorative experiments have been performed to find out if such effects can be observed when operating with an all-metal-wall. A gas gun launcher was developed capable to inject pellets containing about 1.6 x 10{sup 20} Li atoms at 2 Hz. The speed of about 600 m/s is sufficient to achieve core penetration and to create a homogeneous Li concentration of up to 10 %. With a typical sustainment time on the order of 100 ms, only transient Li presence without any pile up was achieved. Deposition of Li on plasma facing components, which remained for several discharges after injection, was observed. This short lived wall conditioning showed beneficial effects during plasma start-up. However, the accompanying surface contamination negatively impacted some diagnostics. The Li impact on the confinement was investigated in a dedicated plasma scenario with a proven sensitivity to nitrogen and helium. In phases with N seeding enhancing the confinement by about 30 %, Li injection resulted in a very modest, transient loss of confinement (about 5 %). No Li impact was found for pure Deuterium plasmas.

Transport analysis of high radiation and high density plasmas in the ASDEX Upgrade tokamak

Directory of Open Access Journals (Sweden)

Casali L.

2014-01-01

Full Text Available Future fusion reactors, foreseen in the “European road map” such as DEMO, will operate under more demanding conditions compared to present devices. They will require high divertor and core radiation by impurity seeding to reduce heat loads on divertor target plates. In addition, DEMO will have to work at high core densities to reach adequate fusion performance. The performance of fusion reactors depends on three essential parameters: temperature, density and energy confinement time. The latter characterizes the loss rate due to both radiation and transport processes. The DEMO foreseen scenarios described above were not investigated so far, but are now addressed at the ASDEX Upgrade tokamak. In this work we present the transport analysis of such scenarios. Plasma with high radiation by impurity seeding: transport analysis taking into account the radiation distribution shows no change in transport during impurity seeding. The observed confinement improvement is an effect of higher pedestal temperatures which extend to the core via stiffness. A non coronal radiation model was developed and compared to the bolometric measurements in order to provide a reliable radiation profile for transport calculations. High density plasmas with pellets: the analysis of kinetic profiles reveals a transient phase at the start of the pellet fuelling due to a slower density build up compared to the temperature decrease. The low particle diffusion can explain the confinement behaviour.

Analysis of toroidal rotation data for the DIII-D tokamak

International Nuclear Information System (INIS)

John, H.St.; Burrell, K.H.; Groebner, R.; DeBoo, J.; Gohil, P.

1989-01-01

Both poloidal and toroidal rotation are observed during routine neutral beam heating operation of the DIII-D tokamak. Poloidal rotation results and the empirical techniques used to measure toroidal and poloidal rotation speeds are described by Groebner et al. Here we concentrate on the analysis of recent measurements of toroidal rotation made during diverted, H-mode operation of the DIII-D tokamak during co- and counter-neutral beam injection of hydrogen into deuterium plasmas. Similar studies have been previously reported for Doublet III, ASDEX, TFTR, JET and other tokamaks. (author) 13 refs., 4 figs

Fluctuation measurements by Langmuir probes during LHCD on ASDEX tokamak

International Nuclear Information System (INIS)

Stoeckel, J.

1991-01-01

The level of edge electrostatic fluctuations decreases and the global particle/energy confinement improves during lower hybrid current drive (LHCD) regimes on ASDEX, when the total power remains below the initial OH power level. For higher powers, the fluctuations increase noticeably, whereas the global confinement is returning to its OH value. The observed increase of fluctuations is poloidally asymmetric and is caused by local power deposition in front of the grill antenna. (orig.)

An overview on plasma disruption mitigation and avoidance in tokamak

International Nuclear Information System (INIS)

He Kaihui; Pan Chuanhong; Feng Kaiming

2002-01-01

Plasma disruption, which seems to be unavoidable in Tokamak operation, occurs very fast and uncontrolled. In order to keep Tokamak plasma from disruption and mitigate the disruption frequency, the research on Tokamak plasma major disruption constitutes one of the main topics in plasma physics. The phenomena and processes of the precursor, thermal quench, current quench, VDE, halo current and runaway electrons generation during plasma disruption are analyzed in detail and systematically based on the data obtained from current Tokamaks such as TFTR, JET, JT-60U and ASDEX-U, etc. The methods to mitigate and avoid disruption in Tokamak are also highlighted schematically. Therefore, it is helpful and instructive for plasma disruption research in next generation large Tokamak such as ITER-FEAT

Investigation of low-frequency fluctuations in the edge plasma of ASDEX

Energy Technology Data Exchange (ETDEWEB)

Rudyj, A; Carlson, A; Giannone, L.; Niedermeyer, H [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany, F.R.); Bengtson, R D; Ritz, Ch P [Texas Univ., Austin, TX (USA); Kraemer, M [Bochum Univ. (Germany, F.R.); Tsois, N [NRS Demokritos, Attiki (Greece)

1989-01-01

Density fluctuations in the edge plasma of tokamaks in the frequency range up to a few 100 kHz have been reported for many years. The fluctuations are easily observed with Langmuir probes and are also visible in the H/sub {alpha}/ emission at locations with sufficient neutral gas density. High speed cine films taken on ASDEX show fluctuating stripes aligned approximately parallel to the magnetic field. It has been shown that these fluctuations, which are electrostatic, cause a major part if not all of the particle transport at the plasma edge. The mechanism driving these instabilities is however not yet clear. Langmuir probe measurements and optical observations were performed on ASDEX and a comparison was made with magnetic fluctuation measurements in order to further clarify the mechanism responsible for the edge turbulence. 5 refs., 3 figs.

Investigation of low-frequency fluctuations in the edge plasma of ASDEX

International Nuclear Information System (INIS)

Rudyj, A.; Carlson, A.; Giannone, L.; Niedermeyer, H.; Bengtson, R.D.; Ritz, Ch.P.; Kraemer, M.; Tsois, N.

1989-01-01

Density fluctuations in the edge plasma of tokamaks in the frequency range up to a few 100 kHz have been reported for many years. The fluctuations are easily observed with Langmuir probes and are also visible in the H α emission at locations with sufficient neutral gas density. High speed cine films taken on ASDEX show fluctuating stripes aligned approximately parallel to the magnetic field. It has been shown that these fluctuations, which are electrostatic, cause a major part if not all of the particle transport at the plasma edge. The mechanism driving these instabilities is however not yet clear. Langmuir probe measurements and optical observations were performed on ASDEX and a comparison was made with magnetic fluctuation measurements in order to further clarify the mechanism responsible for the edge turbulence. 5 refs., 3 figs

Recent Upgrades and Extensions of the ASDEX Upgrade ECRH System

Science.gov (United States)

Wagner, Dietmar; Stober, Jörg; Leuterer, Fritz; Monaco, Francesco; Münich, Max; Schmid-Lorch, Dominik; Schütz, Harald; Zohm, Hartmut; Thumm, Manfred; Scherer, Theo; Meier, Andreas; Gantenbein, Gerd; Flamm, Jens; Kasparek, Walter; Höhnle, Hendrik; Lechte, Carsten; Litvak, Alexander G.; Denisov, Gregory G.; Chirkov, Alexey; Popov, Leonid G.; Nichiporenko, Vadim O.; Myasnikov, Vadim E.; Tai, Evgeny M.; Solyanova, Elena A.; Malygin, Sergey A.

2011-03-01

The multi-frequency Electron Cyclotron Heating (ECRH) system at the ASDEX Upgrade tokamak employs depressed collector gyrotrons, step-tunable in the range 105-140 GHz. The system is equipped with a fast steerable launcher allowing for remote steering of the ECRH RF beam during the plasma discharge. The gyrotrons and the mirrors are fully integrated in the discharge control system. The polarization can be controlled in a feed-forward mode. 3 Sniffer probes for millimeter wave stray radiation detection have been installed.

Contribution of ASDEX Upgrade to disruption studies for ITER

International Nuclear Information System (INIS)

Pautasso, G.; Reiter, B.; Giannone, L.; Gruber, O.; Herrmann, A.; Kardaun, O.; Maraschek, M.; Mlynek, A.; Schneider, W.; Zhang, Y.; Khayrutdinov, K.K.; Lukash, V.E.; Nakamura, Y.; Sias, G.; Sugihara, M.

2011-01-01

This paper describes the most recent contributions of ASDEX Upgrade to ITER in the field of disruption studies. (1) The ITER specifications for the halo current magnitude are based on data collected from several tokamaks and summarized in the plot of the toroidal peaking factor versus the maximum halo current fraction. Even if the maximum halo current in ASDEX Upgrade reaches 50% of the plasma current, the duration of this maximum lasts a fraction of a ms. (2) Long-lasting asymmetries of the halo current are rare and do not give rise to a large asymmetric component of the mechanical forces on the machine. Differently from JET, these asymmetries are neither locked nor exhibit a stationary harmonic structure. (3) Recent work on disruption prediction has concentrated on the search for a simple function of the most relevant plasma parameters, which is able to discriminate between the safe and pre-disruption phases of a discharge. For this purpose, the disruptions of the last four years have been classified into groups and then discriminant analysis is used to select the most significant variables and to derive the discriminant function. (4) The attainment of the critical density for the collisional suppression of the runaway electrons seems to be technically and physically possible on our medium size tokamak. The CO 2 interferometer and the AXUV diagnostic provide information on the highly 3D impurity transport process during the whole plasma quench.

Characterization and interpretation of the Edge Snake in between type-I edge localized modes at ASDEX Upgrade

NARCIS (Netherlands)

Sommer, F.; Günter, S.; Kallenbach, A.; Maraschek, M.; Boom, J.E.; Fischer, R.; Hicks, N.; Luhmann, N.C.; Park, H.K.; Reiter, B.; Wenninger, R.; Wolfrum, E.

2011-01-01

A new magnetohydrodynamic instability called the 'Edge Snake', which was found in 2006 at the tokamak ASDEX Upgrade during type-I ELMy H-modes, is investigated. It is located within the separatrix in the region of high temperature and density gradients and has a toroidal mode number of n = 1. The

Flux surface shaping effects on tokamak edge turbulence and flows

Energy Technology Data Exchange (ETDEWEB)

Kendl, A. [Innsbruck Univ., Institut fuer Theoretische Physik, Association EURATOM (Austria); Scott, B.D. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Garching bei Muenchen (Germany)

2004-07-01

The influence of shaping of magnetic flux surfaces in tokamaks on gyro-fluid edge turbulence is studied numerically. Magnetic field shaping in tokamaks is mainly due to elongation, triangularity, shift and the presence of a divertor X-point. A series of tokamak configurations with varying elongation 1 {<=} {kappa} {>=} 2 and triangularity 0 {<=} {delta} {<=} 0.4, and an actual ASDEX Upgrade divertor configuration are obtained with the equilibrium code HELENA and implemented into the gyro-fluid turbulence code GEM. The study finds minimal impact on the zonal flow physics itself, but strong impact on the turbulence and transport. (authors)

Flux surface shaping effects on tokamak edge turbulence and flows

International Nuclear Information System (INIS)

Kendl, A.; Scott, B.D.

2004-01-01

The influence of shaping of magnetic flux surfaces in tokamaks on gyro-fluid edge turbulence is studied numerically. Magnetic field shaping in tokamaks is mainly due to elongation, triangularity, shift and the presence of a divertor X-point. A series of tokamak configurations with varying elongation 1 ≤ κ ≥ 2 and triangularity 0 ≤ δ ≤ 0.4, and an actual ASDEX Upgrade divertor configuration are obtained with the equilibrium code HELENA and implemented into the gyro-fluid turbulence code GEM. The study finds minimal impact on the zonal flow physics itself, but strong impact on the turbulence and transport. (authors)

Experimental and theoretical investigation of density and potential fluctuations in the scrape-off layer of ASDEX

International Nuclear Information System (INIS)

Endler, M.; Giannone, L.; Niedermeyer, H.; Rudyj, A.; Theimer, G.

1993-01-01

In the divertor tokamak ASDEX density and potential fluctuations in the scrape-off layer were investigated with high temporal and spatial resolution by Langmuir probes and an H α diagnostic. Many results of these measurements were reported and are summarized below. Several of these properties of the fluctuations have also been reported from other experiments. (orig.)

The new centrifuge high-speed pellet injector for ASDEX Upgrade

International Nuclear Information System (INIS)

Lang, P.T.; Andelfinger, C.; Beck, W.; Buchelt, E.; Buechl, K.; Cierpka, P.; Kollotzek, H.; Lang, R.S.; Prausner, G.; Soeldner, F.X.; Ulrich, M.; Weber, G.

1993-04-01

We report on the new pellet injection system for refuelling the ASDEX Upgrade tokamak with cubic H 2 or D 2 pellets having alternative side lengths of 1.5, 1.75 and 2.0 mm and optional Ne doping. The system delivers series of about one hundred pellets at a maximum repetition rate of more than 40 Hz. The pellets are accelerated by means of a centrifuge with an optimized straight acceleration arm. This configuration minimizes the compulsive force acting on the pellet during the acceleration process. Since this also minimizes stresses inside the pellet, high velocities - a maximum of 1211 m/s being achieved - are possible without destroying the hydrogen cubes. A special pellet feed-in technique based on a static stop cylinder interrupting the acceleration path successfully reduced the horizontal scattering angle to values of less than ± 4 degrees. Thus, a high efficiency - with more than 90% of the pellets arriving within the acceptance angle - was achieved without using a guide tube. The whole system was found to work very reliably and reproducibly during the whole test operation period, covering about 10 5 pellet shots. The new centrifuge, now integrated into the ASDEX Upgrade setup, has proved to be a reliable unit even for long operation periods thus affording the possibility of quasicontinuous particle refuelling throughout a plasma discharge in ASDEX Upgrade. (orig.)

Observations on the W-transport in the core plasma of JET and ASDEX Upgrade

Czech Academy of Sciences Publication Activity Database

Pütterich, T.; Dux, R.; Neu, R.; Bernert, M.; Beurskens, M.N.A.; Bobkov, V.; Brezinsek, S.; Challis, C.; Coenen, J.W.; Coffey, I.; Czarnecka, A.; Giroud, C.; Jacquet, P.; Joffrin, E.; Kallenbach, A.; Lehnen, M.; Lerche, E.; De La Luna, E.; Marsen, S.; Matthews, G.; Mayoral, M.-L.; McDermott, R.M.; Meigs, A.; Mlynář, Jan; Sertoli, M.; van Rooij, G.

2013-01-01

Roč. 55, č. 12 (2013), s. 124036-124036 ISSN 0741-3335. [European Physical Society Conference on Plasma Physics/40./. Espoo, 01.07.2013-05.07.2013] Institutional support: RVO:61389021 Keywords : tokamak * impurity transport * core plasma * fusion * tungsten * ASDEX Upgrade Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.386, year: 2013 http://iopscience.iop.org/0741-3335/55/12/124036/pdf/0741-3335_55_12_124036.pdf

Latest investigations on fluctuations, ELM filaments and turbulent transport in the SOL of ASDEX Upgrade

Czech Academy of Sciences Publication Activity Database

Müller, H. W.; Adámek, Jiří; Cavazzana, R.; Conway, G.D.; Fuchs, C.; Gunn, J. P.; Herrmann, A.; Horáček, Jan; Ionita, C.; Kallenbach, A.; Kočan, M.; Maraschek, M.; Maszl, C.; Mehlmann, F.; Nold, B.; Peterka, M.; Rohde, V.; Schweinzer, J.; Schrittwieser, R.; Vianello, N.; Wolfrum, E.; Zuin, M.

2011-01-01

Roč. 51, č. 7 (2011), 073023-073023 ISSN 0029-5515 R&D Projects: GA AV ČR KJB100430901 Institutional research plan: CEZ:AV0Z20430508 Keywords : SOL * ASDEX * ELM * tokamak * ball- pen * H-mode Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 4.090, year: 2011 http://iopscience.iop.org/0029-5515/51/7/073023/pdf/0029-5515_51_7_073023.pdf

The ASDEX Upgrade ICRF system: Operational experience and developments

International Nuclear Information System (INIS)

Faugel, H.; Angene, P.; Becker, W.; Braun, F.; Bobkov, Vl.V.; Eckert, B.; Fischer, F.; Hartmann, D.A.; Heilmaier, G.; Kneidl, J.; Noterdaeme, J.-M.; Siegl, G.; Wuersching, E.

2005-01-01

The ICRF system at the ASDEX Upgrade tokamak is in operation since May 1992. Following some modifications of which the major one was the installation of 3 dB couplers it has become a reliable additional heating system. The maximum power coupled into the plasma has been raised up to 7.2 MW (90% of the installed RF power) for short pulses and up to 6.2 MW for pulses several second long with energy of up to 29 MJ. A power of 5 MW is delivered on a regular basis to replace two NBI sources

Micro-NRA and micro-3HIXE with He-3 microbeam on samples exposed in ASDEX Upgrade and Pilot-PSI machines

NARCIS (Netherlands)

Kelemen, M.; Zaloznik, A.; Vavpetic, P.; Pecovnik, M.; Pelicon, P.; Hakola, A.; Lahtinen, A.; Karhunen, J.; Piip, K.; van der Meiden, H. J.; Paris, P.; Laan, M.; Krieger, K.; Oberkofler, M.; Markelj, S.; ASDEX Upgrade team,

2017-01-01

Micro nuclear reaction analysis (micro-NRA) exploiting the nuclear reaction D(He-3,p)He-4 was used for post-mortem analyses of special marker samples, exposed to deuterium plasma inside ASDEX Upgrade (AUG) tokamak and to the deuterium plasma jet in the Pilot-PSI linear plasma gun. Lateral

Turbulent fluctuations and radial transport in the scrape-off layer of the ASDEX tokamak

Energy Technology Data Exchange (ETDEWEB)

Endler, M [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85740 Garching (Germany); Giannone, L. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85740 Garching (Germany); McCormick, K [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85740 Garching (Germany); Niedermeyer, H [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85740 Garching (Germany); Rudyj, A [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85740 Garching (Germany); Theimer, G [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85740 Garching (Germany); Tsois, N [NCSR ` Demokritos` , Athens (Greece); ASDEX Team

1995-04-01

Electrostatic fluctuations have been measured in the scrape-off layer of ASDEX by Langmuir probes and by observation of H{sub {alpha}} light with high poloidal and temporal resolution. It was demonstrated that these fluctuations contribute a significant, if not dominant, fraction of the ``anomalous`` radial particle transport. A model for an instability mechanism specific to the SOL is presented including density, temperature and electric potential fluctuations. From this model mixing length estimates for the radial transport and resulting density and pressure gradients in the SOL are derived and compared to measurements in the mid-plane and in the divertor of ASDEX. In spite of several simplifications in the model a quantitative agreement up to factors of 1-3 and a qualitative agreement for variations of discharge parameters is achieved between the model predictions and the measurements. ((orig.)).

Turbulent fluctuations and radial transport in the scrape-off layer of the ASDEX tokamak

International Nuclear Information System (INIS)

Endler, M.; Giannone, L.; McCormick, K.; Niedermeyer, H.; Rudyj, A.; Theimer, G.; Tsois, N.

1995-01-01

Electrostatic fluctuations have been measured in the scrape-off layer of ASDEX by Langmuir probes and by observation of H α light with high poloidal and temporal resolution. It was demonstrated that these fluctuations contribute a significant, if not dominant, fraction of the ''anomalous'' radial particle transport. A model for an instability mechanism specific to the SOL is presented including density, temperature and electric potential fluctuations. From this model mixing length estimates for the radial transport and resulting density and pressure gradients in the SOL are derived and compared to measurements in the mid-plane and in the divertor of ASDEX. In spite of several simplifications in the model a quantitative agreement up to factors of 1-3 and a qualitative agreement for variations of discharge parameters is achieved between the model predictions and the measurements. ((orig.))

Scintillator based detector for fast-ion losses induced by magnetohydrodynamic instabilities in the ASDEX upgrade tokamak.

Science.gov (United States)

García-Muñoz, M; Fahrbach, H-U; Zohm, H

2009-05-01

A scintillator based detector for fast-ion losses has been designed and installed on the ASDEX upgrade (AUG) tokamak [A. Herrmann and O. Gruber, Fusion Sci. Technol. 44, 569 (2003)]. The detector resolves in time the energy and pitch angle of fast-ion losses induced by magnetohydrodynamics (MHD) fluctuations. The use of a novel scintillator material with a very short decay time and high quantum efficiency allows to identify the MHD fluctuations responsible for the ion losses through Fourier analysis. A Faraday cup (secondary scintillator plate) has been embedded behind the scintillator plate for an absolute calibration of the detector. The detector is mounted on a manipulator to vary its radial position with respect to the plasma. A thermocouple on the inner side of the graphite protection enables the safety search for the most adequate radial position. To align the scintillator light pattern with the light detectors a system composed by a lens and a vacuum-compatible halogen lamp has been allocated within the detector head. In this paper, the design of the scintillator probe, as well as the new technique used to analyze the data through spectrograms will be described. A last section is devoted to discuss the diagnosis prospects of this method for ITER [M. Shimada et al., Nucl. Fusion 47, S1 (2007)].

Intermittent transport across the scrape-off layer: latest results from ASDEX Upgrade

Czech Academy of Sciences Publication Activity Database

Kočan, M.; Müller, H.W.; Nold, B.; Lunt, T.; Adámek, Jiří; Allan, S.Y.; Bernert, M.; Conway, G.D.; de Marné, P.; Eich, T.; Elmore, S.; Gennrich, F.P.; Herrmann, A.; Horáček, Jan; Huang, Z.; Kallenbach, A.; Komm, Michael; Maraschek, M.; Mehlmann, F.; Müller, S.; Ribeiro, T.T.; Rohde, V.; Schrittwieser, R.; Scott, B.; Stroth, U.; Suttrop, W.; Wolfrum, E.

2013-01-01

Roč. 53, č. 7 (2013), 073047-073047 ISSN 0029-5515 R&D Projects: GA MŠk(CZ) LG11018; GA ČR(CZ) GAP205/12/2327; GA ČR GA202/09/1467 Institutional support: RVO:61389021 Keywords : ASDEX Upgrade scrape-off layer * plasma * tokamak * edge-localized mode (ELM) Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.243, year: 2013 http://iopscience.iop.org/0029-5515/53/7/073047/pdf/0029-5515_53_7_073047.pdf

H-mode confinement properties close to the power threshold in ASDEX Upgrade

International Nuclear Information System (INIS)

Ryter, F; Fuchs, J; Schneider, W; Sips, A; Staebler, A; Stober, J

2008-01-01

Confinement properties close to the H-mode power threshold are studied in the ASDEX Upgrade tokamak. The results show that good confinement can be obtained close to the threshold with Type-I ELMs. The existence of Type-I ELMs does not necessarily require the heating power to be higher than the H-Mode power threshold, but it requires collisionality to be low enough. At higher collisionality Type-III ELMs replace the Type-I ELMs and confinement time is reduced by about 20%

Empiricial scaling of inter-ELM power widths in ASDEX Upgrade and JET

Energy Technology Data Exchange (ETDEWEB)

Eich, T., E-mail: teich@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, EURATOM Association, Boltzmannstr. 2, D-85748 Garching (Germany); Sieglin, B.; Scarabosio, A.; Herrmann, A.; Kallenbach, A. [Max-Planck-Institut für Plasmaphysik, EURATOM Association, Boltzmannstr. 2, D-85748 Garching (Germany); Matthews, G.F. [EURATOM/CCFE Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Jachmich, S. [ERM-KMS, EURATOM Association, Brussels (Belgium); Brezinsek, S.; Rack, M. [IEK-4, Forschungszentrum Jülich, EURATOM Association (Germany); Goldston, R.J. [Princeton Plasma Physics Labaratory, Princeton, NJ 08543 (United States)

2013-07-15

The SOL power decay length (λ{sub q}) deduced from analysis of fully attached divertor heat load profiles from two tokamaks, JET and ASDEX Upgrade with carbon plasma facing components, are presented. Interpretation of the target heat load profiles is performed by using a 1D-fit function which disentangles the upstream λ{sub q} and an effective diffusion in the divertor (S), the latter essentially acting as a power spreading parameter in the divertor volume. It is shown that the so called integral decay length λ{sub int} is approximately given by λ{sub int}≈λ{sub q}+1.64×S. An empirical scaling reveals parametric dependency λ{sub q}/mm≃0.9·B{sub T}{sup -0.7}q{sub cyl}{sup 1.2}P{sub SOL}{sup 0}R{sub geo}{sup 0} for type-I ELMy H-modes. Extrapolation to ITER gives λ{sub q}≃1 mm. Recent measurements in JET-ILW and from ASDEX Upgrade full-W confirm the results. It is shown that a regression for the divertor power spreading parameter S is not yet possible due to the large effect of different divertor geometries of JET and ASDEX Upgrade Divertor-I and Divertor-IIb.

Divertor radiation in the ASDEX upgrade tokamak

Energy Technology Data Exchange (ETDEWEB)

Sehmer, Till; Bernert, Matthias; Koll, Juergen; Meister, Hans; Wischmeier, Marco; Fantz, Ursel [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstrasse 2, 85748 Garching (Germany); Reimold, Felix [Forschungszentrum Juelich GmbH, Institut fuer Energie- und Klimaforschung - Plasmaphysik, 52425 Juelich (Germany); Collaboration: The ASDEX Upgrade Team

2016-07-01

To reduce in ITER the expected power flux density onto the divertor target, the plasma-wall interaction in the divertor needs to be strongly reduced. The fundamental path to achieve this is using radiation from seeded impurities, whereas the localization of this radiation (e.g. inside/outside confined region), which could have an impact onto the power balance, is a key challenge. The absolute radiated power distribution can be measured by foil bolometers. To study at the ASDEX Upgrade tungsten divertor the localization and quantification of radiation, the respective line of sight density of the bolometers has been improved by two additional cameras. The divertor radiation enhanced by nitrogen (N{sub 2}) seeding has been investigated, using variations of (1) the external heating power or (2) the N{sub 2} seeding rate. While in both cases the inner divertor stays fully detached, measurements indicate that the region of dominant radiation moves from the inner divertor through the X-Point into the confined region. In the outer divertor however, the measurements indicate either an immediate upwards shift or a continuous movement of the radiation away from the target, depending on experimental conditions.

Velocity space resolved absolute measurement of fast ion losses induced by a tearing mode in the ASDEX Upgrade tokamak

Science.gov (United States)

Galdon-Quiroga, J.; Garcia-Munoz, M.; Sanchis-Sanchez, L.; Mantsinen, M.; Fietz, S.; Igochine, V.; Maraschek, M.; Rodriguez-Ramos, M.; Sieglin, B.; Snicker, A.; Tardini, G.; Vezinet, D.; Weiland, M.; Eriksson, L. G.; The ASDEX Upgrade Team; The EUROfusion MST1 Team

2018-03-01

Absolute flux of fast ion losses induced by tearing modes have been measured by means of fast ion loss detectors (FILD) for the first time in RF heated plasmas in the ASDEX Upgrade tokamak. Up to 30 MW m-2 of fast ion losses are measured by FILD at 5 cm from the separatrix, consistent with infra-red camera measurements, with energies in the range of 250-500 keV and pitch angles corresponding to large trapped orbits. A resonant interaction between the fast ions in the high energy tail of the ICRF distribution and a m/n  =  5/4 tearing mode leads to enhanced fast ion losses. Around 9.3 +/- 0.7 % of the fast ion losses are found to be coherent with the mode and scale linearly with its amplitude, indicating the convective nature of the transport mechanism. Simulations have been carried out to estimate the contribution of the prompt losses. A good agreement is found between the simulated and the measured velocity space of the losses. The velocity space resonances that may be responsible for the enhanced fast ion losses are identified.

Real-time signal communication between diagnostic and control in ASDEX Upgrade

International Nuclear Information System (INIS)

Treutterer, Wolfgang; Neu, Gregor; Raupp, Gerhard; Zehetbauer, Thomas; Zasche, Dieter; Lueddecke, Klaus; Cole, Richard

2010-01-01

The ASDEX Upgrade tokamak experiment is equipped with a versatile discharge monitoring and control system. It allows to develop and use advanced control algorithms to investigate plasma physics under well-defined conditions with the objective of optimising plasma performance. The achievable quality depends on the accuracy with which the plasma state can be reconstructed from measurements under real-time conditions. Today's advanced algorithms need physics quantities - scalar entities as well as profiles. These are obtained processing huge numbers of raw measurements with complex diagnostic algorithms. Adequate network communication for the resulting signals is crucial to satisfy real-time requirements, especially when several diagnostic systems cooperate in a feedback control loop. Support for the technology of choice, however, is not easily available for all of the diverse, highly specialised diagnostic systems. We give an overview about the methods that have been explored at ASDEX Upgrade for real-time signal transfer. In particular, we investigated reflective shared memory and Ethernet technologies. Our solution strives to combine their strengths. For fast communication on dedicated computing nodes, reflective shared memory is used. For the majority of diagnostic systems producing large data blocks at moderate rates, Ethernet connections with UDP protocol are employed. Following ASDEX Upgrade's framework concept, a software layer hides the networks used from both diagnostic and control applications.

Real-time signal communication between diagnostic and control in ASDEX Upgrade

Energy Technology Data Exchange (ETDEWEB)

Treutterer, Wolfgang, E-mail: Wolfgang.Treutterer@ipp.mpg.d [Max-Planck Institut fuer Plasmaphysik, Garching, EURATOM Association (Germany); Neu, Gregor; Raupp, Gerhard; Zehetbauer, Thomas; Zasche, Dieter [Max-Planck Institut fuer Plasmaphysik, Garching, EURATOM Association (Germany); Lueddecke, Klaus; Cole, Richard [Unlimited Computer Systems, Iffeldorf (Germany)

2010-07-15

The ASDEX Upgrade tokamak experiment is equipped with a versatile discharge monitoring and control system. It allows to develop and use advanced control algorithms to investigate plasma physics under well-defined conditions with the objective of optimising plasma performance. The achievable quality depends on the accuracy with which the plasma state can be reconstructed from measurements under real-time conditions. Today's advanced algorithms need physics quantities - scalar entities as well as profiles. These are obtained processing huge numbers of raw measurements with complex diagnostic algorithms. Adequate network communication for the resulting signals is crucial to satisfy real-time requirements, especially when several diagnostic systems cooperate in a feedback control loop. Support for the technology of choice, however, is not easily available for all of the diverse, highly specialised diagnostic systems. We give an overview about the methods that have been explored at ASDEX Upgrade for real-time signal transfer. In particular, we investigated reflective shared memory and Ethernet technologies. Our solution strives to combine their strengths. For fast communication on dedicated computing nodes, reflective shared memory is used. For the majority of diagnostic systems producing large data blocks at moderate rates, Ethernet connections with UDP protocol are employed. Following ASDEX Upgrade's framework concept, a software layer hides the networks used from both diagnostic and control applications.

High speed cine film studies of plasma behaviour and plasma surface interactions in tokamaks

International Nuclear Information System (INIS)

Goodall, D.H.J.

1982-01-01

High speed cine photography is a useful diagnostic aid for studying plasma behaviour and plasma surface interactions. Several workers have filmed discharges in tokamaks including ASDEX, DITE, DIVA, ISX, JFT2, TFR and PLT. These films are discussed and examples given of the observed phenomena which include plasma limiter interactions, diverted discharges, disruptions, magnetic islands and moving glowing objects often known as 'UFOs'. Examples of plasma structures in ASDEX and DITE not previously published are also given. The paper also reports experiments in DITE to determine the origin of UFOs. (orig.)

Alfven wave heating in ASDEX

International Nuclear Information System (INIS)

Besson, G.; Borg, G.G.; Lister, J.B.; Marmillod, Ph.; Braun, F.; Murphy, A.B.; Noterdaeme, J.M.; Ryter, F.; Wesner, F.

1990-01-01

An experiment has been completed on ASDEX to study the response of the plasma to Alfven wave heating (AWH). Antenna excitation was provided by the old TCA rf generator with an output power capability of 500 kW. Two poloidal loop antennas were installed at the east and west ends of the tokamak allowing either N=1 or N=2 phasings. Since the largest antenna coupling to the Alfven resonance is provided by the m=1 surface wave, the antenna consisted only of a single element on the low field side, whereas in TCA the antennas are located on the top and the bottom of the torus. The antenna elements consisted of 2 parallel bars of inductance 730 nH and, as in TCA, were left unshielded. A typical antenna circulating current of 2 kA peak at 1.80 MHz was provided for the experiments. (author) 3 refs., 4 figs

Zeff from spectroscopic bremsstrahlung measurements at ASDEX Upgrade and JET

International Nuclear Information System (INIS)

Meister, H.; Fischer, R.; Horton, L.D.; Maggi, C.F.; Nishijima, D.; Giroud, C.; Zastrow, K.-D.; Zaniol, B.

2004-01-01

The effective ionic charge Z eff is a means to assess the impurity content of a fusion plasma. It can be derived from measurements of bremsstrahlung intensity. These have been extended at ASDEX Upgrade by the usage of the sight lines for the charge exchange recombination diagnostic. Together with a previously installed sight line array, it is now possible to routinely determine the bremsstrahlung intensity over the whole minor radius purely from spectroscopic measurements. In a tokamak where the plasma facing components are made up of various materials, this is necessary to check if measurements are contaminated by line radiation. The bremsstrahlung background of the respective spectra is determined using Bayesian probability theory, giving consistent and improved error statistics. Using the information for electron temperature and density profiles, the Z eff profile is determined by an integrated method. The same approach to assess the Z eff profile has been demonstrated to be successful also at the JET tokamak

Measurements and modelling of electrostatic fluctuations in the scrape-off layer of ASDEX

Energy Technology Data Exchange (ETDEWEB)

Endler, M; Niedermeyer, H; Giannone, L.; Holzhauer, E; Rudyj, A; Theimer, G; Tsois, N [Association Euratom-Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); ASDEX Team

1995-11-01

In the edge plasma of the ASDEX tokamak, electrostatic fluctuations were observed with Langmuir probes and in H{sub {alpha}} light with high poloidal and temporal resolution. These fluctuations contribute a significant fraction to the `anomalous` radial particle transport in the scrape-off layer (SOL). The basic properties and the dependence of the fluctuations parameters on the discharge conditions are documented. A model for an instability mechanism specific to the SOL is introduced and the experimentally observed fluctuation parameters are compared with the predictions of the linearized version of this model. For plasma temperatures above {approx} 10eV in the SOL the observed parameter dependences of the fluctuations are well reproduced by the model. By mixing length arguments the radial transport and the resulting density and pressure gradients in the SOL are estimated from the model. Their dependence on plasma temperature and density qualitatively agrees with the behaviour observed in ohmic discharges on ASDEX. (author). 54 refs, 25 figs.

Measurements and modelling of electrostatic fluctuations in the scrape-off layer of ASDEX

International Nuclear Information System (INIS)

Endler, M.; Niedermeyer, H.; Giannone, L.; Holzhauer, E.; Rudyj, A.; Theimer, G.; Tsois, N.

1995-01-01

In the edge plasma of the ASDEX tokamak, electrostatic fluctuations were observed with Langmuir probes and in H α light with high poloidal and temporal resolution. These fluctuations contribute a significant fraction to the 'anomalous' radial particle transport in the scrape-off layer (SOL). The basic properties and the dependence of the fluctuations parameters on the discharge conditions are documented. A model for an instability mechanism specific to the SOL is introduced and the experimentally observed fluctuation parameters are compared with the predictions of the linearized version of this model. For plasma temperatures above ∼ 10eV in the SOL the observed parameter dependences of the fluctuations are well reproduced by the model. By mixing length arguments the radial transport and the resulting density and pressure gradients in the SOL are estimated from the model. Their dependence on plasma temperature and density qualitatively agrees with the behaviour observed in ohmic discharges on ASDEX. (author). 54 refs, 25 figs

Analysis of the ion energy transport in ohmic discharges in the ASDEX tokamak

International Nuclear Information System (INIS)

Simmet, E.E.; Fahrbach, H.U.; Herrmann, W.; Stroth, U.

1996-10-01

An analysis of the local ion energy transport is performed for more than one hundred well documented ohmic ASDEX discharges. These are characterized by three different confinement regimes: the linear ohmic confinement (LOC), the saturated ohmic confinement (SOC) and the improved ohmic confinement (IOC). All three are covered by this study. To identify the most important local transport mechanism of the ion heat, the ion power balance equation is analyzed. Two methods are used: straightforward calculation with experimental data only, and a comparison of measured and calculated profiles of the ion temperature and the ion heat conductivity, respectively. A discussion of the power balance shows that conductive losses dominate the ion energy transport in all ohmic discharges of ASDEX. Only inside the q=1-surface losses due to sawtooth activity play a role, while at the edge convective fluxes and CX-losses influence the ion energy transport. Both methods lead to the result that both the ion temperature and the ion heat conductivity are consistent with predictions of the neoclassical theory. Enhanced heat losses as suggested by theories eg. on the basis of η i modes can be excluded. (orig.)

Disruption studies on ASDEX upgrade

International Nuclear Information System (INIS)

Pautasso, G.; Egorov, S.; Finken, K.H.

2003-01-01

Disruptions generate large thermal and mechanical stresses on the tokamak components and are occasionally responsible for damages to the machine. For a future reactor disruptions have a significant impact on the design since all loading conditions must be analyzed in accordance with stricter design criteria (due to safety or difficult maintenance). Therefore the uncertainties affecting the predicted stresses must be reduced as much as possible with a more comprehensive set of measurements and analyses in this generation of experimental machines, and avoidance/predictive methods must be developed further. Disruption studies on ASDEX Upgrade are focused on these subjects, namely on: (1) understanding the physical mechanisms leading to this phenomenon in order to learn to avoid it or to predict its occurrence and to mitigate its effects; (2) analyzing the effects of disruptions on the machine to determine the functional dependence of the thermal and mechanical loads upon the discharge parameters. This allows, firstly, to dimension or reinforce the machine components to withstand these loads and, secondly, to extrapolate them to tokamaks still in the design phase; (3) learning to mitigate the consequence of disruptions, i.e. thermal loads, mechanical forces and runaways with injection of impurity pellets or gas. This paper is focused on most recent results concerning points, i.e. on the analysis of the degree of asymmetry of the forces and on the use of impurity puff for mitigation

Characterization and interpretation of the Edge Snake in between type-I edge localized modes at ASDEX Upgrade

Energy Technology Data Exchange (ETDEWEB)

Sommer, F; Guenter, S; Kallenbach, A; Maraschek, M; Boom, J; Fischer, R; Hicks, N; Reiter, B; Wolfrum, E [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstrasse 2, D-85748 Garching, EURATOM Association (Germany); Luhmann, N C Jr [University of California at Davis, Davis, CA 95616 (United States); Park, H K [POSTECH, Pahang, Gyeongbuk 790-784 (Korea, Republic of); Wenninger, R, E-mail: fabian.sommer@ipp.mpg.de [Universitaetssternwarte der Ludwig-Maximilians-Universitaet, D-81679 Muenchen (Germany)

2011-08-15

A new magnetohydrodynamic instability called the 'Edge Snake', which was found in 2006 at the tokamak ASDEX Upgrade during type-I ELMy H-modes, is investigated. It is located within the separatrix in the region of high temperature and density gradients and has a toroidal mode number of n = 1. The Edge Snake consists of a radially and poloidally strongly localized current wire, in which the temperature and density profiles flatten. This significant reduction in pressure gradient leads to a reduction in the neoclassical Bootstrap current and can plausibly explain the drive of the instability. The experimental observations point towards a magnetic island with a defect current inside the O-point of the island. The Edge Snake is compared with similar instabilities at JET, DIII-D and ASDEX Upgrade.

Fast-ion transport and neutral beam current drive in ASDEX upgrade

DEFF Research Database (Denmark)

Geiger, B.; Weiland, M.; Jacobsen, Asger Schou

2015-01-01

The neutral beam current drive efficiency has been investigated in the ASDEX Upgrade tokamak by replacing on-axis neutral beams with tangential off-axis beams. A clear modification of the radial fast-ion profiles is observed with a fast-ion D-alpha diagnostic that measures centrally peaked profiles...... during on-axis injection and outwards shifted profiles during off-axis injection. Due to this change of the fast-ion population, a clear modification of the plasma current profile is predicted but not observed by a motional Stark effect diagnostic. The fast-ion transport caused by MHD activity has been...

Comparison between 2D turbulence model ESEL and experimental data from AUG and COMPASS tokamaks

DEFF Research Database (Denmark)

Ondac, Peter; Horacek, Jan; Seidl, Jakub

2015-01-01

In this article we have used the 2D fluid turbulence numerical model, ESEL, to simulate turbulent transport in edge tokamak plasma. Basic plasma parameters from the ASDEX Upgrade and COMPASS tokamaks are used as input for the model, and the output is compared with experimental observations obtain...... for an extension of the ESEL model from 2D to 3D to fully resolve the parallel dynamics, and the coupling from the plasma to the sheath....

ICRF hydrogen minority heating in the boronized ASDEX tokamak

International Nuclear Information System (INIS)

Ryter, F.; Braun, F.; Hofmeister, F.; Noterdaeme, J.M.; Steuer, K.H.; Wesner, F.

1990-01-01

Since the divertor of ASDEX has been modified (1986-87) the hydrogen concentration in deuterium plasmas could not be reduced below 10%, although the machine was operated for long periods of time with deuterium injection. This is probably due to desorption in the divertor as indicated by the increasing H-concentration during a deuterium injection pulse. As a consequence for H-minority heating in deuterium, the maximum power into ohmic plasmas without causing a disruption was limited to few hundred kW. A partial solution was ICRH in combination with deuterium injection which allowed us to apply up to 1.5 MW ICRH to the plasma. The beneficial role of the injection is attributed to an improved ICRH absorption and to the higher energy flux and temperature in the divertor. During the last ICRH campaign we operated mainly in helium plasmas for a lower hydrogen concentration and the vessel was boronised. The H-concentration is measured routinely by a mass spectrometer in the divertor chamber. (orig./AH)

Interpretation of fast measurements of plasma potential, temperature and density in SOL of ASDEX Upgrade

Czech Academy of Sciences Publication Activity Database

Horáček, Jan; Adámek, Jiří; Müller, H. W.; Seidl, J.; Nielsen, A.H.; Rohde, V.; Mehlmann, F.; Ionita, C.; Havlíčková, Eva

2010-01-01

Roč. 50, č. 10 (2010), s. 105001-105001 ISSN 0029-5515 R&D Projects: GA AV ČR KJB100430901; GA ČR GA202/09/1467; GA MŠk 7G09042 Institutional research plan: CEZ:AV0Z20430508 Keywords : Ball- pen probe * Langmuir probe * tokamak * plasma * SOL * turbulence * blob * interchange instability * ASDEX Upgrade Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.303, year: 2010 http://iopscience.iop.org/0029-5515/50/10/105001

ICRF heating analysis on ASDEX plasmas

International Nuclear Information System (INIS)

Itoh, Sanae; Itoh, Kimitaka; Fukuyama, Atsushi; Morishita, Takayuki; Steinmetz, K.; Noterdaeme, J.-M.

1988-01-01

ICRF (ion cyclotron range of frequencies) waves heating in an ASDEX tokamak are analyzed. The excitation, propagation and absorption are studied by using a global wave code. This analysis is combined with a Fokker-Planck code. The waveform in the plasma, the loading resistance and the reactance of the antenna are calculated for both the minority ion heating and the second harmonic resonance heating. Attention is given to the change of the antenna loading associated with the L/H transition. Optimum conditions for the loading are discussed. In the minority heating case, the tail generation and thermalization are analyzed. Spatial profiles of the tail-ion temperature and the power transferred to the bulk electrons and ions are obtained. Central as well as off-central heating cases are investigated. The effect of the reactive electric field is discussed in connection with rf losses and impurity production. (author)

First operations with the new Collective Thomson Scattering diagnostic on the Frascati Tokamak Upgrade device

DEFF Research Database (Denmark)

Bin, W.; Bruschi, A.; D'Arcangelo, O.

2015-01-01

Anomalous emissions were found over the last few years in spectra of Collective Thomson Scattering (CTS) diagnostics in tokamak devices such as TEXTOR, ASDEX and FTU, in addition to real CTS signals. The signal frequency, down-shifted with respect to the probing one, suggested a possible origin i...

Comparison of runaway electron generation parameters in small, medium-sized and large tokamaks—A survey of experiments in COMPASS, TCV, ASDEX-Upgrade and JET

Science.gov (United States)

Plyusnin, V. V.; Reux, C.; Kiptily, V. G.; Pautasso, G.; Decker, J.; Papp, G.; Kallenbach, A.; Weinzettl, V.; Mlynar, J.; Coda, S.; Riccardo, V.; Lomas, P.; Jachmich, S.; Shevelev, A. E.; Alper, B.; Khilkevitch, E.; Martin, Y.; Dux, R.; Fuchs, C.; Duval, B.; Brix, M.; Tardini, G.; Maraschek, M.; Treutterer, W.; Giannone, L.; Mlynek, A.; Ficker, O.; Martin, P.; Gerasimov, S.; Potzel, S.; Paprok, R.; McCarthy, P. J.; Imrisek, M.; Boboc, A.; Lackner, K.; Fernandes, A.; Havlicek, J.; Giacomelli, L.; Vlainic, M.; Nocente, M.; Kruezi, U.; COMPASS Team; TCV Team; ASDEX-Upgrade Team; EUROFusion MST1 Team; contributors, JET

2018-01-01

This paper presents a survey of the experiments on runaway electrons (RE) carried out recently in frames of EUROFusion Consortium in different tokamaks: COMPASS, ASDEX-Upgrade, TCV and JET. Massive gas injection (MGI) has been used in different scenarios for RE generation in small and medium-sized tokamaks to elaborate the most efficient and reliable ones for future RE experiments. New data on RE generated at disruptions in COMPASS and ASDEX-Upgrade was collected and added to the JET database. Different accessible parameters of disruptions, such as current quench rate, conversion rate of plasma current into runaways, etc have been analysed for each tokamak and compared to JET data. It was shown, that tokamaks with larger geometrical sizes provide the wider limits for spatial and temporal variation of plasma parameters during disruptions, thus extending the parameter space for RE generation. The second part of experiments was dedicated to study of RE generation in stationary discharges in COMPASS, TCV and JET. Injection of Ne/Ar have been used to mock-up the JET MGI runaway suppression experiments. Secondary RE avalanching was identified and quantified for the first time in the TCV tokamak in RE generating discharges after massive Ne injection. Simulations of the primary RE generation and secondary avalanching dynamics in stationary discharges has demonstrated that RE current fraction created via avalanching could achieve up to 70-75% of the total plasma current in TCV. Relaxations which are reminiscent the phenomena associated to the kinetic instability driven by RE have been detected in RE discharges in TCV. Macroscopic parameters of RE dominating discharges in TCV before and after onset of the instability fit well to the empirical instability criterion, which was established in the early tokamaks and examined by results of recent numerical simulations.

Transforming the ASDEX Upgrade discharge control system to a general-purpose plasma control platform

International Nuclear Information System (INIS)

Treutterer, Wolfgang; Cole, Richard; Gräter, Alexander; Lüddecke, Klaus; Neu, Gregor; Rapson, Christopher; Raupp, Gerhard; Zasche, Dieter; Zehetbauer, Thomas

2015-01-01

Highlights: • Control framework split in core and custom part. • Core framework deployable in other fusion device environments. • Adaptible through customizable modules, plug-in support and generic interfaces. - Abstract: The ASDEX Upgrade Discharge Control System DCS is a modern and mature product, originally designed to regulate and supervise ASDEX Upgrade Tokamak plasma operation. In its core DCS is based on a generic, versatile real-time software framework with a plugin architecture that allows to easily combine, modify and extend control function modules in order to tailor the system to required features and let it continuously evolve with the progress of an experimental fusion device. Due to these properties other fusion experiments like the WEST project have expressed interest in adopting DCS. For this purpose, essential parts of DCS must be unpinned from the ASDEX Upgrade environment by exposure or introduction of generalised interfaces. Re-organisation of DCS modules allows distinguishing between intrinsic framework core functions and device-specific applications. In particular, DCS must be prepared for deployment in different system environments with their own realisations for user interface, pulse schedule preparation, parameter server, time and event distribution, diagnostic and actuator systems, network communication and data archiving. The article explains the principles of the revised DCS structure, derives the necessary interface definitions and describes major steps to achieve the separation between general-purpose framework and fusion device specific components.

Transforming the ASDEX Upgrade discharge control system to a general-purpose plasma control platform

Energy Technology Data Exchange (ETDEWEB)

Treutterer, Wolfgang, E-mail: Wolfgang.Treutterer@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Cole, Richard [Unlimited Computer Systems, Seeshaupter Str. 15, 82393 Iffeldorf (Germany); Gräter, Alexander [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Lüddecke, Klaus [Unlimited Computer Systems, Seeshaupter Str. 15, 82393 Iffeldorf (Germany); Neu, Gregor; Rapson, Christopher; Raupp, Gerhard; Zasche, Dieter; Zehetbauer, Thomas [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany)

2015-10-15

Highlights: • Control framework split in core and custom part. • Core framework deployable in other fusion device environments. • Adaptible through customizable modules, plug-in support and generic interfaces. - Abstract: The ASDEX Upgrade Discharge Control System DCS is a modern and mature product, originally designed to regulate and supervise ASDEX Upgrade Tokamak plasma operation. In its core DCS is based on a generic, versatile real-time software framework with a plugin architecture that allows to easily combine, modify and extend control function modules in order to tailor the system to required features and let it continuously evolve with the progress of an experimental fusion device. Due to these properties other fusion experiments like the WEST project have expressed interest in adopting DCS. For this purpose, essential parts of DCS must be unpinned from the ASDEX Upgrade environment by exposure or introduction of generalised interfaces. Re-organisation of DCS modules allows distinguishing between intrinsic framework core functions and device-specific applications. In particular, DCS must be prepared for deployment in different system environments with their own realisations for user interface, pulse schedule preparation, parameter server, time and event distribution, diagnostic and actuator systems, network communication and data archiving. The article explains the principles of the revised DCS structure, derives the necessary interface definitions and describes major steps to achieve the separation between general-purpose framework and fusion device specific components.

Plasma behavior with molecular beam injection in the HL-1m tokamak

International Nuclear Information System (INIS)

Yao Lianghua; Tang Nianyi; Cui Zhengying; Xu Deming; Deng Zhongchao; Ding Xuantong; Luo Junlin; Dong Jiafu; Guo Gancheng; Yang Shikun; Cui Chenghe; Xiao Zhenggui; Liu Dequan; Chen Xiaoping; Yan Longwen; Yan Donghai; Wang Enyao; Deng Xiwen

1999-01-01

The authors report effect of the new fueling method of high speed molecular beam injection on Tokamak confinement improvement. The present method is an improvement of conventional gas puffing, with performance comparable to the small pellet injection in HL-1M and also to the slow pellet in ASDEX. The fact that a shallower fueling can lead to similar confinement improvement as a deep one suggests that there may exist a critical position in a Tokamak plasma such that any kind of fueling will have a better confinement as long as it can give rise to density peaking at the critical position

Pellet-plasma interaction studies at ASDEX Upgrade

International Nuclear Information System (INIS)

Kocsis, G.; Belonohy, E.; Gal, K.; Kalvin, S.; Veres, G.; Lang, P.T.

2005-01-01

Pellets produced from cryogenic hydrogen isotopes are used for efficient plasma refueling. Beyond this 'classical' application, pellets pacing the frequency of Edge Localized Modes (ELMs) turned out to be a suitable technique to mitigate the power load on plasma facing components. Although pellet pacing is already integrated in the toolkit for plasma control, its underlying physics is still poorly understood. For investigations aiming to resolve where and how an ELM is triggered by the pellet imposed local perturbation precise knowledge of the ablation profile is required. This renewed and even boosted the interest to understand the interaction of pellets with the hot ambient plasma. Both the investigation of the pellet ablation and also its impact on the target plasma were highlighted. Dedicated investigations require precise information both in the space and time domain. E. g. it is necessary to determine the localization of the pellet at the moment it triggers the ELM as well as the actual imposed 3D distribution of the pellet cloud and its mass deposition profile. By these means, a spatial distribution can be mapped out for a local perturbation of the plasma sufficient to release ELMs. High resolution ablation profile and pellet path measurements at different pellet parameters (mass and velocity) could also help to understand the mechanism of the ELM triggering. Recently pellet-plasma interaction is intensively investigated both experimentally at ASDEX Upgrade tokamak and theoretically based on the obtained experimental data. To gain detailed information an observation system was developed at ASDEX Upgrade consisting of digital cameras that detect the pellet cloud distribution and photo diodes that measure the time evolution of the light emission. The great variety of possible combinations of different images, timings and wavelength selections makes the detection sophisticated. Combination of triggered fast camera images and photo diode signals also enables us

Overview of progress in European medium sized tokamaks towards an integrated plasma-edge/wall solution

NARCIS (Netherlands)

Meyer, H.; Eich, T.; Beurskens, M.N.A.; Coda, S.; Hakola, A.; Martin, P.; Adamek, J.; Agostini, M.; Aguiam, D.; Ahn, J.; Aho-Mantila, L.; Akers, R.; Albanese, R.; Aledda, R.; Alessi, E.; Allan, S.; Alves, D.; Ambrosino, R.; Amicucci, L.; Anand, H.; Anastassiou, G.; Andrèbe, Y.; Angioni, C.; Apruzzese, G.; Ariola, M.; Arnichand, H.; Arter, W.; Baciero, A.; Barnes, M.; Barrera, L.; Behn, R.; Bencze, A.; Bernardo, J.; Bernert, M.; Bettini, P.; Bilková, P.; Bin, W.; Birkenmeier, G.; Bizarro, J. P.S.; Blanchard, P.; Blanken, T.; Bluteau, M.; Bobkov, V.; Bogar, O.; Böhm, P.; Bolzonella, T.; Boncagni, L.; Botrugno, A.; Bottereau, C.; Bouquey, F.; Bourdelle, C.; Brémond, S.; Brezinsek, S.; Brida, D.; Brochard, F.; Buchanan, J.; Bufferand, H.; Buratti, P.; Cahyna, P.; Calabrò, G.; Camenen, Y.; Caniello, R.; Cannas, B.; Canton, A.; Cardinali, A.; Carnevale, D.; Carr, M.; Carralero, D.; Carvalho, P.; Casali, L.; Castaldo, C.; Castejón, F.; Castro, R.; Causa, F.; Cavazzana, R.; Cavedon, M.; Cecconello, M.; Ceccuzzi, S.; Cesario, R.; Challis, C.D.; Chapman, I.T.; Chapman, S.; Chernyshova, M.; Choi, D.; Cianfarani, C.; Ciraolo, G.; Citrin, J.; Clairet, F.; Classen, I.; Coelho, R.; Coenen, J. W.; Colas, L.; Conway, G.; Corre, Y.; Costea, S.; Crisanti, F.; Cruz, N.; Cseh, G.; Czarnecka, A.; D'Arcangelo, O.; De Angeli, M.; De Masi, G.; De Temmerman, G.; De Tommasi, G.; Decker, J.; Delogu, R. S.; Dendy, R.; Denner, P.; Di Troia, C.; Dimitrova, M.; D'Inca, R.; Dorić, V.; Douai, D.; Drenik, A.; Dudson, B.; Dunai, D.; Dunne, M.; Duval, B. P.; Easy, L.; Elmore, S.; Erdös, B.; Esposito, B.; Fable, E.; Faitsch, M.; Fanni, A.; Fedorczak, N.; Felici, F.; Ferreira, J.; Février, O.; Ficker, O.; Fietz, S.; Figini, L.; Figueiredo, A.; Fil, A.; Fishpool, G.; Fitzgerald, M.; Fontana, M.; Ford, O.; Frassinetti, L.; Fridström, R.; Frigione, D.; Fuchert, G.; Fuchs, C.; Furno Palumbo, M.; Futatani, S.; Gabellieri, L.; Gałazka, K.; Galdon-Quiroga, J.; Galeani, S.; Gallart, D.; Gallo, A.; Galperti, C.; Gao, Y.; Garavaglia, S.; Garcia, J.; Garcia-Carrasco, A.; Garcia-Lopez, J.; Garcia-Munoz, M.; Gardarein, J. L.; Garzotti, L.; Gaspar, J.; Gauthier, E.; Geelen, P.; Geiger, B.; Ghendrih, P.; Ghezzi, F.; Giacomelli, L.; Giannone, L.; Giovannozzi, E.; Giroud, C.; Gleason González, C.; Gobbin, M.; Goodman, T. P.; Gorini, G.; Gospodarczyk, M.; Granucci, G.; Gruber, M.; Gude, A.; Guimarais, L.; Guirlet, R.; Gunn, J.; Hacek, P.; Hacquin, S.; Hall, S.; Ham, C.; Happel, T.; Harrison, J.; Harting, D.; Hauer, V.; Havlickova, E.; Hellsten, T.; Helou, W.; Henderson, S.; Hennequin, P.; Heyn, M.; Hnat, B.; Hölzl, M.; Hogeweij, D.; Honoré, C.; Hopf, C.; Horáček, J.; Hornung, G.; Horváth, L.; Huang, Z.; Huber, A.; Igitkhanov, J.; Igochine, V.; Imrisek, M.; Innocente, P.; Ionita-Schrittwieser, C.; Isliker, H.; Ivanova-Stanik, I.; Jacobsen, A. S.; Jacquet, P.; Jakubowski, M.; Jardin, A.; Jaulmes, F.; Jenko, F.; Jensen, T.; Jeppe Miki Busk, O.; Jessen, M.; Joffrin, E.; Jones, O.; Jonsson, T.; Kallenbach, A.; Kallinikos, N.; Kálvin, S.; Kappatou, A.; Karhunen, J.; Karpushov, A.; Kasilov, S.; Kasprowicz, G.; Kendl, A.; Kernbichler, W.; Kim, D.; Kirk, A.; Kjer, S.; Klimek, I.; Kocsis, G.; Kogut, D.; Komm, M.; Korsholm, S. B.; Koslowski, H. R.; Koubiti, M.; Kovacic, J.; Kovarik, K.; Krawczyk, N.; Krbec, J.; Krieger, K.; Krivska, A.; Kube, R.; Kudlacek, O.; Kurki-Suonio, T.; Labit, B.; Laggner, F. M.; Laguardia, L.; Lahtinen, A.; Lalousis, P.; Lang, P.; Lauber, P.; Lazányi, N.; Lazaros, A.; Le, H.B.; Lebschy, A.; Leddy, J.; Lefévre, L.; Lehnen, M.; Leipold, F.; Lessig, A.; Leyland, M.; Li, L.; Liang, Y.; Lipschultz, B.; Liu, Y.Q.; Loarer, T.; Loarte, A.; Loewenhoff, T.; Lomanowski, B.; Loschiavo, V. P.; Lunt, T.; Lupelli, I.; Lux, H.; Lyssoivan, A.; Madsen, J.; Maget, P.; Maggi, C.; Maggiora, R.; Magnussen, M. L.; Mailloux, J.; Maljaars, B.; Malygin, A.; Mantica, P.; Mantsinen, M.; Maraschek, M.; Marchand, B.; Marconato, N.; Marini, C.; Marinucci, M.; Markovic, T.; Marocco, D.; Marrelli, L.; Martin, Y.; Martin Solis, J. R.; Martitsch, A.; Mastrostefano, S.; Mattei, M.; Matthews, G.; Mavridis, M.; Mayoral, M. L.; Mazon, D.; McCarthy, P.; McAdams, R.; McArdle, G.; McCarthy, P.; McClements, K.; McDermott, R.; McMillan, B.; Meisl, G.; Merle, A.; Meyer, O.; Milanesio, D.; Militello, F.; Miron, I. G.; Mitosinkova, K.; Mlynar, J.; Mlynek, A.; Molina, D.; Molina, P.; Monakhov, I.; Morales, J.; Moreau, D.; Morel, P.; Moret, J. M.; Moro, A.; Moulton, D.; Müller, H. W.; Nabais, F.; Nardon, E.; Naulin, V.; Nemes-Czopf, A.; Nespoli, F.; Neu, R.; Nielsen, A. H.; Nielsen, S. K.; Nikolaeva, V.; Nimb, S.; Nocente, M.; Nouailletas, R.; Nowak, S.; Oberkofler, M.; Oberparleiter, M.; Ochoukov, R.; Odstrčil, T.; Olsen, J.; Omotani, J.; O'Mullane, M. G.; Orain, F.; Osterman, N.; Paccagnella, R.; Pamela, S.; Pangione, L.; Panjan, M.; Papp, G.; Papřok, R.; Parail, V.; Parra, F. I.; Pau, A.; Pautasso, G.; Pehkonen, S. P.; Pereira, A.; Perelli Cippo, E.; Pericoli Ridolfini, V.; Peterka, M.; Petersson, P.; Petrzilka, V.; Piovesan, P.; Piron, C.; Pironti, A.; Pisano, F.; Pisokas, T.; Pitts, R.; Ploumistakis, I.; Plyusnin, V.; Pokol, G.; Poljak, D.; Pölöskei, P.; Popovic, Z.; Pór, G.; Porte, L.; Potzel, S.; Predebon, I.; Preynas, M.; Primc, G.; Pucella, G.; Puiatti, M. E.; Pütterich, T.; Rack, M.; Ramogida, G.; Rapson, C.; Rasmussen, J. Juul; Rasmussen, J.; Rattá, G. A.; Ratynskaia, S.; Ravera, G.; Réfy, D.; Reich, M.; Reimerdes, H.; Reimold, F.; Reinke, M.; Reiser, D.; Resnik, M.; Reux, C.; Ripamonti, D.; Rittich, D.; Riva, G.; Rodriguez-Ramos, M.; Rohde, V.; Rosato, J.; Ryter, F.; Saarelma, S.; Sabot, R.; Saint-Laurent, F.; Salewski, M.; Salmi, A.; Samaddar, D.; Sanchis-Sanchez, L.; Santos, J.; Sauter, O.; Scannell, R.; Scheffer, M.; Schneider, M.; Schneider, B.; Schneider, P.; Schneller, M.; Schrittwieser, R.; Schubert, M.; Schweinzer, J.; Seidl, J.; Sertoli, M.; Šesnić, S.; Shabbir, A.; Shalpegin, A.; Shanahan, B.; Sharapov, S.; Sheikh, U.; Sias, G.; Sieglin, B.; Silva, C.; Silva, A.; Silva Fuglister, M.; Simpson, J.; Snicker, A.; Sommariva, C.; Sozzi, C.; Spagnolo, S.; Spizzo, G.; Spolaore, M.; Stange, T.; Stejner Pedersen, M.; Stepanov, I.; Stober, J.; Strand, P.; Šušnjara, A.; Suttrop, W.; Szepesi, T.; Tál, B.; Tala, T.; Tamain, P.; Tardini, G.; Tardocchi, M.; Teplukhina, A.; Terranova, D.; Testa, D.; Theiler, C.; Thornton, A.; Tolias, P.; Tophj, L.; Treutterer, W.; Trevisan, G. L.; Tripsky, M.; Tsironis, C.; Tsui, C.; Tudisco, O.; Uccello, A.; Urban, J.; Valisa, M.; Vallejos, P.; Valovic, M.; Van Den Brand, H.; Vanovac, B.; Varoutis, S.; Vartanian, S.; Vega, J.; Verdoolaege, G.; Verhaegh, K.; Vermare, L.; Vianello, N.; Vicente, J.; Viezzer, E.; Vignitchouk, L.; Vijvers, W.A.J.; Villone, F.; Viola, B.; Vlahos, L.; Voitsekhovitch, I.; Vondráček, P.; Vu, N. M.T.; Wagner, D.; Walkden, N.; Wang, N.; Wauters, T.; Weiland, M.; Weinzettl, V.; Westerhof, E.; Wiesenberger, M.; Willensdorfer, M.; Wischmeier, M.; Wodniak, I.; Wolfrum, E.; Yadykin, D.; Zagórski, R.; Zammuto, I.; Zanca, P.; Zaplotnik, R.; Zestanakis, P.; Zhang, W.; Zoletnik, S.; Zuin, M.

2017-01-01

Integrating the plasma core performance with an edge and scrape-off layer (SOL) that leads to tolerable heat and particle loads on the wall is a major challenge. The new European medium size tokamak task force (EU-MST) coordinates research on ASDEX Upgrade (AUG), MAST and TCV. This multi-machine

Overview of progress in European medium sized tokamaks towards an integrated plasma-edge/wall solution

DEFF Research Database (Denmark)

Meyer, H.; Eich, T.; Beurskens, M.

2017-01-01

Integrating the plasma core performance with an edge and scrape-off layer (SOL) that leads to tolerable heat and particle loads on the wall is a major challenge. The new European medium size tokamak task force (EU-MST) coordinates research on ASDEX Upgrade (AUG), MAST and TCV. This multi-machine ...

The ASDEX upgrade toroidal field magnet and poloidal divertor field coil system adapted to reactor requirements

International Nuclear Information System (INIS)

Koeppendoerfer, W.; Blaumoser, M.; Ennen, K.; Gruber, J.; Gruber, O.; Jandl, O.; Kaufmann, M.; Kollotzek, H.; Kotzlowski, H.; Lackner, E.; Lackner, K.; Larcher, T. von; Noterdaeme, J.M.; Pillsticker, M.; Poehlchen, R.; Preis, H.; Schneider, H.; Seidel, U.; Sombach, B.; Speth, E.; Streibl, B.; Vernickel, H.; Werner, F.; Wesner, F.; Wieczorek, A.

1986-01-01

ASDEX Upgrade is a tokamak experiment with external poloidal field coils that is now under construction at IPP Garching. It can produce elongated single-null (SN), double-null (DN) , and limiter (L) configurations. The SN is the reference configuration with asymmetric load distributions in the poloidal field (PF) system and the toroidal field (TF) magnet. Plasma control and stabilization require a rigid passive conductor close to the plasma. The design principles of the coils and support structure are described. (orig.)

Investigation of density fluctuations in the ASDEX tokamak via collective laser scattering

International Nuclear Information System (INIS)

Dodel, G.; Holzhauer, E.

1990-01-01

A 119μm laser scattering experiment is used on ASDEX to investigate wavenumber and frequency spectra of the density fluctuations occurring in the different operational modes of the machine. The aim of the measurements is to get insight in the physical nature of the fluctuations and their possible role in connection with anomalous transport. Since no complete theory exists, the simple guidelines of gyroradius-scaling and mixinglength level are used in the choice of parameters to be varied. Particular emphasis has been placed on the investigation of the fluctuations in the ohmic phase. (author) 1 ref., 3 figs

Investigation of density fluctuations in the ASDEX tokamak via collective laser scattering

International Nuclear Information System (INIS)

Dodel, G.; Holzhauer, E.

1990-01-01

A 119 μm laser scattering experiment is used on ASDEX to investigate wavenumber and frequency spectra of the density fluctuations occurring in the different operational modes of the machine. The aim of the measurements is to get insight in the physical nature of the fluctuations and their possible role in connection with anomalous transport. Since no complete theory exists, the simple guidelines of gyroradius-scaling and mixinglength level are used in the choice of parameters to be varied. Particular emphasis has been placed on the investigation of the fluctuations in the ohmic phase. (orig./AH)

Experiences with tungsten coatings in high heat flux tests and under plasma load in ASDEX Upgrade

International Nuclear Information System (INIS)

Herrmann, A; Greuner, H; Fuchs, J C; Marne, P de; Neu, R

2009-01-01

ASDEX Upgrade was operated with about 6400 s plasma discharge during the scientific program in 2007/2008 exploring tungsten as a first wall material in tokamaks. In the first phase, the heating power was restricted to 10 MW. It was increased to 15 MW in the second phase. During this operational period, a delamination of the 200 μm W-VPS coating happened at 2 out of 128 tiles of the outer divertor and an unscheduled opening was required. In the third phase, ASDEX Upgrade was operated with partly predamaged tiles and up to 15 MW heating power. The target load was actively controlled by N 2 -seeding. This paper presents the screening test of target tiles in the high heat flux test facility GLADIS, experiences with operation and detected damages of the outer divertor as well as the heat load to the outer divertor and the reasons for the toroidal asymmetry of the divertor load.

Density limit in ASDEX discharges with peaked density profiles

International Nuclear Information System (INIS)

Staebler, A.; Niedermeyer, H.; Loch, R.; Mertens, V.; Mueller, E.R.; Soeldner, F.X.; Wagner, F.

1989-01-01

Results concerning the density limit in OH and NI-heated ASDEX discharges with the usually observed broad density profiles have been reported earlier: In ohmic discharges with high q a (q-cylindrical is used throughout this paper) the Murakami parameter (n e R/B t ) is a good scaling parameter. At the high densities edge cooling is observed causing the plasma to shrink until an m=2-instability terminates the discharge. When approaching q a =2 the density limit is no longer proportional to I p ; a minimum exists in n e,max (q a ) at q a ∼2.15. With NI-heating the density limit increases less than proportional to the heating power; the behaviour during the pre-disruptive phase is rather similar to the one of OH discharges. There are specific operating regimes on ASDEX leading to discharges with strongly peaked density profiles: the improved ohmic confinement regime, counter neutral injection, and multipellet injection. These regimes are characterized by enhanced energy and particle confinement. The operational limit in density for these discharges is, therefore, of great interest having furthermore in mind that high central densities are favourable in achieving high fusion yields. In addition, further insight into the mechanisms of the density limit observed in tokamaks may be obtained by comparing plasmas with rather different density profiles at their maximum attainable densities. 7 refs., 2 figs

Interpretation of fast measurements of plasma potential, temperature and density in SOL of ASDEX Upgrade

DEFF Research Database (Denmark)

Horacek, J.; Adamek, J.; Müller, H.W.

2010-01-01

This paper focuses on interpretation of fast (1 µs) and local (2–4 mm) measurements of plasma density, potential and electron temperature in the edge plasma of tokamak ASDEX Upgrade. Steady-state radial profiles demonstrate the credibility of the ball-pen probe. We demonstrate that floating...... potential fluctuations measured by a Langmuir probe are dominated by plasma electron temperature rather than potential. Spatial and temporal scales are found consistent with expectations based on interchange-driven turbulence. Conditionally averaged signals found for both potential and density are also...

Density fluctuations in ohmic-, L-mode an H-mode discharges of ASDEX

Energy Technology Data Exchange (ETDEWEB)

Dodel, G; Holzhauer, E [Stuttgart Univ. (Germany). Inst. fuer Plasmaforschung; Niedermeyer, H; Endler, M; Gerhardt, J; Giannone, L.; Wagner, F; Zohm, H [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

1991-01-01

The 119 [mu]m laser scattering device ASDEX was used to investigate the direction of propagation and temporal development of density fluctuations. In ohmic discharges the density fluctuations propagate predominantly in the electron-diamagnetic direction and change direction with NI co-injection. A strong drop in total scattered power together with a further increase in the frequency shift is observed after the build-up of the transport barrier. Similar observations have been reported on other tokamaks. Due to the finite spatial resolution of the scattering system the variation of the fluctuations with local parameters cannot be sufficiently resolved to confirm their nature. (author) 5 refs., 3 figs.

Density fluctuations in ohmic-, L-mode an H-mode discharges of ASDEX

International Nuclear Information System (INIS)

Dodel, G.; Holzhauer, E.

1991-01-01

The 119 μm laser scattering device ASDEX was used to investigate the direction of propagation and temporal development of density fluctuations. In ohmic discharges the density fluctuations propagate predominantly in the electron-diamagnetic direction and change direction with NI co-injection. A strong drop in total scattered power together with a further increase in the frequency shift is observed after the build-up of the transport barrier. Similar observations have been reported on other tokamaks. Due to the finite spatial resolution of the scattering system the variation of the fluctuations with local parameters cannot be sufficiently resolved to confirm their nature. (author) 5 refs., 3 figs

Tokamaks - Third Edition

International Nuclear Information System (INIS)

Rogister, A L

2004-01-01

an introduction to diagnostics for tokamaks. The complexity of fusion plasmas is attested to by the discovery of new phenomena and new operational regimes as machine size and power increased and the diagnostic tools improved over the forty years of research on magnetic confinement. The history of those discoveries in the devices which have been built worldwide after the results obtained on the first tokamaks at the Kurchatov Institute had been confirmed is outlined in chapters 11-12. Particular emphasis is naturally given to the results from the larger tokamaks: ASDEX Upgrade, DIII-D, TFTR, JT-60/JT-60U and JET. Chapter 13 is devoted to the International Tokamak Experimental Reactor and prospects beyond ITER. Examples of operational regimes and of often unexpected phenomena are the linear and saturated ohmic confinement modes, confinement degradation when auxiliary heating is applied, the high energy confinement mode, the formation of internal transport barriers in weak or negative central shear discharges, sawtooth relaxations, disruptions, multifaceted asymmetric radiation from the edge, edge localised modes, etc. The relevant observations are described very thoroughly with the support of numerous selected figures and their physical interpretation, a major topic of the book, is carefully discussed on the basis of simplified but convincing mathematical models. With respect to the previous edition (1997), a few additions have been introduced; those concern plasma rotation (section 3.13), internal transport barriers (4.14), the role of radial electric field shear (4.19), turbulence simulations (4.21), impurity transport (4.22) and neoclassical drive of tearing modes (7.3). It is my personal feeling that some of those additions should have been somewhat more elaborated. A few pages have finally been added concerning the TCV, START, MAST, NSTX and ASDEX Upgrade tokamaks. With this book, John Wesson offers the fusion community a very precious and thorough survey of

First 50 pps Thomson scattering diagnostics in a tokamak

International Nuclear Information System (INIS)

Roehr, H.; Schramm, G.; Steuer, K.H.; Hirsch, K.; Salzmann, H.

1981-12-01

Electron temperature and density measurements by Thomson scattering were performed for the first time for the whole duration of a tokamak discharge. A 50 pps Nd:YAG laser at 1.06 μm was used in ASDEX in combination with Si avalanche photodiode detectors. Density calibration was done by rotational anti-Stokes Raman scattering from hydrogen. The system is used for measurements at electron densities of as low as 2 x 10 12 cm -3 . (orig.)

Application of AXUV diode detectors at ASDEX Upgrade

Science.gov (United States)

Bernert, M.; Eich, T.; Burckhart, A.; Fuchs, J. C.; Giannone, L.; Kallenbach, A.; McDermott, R. M.; Sieglin, B.

2014-03-01

In the ASDEX Upgrade tokamak, a radiation measurement for a wide spectral range, based on semiconductor detectors, with 256 lines of sight and a time resolution of 5μs was recently installed. In combination with the foil based bolometry, it is now possible to estimate the absolutely calibrated radiated power of the plasma on fast timescales. This work introduces this diagnostic based on AXUV (Absolute eXtended UltraViolet) n-on-p diodes made by International Radiation Detectors, Inc. The measurement and the degradation of the diodes in a tokamak environment is shown. Even though the AXUV diodes are developed to have a constant sensitivity for all photon energies (1 eV-8 keV), degradation leads to a photon energy dependence of the sensitivity. The foil bolometry, which is restricted to a time resolution of less than 1 kHz, offers a basis for a time dependent calibration of the diodes. The measurements of the quasi-calibrated diodes are compared with the foil bolometry and found to be accurate on the kHz time scale. Therefore, it is assumed, that the corrected values are also valid for the highest time resolution (200 kHz). With this improved diagnostic setup, the radiation induced by edge localized modes is analyzed on fast timescales.

Application of AXUV diode detectors at ASDEX Upgrade

International Nuclear Information System (INIS)

Bernert, M.; Eich, T.; Burckhart, A.; Fuchs, J. C.; Giannone, L.; Kallenbach, A.; McDermott, R. M.; Sieglin, B.

2014-01-01

In the ASDEX Upgrade tokamak, a radiation measurement for a wide spectral range, based on semiconductor detectors, with 256 lines of sight and a time resolution of 5μs was recently installed. In combination with the foil based bolometry, it is now possible to estimate the absolutely calibrated radiated power of the plasma on fast timescales. This work introduces this diagnostic based on AXUV (Absolute eXtended UltraViolet) n-on-p diodes made by International Radiation Detectors, Inc. The measurement and the degradation of the diodes in a tokamak environment is shown. Even though the AXUV diodes are developed to have a constant sensitivity for all photon energies (1 eV-8 keV), degradation leads to a photon energy dependence of the sensitivity. The foil bolometry, which is restricted to a time resolution of less than 1 kHz, offers a basis for a time dependent calibration of the diodes. The measurements of the quasi-calibrated diodes are compared with the foil bolometry and found to be accurate on the kHz time scale. Therefore, it is assumed, that the corrected values are also valid for the highest time resolution (200 kHz). With this improved diagnostic setup, the radiation induced by edge localized modes is analyzed on fast timescales

Low frequency sawtooth precursor activity in ASDEX Upgrade

Energy Technology Data Exchange (ETDEWEB)

Papp, G; Pokol, G I; Por, G; Magyarkuti, A; Lazanyi, N; Horvath, L [Department of Nuclear Techniques, Budapest University of Technology and Economics, Association EURATOM, Pf 91, H-1521 Budapest (Hungary); Igochine, V; Maraschek, M, E-mail: papp@reak.bme.h [Max-Planck-Institut fuer Plasmaphysik, Association EURATOM, D-85748 Garching (Germany)

2011-06-15

This paper describes the precursor activity observed in the ASDEX Upgrade tokamak before sawtooth crashes in various neutral beam heated plasmas, utilizing the soft x-ray diagnostic. In addition to the well-known (m, n) = (1,1) internal kink mode and its harmonics, a lower frequency mode is studied in detail. Power modulation of this mode is found to correlate with the power modulation of the (1, 1) kink mode in the quasistationary intervals indicating possible nonlinear interaction. Throughout the studied sawtooth crashes, the power of the lower frequency mode rose by several orders of magnitude just before the crash. In addition to its temporal behaviour, its spatial structure was estimated and the most likely value was found to be (1, 1). A possible role of this mode in the mechanism of the sawtooth crash is discussed.

Low frequency sawtooth precursor activity in ASDEX Upgrade

International Nuclear Information System (INIS)

Papp, G; Pokol, G I; Por, G; Magyarkuti, A; Lazanyi, N; Horvath, L; Igochine, V; Maraschek, M

2011-01-01

This paper describes the precursor activity observed in the ASDEX Upgrade tokamak before sawtooth crashes in various neutral beam heated plasmas, utilizing the soft x-ray diagnostic. In addition to the well-known (m, n) = (1,1) internal kink mode and its harmonics, a lower frequency mode is studied in detail. Power modulation of this mode is found to correlate with the power modulation of the (1, 1) kink mode in the quasistationary intervals indicating possible nonlinear interaction. Throughout the studied sawtooth crashes, the power of the lower frequency mode rose by several orders of magnitude just before the crash. In addition to its temporal behaviour, its spatial structure was estimated and the most likely value was found to be (1, 1). A possible role of this mode in the mechanism of the sawtooth crash is discussed.

Non-linear coupling of the lower hybrid grill in ASDEX

International Nuclear Information System (INIS)

Petrzilka, V.A.

1991-01-01

Computations of the reflection coefficient based on a non-linear lower hybrid (LH) coupling theory are presented and compared with the measurements of the reflection coefficient of the ASDEX tokamak LH grill, where powers up to 4 kW/cm 2 have been launched. This high LH power density modifies the electron density in front of the grill because of ponderomotive forces. Thus, the coupling and the power reflection coefficient change. To explain the observed saturation of the growth of the reflection coefficient with power, it is necessary to take into account some heating of the plasma in front of the grill by the transmitted LH power, which also leads to a poloidally inhomogeneous edge electron density. (author). Letter-to-the-editor. 14 refs, 13 figs

Non-linear coupling of the lower hybrid grill in ASDEX

Energy Technology Data Exchange (ETDEWEB)

Petrzilka, V A [Ceskoslovenska Akademie Ved, Prague (Czechoslovakia). Ustav Fyziky Plazmatu; Leuterer, F; Soeldner, F X; Giannone, L.; Schubert, R [Association Euratom-Max-Planck-Institut fuer Plasmaphysik, Garching (Germany, F.R.)

1991-09-01

Computations of the reflection coefficient based on a non-linear lower hybrid (LH) coupling theory are presented and compared with the measurements of the reflection coefficient of the ASDEX tokamak LH grill, where powers up to 4 kW/cm{sup 2} have been launched. This high LH power density modifies the electron density in front of the grill because of ponderomotive forces. Thus, the coupling and the power reflection coefficient change. To explain the observed saturation of the growth of the reflection coefficient with power, it is necessary to take into account some heating of the plasma in front of the grill by the transmitted LH power, which also leads to a poloidally inhomogeneous edge electron density. (author). Letter-to-the-editor. 14 refs, 13 figs.

Numerical Simulation of Neoclassical Currents, Parallel Viscosity, and Radial Current Balance in Tokamak Plasmas

International Nuclear Information System (INIS)

Kiviniemi, T.

2001-01-01

One of the principal problems en route to a fusion reactor is that of insufficient plasma confinement, which has lead to both theoretical and experimental research into transport processes in the parameter range relevant for fusion energy production. The neoclassical theory of tokamak transport is well-established unlike the theory of turbulence driven anomalous transport in which extensive progress has been made during last few years. So far, anomalous transport has been dominant in experiments, but transport may be reduced to the neoclassical level in advanced tokamak scenarios. This thesis reports a numerical study of neoclassical fluxes, parallel viscosity, and neoclassical radial current balance in tokamaks. Neoclassical parallel viscosity and particle fluxes are simulated over a wide range of collisionalities, using the fully kinetic five-dimensional neoclassical orbit-following Monte Carlo code ASCOT. The qualitative behavior of parallel viscosity derived in earlier analytic models is shown to be incorrect for high poloidal Mach numbers. This is because the poloidal dependence of density was neglected. However, in high Mach number regime, it is the convection and compression terms, rather than the parallel viscosity term, that are shown to dominate the momentum balance. For fluxes, a reasonable agreement between numerical and analytical results is found in the collisional parameter regime. Neoclassical particle fluxes are additionally studied in the banana regime using the three-dimensional Fokker-Planck code DEPORA, which solves the drift-kinetic equation with finite differencing. Limitations of the small inverse aspect ratio approximation adopted in the analytic theory are addressed. Assuming that the anomalous transport is ambipolar, the radial electric field and its shear at the tokamak plasma edge can be solved from the neoclassical radial current balance. This is performed both for JET and ASDEX Upgrade tokamaks using the ASCOT code. It is shown that

Visible spectroscopy on ASDEX

International Nuclear Information System (INIS)

Hofmann, J.V.

1991-12-01

In this report visible spectroscopy and impurity investigations on ASDEX are reviewed and several sets of visible spectra are presented. As a basis for identification of metallic impurity lines during plasma discharges spectra from a stainless steel - Cu arc have been recorded. In a next step a spectrum overview of ASDEX discharges is shown which reveals the dominating role of lines from light impurities like carbon and oxygen throughout the UV and visible range (2000 A ≤ λ ≤ 8000 A). Metallic impurity lines of neutrals or single ionized atoms are observed near localized surfaces. The dramatic effect of impurity reduction by boronization of the vessel walls is demonstrated in a few examples. In extension to some ivesti-gations already published, further diagnostic applications of visible spectroscopy are presented. Finally, the hardware and software system used on ASDEX are described in detail. (orig.)

Ammonia production in nitrogen seeded plasma discharges in ASDEX Upgrade

Energy Technology Data Exchange (ETDEWEB)

Rohde, V., E-mail: Volker.Rohde@ipp.mpg.de; Oberkofler, M.

2015-08-15

In present tokamaks nitrogen seeding is used to reduce the power load onto the divertor tiles. Some fraction of the seeded nitrogen reacts with hydrogen to form ammonia. The behaviour of ammonia in ASDEX Upgrade is studied by mass spectrometry. Injection without plasma shows strong absorption at the inner walls of the vessel and isotope exchange reactions. During nitrogen seeding in H-mode discharges the onset of a saturation of the nitrogen retention is observed. The residual gas consists of strongly deuterated methane and ammonia with almost equal amounts of deuterium and protium. This confirms the role of surface reactions in the ammonia formation. The results are consistent with findings in previous investigations. A numerical decomposition of mass spectra is under development and will be needed for quantitative evaluation of the results obtained.

Characteristics of equilibrium and perturbed transport coefficients in tokamaks

International Nuclear Information System (INIS)

Gentle, K.W.

1995-01-01

Although the evolution of a perturbation to a tokamak equilibrium can generally be described by local transport coefficients modestly enhanced above the equilibrium values, there are some significant cases for which this is inadequate. The density profile evolution in ASDEX-U occurs far more rapidly than is consistent with reasonable particle confinement times, and the evolution of cold pulses in TEXT requires nonlocal behavior in the core and some kind of anomaly near the periphery. The experiments are suggesting effects beyond standard local turbulent transport models. (orig.)

Analysis of the H-mode density limit in the ASDEX upgrade tokamak using bolometry

Energy Technology Data Exchange (ETDEWEB)

Bernert, Matthias

2013-10-23

The high confinement mode (H-mode) is the operational scenario foreseen for ITER, DEMO and future fusion power plants. At high densities, which are favourable in order to maximize the fusion power, a back transition from the H-mode to the low confinement mode (L-mode) is observed. This H-mode density limit (HDL) occurs at densities on the order of, but below, the Greenwald density. In this thesis, the HDL is revisited in the fully tungsten walled ASDEX Upgrade tokamak (AUG). In AUG discharges, four distinct operational phases were identified in the approach towards the HDL. First, there is a stable H-mode, where the plasma density increases at steady confinement, followed by a degrading H-mode, where the core electron density is fixed and the confinement, expressed as the energy confinement time, reduces. In the third phase, the breakdown of the H-mode and transition to the L-mode, the overall electron density is fixed and the confinement decreases further, leading, finally, to an L-mode, where the density increases again at a steady confinement at typical L-mode values until the disruptive Greenwald limit is reached. These four phases are reproducible, quasi-stable plasma regimes and provide a framework in which the HDL can be further analysed. Radiation losses and several other mechanisms, that were proposed as explanations for the HDL, are ruled out for the current set of AUG experiments with tungsten walls. In addition, a threshold of the radial electric field or of the power flux into the divertor appears to be responsible for the final transition back to L-mode, however, it does not determine the onset of the HDL. The observation of the four phases is explained by the combination of two mechanisms: a fueling limit due to an outward shift of the ionization profile and an additional energy loss channel, which decreases the confinement. The latter is most likely created by an increased radial convective transport at the edge of the plasma. It is shown that the

Analysis of the H-mode density limit in the ASDEX upgrade tokamak using bolometry

International Nuclear Information System (INIS)

Bernert, Matthias

2013-01-01

The high confinement mode (H-mode) is the operational scenario foreseen for ITER, DEMO and future fusion power plants. At high densities, which are favourable in order to maximize the fusion power, a back transition from the H-mode to the low confinement mode (L-mode) is observed. This H-mode density limit (HDL) occurs at densities on the order of, but below, the Greenwald density. In this thesis, the HDL is revisited in the fully tungsten walled ASDEX Upgrade tokamak (AUG). In AUG discharges, four distinct operational phases were identified in the approach towards the HDL. First, there is a stable H-mode, where the plasma density increases at steady confinement, followed by a degrading H-mode, where the core electron density is fixed and the confinement, expressed as the energy confinement time, reduces. In the third phase, the breakdown of the H-mode and transition to the L-mode, the overall electron density is fixed and the confinement decreases further, leading, finally, to an L-mode, where the density increases again at a steady confinement at typical L-mode values until the disruptive Greenwald limit is reached. These four phases are reproducible, quasi-stable plasma regimes and provide a framework in which the HDL can be further analysed. Radiation losses and several other mechanisms, that were proposed as explanations for the HDL, are ruled out for the current set of AUG experiments with tungsten walls. In addition, a threshold of the radial electric field or of the power flux into the divertor appears to be responsible for the final transition back to L-mode, however, it does not determine the onset of the HDL. The observation of the four phases is explained by the combination of two mechanisms: a fueling limit due to an outward shift of the ionization profile and an additional energy loss channel, which decreases the confinement. The latter is most likely created by an increased radial convective transport at the edge of the plasma. It is shown that the

Experimental and theoretical investigation of density and potential fluctuations in the scrape-off layer of ASDEX

Energy Technology Data Exchange (ETDEWEB)

Endler, M; Giannone, L.; Niedermeyer, H; Rudyj, A; Theimer, G [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

1994-12-31

Electrostatic fluctuations (i.e. the magnetic field is assumed constant) are candidates for the explanation of the anomalous transport of particles and energy in both tokamaks and stellarators. While most theoretical effort has been directed to an explanation of the anomalous transport in the bulk plasma, it is now widely being realized that the anomalous radial transport in the scrape-off layer, determining the width of the power flow channel at limiter or divertor plates, may be equally important to a future reactor experiment. In the divertor tokamak ASDEX density and potential fluctuations in the scrape-off layer were investigated with high temporal and spatial resolution by Langmuir probes and an H{sub {alpha}} diagnostic. Many results of these measurements were reported and are summarized below. Several of these properties of the fluctuations have also been reported from other experiments. (author) 3 refs., 4 figs.

Regime identification in ASDEX Upgrade

International Nuclear Information System (INIS)

Giannone, L; Sips, A C C; Kardaun, O; Spreitler, F; Suttrop, W

2004-01-01

The ability to recognize the transition from the L-mode to the H-mode or from the H-mode to the improved H-mode reliably from a conveniently small number of measurements in real time is of increasing importance for machine control. Discriminant analysis has been applied to regime identification of plasma discharges in the ASDEX Upgrade tokamak. An observation consists of a set of plasma parameters averaged over a time slice in a discharge. The data set consists of all observations over different discharges and time slices. Discriminant analysis yields coefficients allowing the classification of a new observation. The results of a frequentist and a formal Bayesian approach to discriminant analysis are compared. With five plasma variables, a failure rate of 1.3% for predicting the L-mode and the H-mode confinement regime was achieved. With five plasma variables, a failure rate of 5.3% for predicting the H-mode and the improved H-mode confinement regime was achieved. The coefficients derived by discriminant analysis have been applied subsequently to discharges to illustrate the operation of regime identification in a real time control system

Fluctuation measurements by Langmuir probes during LHCD on ASDEX tokamak. [LHCD (Lower Hybrid Current Drive)

Energy Technology Data Exchange (ETDEWEB)

Stoeckel, J [Ceskoslovenska Akademie Ved, Prague (Czech Republic). Ustav Fyziky Plazmatu; Soeldner, F; Giannone, L.; Leuterer, F [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

1991-01-01

The level of edge electrostatic fluctuations decreases and the global particle/energy confinement improves during lower hybrid current drive (LHCD) regimes on ASDEX, when the total power remains below the initial OH power level. For higher powers, the fluctuations increase noticeably, whereas the global confinement is returning to its OH value. The observed increase of fluctuations is poloidally asymmetric and is caused by local power deposition in front of the grill antenna. (author) 5 refs., 4 figs.

Thermal ions dilution and ITG suppression in ASDEX Upgrade ion ITBs

International Nuclear Information System (INIS)

Tardini, G.; Hobirk, J.; Igochine, V.G.; Maggi, C.F.; Martin, P.; McCune, D.; Peeters, A.G.; Sips, A.C.C.; Staebler, A.; Stober, J.

2007-01-01

Internal transport barriers (ITBs) in the ion channel in the tokamak ASDEX Upgrade allow for high energy confinement but collapse after only several energy confinement times. In this paper we show that in most cases the ITB phase is terminated clearly before the first ELM burst, thereby ruling out the ELMs as the main trigger of the ITB collapse. For the first time, the ITB formation and sustainment are found to be associated with a mechanism of transport suppression based on thermal ions dilution by the injected fast ions. Interestingly, such ITBs do not require reversed magnetic shear. The linear growth rate of the ion temperature gradient driven mode is computed as a function of the fast ion fraction with gyrokinetic stability analysis. Monte Carlo simulations predict the fast ion population to be above the gyrokinetic critical fraction in a region consistent with the experimental ITB width. The density threshold documented for the onset of ASDEX Upgrade ion ITBs is explained. The role of T i /T e and of the plasma sheared rotation for ITB sustainment are analysed. The stabilization mechanism presented here is consistent with the observed ITB lifetime of the order of the beam slowing down time. A possible runaway mechanism leading to ITB collapse is described. Finally, the relevance of this particular ITB scheme for ITER is discussed

Simulation of feedback control system for NTM stabilisation in ASDEX Upgrade

Energy Technology Data Exchange (ETDEWEB)

Rapson, Christopher, E-mail: chris.rapson@ipp.mpg.de [Max Planck Institute for Plasma Physics, EURATOM Association, Boltzmannstrasse 2, 85748 Garching (Germany); Monaco, Francesco; Reich, Matthias; Stober, Joerg; Treutterer, Wolfgang [Max Planck Institute for Plasma Physics, EURATOM Association, Boltzmannstrasse 2, 85748 Garching (Germany)

2013-10-15

Highlights: ► Feedback loop to control the ECRH deposition location is modelled in Simulink. Controller optimised using simulation results. ► Apart from optimising the PID gain values, alternative architectures were trialed without risk to hardware. ► Off-normal events could be simulated, and the controller response improved. ► Optimised controller applied in experiment. Even for the low power used, partial stabilisation of NTM was observed. ► The simulation is useful outside its intended application, and for future developments of the NTM feedback control system. -- Abstract: Neoclassical Tearing Modes (NTMs) are a class of MHD instability in high beta tokamak plasmas which significantly increase radial transport, thus capping the performance of fusion plasmas. More importantly, NTMs can lead to disruptions which compromise the lifetime of structural components. Several tokamaks have demonstrated that Electron Cyclotron Resonant Heating (ECRH) can stabilise NTMs if the power deposition is aligned with the mode location. The deposition location depends on the toroidal magnetic field, flux and density profiles, and can be controlled by tilting the mirror in the ECRH launcher. Until recently, the mirror angle was set by feedforward control at ASDEX Upgrade. In order to adapt automatically to different discharge scenarios, the system at ASDEX Upgrade has been extended to steer the mirror using feedback control. The mirror must react on the current diffusion time scale, on the order of 100 ms. This is within the capabilities of the mechanical subsystem and real-time plasma diagnostics, but requires careful interfacing between these components. For example, asynchronous data transfer and non-linearities make it difficult to design an analytically optimal controller. Therefore a simulation has been used to test and tune different controller architectures. This simulation is the subject of the current contribution. Performing the optimisation process offline

ICRF induced edge plasma convection in ASDEX upgrade

Energy Technology Data Exchange (ETDEWEB)

Zhang, Wei [Max Planck Institute for Plasma Physics, Garching/Greifswald (Germany); University of Ghent, Ghent (Belgium); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Feng, Yuehe; Lunt, Tilmann; Jacquot, Jonathan; Coster, David; Bilato, Roberto; Bobkov, Volodymyr; Ochoukov, Roman [Max Planck Institute for Plasma Physics, Garching/Greifswald (Germany); Noterdaeme, Jean-Marie [Max Planck Institute for Plasma Physics, Garching/Greifswald (Germany); University of Ghent, Ghent (Belgium); Colas, Laurent [CEA, IRFM, Saint-Paul-Lez-Durance (France); Collaboration: ASDEX Upgrade Team

2016-07-01

Ion Cyclotron Range of Frequency (ICRF) heating is one of the main auxiliary plasma heating methods in tokamaks. It relies on the fast wave to heat the plasma. However the slow wave can also be generated parasitically. The parallel electric field of the slow wave can induce large biased plasma potential through sheath rectification. The rapid variation of this rectified potential across the magnetic field can cause significant E x B convection in the Scrape-Off Layer (SOL). The ICRF induced convection can affect the SOL density, influence the ICRF power coupling and enhance the strength of plasma-wall interactions. To explore these physics, we not only show the experimental evidences in ASDEX Upgrade, but also present the associated simulation results with the 3D edge plasma fluid code EMC3-Eirene. Further simulations via combination of EMC3-Eirene and a sheath code SSWICH in an iterative and quasi self-consistent way can give good predictions for future experiments.

Characteristics of toroidal energy deposition asymmetries in ASDEX

International Nuclear Information System (INIS)

Evans, T.E.; Neuhauser, J.; Leuterer, F.; Mueller, E.R.

1990-01-01

Large toroidal and poloidal asymmetries with characteristics which are sensitively dependent on q a , the vertical position of the plasma, and the type of additional heating are observed in the energy flow to the ASDEX divertor target plates. The largest asymmetries and total energy depositions are observed during lower hybrid wave injection experiments with approximately 50% of the input energy going to the combined divertor targets and shields. A maximum localized energy density loading of 10 MJ/m 2 is typical under these conditions. Measurements of the asymmetries are consistent with a model in which magnetic islands and ergodicity due to intrinsic magnetic perturbations dominate the energy transpot across the primary magnetic separatrix. The results emphasize the essential role of resonant magnetic perturbations in determining the performance of tokamaks and demonstrate that non-axisymmetric effects caused by small perturbations become increasingly important in determining the transport properties as the injected power is increased. (orig.)

Investigation of the density turbulence in ohmic ASDEX plasmas

International Nuclear Information System (INIS)

Dodel, G.; Holtzhauer, E.

1989-01-01

A 119 μm homodyne laser scattering experiment is used on ASDEX to investigate wavenumber and frequency of the density fluctuations occuring in the different operational conditions of the machine. The changes of the density turbulence caused by additional heating are of primary interest with regard to a possible correlation to anomalous transport. Therefore, in the current experiment particular emphasis is placed on these investigations. On the other hand it is the ohmic phase which constitutes the least complicated physical situation in a tokamak and is therefore best suited to reveal the basic physical nature of the density turbulence. In the following we present a summary of our findings in the ohmic phase and make an attempt to compare these findings with what would be expected from the simplest model of density-gradient-driven driftwave turbulence saturated at the mixing-length level. (author) 3 refs., 4 figs

Investigation of the density turbulence in ohmic ASDEX plasmas

Energy Technology Data Exchange (ETDEWEB)

Dodel, G; Holtzhauer, E [Stuttgart Univ. (Germany, F.R.). Inst. fuer Plasmaforschung; Giannone, L.; Niedermeyer, H [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany, F.R.)

1989-01-01

A 119 {mu}m homodyne laser scattering experiment is used on ASDEX to investigate wavenumber and frequency of the density fluctuations occuring in the different operational conditions of the machine. The changes of the density turbulence caused by additional heating are of primary interest with regard to a possible correlation to anomalous transport. Therefore, in the current experiment particular emphasis is placed on these investigations. On the other hand it is the ohmic phase which constitutes the least complicated physical situation in a tokamak and is therefore best suited to reveal the basic physical nature of the density turbulence. In the following we present a summary of our findings in the ohmic phase and make an attempt to compare these findings with what would be expected from the simplest model of density-gradient-driven driftwave turbulence saturated at the mixing-length level. (author) 3 refs., 4 figs.

A new beam emission polarimetry diagnostic for measuring the magnetic field line angle at the plasma edge of ASDEX Upgrade.

Science.gov (United States)

Viezzer, E; Dux, R; Dunne, M G

2016-11-01

A new edge beam emission polarimetry diagnostic dedicated to the measurement of the magnetic field line angle has been installed on the ASDEX Upgrade tokamak. The new diagnostic relies on the motional Stark effect and is based on the simultaneous measurement of the polarization direction of the linearly polarized π (parallel to the electric field) and σ (perpendicular to the electric field) lines of the Balmer line D α . The technical properties of the system are described. The calibration procedures are discussed and first measurements are presented.

Generation and analysis of plasmas with centrally reduced helicity in full-tungsten ASDEX upgrade

Energy Technology Data Exchange (ETDEWEB)

Bock, Alexander

2016-03-01

The most promising concepts for harnessing nuclear fusion are toroidal devices like tokamaks, where a plasma is confined by helically twisted magnetic field lines. To provide the twisting of the field lines, a tokamak relies on a toroidal current in the plasma, which is largely generated by a transformer. As such, conventional tokamaks are limited to pulsed operation. Moreover, this current makes tokamak plasmas prone to numerous confinement degrading magnetohydrodynamic (MHD) instabilities that can emerge at locations where the field line helicity q takes on rational values like 1/1, 3/2 or 2/1, i.e. sawteeth or neoclassical tearing modes (NTMs). This thesis presents studies of plasmas with centrally elevated q-profiles created by external electron-cyclotron and neutral beam current drive (ECCD/NBCD) under steady-state conditions in the full-tungsten tokamak ASDEX Upgrade. Without the usually monotonic q-profile, instabilities of low helicity disappear, thereby improving the plasma stability. Furthermore, elevating q increases the amount of so-called (toroidal) bootstrap current, which the plasma drives by itself in the presence of pressure gradients, thereby reducing the reliance on the transformer. In the best case, an advanced tokamak (AT) could thus run in steady state. Additionally, an elevated and thus flat/slightly reversed q-profile is thought to improve confinement by impeding turbulent transport. Reconstruction of the tailored q-profile is accomplished with the new integrated data equilibrium (IDE) code and information from a key diagnostic that is based on the Motional Stark Effect (MSE). During the course of this work it was discovered that the MSE diagnostic suffers from interference from polarised background light. A prototype mitigation system was successfully tested. Also, non-linearities in the diagnostic's optical relay system were found and a calibration scheme devised to take them into account. Both the conventional approach of AT

Characterization of type-I ELM induced filaments in the far scrape-off layer of ASDEX upgrade

Energy Technology Data Exchange (ETDEWEB)

Schmid, Andreas

2008-03-18

This thesis focuses on the characterization of filaments and their propagation in the ASDEX Upgrade tokamak. The aim is to provide experimental measurements for understanding the filament formation process and their temporal evolution, and to provide a comprehensive database for an extrapolation to future fusion devices. For this purpose, a new magnetically driven probe for filament measurements has been developed and installed in ASDEX Upgrade. The probe carries several Langmuir probes and a magnetic coil in between. The Langmuir probes allow for measurements of the radial and poloidal/toroidal propagation of filaments as well as for measurements of filament size, density, and their radial (or temporal) evolution. The magnetic coil on the filament probe allows for measurements of currents in the filaments. A set of 7 coils, measuring 3 field components at different positions along the filament, has been used to measure the magnetic signature during an ELM. The aim was, on the one hand, to study which role filaments play for the magnetic structure, and on the other hand if the parallel currents predicted by the sheath damped model could be verified. Filament temperatures have been derived and the corresponding heat transport mechanisms have been studied. (orig.)

Characterization of type-I ELM induced filaments in the far scrape-off layer of ASDEX upgrade

International Nuclear Information System (INIS)

Schmid, Andreas

2008-01-01

This thesis focuses on the characterization of filaments and their propagation in the ASDEX Upgrade tokamak. The aim is to provide experimental measurements for understanding the filament formation process and their temporal evolution, and to provide a comprehensive database for an extrapolation to future fusion devices. For this purpose, a new magnetically driven probe for filament measurements has been developed and installed in ASDEX Upgrade. The probe carries several Langmuir probes and a magnetic coil in between. The Langmuir probes allow for measurements of the radial and poloidal/toroidal propagation of filaments as well as for measurements of filament size, density, and their radial (or temporal) evolution. The magnetic coil on the filament probe allows for measurements of currents in the filaments. A set of 7 coils, measuring 3 field components at different positions along the filament, has been used to measure the magnetic signature during an ELM. The aim was, on the one hand, to study which role filaments play for the magnetic structure, and on the other hand if the parallel currents predicted by the sheath damped model could be verified. Filament temperatures have been derived and the corresponding heat transport mechanisms have been studied. (orig.)

Fuzzy-neural approaches to the prediction of disruptions in ASDEX Upgrade

International Nuclear Information System (INIS)

Morabito, F.C.; Versaci, M.; Pautasso, G.; Tichmann, C.

2001-01-01

Disruption is a sudden loss of magnetic confinement that can cause damage to the machine walls and support structures. For this reason, it is of practical interest to be able to detect the onset of such an event early. A novel technique is presented of early prediction of plasma disruption in tokamak reactors which uses neural networks and 'fuzzy' inference. The studies carried out in the work make use of an experimental database of disruptive shots made available by the ASDEX Upgrade Team. The main result of the work is that, in the limit of the available database, it is possible to predict the onset of the disruptive event sufficiently in advance in order to put the control system into action. The proposed system is a modular scheme that exploits a decomposition of the original database carried out in a proper way. (author)

Migration and deposition of 13C in the full-tungsten ASDEX Upgrade tokamak

International Nuclear Information System (INIS)

Hakola, A; Aho-Mantila, L; Groth, M; Kurki-Suonio, T; Makkonen, T; Likonen, J; Koivuranta, S; Krieger, K; Mayer, M; Mueller, H W; Neu, R; Rohde, V

2010-01-01

The migration of carbon in low-density, low-confinement plasmas of ASDEX Upgrade was studied by injecting 13 C into the main chamber of the torus at the end of the 2007 experimental campaign. A selection of standard tungsten-coated lower-divertor and main-chamber tiles as well as a complete set of lower-divertor tiles with an uncoated poloidal marker stripe were removed from one poloidal cross section and analysed using secondary ion mass spectrometry. The poloidal deposition profiles of 13 C on both the tungsten-coated tiles and on the uncoated graphite areas of the marker tiles were measured and compared. For the W-coated lower-divertor tiles, 13 C was deposited mainly on the high-field side tiles, while barely detectable amounts of 13 C were observed on low-field side samples. In contrast, on the uncoated marker stripes the deposition was equally pronounced in the high-field and low-field side divertor. The marker-tile results are in agreement with those obtained from graphite tiles after the 2003 and 2005 13 C experiments in ASDEX Upgrade. In the case of W-coated tiles, the 13 C measurements were complemented by determining the total amount of deposited carbon ( 12 C) on the tiles, which also shows strong deposition at the inner parts of the lower divertor. The estimated deposition of 13 C on W at the divertor areas was less than 1.5% of the injected amount of 13 C atoms. The 13 C analyses of the main-chamber tiles and small silicon samples mounted in remote areas revealed significant deposition in the upper divertor, in upper parts of the heat shield, in the limiter region close to the injection valve, and below the roof baffle. Approximately 8% of the injected 13 C is estimated to have accumulated in these regions. Possible reasons for the different deposition patterns on W and on graphite in different regions of the torus are discussed.

Frequency spectral broadening of lower hybrid waves in tokamak plasmas - causes and effects

Energy Technology Data Exchange (ETDEWEB)

Pericoli Ridolfini, V; Giannone, L.; Bartiromo, R [Associazione Euratom-ENEA sulla Fusione, Rome (Italy). Centro Ricerche Energia Frascati

1994-04-01

The frequency spectral broadening of lower hybrid (LH) waves injected into tokamak plasmas is extensively analyzed with reference mostly to experimental data from the ASDEX tokamak. The link between the magnitude of the pump spectral width and the degradation of the LH current drive efficiency (up to a factor of 2), pointed out in previous works, is explained. The experimental behaviour of LH power absorption is also well reproduced, even in situations when the access of the launched LH waves to the core plasma should be largely forbidden. Experiments are described that are aimed at determined whether parametric decay instabilities (PDIs) or scattering of LH waves by density fluctuations in the plasma edge are causes of the broadening of the LH pump frequency spectrum. Fluctuations emerge as the largely dominant process, while no signature of PDI processes is observed. Careful measurements of the density fluctuations in the ASDEX scrape-off layer plasma allow the analytical description given by Andrews and Perkins to be assumed as the appropriate model of LH scattering. Indeed, it supplies the correct magnitude for the frequency spectral width of the LH pump, and explains quantitatively, together with a ray tracing code, why the CD efficiency decreases with the broadening of the pump spectrum. It can also account for the observed LH power absorption coefficient. (author). 48 refs, 13 figs, 2 tabs.

Characterization of axisymmetric disruption dynamics toward VDE avoidance in tokamaks

International Nuclear Information System (INIS)

Nakamura, Y.

2002-01-01

Disruption experiments on Alcator C-Mod and ASDEX-Upgrade tokamaks and axisymmetric MHD simulations using the TSC have explicated the underlying mechanisms of Vertical Displacement Events (VDEs) and a diversity of disruption dynamics. First, the neutral point, which is known as an initial vertical plasma position advantageous to VDE avoidance, is shown to be fairly insensitive to plasma shape and current profile parameters, while the VDE rate significantly depends on those parameters. Secondly, it is clarified that a rapid flattening of the plasma current profile frequently seen at the thermal quench drags a single null-diverted, up-down asymmetric plasma vertically toward divertor, whereas the dragging effect is absent in up-down symmetric limiter discharges. As a consequence, the occurrence of downward-going VDEs predominates over the upward-going ones in bottom-diverted discharges, being consistent with experiments in ASDEX-Upgrade. Together with the attractive force that arises from passive shell currents induced by the current quench and vanishes at the neutral point, the dragging effect explains many details of the VDE dynamics over the whole period of disruptive termination. (author)

Major upgrades of the high frequency B-dot probe diagnostic suite on ASDEX Upgrade

Directory of Open Access Journals (Sweden)

Ochoukov Roman

2017-01-01

Full Text Available The high frequency B-dot (HFB probe diagnostic on the ASDEX Upgrade tokamak has undergone a considerable upgrade during the 2016 opening of the torus. The probe coverage is now greatly expanded toroidally, as well as radially with the addition of probes on the high field side and the removable manipulator head. A new 2-channel fast digitizer now allows to examine and record radio frequency (RF wave emissions emanating from the plasma in the ion cyclotron range of frequencies (ICRF. Possible studies that can be achieved now include: a study of core ICRF power absorption efficiency; a study of ion cyclotron emissions from the plasma generated by energetic ions; and study of ICRF wave/plasma turbulence interactions in the scrape-off layer region.

Auxiliary radiofrequency heating of tokamaks, Task 3

International Nuclear Information System (INIS)

Scharer, J.E.

1991-07-01

The research performed under this grant during the past three years has been concentrated on the following several key tokamak ICRF (Ion Cyclotron Range of Frequencies) coupling and heating issues: efficient coupling during the L- to H-mode transition by analysis and computer simulation of ICRF antennas edge plasma profiles; analysis of both dielectric-filled waveguide and coil ICRF antenna coupling to plasma edge profiles; benchmarking the codes to compare with current JET, D-IIID and ASDEX experimental results; ICRF full-wave field solutions, power conservation and heating analyses; and the effects of fusion alpha particle or ion tail populations on the ICRF absorption. Research progress, publications, and conference and workshop presentations are summarized in this report. 15 refs

New operational spaces for the electron cyclotron resonance heating at ASDEX upgrade

International Nuclear Information System (INIS)

Hoehnle, Hendrik Sebastian

2012-01-01

In this thesis, new electron cyclotron resonance heating (ECRH) scenarios were developed for an extension of the operational space at the tokamak ASDEX Upgrade in view of ITER compatibility. In the last years, the first wall material at ASDEX Upgrade was changed from graphite to tungsten, and the ECRH is needed to control the tungsten concentration in the plasma core. But, in ITER-like plasma discharges at ASDEX Upgrade, the usage of the ECRH in the typically used second harmonic extraordinary polarised mode (X2 mode) is limited. In these ITER-scenarios a small safety factor should be achieved, which implements an increase of the plasma current at ASDEX Upgrade. A higher plasma current and a high confinement lead to a raised density and for the ITER scenario to an electron density above the cutoff of the X2 mode at ASDEX Upgrade. Therefore, the X2 mode is reflected at the cutoff layer and cannot be used for central heating and the control of the tungsten concentration. One possibility to overcome this problem is to apply the third harmonic mode at reduced magnetic field. Here the cutoff is increased by 33% due to the dependence on the magnetic field. However, at the reachable plasma parameters at the reduced field the absorption of the X3 mode is incomplete (60-70 %) and the shine-trough power can destroy microwave sensitive components in ASDEX Upgrade. To solve this problem the magnetic field has to be optimized. A slightly increased magnetic field from 1.7 T to 1.8 T moves the second harmonic resonance in the region of confined plasma with high temperatures and density, so that this resonance can act as beam dump. The deposition in the plasma core is still central enough for the tungsten control ability of the ECRH. The benefit of the beam dump was verified in experiments with two different magnetic fields (1.7 T and 1.8 T). In case of the higher magnetic field, the stray radiation was reduced; simultaneously the electron temperature was increased. In addition

Feedback controlled, reactor relevant, high-density, high-confinement scenarios at ASDEX Upgrade

Science.gov (United States)

Lang, P. T.; Blanken, T. C.; Dunne, M.; McDermott, R. M.; Wolfrum, E.; Bobkov, V.; Felici, F.; Fischer, R.; Janky, F.; Kallenbach, A.; Kardaun, O.; Kudlacek, O.; Mertens, V.; Mlynek, A.; Ploeckl, B.; Stober, J. K.; Treutterer, W.; Zohm, H.; ASDEX Upgrade Team

2018-03-01

One main programme topic at the ASDEX Upgrade all-metal-wall tokamak is development of a high-density regime with central densities at reactor grade level while retaining high-confinement properties. This required development of appropriate control techniques capable of coping with the pellet tool, a powerful means of fuelling but one which presented challenges to the control system for handling of related perturbations. Real-time density profile control was demonstrated, raising the core density well above the Greenwald density while retaining the edge density in order to avoid confinement losses. Recently, a new model-based approach was implemented that allows direct control of the central density. Investigations focussed first on the N-seeding scenario owing to its proven potential to yield confinement enhancements. Combining pellets and N seeding was found to improve the divertor buffering further and enhance the operational range accessible. For core densities up to about the Greenwald density, a clear improvement with respect to the non-seeding reference was achieved; however, at higher densities this benefit is reduced. This behaviour is attributed to recurrence of an outward shift of the edge density profile, resulting in a reduced peeling-ballooning stability. This is similar to the shift seen during strong gas puffing, which is required to prevent impurity influx in ASDEX Upgrade. First tests indicate that highly-shaped plasma configurations like the ITER base-line scenario, respond very well to pellet injection, showing efficient fuelling with no measurable impact on the edge density profile.

Gaseous electron multiplier-based soft x-ray plasma diagnostics development: Preliminary tests at ASDEX Upgrade

Energy Technology Data Exchange (ETDEWEB)

Chernyshova, M., E-mail: maryna.chernyshova@ipplm.pl; Malinowski, K.; Czarski, T.; Kowalska-Strzęciwilk, E. [Institute of Plasma Physics and Laser Microfusion, Hery 23, 01-497 Warsaw (Poland); Wojeński, A.; Poźniak, K. T.; Kasprowicz, G.; Krawczyk, R.; Kolasiński, P.; Zabołotny, W.; Zienkiewicz, P. [Institute of Electronic Systems, Warsaw University of Technology, Nowowiejska 15/19, 00-665 Warsaw (Poland); Vezinet, D.; Herrmann, A. [Max Planck Institute for Plasma Physics, Boltzmannstr. 2, 85748 Garching (Germany); Mazon, D.; Jardin, A. [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France)

2016-11-15

A Gaseous Electron Multiplier (GEM)-based detector is being developed for soft X-ray diagnostics on tokamaks. Its main goal is to facilitate transport studies of impurities like tungsten. Such studies are very relevant to ITER, where the excessive accumulation of impurities in the plasma core should be avoided. This contribution provides details of the preliminary tests at ASDEX Upgrade (AUG) with a focus on the most important aspects for detector operation in harsh radiation environment. It was shown that both spatially and spectrally resolved data could be collected, in a reasonable agreement with other AUG diagnostics. Contributions to the GEM signal include also hard X-rays, gammas, and neutrons. First simulations of the effect of high-energy photons have helped understanding these contributions.

Electrical conductivity in tokamaks and extended neoclassical theory

International Nuclear Information System (INIS)

Segre, S.E.; Zanza, V.

1992-01-01

The electrical conductivity measurements reported from various tokamaks (D-III, PLT, TEXT, ASDEX, JT-60, TEXTOR, JET, TFTR) and compared with the usual neoclassical theory are here also compared with the extended neoclassical theory where the electron-electron collision rate is anomalous while the electron-ion collision rate remains Coulombian. It is found that, out of the 14 experiments considered, three are consistent with both the neoclassical and the extended neoclassical theories, four are consistent only with the extended neoclassical theory, and four are consistent with the neoclassical theory and also, within the experimental errors, not inconsistent with the extended neoclassical theory; the remaining three experiments appear to be incompatible with both theories. It is concluded that the extended neoclassical theory is in better agreement with conductivity experiments than the conventional neoclassical theory and, indeed, the extended theory is a serious candidate for explaining tokamak behaviour, since it accommodates naturally an anomalous electron thermal transport, which the conventional neoclassical theory is unable to do. (author). 31 refs, 1 fig

Overview of progress in European medium sized tokamaks towards an integrated plasma-edge/wall solution

OpenAIRE

Meyer, H.; Eich, T.; Beurskens, M.N.A.; Coda, S.; Hakola, A.; Martin, P.; Adamek, J.; Agostini, M.; Aguiam, D.; Ahn, J.; Aho-Mantila, L.; Akers, R.; Albanese, R.; Aledda, R.; Alessi, E.

2017-01-01

Integrating the plasma core performance with an edge and scrape-off layer (SOL) that leads to tolerable heat and particle loads on the wall is a major challenge. The new European medium size tokamak task force (EU-MST) coordinates research on ASDEX Upgrade (AUG), MAST and TCV. This multi-machine approach within EU-MST, covering a wide parameter range, is instrumental to progress in the field, as ITER and DEMO core/pedestal and SOL parameters are not achievable simultaneously in present day de...

ICRF hydrogen minority heating in the boronized ASDEX tokamak

International Nuclear Information System (INIS)

Ryter, F.; Braun, F.; Hofmeister, F.; Noterdaeme, J.M.; Steuer, K.H.; Wesner, F.

1990-01-01

Since the divertor of ASDEX has been modified (1986-87) the hydrogen concentration in deuterium plasmas could not be reduced below 10%, although the machine was operated for long periods of time with deuterium injection. This is probably due to desorption in the divertor as indicated by the increasing H-concentration during a deuterium injection pulse. As a consequence for H-minority heating in deuterium, the maximum power into ohmic plasmas without causing a disruption was limited to few hundred kW. A partial solution was ICRH in combination with deuterium injection which allowed us to apply up to 1.5 MW ICRH to the plasma. The beneficial role of the injection is attributed to an improved ICRH absorption and to the higher energy flux and temperature in the divertor. During the last ICRH campaign we operated mainly in helium plasmas for a lower hydrogen concentration and the vessel was boronised. The H-concentration is measured routinely by a mass spectrometer in the divertor chamber. This measurement does not give a fast response to eventual changes and also no absolute concentrations in the main plasma, but it gives a reliable estimate of the time evolution during one discharge or from shot to shot. The data from the mass spectrometer were often cross-checked with charge exchange measurements from the main plasma. In helium discharges the hydrogen concentration is around 2% in the ohmic phase but it increases up to 8% as ICRH is applied. Under these conditions the maximum available power (2.7 MW) could be applied to the plasma without causing a disruption. This is partly due to the low H-concentration in helium at the beginning of the ICRH pulse but also to the boronisation, as discussed in a later. (author) 4 refs., 6 figs

Sawtooth-free Ohmic discharges in ASDEX and the aspects of neoclassical ion transport

International Nuclear Information System (INIS)

Stroth, U.; Fussmann, G.; Krieger, K.; Mertens, V.; Wagner, F.; Bessenrodt-Weberpals, M.; Buechse, R.; Giannone, L.; Herrmann, H.; Simmet, E.; Steuer, K.H.

1991-05-01

Sawtooth-free Ohmic discharges can serve as a model case for a quiescent Tokamak plasma. We report on the properties and the global parameters of these discharges observed in ASDEX and make comments on the mechanism which seems to be responsible for the stabilization of the sawtooth instability. Stationary Ohmic discharge were used to investigate particle, impurity and energy transport in the absence of the sawtooth instability. Particular emphasis has been devoted to a comparison with the predictions of neoclassical theories. We find that the ion energy transport is on the level predicted by neoclassical theory and can explain particle and impurity transport with neoclassical inward drift velocities and diffusion coefficients with the same small anomalous contribution. In the central region of the plasma, where the power flux is low, very small values were found for the electron heat conductivity. (orig.)

Phase-space resolved measurement of 2nd harmonic ion cyclotron heating using FIDA tomography at the ASDEX Upgrade tokamak

DEFF Research Database (Denmark)

Weiland, M.; Bilato, R.; Geiger, B.

2017-01-01

Recent upgrades to the FIDA (fast-ion D-alpha) diagnostic at ASDEX Upgrade allow to reconstruct the fast-ion phase space at several radial positions with decent energy and pitch resolution. These new diagnostic capabilities are applied to study the physics of 2nd harmonic ion cyclotron heating, w....... Furthermore, comparisons to other fast-ion diagnostics (neutron yield and neutral particle analyzers) are discussed....

Non-linear simulations of ELMs in ASDEX upgrade

Energy Technology Data Exchange (ETDEWEB)

Lessig, Alexander; Hoelzl, Matthias; Orain, Francois; Guenter, Sibylle [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstrasse 2, 85748 Garching (Germany); Becoulet, Marina; Huysmans, Guido [CEA-IRFM, Cadarache, 13108 Saint-Paul-Lez-Durance (France); Collaboration: the ASDEX Upgrade Team

2016-07-01

Large edge localized modes (ELMs) are a severe concern for the operation of future tokamak devices like ITER or DEMO due to the high transient heat loads induced on divertor targets and wall structures. It is therefore important to study ELMs both theoretically and experimentally in order to obtain a comprehensive understanding of the underlying mechanisms which is necessary for the prediction of ELM properties and the design of ELM mitigation systems. Using the non-linear MHD code JOREK, we have performed first simulations of full ELM crashes in ASDEX Upgrade, taking into account a large number of toroidal Fourier harmonics. The evolution of the toroidal mode spectrum has been investigated. In particular, we confirm the previously observed non-linear drive of linearly sub-dominant low-n components in the early non-linear phase of the ELM crash. Preliminary comparisons of the simulations with experimental observations regarding heat and particle losses, pedestal evolution and heat deposition patterns are shown. On the long run we aim at code validation as well as an improved understanding of the ELM dynamics and possibly a better characterization of different ELM types.

Real-time diamagnetic flux measurements on ASDEX Upgrade.

Science.gov (United States)

Giannone, L; Geiger, B; Bilato, R; Maraschek, M; Odstrčil, T; Fischer, R; Fuchs, J C; McCarthy, P J; Mertens, V; Schuhbeck, K H

2016-05-01

Real-time diamagnetic flux measurements are now available on ASDEX Upgrade. In contrast to the majority of diamagnetic flux measurements on other tokamaks, no analog summation of signals is necessary for measuring the change in toroidal flux or for removing contributions arising from unwanted coupling to the plasma and poloidal field coil currents. To achieve the highest possible sensitivity, the diamagnetic measurement and compensation coil integrators are triggered shortly before plasma initiation when the toroidal field coil current is close to its maximum. In this way, the integration time can be chosen to measure only the small changes in flux due to the presence of plasma. Two identical plasma discharges with positive and negative magnetic field have shown that the alignment error with respect to the plasma current is negligible. The measured diamagnetic flux is compared to that predicted by TRANSP simulations. The poloidal beta inferred from the diamagnetic flux measurement is compared to the values calculated from magnetic equilibrium reconstruction codes. The diamagnetic flux measurement and TRANSP simulation can be used together to estimate the coupled power in discharges with dominant ion cyclotron resonance heating.

The deuterium inventory in ASDEX Upgrade

International Nuclear Information System (INIS)

Mayer, M.; Rohde, V.; Ramos, G; Vainonen-Ahlgren, E.; Likonen, J.; Herrmann, A.; Neu, R.

2007-01-01

The deuterium inventory in ASDEX Upgrade was determined by quantitative ion beam analysis techniques and SIMS for different discharge campaigns between the years 2002 and 2005. ASDEX Upgrade was a carbon dominated machine during this phase. Full poloidal sections of the lower and upper divertor tile surfaces, limiter tiles, gaps between divertor tiles, gaps between inner heat shield tiles and samples from remote areas below the roof baffle and in pump ducts were analysed, thus offering an exhaustive survey of all relevant areas in ASDEX Upgrade. Deuterium is mainly trapped on plasma-exposed surfaces of inner divertor tiles, where about 70% of the retained deuterium inventory is found. About 20% of the inventory is retained at or below the divertor roof baffle, and about 10% is observed in other areas, such as the outer divertor and in gaps between tiles. The long term deuterium retention is 3-4% of the total deuterium input. The obtained results are compared with gas balance measurements, and conclusions about tritium retention in ITER are made

Amplitude based feedback control for NTM stabilisation at ASDEX Upgrade

Energy Technology Data Exchange (ETDEWEB)

Rapson, Christopher, E-mail: chris.rapson@ipp.mpg.de; Giannone, Louis; Maraschek, Marc; Reich, Matthias; Stober, Joerg; Treutterer, Wolfgang

2014-05-15

Highlights: • Two algorithms have been developed which use the NTM amplitude to control ECCD deposition and stabilise NTMs. • Both algorithms were tested and tuned in a simulation of the full feedback loop including an MRE. • Both algorithms have been successfully deployed in ASDEX Upgrade experiments. • Use of the NTM amplitude adds considerable robustness, which is necessary when trying to target ECCD to within 1 cm of the island location. • This is part of ongoing work to reliably and quickly stabilise NTMs in any plasma scenario. - Abstract: Neoclassical Tearing Modes (NTMs) degrade the confinement in tokamak plasmas at high beta, placing a major limitation on the projected fusion performance. Furthermore, NTMs can lead to disruptions with even more severe consequences. Therefore methods to stabilise NTMs are being developed with high priority at several research institutes worldwide. The favoured method is to deposit Electron Cyclotron Current Drive (ECCD) precisely at the mode location by controlling a movable mirror in the ECCD launcher. This method requires both the mode location and the deposition location to be known with high accuracy in real time. The required accuracy is given by half of the marginal island width, or approximately 1 cm for a m/n = 3/2 NTM at ASDEX Upgrade. Despite considerable development on a range of diagnostics, it remains challenging to provide the necessary accuracy reliably and in real time. To relax the accuracy requirements and add robustness, the feedback controller can additionally consider the effect of ECCD on the NTM amplitude directly. Then the optimal deposition location is simply where the NTM amplitude is minimised. The simplest implementation sweeps the ECCD beam across the expected NTM location. After the sweep, the beam can be returned to the optimal location and held there to stabilise the NTM. Unfortunately, waiting for a full sweep takes too long. Therefore a second method assesses the NTM growth every

Experimental studies of high-confinement mode plasma response to non-axisymmetric magnetic perturbations in ASDEX Upgrade

Science.gov (United States)

Suttrop, W.; Kirk, A.; Nazikian, R.; Leuthold, N.; Strumberger, E.; Willensdorfer, M.; Cavedon, M.; Dunne, M.; Fischer, R.; Fietz, S.; Fuchs, J. C.; Liu, Y. Q.; McDermott, R. M.; Orain, F.; Ryan, D. A.; Viezzer, E.; The ASDEX Upgrade Team; The DIII-D Team; The Eurofusion MST1 Team

2017-01-01

The interaction of externally applied small non-axisymmetric magnetic perturbations (MP) with tokamak high-confinement mode (H-mode) plasmas is reviewed and illustrated by recent experiments in ASDEX Upgrade. The plasma response to the vacuum MP field is amplified by stable ideal kink modes with low toroidal mode number n driven by the H-mode edge pressure gradient (and associated bootstrap current) which is experimentally evidenced by an observable shift of the poloidal mode number m away from field alignment (m  =  qn, with q being the safety factor) at the response maximum. A torque scan experiment demonstrates the importance of the perpendicular electron flow for shielding of the resonant magnetic perturbation, as expected from a two-fluid MHD picture. Two significant effects of MP occur in H-mode plasmas at low pedestal collisionality, ν \\text{ped}\\ast≤slant 0.4 : (a) a reduction of the global plasma density by up to 61 % and (b) a reduction of the energy loss associated with edge localised modes (ELMs) by a factor of up to 9. A comprehensive database of ELM mitigation pulses at low {ν\\ast} in ASDEX Upgrade shows that the degree of ELM mitigation correlates with the reduction of pedestal pressure which in turn is limited and defined by the onset of ELMs, i. e. a modification of the ELM stability limit by the magnetic perturbation.

Surface composition and topography of the graphite limiter in ASDEX

International Nuclear Information System (INIS)

Behrisch, R.; Boergesen, P.; Ehrenberg, J.; Scherzer, B.M.U.; Sawicka, B.D.; Sawicki, J.A.

1984-01-01

After having been in use during about 150 discharges the graphite limiter of the ASDEX tokamak was removed and analyzed by means of SEM, EIXE, PIXE, RBS and Moessbauer Spectroscopy. The surface was seen to be covered by a nonuniform layer of 10 16 -10 18 metal at/cm 2 , primarily in the form of droplets with diameters up to a few μm. Simulation experiments suggest that the formation of these droplets is caused by heating. The relative metal concentrations in the droplets are close to those for stainless steel. The Moessbauer spectroscopy analysis indicates, however, that the major part of the metal is present in carbide phases, most probably (Fe, Cr, Ni) 3 C. On parts of the limiter the metal coverage shows a sharp maximum at the points nearest the main plasma, flanked by two minima. The minimum on the ion drift side is much broader than the one on the electron drift side. However, these features are not common to all parts of the limiter. (orig.)

A phased array antenna for Doppler reflectometry in ASDEX upgrade

Energy Technology Data Exchange (ETDEWEB)

Wolf, Stefan; Lechte, Carsten; Kasparek, Walter [IGVP, Universitaet Stuttgart, D-70569 Stuttgart (Germany); Hennequin, Pascale [Laboratoire de Physique des Plasmas, CNRS, Ecole Polytech., F-91128 Palaiseau (France); Conway, Garrard; Happel, Tim [Max-Planck-Institut fuer Plasmaphysik, D-85748 Garching (Germany); Collaboration: ASDEX Upgrade Team

2016-07-01

In a toroidal plasma, Doppler reflectometry (DR) allows investigating electron density fluctuations with finite k {sub perpendicular} {sub to}. The injected microwave beam's frequency determines the radial position of the probed region, its tilt angle selects the wavenumber satisfying the Bragg condition for backscattering. The rotation velocity can be calculated from the Doppler shift of the backscattered signal's frequency. By varying the injected frequency, radial profiles can be reconstructed. Varying the tilt angle resolves the k {sub perpendicular} {sub to} -spectrum of the fluctuations. For DR, a pair of phased array antennas (PAAs) has been designed, built, and installed in the ASDEX Upgrade tokamak. Beam steering is done by slightly changing the injected frequency, thus, the PAAs do not need any movable parts or electronics inside the vacuum vessel. From 75 to 105 GHz, the PAAs feature 13 frequency bands, each with an angular scan range of -20 to +20 {sup circle}. So, for each angle, there are 13 radial positions to be probed. The results from PAA characterisation, commissioning, and first DR measurements are presented.

Fast-ion losses induced by ELMs and externally applied magnetic perturbations in the ASDEX Upgrade tokamak

International Nuclear Information System (INIS)

Garcia-Munoz, M; Rodriguez-Ramos, M; Äkäslompolo, S; De Marne, P; Dunne, M G; Dux, R; Fietz, S; Fuchs, C; Geiger, B; Herrmann, A; Hoelzl, M; Kurzan, B; McDermott, R M; Strumberger, E; Evans, T E; Ferraro, N M; Pace, D C; Lazanyi, N; Nocente, M; Shinohara, K

2013-01-01

Phase-space time-resolved measurements of fast-ion losses induced by edge localized modes (ELMs) and ELM mitigation coils have been obtained in the ASDEX Upgrade tokamak by means of multiple fast-ion loss detectors (FILDs). Filament-like bursts of fast-ion losses are measured during ELMs by several FILDs at different toroidal and poloidal positions. Externally applied magnetic perturbations (MPs) have little effect on plasma profiles, including fast-ions, in high collisionality plasmas with mitigated ELMs. A strong impact on plasma density, rotation and fast-ions is observed, however, in low density/collisionality and q 95 plasmas with externally applied MPs. During the mitigation/suppression of type-I ELMs by externally applied MPs, the large fast-ion bursts observed during ELMs are replaced by a steady loss of fast-ions with a broad-band frequency and an amplitude of up to an order of magnitude higher than the neutral beam injection (NBI) prompt loss signal without MPs. Multiple FILD measurements at different positions, indicate that the fast-ion losses due to static 3D fields are localized on certain parts of the first wall rather than being toroidally/poloidally homogeneously distributed. Measured fast-ion losses show a broad energy and pitch-angle range and are typically on banana orbits that explore the entire pedestal/scrape-off-layer (SOL). Infra-red measurements are used to estimate the heat load associated with the MP-induced fast-ion losses. The heat load on the FILD detector head and surrounding wall can be up to six times higher with MPs than without 3D fields. When 3D fields are applied and density pump-out is observed, an enhancement of the fast-ion content in the plasma is typically measured by fast-ion D-alpha (FIDA) spectroscopy. The lower density during the MP phase also leads to a deeper beam deposition with an inward radial displacement of ≈2 cm in the maximum of the beam emission. Orbit simulations are used to test different models for 3D

Correlation electron cyclotron emission diagnostic and improved calculation of turbulent temperature fluctuation levels on ASDEX Upgrade

Science.gov (United States)

Creely, A. J.; Freethy, S. J.; Burke, W. M.; Conway, G. D.; Leccacorvi, R.; Parkin, W. C.; Terry, D. R.; White, A. E.

2018-05-01

A newly upgraded correlation electron cyclotron emission (CECE) diagnostic has been installed on the ASDEX Upgrade tokamak and has begun to perform experimental measurements of electron temperature fluctuations. CECE diagnostics measure small amplitude electron temperature fluctuations by correlating closely spaced heterodyne radiometer channels. This upgrade expanded the system from six channels to thirty, allowing simultaneous measurement of fluctuation level radial profiles without repeat discharges, as well as opening up the possibility of measuring radial turbulent correlation lengths. Newly refined statistical techniques have been developed in order to accurately analyze the fluctuation data collected from the CECE system. This paper presents the hardware upgrades for this system and the analysis techniques used to interpret the raw data, as well as measurements of fluctuation spectra and fluctuation level radial profiles.

ICRF power limitation relation to density limit in ASDEX

International Nuclear Information System (INIS)

Ryter, F.

1992-01-01

Launching high ICRF power into ASDEX plasmas required good antenna-plasma coupling. This could be achieved by sufficient electron density in front of the antennas i.e. small antenna-plasma distance (1-2 cm) and moderate to high line-averaged electron density compared to the density window in ASDEX. These are conditions eventually close to the density limit. ICRF heated discharges terminated by plasma disruptions caused by the RF pulse limited the maximum RF power which can be injected into the plasma. The disruptions occurring in these cases have clear phenomenological similarities with those observed in density limit discharges. We show in this paper that the ICRF-power limitation by plasma disruptions in ASDEX was due to reaching the density limit. (orig.)

ICRF power limitation relation to density limit in ASDEX

International Nuclear Information System (INIS)

Ryter, F.

1992-01-01

Launching high ICRF power into ASDEX plasmas required good antenna-plasma coupling. This could be achieved by sufficient electron density in front of the antennas i.e. small antenna-plasma distance (1-2 cm) and moderate to high line-averaged electron density compared to the density window in ASDEX. These are conditions eventually close to the density limit. ICRF heated discharges terminated by plasma disruptions caused by the RF pulse limited the maximum RF power which can be injected into the plasma. The disruptions occurring in these cases have clear phenomenological similarities with those observed in density limit discharges. We show in this paper that the ICRF-power limitation by plasma disruptions in ASDEX was due to reaching the density limit. (author) 3 refs., 3 figs

Characterisation of the core poloidal flow at ASDEX upgrade

Energy Technology Data Exchange (ETDEWEB)

Lebschy, Alexander [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstr. 2, D-85748 Garching (Germany); Physik-Department E28, Technische Universitaet Muenchen, D-85748 Garching (Germany); McDermott, Rachael M.; Geiger, Benedikt; Cavedon, Marco; Dunne, Michael G.; Dux, Ralph; Fischer, Rainer; Kappatou, Athina; McCarthy, Patrick J.; Viezzer, Eleonora [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstr. 2, D-85748 Garching (Germany); Collaboration: the ASDEX Upgrade Team

2016-07-01

Plasma rotation has a strong influence on the transport of heat, particles, and momentum in fusion plasmas via a variety of mechanisms, for example, by the stabilization of modes and the suppression of plasma turbulence. In tokamaks, the toroidal rotation (u{sub tor}) is essentially a free parameter that is usually dominated by the external momentum input from neutral beams used to heat the plasma. The poloidal rotation (u{sub pol}), on the other hand, is strongly damped and is predicted to remain at Neoclassical (NC) levels of a few km/s. Measuring the inboard-outboard asymmetry of u{sub tor} with charge exchange recombination spectroscopy enables an indirect measurement of u{sub pol} and, hence, the measurement of the complete plasma flow on a flux surface. In order to characterise the nature of u{sub pol} at ASDEX Upgrade a poloidal rotation database has been built that contains a large variation in the parameters that, according to NC theory, drive u{sub pol}; namely, the main ion temperature and density gradients and collisionality. Initial results from this database and a detailed comparison of u{sub pol} to NC theory in interesting plasma scenarios, are presented in this poster.

Direct measurements of the plasma potential in ELMy H-mode plasma with ball-pen probes on ASDEX Upgrade tokamak

Czech Academy of Sciences Publication Activity Database

Adámek, Jiří; Stöckel, Jan; Brotánková, Jana; Horáček, Jan; Rohde, V.; Müller, H. W.; Herrmann, A.; Schrittwieser, R.; Mehlmann, F.; Ionita, C.

390-391, - (2009), s. 1114-1117 ISSN 0022-3115. [International Conference on Plasma-Surface Interactions in Controlled Fusion Device/18th./. Toledo, 26.05.2008-30.05.2008] R&D Projects: GA AV ČR KJB100430601 Institutional research plan: CEZ:AV0Z20430508 Keywords : Edge plasma * Electric field * ELMs * H-mode * ASDEX-Upgrade Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.933, year: 2009 http://dx.doi.org/10.1016/j.jnucmat.2009.01.286

Density profile evolution during dynamic processes in ASDEX Upgrade

International Nuclear Information System (INIS)

Nunes, I.; Santos, J.; Salzedas, F.; Manso, M.; Serra, F.; Conway, G.D.; Horton, L.D.; Neuhauser, J.; Suttrop, W.

2005-01-01

The current understanding of edge localized modes (ELMs) and the trigger of major disruptions is largely based on phenomenology. The need to better understand the processes underlying these phenomena requires high temporal and spatial resolution diagnostics. Fast diagnostics for the temperature measurements exist, such as the ECE radiometer but, for the plasma density, the existing diagnostics such as Lithium Beam and Thomson Scattering do not have the required high temporal resolution for a period long enough to characterize the entire ELM event. The microwave reflectometry system on ASDEX Upgrade has the capability to measure electron density profiles simultaneously at the low-field and high-field sides, in broadband swept ultrafast (35μs) operation with a spatial resolution of 5mm. In this paper we report on recent results on the effects of type I ELMs on density profiles and on the density pedestal width and ELM affected depth. During the ELM event, three phases are identified: precursor, collapse and recovery. The density pedestal width is found to be approximately constant for all the ELMy H-mode discharges analyzed here, except for high input power discharges, where an increase of the density pedestal width is observed. Major disruptions limit the range of parameters used in the operation of a tokamak, especially density limit disruptions, that limit the maximum usable density. Very abrupt increases of density are observed before the onset of the electron temperature profile erosion, supporting the hypothesis that this erosion is due to convection of the magnetic field. In ITER, during the long steady state flat-top phase of the discharges magnetic measurements may accumulate significant drifts. Plasma position and shape control using reflectometry is being assessed in ASDEX Upgrade for ITER like scenarios with successful results, where it is shown that position measurements from reflectometry compared to magnetic data satisfy the ITER requirements

The H-mode of ASDEX

International Nuclear Information System (INIS)

1989-01-01

The paper is a review of investigations of the H-mode on ASDEX performed since its discovery in 1982. The topics discussed are: (1) the development of the plasma profiles, with steep gradients in the edge region and flat profiles in the bulk plasma, (2) the MHD properties resulting from the profile changes, including an extensive stability analysis, (3) the impurity development, with special emphasis on the MHD aspects and on neoclassical impurity transport effects in quiescent H-phases, and (4) the properties of the edge plasma, including the evidence of three-dimensional distortions at the edge. The part on confinement includes scaling studies and the results of transport analysis. The power threshold of the H-mode is found to depend weakly on the density, but there is probably no dependence on the toroidal field or the current. For the operational range of the H-mode, new results for the limiter H-mode on ASDEX and the development of the H-mode under beam current drive conditions are included. A number of experiments are described which demonstrate the crucial role of the edge electron temperature in the L-H transition. New results of magnetic and density fluctuation studies at the plasma edge within the edge transport barrier are presented. Finally, the findings on ASDEX are compared with results obtained on other machines and are used to test various H-mode theories. (author). 131 refs, 103 figs, 1 tab

Stochastic sawtooth reconnection in ASDEX Upgrade

International Nuclear Information System (INIS)

Igochine, V.; Dumbrajs, O.; Zohm, H.; Flaws, A.

2007-01-01

In this paper we investigate non-complete sawtooth reconnection in the ASDEX Upgrade tokamak. Such reconnection phenomena are associated with internal m/n = 1/1 kink mode which does not vanish after the crash phase (as would be the case for complete reconnection). It is shown that this sawtooth cannot be fully described by pure m/n = 1/1 mode and that higher harmonics play an important role during the sawtooth crash phase. We employ the Hamiltonian formalism and reconstructed perturbations to model incomplete sawtooth reconnection. It is demonstrated that stochastization appears due to the excitation of low-order resonances which are present in the corresponding q-profiles inside the q = 1 surface which reflects the key role of the q 0 value. Depending on this value two completely different situations are possible for one and the same mode perturbations: (i) the resonant surfaces are present in the q-profile leading to stochasticity and sawtooth crash (q 0 ∼ 0.7 ± 0.1); (ii) the resonant surfaces are not present, which means no stochasticity in the system and no crash event (q 0 ∼ 0.9 ± 0.05). Accordingly the central safety factor value is always less than unity in the case of a non-complete sawtooth reconnection. Our investigations show that the stochastic model agrees well with the experimental observations and can be proposed as a promising candidate for an explanation of the sawtooth reconnection

Multi-channel Langmuir-probe and H[alpha]-measurements of edge fluctuations on ASDEX

Energy Technology Data Exchange (ETDEWEB)

Niedermeyer, H; Carlson, A; Endler, M; Giannone, L.; Rudyj, A; Theimer, G [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

1991-01-01

The anomalous transport observed in tokamaks is caused by turbulent fluctuations, the nature of which is still poorly understood. Diagnostic difficulties are one major reason for this lack of understanding, at least with respect to the bulk plasma. The plasma edge, however, is accessible by several diagnostics permitting localized measurements of different parameters with good spatial and temporal resolution. For this reason one can hope to obtain enough information about edge fluctuations to permit the development of theoretical models. Different ranges of plasma parameters and the lack of closed magnetic surfaces distinguish this plasma zone from the bulk plasma. Edge turbulence might well involve other mechanisms than the turbulence in the bulk. Although transport in the bulk plasma receives more attention transport in the edge plasma and edge physics are very relevant for reactor design. The realistic hope to find a solution and the importance of the problem for the next step in fusion research are reasons for the strong effort in this field on many tokamaks. Like in many limiter tokamaks Langmuir probes were used in the ASDEX divertor device for measurements of the floating potential and of the ion saturation current. Under certain assumptions the electron density and the plasma potential can be derived from these data. Observation of the H[alpha]-light emitted from the edge in the vicinity of a neutral gas source yields information about the electron density. While probe measurements are more suitable for quantitative evaluations including the calculation of the local particle flux the H[alpha]-method is not calibrated and integrates radially over the edge. It is however applicable in situations where probes fail because of excessive heat load. With 16-channel arrays both methods permit spatial correlations and wavenumber spectra to be determined without any further assumptions. (author) 4 refs., 2 figs.

Direct measurements of the plasma potential in ELMy H-mode plasma with ball-pen probes on ASDEX Upgrade tokamak

Energy Technology Data Exchange (ETDEWEB)

Adamek, J., E-mail: adamek@ipp.cas.c [Institute of Plasma Physics, Association EURATOM/IPP.CR, Prague, Za Slovankou 3, 182 00, Prague 8 (Czech Republic); Rohde, V.; Mueller, H.W.; Herrmann, A. [Institute of Plasma Physics, Association EURATOM/IPP, Garching (Germany); Ionita, C.; Schrittwieser, R.; Mehlmann, F. [Institute for Ion Physics and Applied Physics, University of Innsbruck, Association EURATOM/OAW (Austria); Stoeckel, J.; Horacek, J.; Brotankova, J. [Institute of Plasma Physics, Association EURATOM/IPP.CR, Prague, Za Slovankou 3, 182 00, Prague 8 (Czech Republic)

2009-06-15

Experimental investigations of the plasma potential and electric field were performed for ELMy H-mode plasmas in the vicinity of the limiter shadow of ASDEX Upgrade. A fast reciprocating probe with a probe head containing four ball-pen probes (BPPs) [J. Adamek et al., Czech. J. Phys. 54 (2004) C95 - C99.] was used on the midplane manipulator. Different gradients of the plasma potential were observed during ELMs and in between them. The temporal resolution of the direct plasma potential measurements with BPP was determined by using power-spectra analysis.

Development of a flexible Doppler reflectometry system and its application to turbulence characterization in the ASDEX Upgrade tokamak

Energy Technology Data Exchange (ETDEWEB)

Troester, Carolin Helma

2008-04-15

An essential challenge in present fusion plasma research is the study of plasma turbulence. The turbulence behavior is investigated experimentally on the ASDEX Upgrade tokamak using Doppler reflectometry, a diagnostic technique sensitive to density fluctuations at a specific wavenumber k {sub perpendicular} {sub to}. This microwave radar diagnostic utilizes localized Bragg backscattering of the launched beam (k{sub 0}) by the density fluctuations at the plasma cutoff layer. The incident angle {theta} selects the probed k {sub perpendicular} {sub to} via the Bragg condition k {sub perpendicular} {sub to} {approx} 2k{sub 0}sin{theta}. The measured Doppler shifted frequency spectrum allows the determination of the perpendicular plasma rotation velocity, u {sub perpendicular} {sub to} =v{sub E} {sub x} {sub B}+v{sub turb}, directly from the Doppler frequency shift(f{sub D} = u {sub perpendicular} {sub to} k {sub perpendicular} {sub to} /2{pi}), and the turbulence amplitude from the backscattered power level. This thesis work presents a survey of u {sub perpendicular} {sub to} radial profiles and k {sub perpendicular} {sub to} spectrum measurements for a variety of plasma conditions obtained by scanning the antenna tilt angle. This was achieved by extending the existing V-band Doppler reflectometry system (50 - 75 GHz) with a new W-band system (75 - 110 GHz), which was especially designed for measuring the k {sub perpendicular} {sub to} spectrum and additionally expands the radial coverage into the plasma core region. It consists of a remote steerable antenna with an adjustable line of sight allowing for dynamic wavenumber selection up to 25 cm {sup -1} and a reflectometer with a 'phase locked loop' stabilized transmitter allowing for the precise determination of the instrument response function. The proper system functionality was demonstrated by laboratory testing and benckmarking against the V-band system. The new profile measurements obtained show a

Feedback-controlled NTM stabilization on ASDEX Upgrade

Directory of Open Access Journals (Sweden)

Stober J.

2015-01-01

Full Text Available On ASDEX Upgrade a concept for real-time stabilization of NTMs has been realized and successfully applied to (3,2- and (2,1-NTMs. Since most of the work has meanwhile been published elsewhere, a short summary with the appropriate references is given. Limitations, deficits and future extensions of the system are discussed. In a second part the recent work on using modulated ECCD for NTM stabilisation is described in some detail. In these experiments ECCD power is modulated according to a magnetic footprint of the rotating NTM. In agreement with earlier results it could be shown that O-point heating reduces the necessary average power for stabilisation whereas X-point heating hampers stabilisation. Although this modulated scheme is not relevant for routine NTM stabilisation on ASDEX Upgrade it may be mandatory for ITER or DEMO. On ASDEX Upgrade it has been re-developed to demonstrate the usage of a FAst DIrectional Switch to continously heat the O-point of the rotating island with only one gyrotron switching between two launchers which target the mode at locations separated in phase by 180 degrees as described in [1].

Proposal of an alternative upper divertor in ASDEX Upgrade supported by EMC3-EIRENE simulations

Directory of Open Access Journals (Sweden)

T. Lunt

2017-08-01

Full Text Available We discuss the benefits of installing a pair of in-vessel coils with currents |Ifx| ≲ 50 kAt in the upper divertor of ASDEX Upgrade (AUG to study a series of ‘alternative’ divertor configurations, like the Snowflake (SF and the X-divertor (XD, that are currently considered as alternative solutions for the power exhaust problem. The possibility of operating the standard lower single-null (SN and double-null (DN would be preserved. Potential effects to reduce the peak parallel- and/or perpendicular heat flux are predicted from a simple geometrical-diffusive model as well as by numerical EMC3-EIRENE simulations for pure deuterium attached conditions with spatially constant diffusion coefficients. Beyond that a series of other potential transport- and radiation related heat flux mitigation effects are identified and could be studied experimentally with the modified upper divertor in the high-power divertor Tokamak AUG.

Electron heat transport studies using transient phenomena in ASDEX Upgrade

International Nuclear Information System (INIS)

Jacchia, A.; Angioni, C.; Manini, A.; Ryter, F.; Apostoliceanu, M.; Conway, G.; Fahrbach, H.-U.; Kirov, K.K.; Leuterer, F.; Reich, M.; Sutttrop, W.; Cirant, S.; Mantica, P.; De Luca, F.; Weiland, J.

2005-01-01

Experiments in tokamaks suggest that a critical gradient length may cause the resilient behavior of T e profiles, in the absence of ITBs. This agrees in general with ITG/TEM turbulence physics. Experiments in ASDEX Upgrade using modulation techniques with ECH and/or cold pulses demonstrate the existence of a threshold in R/L Te when T e >T i and T e ≤T i . For T e >T i linear stability analyses indicate that electron heat transport is dominated by TEM modes. They agree in the value of the threshold (both T e and n e ) and for the electron heat transport increase above the threshold. The stabilization of TEM modes by collisions yielded by gyro-kinetic calculations, which suggests a transition from TEM to ITG dominated transport at high collisionality, is experimentally demonstrated by comparing heat pulse and steady-state diffusivities. For the T e ∼T i discharges above the threshold the resilience, normalized by T e 3/2 , is similar to that of the TEM dominated cases, despite very different conditions. The heat pinch predicted by fluid modeling of ITG/TEM turbulence is investigated by perturbative transport in off-axis ECH-heated discharges. (author)

H-mode transition physics close to double null on MAST and its applications to other tokamaks

International Nuclear Information System (INIS)

Meyer, H.; Carolan, P.G.; Cunningham, G.; Kirk, A.; Lloyd, B.; Saarelma, S.; Wilson, H.R.; Conway, G.D.; Horton, L.D.; Ryter, F.; Schirmer, J.; Suttrop, W.; Maingi, R.

2005-01-01

By accessing extreme parameter regimes combined with well diagnosed edge MAST data contribute towards the understanding of H-mode physics. The first inter-machine comparisons with respect to the influence of the magnetic topology on the power threshold with ASDEX Upgrade and NSTX reveal a reduction of the power threshold in true double null (C-DN) configuration opening new operation regimes in both devices. In L-mode, the negative radial electric field close to the separatrix was found to be more negative in C-DN than in single null (SN), whilst most of the other edge parameters are similar. Pedestal temperatures in MAST are lower than in ASDEX Upgrade in MAST-equivalent discharges, whereas the pedestal densities can be similar, although in long inter ELM periods the MAST density pedestal is higher than on ASDEX Upgrade. In order to test four leading H-mode theories MAST data are compared statistically to their H-mode access criteria. The usual DN operating regime with co current NBI in MAST has been extended to include single null (SN) configurations, to provide more direct comparisons with conventional tokamaks. The plasma edge in SN on MAST is more stable to ELMs and the typical type-III ELMs, often observed in C-DN, are absent, despite input powers close to the H-mode threshold power. In this respect, the stability of measured plasma edge profiles in SN and DN against ideal peeling-ballooning modes will be discussed. (author)

Neutral beam power measurements inside the ASDEX torus

International Nuclear Information System (INIS)

Zengliang, Y.; Staebler, A.; Vollmer, O.

1982-11-01

Neutral beam power measurements inside the ASDEX torus are done with a retractable calorimeter which is only radiation cooled. The calorimeter plate made from Molybdenum is subdivided into nine segments whose increase in energy content due to a shot yields the absorbed beam power. Different models for the backward extrapolation of the measured temperature curves are examined for a series of low energy shots with the result that pure radiation cooling is a valid assumption. Furthermore, a temperature correction to the measured power is derived from these experiments. The evaluation of the shots onto this calorimeter is done by a computer program. The application of this program to a few full power shots shows that a neutral power up to 3.2 MW has been injected into the ASDEX vessel by the two injectors with an overall efficiency of up to 40%. Reionization losses due to the ASDEX stray field are less than 10%; they do not show any dependence upon the pulse length for shots up to 200 ms. (orig.)

Overview of ASDEX Upgrade results

NARCIS (Netherlands)

Kallenbach, A.; Adamek, J.; Aho-Mantila, L.; Akaslompolo, S.; Angioni, C.; Atanasiu, C. V.; Balden, M.; Behler, K.; Belonohy, E.; Bergmann, A.; Bernert, M.; Bilato, R.; Bobkov, V.; Boom, J.; Bottino, A.; Braun, F.; Brudgam, M.; Buhler, A.; Burckhart, A.; Chankin, A.; Classen, I.G.J.; Conway, G. D.; Coster, D. P.; de Marne, P.; D' Inca, R.; Drube, R.; Dux, R.; Eich, T.; Endstrasser, N.; Engelhardt, K.; Esposito, B.; Fable, E.; Fahrbach, H. U.; Fattorini, L.; Fischer, R.; Flaws, A.; Funfgelder, H.; Fuchs, J. C.; Gal, K.; Munoz, M. G.; Geiger, B.; Adamov, M. G.; Giannone, L.; Giroud, C.; Gorler, T.; da Graca, S.; Greuner, H.; Gruber, O.; Gude, A.; Gunter, S.; Haas, G.; Hakola, A. H.; Hangan, D.; Happel, T.; Hauff, T.; Heinemann, B.; Herrmann, A.; Hicks, N.; Hobirk, J.; Hohnle, H.; Holzl, M.; Hopf, C.; Horton, L.; Huart, M.; Igochine, V.; Ionita, C.; Janzer, A.; Jenko, F.; Kasemann, C. P.; Kalvin, S.; Kardaun, O.; Kaufmann, M.; Kirk, A.; Klingshirn, H. J.; Kocan, M.; Kocsis, G.; Kollotzek, H.; Konz, C.; Koslowski, R.; Krieger, K.; Kurki-Suonio, T.; Kurzan, B.; Lackner, K.; Lang, P. T.; Lauber, P.; Laux, M.; Leipold, F.; Leuterer, F.; Lohs, A.; N C Luhmann Jr.,; Lunt, T.; Lyssoivan, A.; Maier, H.; Maggi, C.; Mank, K.; Manso, M. E.; Maraschek, M.; Martin, P.; Mayer, M.; McCarthy, P. J.; McDermott, R.; Meister, H.; Menchero, L.; Meo, F.; Merkel, P.; Merkel, R.; Mertens, V.; Merz, F.; Mlynek, A.; Monaco, F.; Muller, H. W.; Munich, M.; Murmann, H.; Neu, G.; Neu, R.; Nold, B.; Noterdaeme, J. M.; Park, H. K.; Pautasso, G.; Pereverzev, G.; Podoba, Y.; Pompon, F.; Poli, E.; Polochiy, K.; Potzel, S.; Prechtl, M.; Puschel, M. J.; Putterich, T.; Rathgeber, S. K.; Raupp, G.; Reich, M.; Reiter, B.; Ribeiro, T.; Riedl, R.; Rohde, V.; Roth, J.; Rott, M.; Ryter, F.; Sandmann, W.; Santos, J.; Sassenberg, K.; Sauter, P.; Scarabosio, A.; Schall, G.; Schmid, K.; Schneider, P. A.; Schneider, W.; Schramm, G.; Schrittwieser, R.; Schweinzer, J.; Scott, B.; Sempf, M.; Serra, F.; Sertoli, M.; Siccinio, M.; Sigalov, A.; Silva, A.; Sips, A.C.C.; Sommer, F.; Stabler, A.; Stober, J.; Streibl, B.; Strumberger, E.; Sugiyama, K.; Suttrop, W.; Szepesi, T.; Tardini, G.; Tichmann, C.; Told, D.; Treutterer, W.; Urso, L.; Varela, P.; Vincente, J.; Vianello, N.; Vierle, T.; Viezzer, E.; Vorpahl, C.; Wagner, D.; Weller, A.; Wenninger, R.; Wieland, B.; Wigger, C.; Willensdorfer, M.; Wischmeier, M.; Wolfrum, E.; Wursching, E.; Yadikin, D.; Yu, Q.; Zammuto, I.; Zasche, D.; Zehetbauer, T.; Zhang, Y.; Zilker, M.; Zohm, H.

2011-01-01

The ASDEX Upgrade programme is directed towards physics input to critical elements of the ITER design and the preparation of ITER operation, as well as addressing physics issues for a future DEMO design. After the finalization of the tungsten coating of the plasma facing components, the

Overview of ASDEX Upgrade results

DEFF Research Database (Denmark)

Kallenbach, A.; Adamek, J.; Aho-Mantila, L.

2011-01-01

The ASDEX Upgrade programme is directed towards physics input to critical elements of the ITER design and the preparation of ITER operation, as well as addressing physics issues for a future DEMO design. After the finalization of the tungsten coating of the plasma facing components, the re-availa...

Overview of ASDEX Upgrade results

DEFF Research Database (Denmark)

Kallenbach, A.; Aguiam, D.; Aho-Mantila, L.

2017-01-01

The ASDEX Upgrade (AUG) programme is directed towards physics input to critical elements of the ITER design and the preparation of ITER operation, as well as addressing physics issues for a future DEMO design. Since 2015, AUG is equipped with a new pair of 3-strap ICRF antennas, which were design...

Status of the COMPASS tokamak and characterization of the first H-mode

Science.gov (United States)

Pánek, R.; Adámek, J.; Aftanas, M.; Bílková, P.; Böhm, P.; Brochard, F.; Cahyna, P.; Cavalier, J.; Dejarnac, R.; Dimitrova, M.; Grover, O.; Harrison, J.; Háček, P.; Havlíček, J.; Havránek, A.; Horáček, J.; Hron, M.; Imríšek, M.; Janky, F.; Kirk, A.; Komm, M.; Kovařík, K.; Krbec, J.; Kripner, L.; Markovič, T.; Mitošinková, K.; Mlynář, J.; Naydenkova, D.; Peterka, M.; Seidl, J.; Stöckel, J.; Štefániková, E.; Tomeš, M.; Urban, J.; Vondráček, P.; Varavin, M.; Varju, J.; Weinzettl, V.; Zajac, J.; the COMPASS Team

2016-01-01

This paper summarizes the status of the COMPASS tokamak, its comprehensive diagnostic equipment and plasma scenarios as a baseline for the future studies. The former COMPASS-D tokamak was in operation at UKAEA Culham, UK in 1992-2002. Later, the device was transferred to the Institute of Plasma Physics of the Academy of Sciences of the Czech Republic (IPP AS CR), where it was installed during 2006-2011. Since 2012 the device has been in a full operation with Type-I and Type-III ELMy H-modes as a base scenario. This enables together with the ITER-like plasma shape and flexible NBI heating system (two injectors enabling co- or balanced injection) to perform ITER relevant studies in different parameter range to the other tokamaks (ASDEX-Upgrade, DIII-D, JET) and to contribute to the ITER scallings. In addition to the description of the device, current status and the main diagnostic equipment, the paper focuses on the characterization of the Ohmic as well as NBI-assisted H-modes. Moreover, Edge Localized Modes (ELMs) are categorized based on their frequency dependence on power density flowing across separatrix. The filamentary structure of ELMs is studied and the parallel heat flux in individual filaments is measured by probes on the outer mid-plane and in the divertor. The measurements are supported by observation of ELM and inter-ELM filaments by an ultra-fast camera.

Investigation of the impurity transport in the ASDEX tokamak by spectroscopical methods

International Nuclear Information System (INIS)

Krieger, K.W.

1990-12-01

Plasma impurities: a central problem of controlled thermonuclear fusion; magnetic plasma confinement in a Tokamak; methods to the determination of plasma impurity transport coefficients - by temporally modulated gas admission; the transport equation for impurities; neoclassical and anomalous transport; harmonic analysis of time-dependent signals; solutions of the transport equation; experimental equipment and measurements; measuring results - consistency of simple transport models with radial phase measurements; linearity of the transport processes; plasma disturbance by impurity injection; determination of the diffusion coefficient by simplified transport models; comparison of transport models for impurities and background plasma; measurements of the impurity transport at the plasma edge by high modulation frequencies. (AH)

Actuator management for ECRH at ASDEX Upgrade

Energy Technology Data Exchange (ETDEWEB)

Rapson, Christopher J., E-mail: chris.rapson@ipp.mpg.de; Reich, Matthias; Stober, Joerg; Treutterer, Wolfgang

2015-10-15

Highlights: • Real-time actuator management algorithm developed for ECRH at ASDEX Upgrade. • First use of a control hierarchy in a fusion experiment. • Cost function evaluates optimal combination of all gyrotrons to all possible targets. • Considers many factors e.g. mirror movement, power available, presence and mode number of NTMs. • Configurable, robust algorithm is ready for online testing. - Abstract: Automated actuator management will be necessary on long pulse fusion experiments to adjust to unforeseen plasma events and unpredictable actuator availability. However, as a control problem, actuator management is underdeveloped in the fusion community. This contribution proposes an algorithm based on a control hierarchy and a cost function to optimally allocate scarce actuator resources to various objectives in real-time. Details are given on the development and offline testing which have been completed ready for deployment at ASDEX Upgrade. Electron Cyclotron Resonance Heating (ECRH) is particularly relevant for actuator management due to its localised deposition which can flexibly target specific regions of the plasma for different effects such as non-inductive current drive, impurity regulation, control of MHD modes and of course heating. A further motivation is that automated actuator management will simplify the setup of ECRH, in keeping with the long term goal of integrating MHD control as a routine part of ASDEX Upgrade experiments.

Actuator management for ECRH at ASDEX Upgrade

International Nuclear Information System (INIS)

Rapson, Christopher J.; Reich, Matthias; Stober, Joerg; Treutterer, Wolfgang

2015-01-01

Highlights: • Real-time actuator management algorithm developed for ECRH at ASDEX Upgrade. • First use of a control hierarchy in a fusion experiment. • Cost function evaluates optimal combination of all gyrotrons to all possible targets. • Considers many factors e.g. mirror movement, power available, presence and mode number of NTMs. • Configurable, robust algorithm is ready for online testing. - Abstract: Automated actuator management will be necessary on long pulse fusion experiments to adjust to unforeseen plasma events and unpredictable actuator availability. However, as a control problem, actuator management is underdeveloped in the fusion community. This contribution proposes an algorithm based on a control hierarchy and a cost function to optimally allocate scarce actuator resources to various objectives in real-time. Details are given on the development and offline testing which have been completed ready for deployment at ASDEX Upgrade. Electron Cyclotron Resonance Heating (ECRH) is particularly relevant for actuator management due to its localised deposition which can flexibly target specific regions of the plasma for different effects such as non-inductive current drive, impurity regulation, control of MHD modes and of course heating. A further motivation is that automated actuator management will simplify the setup of ECRH, in keeping with the long term goal of integrating MHD control as a routine part of ASDEX Upgrade experiments.

Overview of ASDEX Upgrade results

DEFF Research Database (Denmark)

Zohm, H.; Adamek, J.; Angioni, C.

2009-01-01

ASDEX Upgrade was operated with a fully W-covered wall in 2007 and 2008. Stationary H-modes at the ITER target values and improved H-modes with H up to 1.2 were run without any boronization. The boundary conditions set by the full W wall (high enough ELM frequency, high enough central heating and...

The Role of an Electric Field in the Formation of a Detached Regime in Tokamak Plasma

Science.gov (United States)

Senichenkov, I.; Kaveeva, E.; Rozhansky, V.; Sytova, E.; Veselova, I.; Voskoboynikov, S.; Coster, D.

2018-03-01

Modeling of the transition to the detachment of ASDEX Upgrade tokamak plasma with increasing density is performed using the SOLPS-ITER numerical code with a self-consistent account of drifts and currents. Their role in plasma redistribution both in the confinement region and in the scrape-off layer (SOL) is investigated. The mechanism of high field side high-density formation in the SOL in the course of detachment is suggested. In the full detachment regime, when the cold plasma region expands above the X-point and reaches closed magnetic-flux surfaces, plasma perturbation in a confined region may lead to a change in the confinement regime.

Characterization of dust particles produced in an all-tungsten wall tokamak and potentially mobilized by airflow

Energy Technology Data Exchange (ETDEWEB)

Rondeau, A., E-mail: anthony.rondeau@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES, SCA, 91192 Gif-sur-Yvette (France); Peillon, S.; Roynette, A.; Sabroux, J.-C.; Gelain, T.; Gensdarmes, F. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES, SCA, 91192 Gif-sur-Yvette (France); Rohde, V. [Max-Planck-Institut für Plasmaphysik, Boltzmannstraße 2, 85748 Garching (Germany); Grisolia, C. [CEA, IRFM, 13108 Saint-Paul-lez-Durance (France); Chassefière, E. [Laboratoire Géosciences Paris Sud (GEOPS), UMR 8148, Université Paris Sud, 91403 Orsay Cedex (France)

2015-08-15

At the starting of the shutdown of the AUG (ASDEX Upgrade: Axially Symmetric Divertor EXperiment) German tokamak, we collected particles deposited on the divertor surfaces by means of a dedicated device called “Duster Box”. This device allows to collect the particles using a controlled airflow with a defined shear stress. Consequently, the particles collected correspond to a potentially mobilizable fraction, by an airflow, of deposited dust. A total of more than 70,000 tungsten particles was, analysed showing a bimodal particle size distribution with a mode composed of flakes at 0.6 μm and a mode composed of spherical particles at 1.8 μm.

The physics of W transport illuminated by recent progress in W density diagnostics at ASDEX Upgrade

Science.gov (United States)

Odstrcil, T.; Pütterich, T.; Angioni, C.; Bilato, R.; Gude, A.; Odstrcil, M.; ASDEX Upgrade Team; the EUROfusion MST1 Team

2018-01-01

Due to the high mass and charge of the heavy ions, centrifugal and electrostatic forces cause a significant variation in their poloidal density. The impact of these forces on the poloidal density profile of tungsten was investigated utilizing the detailed two-dimensional SXR emissivity profiles from the ASDEX Upgrade tokamak. The perturbation in the electrostatic potential generated by magnetic trapping of the non-thermal ions from neutral beam injection was found to be responsible for significant changes in the poloidal distribution of tungsten ions. An excellent match with the results from fast particle modeling was obtained, validating the model for the poloidal fast particle distribution. Additionally, an enhancement of the neoclassical transport due to an outboard side impurity localization was measured in the experiment when analyzing the tungsten flux between sawtooth crashes. A qualitative match with neoclassical modeling was found, demonstrating the possibility of minimizing neoclassical transport by an optimization of the poloidal asymmetry profile of the impurity.

Impact of magnetic perturbation fields on tokamak plasmas

Energy Technology Data Exchange (ETDEWEB)

Fietz, Sina; Maraschek, Marc; Suttrop, Wolfgang; Zohm, Hartmut [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Classen, Ivo [FOM-Institute DIFFER, Nieuwegein (Netherlands); Collaboration: the ASDEX Upgrade Team

2015-05-01

Non-axisymmetric external magnetic perturbation (MP) fields arise in every tokamak e.g. due to not perfectly positioned external coils. Additionally many tokamaks, like ASDEX Upgrade (AUG), are equipped with a set of external coils, which produce a 3D MP field in addition to the equilibrium field. This field is used to either compensate for the intrinsic MP field or to influence MHD instabilities such as Edge Localised Modes (ELMs) or Neoclassical Tearing Modes (NTMs). But these MP fields can also give rise to a more global plasma response. The resonant components can penetrate the plasma and influence the stability of existing NTMs or even lead to their formation via magnetic reconnection. In addition they exert a local torque on the plasma. These effects are less pronounced at high plasma rotation where the resonant field components are screened. The non-resonant components do not influence NTMs directly but slow down the plasma rotation globally via the neoclassical toroidal viscous torque. The island formation caused by the MP field as well as the interaction of pre-existing islands with the MP field at AUG is presented. It is shown that these effects can be modelled using a simple forced reconnection theory. Also the effect of resonant and non-resonant MPs on the plasma rotation at AUG is discussed.

Radiation losses and global energy balance for Ohmically heated discharges in ASDEX

International Nuclear Information System (INIS)

Mueller, E.R.; Behringer, K.; Niedermeyer, H.

1982-01-01

Global energy balance, radiation profiles and dominant impurity radiation sources are compared for Ohmically heated limiter and divertor discharges in the ASDEX tokamak. In discharges with a poloidal stainless-steel limiter, total radiation from the plasma is the dominant energy loss channel. The axisymmetric divertor reduces this volume-integrated radiation to 30-35% of the heating power and additional Ti-gettering halves it again to 10-15%. Local radiation losses in the plasma centre, which are mainly due to the presence of iron impurity ions, are reduced by about one order of magnitude. In high-current (Isub(p) = 400 kA) and high-density (nsub(e)-bar = 6 x 10 13 cm -3 ) ungettered divertor discharges, up to 55% of the heating power is dumped into a cold-gas target inside the divertor chambers. The bolometrically detected volume power losses in the chambers can mainly be attributed to neutral hydrogen atoms with kinetic energies of a few eV. In this parameter range, the divertor plasma is dominated by inelastic molecular and atomic processes, the main process being Franck-Condon dissociation of H 2 molecules. (author)

H-mode transition physics close to DN on MAST and its applications to other tokamaks

International Nuclear Information System (INIS)

Meyer, H.

2004-01-01

Full text: ELMy H-mode is the base-line operating scenario for the next step fusion device ITER. To improve active and passive pedestal control a deeper understanding of H- mode physics is desirable. MAST contributes towards this understanding with good edge diagnostics, and by accessing extreme parameter regimes. The first inter-machine comparisons with respect to the influence of the magnetic topology on the power threshold with ASDEX-Upgrade and NSTX reveal a reduction of the power threshold in true double null (C-DN) configuration opening new operation regimes in both devices. The 30% reduction in threshold power close to C-DN observed on ASDEX-Upgrade, though significant, is less than the factor of two or more observed in both large spherical tokamaks, MAST and NSTX. This points towards the importance of field line curvature for this effect. The power thresholds measured in C-DN on MAST and NSTX are very similar. Despite this strong effect on the power threshold, changes in most edge parameters in L-mode due to the different magnetic configurations are small. However, significant changes are seen in the toroidal impurity flow velocity, related to the radial electric field, and in the scrape-off-layer temperature decay length at the high field side. The statistical comparison of MAST data with various H-mode theories suggests that different instabilities need to be stabilised at different spatial positions in the region where the pedestal forms to access H-mode. Pedestal temperatures observed on MAST are two to five times lower than in MAST equivalent discharges at ASDEX-Upgrade. However, the pedestal densities are similar. The differences in L-mode are less significant. The usual DN operating regime with co current NBI in MAST has been extended to include single null (SN) configurations, to provide more direct comparisons with conventional tokamaks. The plasma edge in SN on MAST is more stable to ELMs and the typical type-III ELMs, often observed in C-DN, are

A data acquisition system for real-time magnetic equilibrium reconstruction on ASDEX Upgrade and its application to NTM stabilization experiments

Energy Technology Data Exchange (ETDEWEB)

Giannone, L., E-mail: Louis.Giannone@ipp.mpg.de [Max Planck Institute for Plasma Physics, EURATOM Association, 85748 Garching (Germany); Reich, M.; Maraschek, M.; Poli, E.; Rapson, C.; Barrera, L.; McDermott, R.; Mlynek, A. [Max Planck Institute for Plasma Physics, EURATOM Association, 85748 Garching (Germany); Ruan, Q. [National Instruments, Austin, TX 78759-3504 (United States); Treutterer, W. [Max Planck Institute for Plasma Physics, EURATOM Association, 85748 Garching (Germany); Wenzel, L. [National Instruments, Austin, TX 78759-3504 (United States); Bock, A.; Conway, G.; Fischer, R.; Fuchs, J.C.; Lackner, K. [Max Planck Institute for Plasma Physics, EURATOM Association, 85748 Garching (Germany); McCarthy, P.J. [Department of Physics, University College Cork, Association EURATOM-DCU, Cork (Ireland); Preuss, R. [Max Planck Institute for Plasma Physics, EURATOM Association, 85748 Garching (Germany); Rampp, M. [Computing Centre (RZG) of the Max Planck Society and the Max Planck Institute for Plasma Physics, 85748 Garching (Germany); Schuhbeck, K.H. [Max Planck Institute for Plasma Physics, EURATOM Association, 85748 Garching (Germany); and others

2013-12-15

Highlights: • Calculation of real-time tokamak magnetic equilibrium with constraints from magnetic probes. • Parallel equilibrium calculation including the Motional Stark Effect diagnostic as additional constraints. • Feedback control of mirror for pre-emptive ECCD stabilization of neo-classical tearing modes. • Probe calibration by individual poloidal field coil currents. • Optimized parameters for poloidal field coil location, integrator gains and the location and orientation of magnetic probes. -- Abstract: The pre-emptive stabilization of a neoclassical tearing mode, NTM, requires the calculation of the tokamak magnetic equilibrium in real-time. A launcher mirror is positioned to deposit electron cyclotron current drive on the rational surface where the NTM should appear. A real-time Grad–Shafranov solver using constraints from magnetic probe, flux loop and Motional Stark Effect measurements has been developed to locate these rational surfaces and deliver this information to the mirror controller in real-time. A novel algorithm significantly reduces the number of operations required in the first and second step of the solver. Contour integrals are carried out to calculate the q profile as a function of normalized radius and the rational surfaces are found by spline interpolation. A cycle time of 0.6 ms for calculating two tokamak equilibria in parallel using four current basis functions with magnetic constraints only and using six current basis functions with magnetic and MSE constraints has been achieved. Using these tools, pre-emptive stabilization of a m/n = 3/2 NTM mode in ASDEX Upgrade could be demonstrated.

Investigation of transient melting of tungsten by ELMs in ASDEX Upgrade

International Nuclear Information System (INIS)

Krieger, K; Sieglin, B; Balden, M; De Marne, P; Nille, D; Rohde, V; Faitsch, M; Giannone, L; Herrmann, A; Coenen, J W; Göths, B; Laggner, F; Matthews, G F; Dejarnac, R; Horacek, J; Komm, M; Pitts, R A; Ratynskaia, S; Thoren, E; Tolias, P

2017-01-01

Repetitive melting of tungsten by power transients originating from edge localized modes (ELMs) has been studied in the tokamak experiment ASDEX Upgrade. Tungsten samples were exposed to H-mode discharges at the outer divertor target plate using the Divertor Manipulator II system. The exposed sample was designed with an elevated sloped surface inclined against the incident magnetic field to increase the projected parallel power flux to a level were transient melting by ELMs would occur. Sample exposure was controlled by moving the outer strike point to the sample location. As extension to previous melt studies in the new experiment both the current flow from the sample to vessel potential and the local surface temperature were measured with sufficient time resolution to resolve individual ELMs. The experiment provided for the first time a direct link of current flow and surface temperature during transient ELM events. This allows to further constrain the MEMOS melt motion code predictions and to improve the validation of its underlying model assumptions. Post exposure ex situ analysis of the retrieved samples confirms the decreased melt motion observed at shallower magnetic field line to surface angles compared to that at leading edges exposed to the parallel power flux. (paper)

Spectroscopic measurement of target plate erosion in the ASDEX Upgrade divertor

Energy Technology Data Exchange (ETDEWEB)

Filed, A R; Garcia-Rosales, C; Lieder, G; Pitcher, C S; Radtke, R [Association Euratom-Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); ASDEX Upgrade Team

1996-02-01

The erosion of the graphite divertor plates in the ASDEX Upgrade tokamak is measured spectroscopically. Spatial profiles of the D{sup 0} and C{sup +} influxes across the outer target plate are determined from measured absolute line intensities. Plasma parameters (n{sub e}, T{sub e}) at the target, which are required to determine the appropriate photon emission efficiencies for these lines, are obtained from an in-vessel reciprocating Langmuir probe above the target plate. Yields for the erosion of the graphite by the incident D{sup +} flux are determined from the ratio of the measured C{sup +} to D{sup 0} fluxes. Over a range of moderate densities the measured yields of {<=} 4% are explicable in terms of physical sputtering alone. Chemical sputtering by low energy Franck-Condon neutrals probably contributes, however, to the total erosion. At higher densities detachment of the plasma from the targets occurs owing to formation of a MARFE near the X point. Under these conditions localized physical sputtering of the targets ceases. The impurity level (Z{sub eff}) is, however, maintained following detachment, indicating a corresponding maintenance of carbon influx, perhaps due to chemical erosion of the total graphite surface and/or an improvement in particle confinement in the detached state. (author). 26 refs, 14 figs, 1 tab.

Influence of plasma rotation on tearing mode stability on the ASDEX upgrade tokamak

Energy Technology Data Exchange (ETDEWEB)

Fietz, Sina Marie Ariane

2013-12-16

Neoclassical tearing modes (NTM) are one of the most serious performance limiting instabilities in next-step fusion devices like ITER. NTMs are destabilised as a consequence of a seed perturbation (trigger) and are driven by a loss of helical bootstrap current inside the island. The appearance of these instabilities is accompanied with a loss of confined plasma energy. Additionally, these modes can stop the plasma rotation, lock to the vessel wall, flush out all plasma energy and terminate a discharge via a disruption. In ITER the confinement reduction will limit the achievable fusion power, whereas a disruption is likely to damage the vessel wall. In order to mitigate and control NTMs in ITER, extrapolations based on the present understanding and observations must be made. One key issue is the rotation dependence of NTMs, especially at the NTM onset. ITER will be operated at low plasma rotation, which is different from most present day experiments. No theory is currently available to describe this dependence. Experiments are therefore required to provide a basis for the theory to describe the physics. Additionally from the experiments scalings can be developed and extrapolated in order to predict the NTM behaviour in the parameter range relevant for ITER. Another important issue is the influence of externally applied magnetic perturbation (MP) fields on the NTM stability and frequency. These fields will be used in ITER primarily for the mitigation of edge instabilities. As a side effect they can slow down an NTM and the plasma rotation, which supports the appearance of locked modes. Additionally, they can also influence the stability of an NTM. This interaction has to be predicted for ITER, based on models validated at present day devices. In this work the influence of plasma rotation on the NTM onset at the ASDEX Upgrade tokamak (AUG) is investigated. An onset database has been created in which the different trigger mechanisms have been identified. Based on this

Influence of plasma rotation on tearing mode stability on the ASDEX upgrade tokamak

International Nuclear Information System (INIS)

Fietz, Sina Marie Ariane

2013-01-01

Neoclassical tearing modes (NTM) are one of the most serious performance limiting instabilities in next-step fusion devices like ITER. NTMs are destabilised as a consequence of a seed perturbation (trigger) and are driven by a loss of helical bootstrap current inside the island. The appearance of these instabilities is accompanied with a loss of confined plasma energy. Additionally, these modes can stop the plasma rotation, lock to the vessel wall, flush out all plasma energy and terminate a discharge via a disruption. In ITER the confinement reduction will limit the achievable fusion power, whereas a disruption is likely to damage the vessel wall. In order to mitigate and control NTMs in ITER, extrapolations based on the present understanding and observations must be made. One key issue is the rotation dependence of NTMs, especially at the NTM onset. ITER will be operated at low plasma rotation, which is different from most present day experiments. No theory is currently available to describe this dependence. Experiments are therefore required to provide a basis for the theory to describe the physics. Additionally from the experiments scalings can be developed and extrapolated in order to predict the NTM behaviour in the parameter range relevant for ITER. Another important issue is the influence of externally applied magnetic perturbation (MP) fields on the NTM stability and frequency. These fields will be used in ITER primarily for the mitigation of edge instabilities. As a side effect they can slow down an NTM and the plasma rotation, which supports the appearance of locked modes. Additionally, they can also influence the stability of an NTM. This interaction has to be predicted for ITER, based on models validated at present day devices. In this work the influence of plasma rotation on the NTM onset at the ASDEX Upgrade tokamak (AUG) is investigated. An onset database has been created in which the different trigger mechanisms have been identified. Based on this

Recent advances in gyrokinetic full-f particle simulation of medium sized Tokamaks with ELMFIRE

International Nuclear Information System (INIS)

Janhunen, S.J.; Kiviniemi, T.P.; Korpio, T.; Leerink, S.; Nora, M.; Heikkinen, J.A.; Ogando, F.

2010-01-01

Large-scale kinetic simulations of toroidal plasmas based on first principles are called for in studies of transition from low to high confinement mode and internal transport barrier formation in the core plasma. Such processes are best observed and diagnosed in detached plasma conditions in mid-sized tokamaks, so gyrokinetic simulations for these conditions are warranted. A first principles test-particle based kinetic model ELMFIRE[1] has been developed and used in interpretation[1,2] of FT-2 and DIII-D experiments. In this work we summarize progress in Cyclone (DIII-D core) and ASDEX Upgrade pedestal region simulations, and show that in simulations the choice of adiabatic electrons results in quenching of turbulence (copyright 2010 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

Recent advances in gyrokinetic full-f particle simulation of medium sized Tokamaks with ELMFIRE

Energy Technology Data Exchange (ETDEWEB)

Janhunen, S.J.; Kiviniemi, T.P.; Korpio, T.; Leerink, S.; Nora, M. [Helsinki University of Technology, Euratom-Tekes Association, Espoo (Finland); Heikkinen, J.A. [VTT, Euratom-Tekes Association, Espoo (Finland); Ogando, F. [Helsinki University of Technology, Euratom-Tekes Association, Espoo (Finland); Universidad Nacional de Educacion a Distancia, Madrid (Spain)

2010-05-15

Large-scale kinetic simulations of toroidal plasmas based on first principles are called for in studies of transition from low to high confinement mode and internal transport barrier formation in the core plasma. Such processes are best observed and diagnosed in detached plasma conditions in mid-sized tokamaks, so gyrokinetic simulations for these conditions are warranted. A first principles test-particle based kinetic model ELMFIRE[1] has been developed and used in interpretation[1,2] of FT-2 and DIII-D experiments. In this work we summarize progress in Cyclone (DIII-D core) and ASDEX Upgrade pedestal region simulations, and show that in simulations the choice of adiabatic electrons results in quenching of turbulence (copyright 2010 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

A new B-dot probe-based diagnostic for amplitude, polarization, and wavenumber measurements of ion cyclotron range-of frequency fields on ASDEX Upgrade

International Nuclear Information System (INIS)

Ochoukov, R.; Bobkov, V.; Faugel, H.; Fünfgelder, H.; Noterdaeme, J.-M.

2015-01-01

A new B-dot probe-based diagnostic has been installed on an ASDEX Upgrade tokamak to characterize ion cyclotron range-of frequency (ICRF) wave generation and interaction with magnetized plasma. The diagnostic consists of a field-aligned array of B-dot probes, oriented to measure fast and slow ICRF wave fields and their field-aligned wavenumber (k // ) spectrum on the low field side of ASDEX Upgrade. A thorough description of the diagnostic and the supporting electronics is provided. In order to compare the measured dominant wavenumber of the local ICRF fields with the expected spectrum of the launched ICRF waves, in-air near-field measurements were performed on the newly installed 3-strap ICRF antenna to reconstruct the dominant launched toroidal wavenumbers (k tor ). Measurements during a strap current phasing scan in tokamak discharges reveal an upshift in k // as strap phasing is moved away from the dipole configuration. This result is the opposite of the k tor trend expected from in-air near-field measurements; however, the near-field based reconstruction routine does not account for the effect of induced radiofrequency (RF) currents in the passive antenna structures. The measured exponential increase in the local ICRF wave field amplitude is in agreement with the upshifted k // , as strap phasing moves away from the dipole configuration. An examination of discharges heated with two ICRF antennas simultaneously reveals the existence of beat waves at 1 kHz, as expected from the difference of the two antennas’ operating frequencies. Beats are observed on both the fast and the slow wave probes suggesting that the two waves are coupled outside the active antennas. Although the new diagnostic shows consistent trends between the amplitude and the phase measurements in response to changes applied by the ICRF antennas, the disagreement with the in-air near-field measurements remains. An electromagnetic model is currently under development to address this issue

A new B-dot probe-based diagnostic for amplitude, polarization, and wavenumber measurements of ion cyclotron range-of frequency fields on ASDEX Upgrade

Science.gov (United States)

Ochoukov, R.; Bobkov, V.; Faugel, H.; Fünfgelder, H.; Noterdaeme, J.-M.

2015-11-01

A new B-dot probe-based diagnostic has been installed on an ASDEX Upgrade tokamak to characterize ion cyclotron range-of frequency (ICRF) wave generation and interaction with magnetized plasma. The diagnostic consists of a field-aligned array of B-dot probes, oriented to measure fast and slow ICRF wave fields and their field-aligned wavenumber (k//) spectrum on the low field side of ASDEX Upgrade. A thorough description of the diagnostic and the supporting electronics is provided. In order to compare the measured dominant wavenumber of the local ICRF fields with the expected spectrum of the launched ICRF waves, in-air near-field measurements were performed on the newly installed 3-strap ICRF antenna to reconstruct the dominant launched toroidal wavenumbers (ktor). Measurements during a strap current phasing scan in tokamak discharges reveal an upshift in k// as strap phasing is moved away from the dipole configuration. This result is the opposite of the ktor trend expected from in-air near-field measurements; however, the near-field based reconstruction routine does not account for the effect of induced radiofrequency (RF) currents in the passive antenna structures. The measured exponential increase in the local ICRF wave field amplitude is in agreement with the upshifted k//, as strap phasing moves away from the dipole configuration. An examination of discharges heated with two ICRF antennas simultaneously reveals the existence of beat waves at 1 kHz, as expected from the difference of the two antennas' operating frequencies. Beats are observed on both the fast and the slow wave probes suggesting that the two waves are coupled outside the active antennas. Although the new diagnostic shows consistent trends between the amplitude and the phase measurements in response to changes applied by the ICRF antennas, the disagreement with the in-air near-field measurements remains. An electromagnetic model is currently under development to address this issue.

Experimental investigation of heat transport and divertor loads of fusion plasmas in all metal ASDEX upgrade and JET

International Nuclear Information System (INIS)

Sieglin, Bernhard A.

2014-01-01

This work presents divertor heat load studies conducted at two of the largest tokamaks currently in operation, ASDEX Upgrade and the Joint European Torus (JET). A commonly agreed empirical scaling for the power fall-off length in H-mode obtained in carbon devices is validated in JET with the ILW. Bohm and Gyro-Bohm like models are identified as possible candidates describing the divertor broadening. Quantities for the assessment of the thermal load induced by transient heat loads are defined. JET with the ILW exhibits an on average longer ELM duration as compared to the carbon wall. For identical pedestal conditions the ELM durations in both cases are found to be the same within error bars. The energy fluency is found to depend mainly on the pedestal pressure with a weak dependence on the relative loss in stored energy. This is noteworthy since the current extrapolation to ITER assumes a linear dependence on the relative ELM size.

Influence of neutron scattering and source extent on the measurement of neutron energy spectra at ASDEX

International Nuclear Information System (INIS)

Huebner, K.; Baetzner, R.; Roos, M.; Robouch, B.V.; Ingrosso, L.; Wurz, H.

1987-08-01

The problem of nuclear emulsion measurements at ASDEX is considered. Besides the application of the VINIA-3DAMC software, this needs a description of the plasma neutron source, a model of the ASDEX structure, and calculation of the response of the nuclear emulsion to the incoming spectral neutron fluence. The latter is essential for comparing the numerical results with measurements at ASDEX. To treat this part, the NEPMC software was developed. The aim of the present work is to demonstrate the feasibility, reliability and usefulness of the method. Therefore simplified treatments for the ASDEX model, the plasma neutron source and the track statistics in the NEPMC software were used. Such calculations are of interest not only for nuclear emulsion measurements as well as any other neutron diagnostics, but also for all problems of neutron shielding for other diagnostics. (orig./GG)

Real time magnetic field and flux measurements for tokamak control using a multi-core PCI Express system

International Nuclear Information System (INIS)

Giannone, L.; Schneider, W.; McCarthy, P.J.; Sips, A.C.C.; Treutterer, W.; Behler, K.; Eich, T.; Fuchs, J.C.; Hicks, N.; Kallenbach, A.; Maraschek, M.; Mlynek, A.; Neu, G.; Pautasso, G.; Raupp, G.; Reich, M.; Schuhbeck, K.H.; Stober, J.; Volpe, F.; Zehetbauer, T.

2009-01-01

The existing real time system for the position and shape control in ASDEX Upgrade has been extended to calculate magnetic flux surfaces in real time using a multi-core PCI Express system running LabVIEW RT. The availability of reflective memory for LabVIEW RT will allow this system to be connected to the control system and other diagnostics in a multi-platform real time network. The measured response of each magnetic probe to the individual poloidal field coil currents in the absence of plasma current is compared to the calculated value. Prior to a tokamak discharge this comparison can be used to check for failure of the magnetic probe, flux loop or integrator.

Piezoelectric valve for massive gas injection in ASDEX upgrade

Energy Technology Data Exchange (ETDEWEB)

Dibon, Mathias; Neu, Rudolf [Max-Planck-Institute for Plasmaphysics, Boltzmannstr. 2, 85748 Garching (Germany); Technical University Munich, Boltzmannstr. 15, 85748 Garching (Germany); Herrmann, Albrecht; Mank, Klaus; Mertens, Vitus; Pautasso, Gabriella; Ploeckl, Bernhard [Max-Planck-Institute for Plasmaphysics, Boltzmannstr. 2, 85748 Garching (Germany)

2016-07-01

A sudden loss of plasma temperature can cause a disruption, which poses a significant problem for current Tokamaks and future fusion devices. Hence, mitigating forces and thermal loads during disruptions is important for the integrity of the vessel and first wall components. Therefore, high speed gas valves are used to deliver a pulse of noble gas onto the plasma, which irradiates the thermal energy quickly, avoiding localized heat loads and mechanical stress due to induced currents. A new design for such a valve is currently under development. The valve plate is driven by two piezoelectric stack actuators. The stroke of the actuators (0.07 mm) is amplified by a monolithic titanium frame and reaches 2 mm. The frame also serves as spring to pre-load the actuators. In the idle state, it also presses the conical valve plate into the seal, closing the gas chamber (42 cm{sup 3}). The actuators accelerate the stem and the valve plate until it is fully opened after 2 ms. The orifice of the valve has a diameter of 14 mm. This allows a peak mass flow rate of the gas up to 8 . 10{sup 4} (Pa.m)/(s) after 1.8 ms and an average mass flow rate of 2 . 10{sup 4} (Pa.m)/(s) over the evacuation time of 10 ms. Therefore, one valve would be sufficient to deliver the required amount of gas to mitigate disruptions at ASDEX Upgrade.

Investigations of MHD activity in ASDEX discharges

International Nuclear Information System (INIS)

Stambaugh, R.; Gernhardt, J.; Klueber, O.; Wagner, F.

1984-06-01

This report makes a strong attempt to relate some specific observations of MHD activity in ADEX discharges to observations made on the Doublet III and PDX tokamaks and to theoretical work on high β MHD modes at GA and PPPL. Three topics are discussed. The first topic is the detailed analysis of the time history of MHD activity in a β discharge. The β limit discharge in ASDEX is identified as a discharge in which, during constant neutral beam power, β reaches a maximum and then decreases, often to a lower steady level if the heating pulse is long enough. During the L phase of this discharge, the MHD activity observed in the B coils is both a continuous and bursting coupled m >= 1 mode of the 'fishbone' type. When β is rising in the H phase, this mode disappears; only ELMs are present. At βsub(max), a different mode appears, the m=2, n=1 tearing mode, which grows rapidly as β decreases. The second topic is the very new observation of the fishbone-like mode in a discharge heated by combined neutral beam and ion cyclotron heating power. The mode characteristics are modulated by sawtooth oscillations in a manner consistent with the importance of q(0) in the stability of this mode. The third topic is the search for ELM precursors in discharges designed to have no other competing and complicating MHD activity. In these cases nonaxisymmetric precursors to the Hsub(α) spike were observed. Hence, it appears that an MHD mode, rather than an energy balance problem, must be the origin of the ELM. (orig./GG)

Upgrade to the control system of the reflectometry diagnostic of ASDEX upgrade

International Nuclear Information System (INIS)

Graca, S.; Santos, J.; Manso, M.E.

2004-01-01

The broadband frequency modulation-continuous wave microwave/millimeter wave reflectometer of ASDEX upgrade tokamak (Institut fuer Plasma Physik (IPP), Garching, Germany) developed by Centro de Fusao Nuclear (Lisboa, Portugal) with the collaboration of IPP, is a complex system with 13 channels (O and X modes) and two types of operation modes (swept and fixed frequency). The control system that ensures remote operation of the diagnostic incorporates VME and CAMAC bus based acquisition/timing systems. Microprocessor input/output boards are used to control and monitor the microwave circuitry and associated electronic devices. The implementation of the control system is based on an object-oriented client/server model: a centralized server manages the hardware and receives input from remote clients. Communication is handled through transmission control protocol/internet protocol sockets. Here we describe recent upgrades of the control system aiming to: (i) accommodate new channels; (ii) adapt to the heterogeneity of computing platforms and operating systems; and (iii) overcome remote access restrictions. Platform and operating system independence was achieved by redesigning the graphical user interface in JAVA. As secure shell is the standard remote access protocol adopted in major fusion laboratories, secure shell tunneling was implemented to allow remote operation of the diagnostic through the existing firewalls

Upgrade to the control system of the reflectometry diagnostic of ASDEX upgrade

Science.gov (United States)

Graça, S.; Santos, J.; Manso, M. E.

2004-10-01

The broadband frequency modulation-continuous wave microwave/millimeter wave reflectometer of ASDEX upgrade tokamak (Institut für Plasma Physik (IPP), Garching, Germany) developed by Centro de Fusão Nuclear (Lisboa, Portugal) with the collaboration of IPP, is a complex system with 13 channels (O and X modes) and two types of operation modes (swept and fixed frequency). The control system that ensures remote operation of the diagnostic incorporates VME and CAMAC bus based acquisition/timing systems. Microprocessor input/output boards are used to control and monitor the microwave circuitry and associated electronic devices. The implementation of the control system is based on an object-oriented client/server model: a centralized server manages the hardware and receives input from remote clients. Communication is handled through transmission control protocol/internet protocol sockets. Here we describe recent upgrades of the control system aiming to: (i) accommodate new channels; (ii) adapt to the heterogeneity of computing platforms and operating systems; and (iii) overcome remote access restrictions. Platform and operating system independence was achieved by redesigning the graphical user interface in JAVA. As secure shell is the standard remote access protocol adopted in major fusion laboratories, secure shell tunneling was implemented to allow remote operation of the diagnostic through the existing firewalls.

Internal transport barrier triggering by rational magnetic flux surfaces in tokamaks

International Nuclear Information System (INIS)

Joffrin, E.; Challis, C.D.; Conway, G.D.

2003-01-01

The formation of Internal Transport Barriers (ITBs) has been experimentally associated with the presence of rational q-surfaces in both JET and ASDEX Upgrade. The triggering mechanisms are related to the occurrence of magneto-hydrodynamic (MHD) instabilities such as mode coupling or fishbone activity. These events could locally modify the poloidal velocity and increase transiently the shearing rate to values comparable to the linear growth rate of ITG modes. For JET reversed magnetic shear scenarios, ITB emergence occurs preferentially when the minimum q reaches an integer value. In this case, transport effects localised in the vicinity of zero magnetic shear and close to rational q values may also contribute to the formation of ITBs.The role of rational q surfaces on ITB triggering stresses the importance of q profile control for advanced tokamak scenario and could contribute to lower substantially the access power to these scenarios in next step facilities. (author)

Internal Transport Barrier triggering by rational magnetic flux surfaces in tokamaks

International Nuclear Information System (INIS)

Joffrin, E.H.

2002-01-01

The formation of Internal Transport Barriers (ITBs) has been experimentally associated with the presence of rational q-surfaces in both JET and ASDEX Upgrade. The triggering mechanisms are related to the occurrence of magneto-hydrodynamic (MHD) instabilities such as mode coupling or fishbone activity. These events could locally modify the poloidal velocity and increase transiently the shearing rate to values comparable to the linear growth rate of ITG modes. For reversed magnetic shear scenario, ITB emergence occurs preferentially when the minimum q reaches an integer value. In this case, transport effects localised in the vicinity of zero magnetic shear and close to rational q values may also contribute to the formation of ITBs. The role of rational q surfaces on ITB triggering stresses the importance of q profile control for advanced tokamak scenario and could contribute to lower substantially the access power to these scenarios in next step facilities. (author)

Numerically derived parametrisation of optimal RMP coil phase as a guide to experiments on ASDEX Upgrade

Science.gov (United States)

Ryan, D. A.; Liu, Y. Q.; Li, L.; Kirk, A.; Dunne, M.; Dudson, B.; Piovesan, P.; Suttrop, W.; Willensdorfer, M.; the ASDEX Upgrade Team; the EUROfusion MST1 Team

2017-02-01

Edge localised modes (ELMs) are a repetitive MHD instability, which may be mitigated or suppressed by the application of resonant magnetic perturbations (RMPs). In tokamaks which have an upper and lower set of RMP coils, the applied spectrum of the RMPs can be tuned for optimal ELM control, by introducing a toroidal phase difference {{Δ }}{{Φ }} between the upper and lower rows. The magnitude of the outermost resonant component of the RMP field | {b}{{res}}1| (other proposed criteria are discussed herein) has been shown experimentally to correlate with mitigated ELM frequency, and to be controllable by {{Δ }}{{Φ }} (Kirk et al 2013 Plasma Phys. Control. Fusion 53 043007). This suggests that ELM mitigation may be optimised by choosing {{Δ }}{{Φ }}={{Δ }}{{{Φ }}}{{opt}}, such that | {b}{{res}}1| is maximised. However it is currently impractical to compute {{Δ }}{{{Φ }}}{{opt}} in advance of experiments. This motivates this computational study of the dependence of the optimal coil phase difference {{Δ }}{{{Φ }}}{{opt}}, on global plasma parameters {β }N and q 95, in order to produce a simple parametrisation of {{Δ }}{{{Φ }}}{{opt}}. In this work, a set of tokamak equilibria spanning a wide range of ({β }N, q 95) is produced, based on a reference equilibrium from an ASDEX Upgrade experiment. The MARS-F code (Liu et al 2000 Phys. Plasmas 7 3681) is then used to compute {{Δ }}{{{Φ }}}{{opt}} across this equilibrium set for toroidal mode numbers n = 1-4, both for the vacuum field and including the plasma response. The computational scan finds that for fixed plasma boundary shape, rotation profiles and toroidal mode number n, {{Δ }}{{{Φ }}}{{opt}} is a smoothly varying function of ({β }N, q 95). A 2D quadratic function in ({β }N, q 95) is used to parametrise {{Δ }}{{{Φ }}}{{opt}}, such that for given ({β }N, q 95) and n, an estimate of {{Δ }}{{{Φ }}}{{opt}} may be made without requiring a plasma response computation. To quantify the uncertainty

Development and application of poloidal correlation reflectometry to study turbulent structures in the ASDEX Upgrade tokamak

Energy Technology Data Exchange (ETDEWEB)

Prisiazhniuk, Dmitrii

2017-06-05

One of the key question of high temperature plasma confinement in a magnetic field is how plasma turbulence influences the radial transport of particles and energy. A better understanding of transport processes caused by turbulence would allow to improve the plasma confinement in fusion devices. To this end a deeper understanding of the mechanisms controlling the development, saturation and stabilization of turbulence is needed. From the experimental point of view a main challenge in these investigations is the measurement of plasma parameters on both small temporal (μs) and spatial (mm) scales. In this thesis a new microwave heterodyne poloidal correlation reflectometry diagnostic has been developed and installed at the ASDEX Upgrade tokamak to investigate the cross-correlation of turbulent density fluctuations. This diagnostic yields information on fundamental turbulence parameters such as the perpendicular propagation velocity v {sub perpendicular} {sub to}, the perpendicular correlation length l {sub perpendicular} {sub to} (characteristic size of the turbulent eddies) and the decorrelation time τ{sub d} (characteristic life time of the turbulent eddies) over a wide range of plasma densities. The inclination of the turbulent eddies α in the poloidal-toroidal plane spanned by the magnetic flux surfaces of a tokamak, being a measure of the magnetic field pitch angle, can also be obtained. The turbulence investigations were performed in low confinement mode (L-mode) plasmas for a range of plasma parameters. All measurements were interpreted taking into account the transfer function of reflectometry in the Born approximation. The results are compared with theoretical predictions and simulations. In the first part of this thesis the inclination and the propagation of turbulent structures are investigated. It is shown that eddies are nearly aligned to the magnetic field line and, therefore, the magnetic field pitch angle can be measured with a precision of about 1

Statistical analyses of local transport coefficients in Ohmic ASDEX discharges

International Nuclear Information System (INIS)

Simmet, E.; Stroth, U.; Wagner, F.; Fahrbach, H.U.; Herrmann, W.; Kardaun, O.J.W.F.; Mayer, H.M.

1991-01-01

Tokamak energy transport is still an unsolved problem. Many theoretical models have been developed, which try to explain the anomalous high energy-transport coefficients. Up to now these models have been applied to global plasma parameters. A comparison of transport coefficients with global confinement time is only conclusive if the transport is dominated by one process across the plasma diameter. This, however, is not the case in most Ohmic confinement regimes, where at least three different transport mechanisms play an important role. Sawtooth activity leads to an increase in energy transport in the plasma centre. In the intermediate region turbulent transport is expected. Candidates here are drift waves and resistive fluid turbulences. At the edge, ballooning modes or rippling modes could dominate the transport. For the intermediate region, one can deduce theoretical scaling laws for τ E from turbulent theories. Predicted scalings reproduce the experimentally found density dependence of τ E in the linear Ohmic confinement regime (LOC) and the saturated regime (SOC), but they do not show the correct dependence on the isotope mass. The relevance of these transport theories can only be tested in comparing them to experimental local transport coefficients. To this purpose we have performed transport calculations on more than a hundred Ohmic ASDEX discharges. By Principal Component Analysis we determine the dimensionless components which dominate the transport coefficients and we compare the results to the predictions of various theories. (author) 6 refs., 2 figs., 1 tab

Density fluctuations in ohmic Asdex discharges

International Nuclear Information System (INIS)

Dodel, G.; Holzhauer, E.

1989-01-01

The investigations on the wave-number and frequency spectra of the density fluctuations, occurring in the different operational modes of ASDEX, are summarized. The aim of the experiments is to study the physical nature of fluctuations and their influence on anomalous transport. The scattering system is described. The results reported were obtained using a 100 mW, λ = 119 μm CW CH-30H laser and homodyne detection

Control processes and machine protection on ASDEX Upgrade

International Nuclear Information System (INIS)

Raupp, G.; Treutterer, W.; Mertens, V.; Neu, G.; Sips, A.; Zasche, D.; Zehetbauer, Th.

2007-01-01

Safe operation of ASDEX Upgrade is guaranteed by a conventional hierarchy of simple and robust hard-wired systems for personnel and machine protection featuring standardized switch-off procedures. Machine protection and handling of off-normal events is further enhanced and peak and lifetime stress minimized through the plasma control system. Based on a real-time process model supporting safety critical applications with data quality tagging, process self-monitoring, watchdog monitoring and alarm propagation, processes detect complex and critical failures and reliably perform case-sensitive counter measures. Intelligent real-time failure handling is done with hardware or software redundancy and performance degradation, or modification of reference values to continue or terminate discharges with reduced machine stress. Examples implemented so far on ASDEX Upgrade are given, such as recovery from measurement failures, switch-over of redundant actuators, handling of actuator limitations, detection of plasma instabilities, plasma state dependent soft landing, or handling of failed switch-off procedures through breakers disconnecting the machine from grid

Improvements in disruption prediction at ASDEX Upgrade

Energy Technology Data Exchange (ETDEWEB)

Aledda, R., E-mail: raffaele.aledda@diee.unica.it; Cannas, B., E-mail: cannas@diee.unica.it; Fanni, A., E-mail: fanni@diee.unica.it; Pau, A., E-mail: alessandro.pau@diee.unica.it; Sias, G., E-mail: giuliana.sias@diee.unica.it

2015-10-15

Highlights: • A disruption prediction system for AUG, based on a logistic model, is designed. • The length of the disruptive phase is set for each disruption in the training set. • The model is tested on dataset different from that used during the training phase. • The generalization capability and the aging of the model have been tested. • The predictor performance is compared with the locked mode detector. - Abstract: In large-scale tokamaks disruptions have the potential to create serious damage to the facility. Hence disruptions must be avoided, but, when a disruption is unavoidable, minimizing its severity is mandatory. A reliable detection of a disruptive event is required to trigger proper mitigation actions. To this purpose machine learning methods have been widely studied to design disruption prediction systems at ASDEX Upgrade. The training phase of the proposed approaches is based on the availability of disrupted and non-disrupted discharges. In literature disruptive configurations were assumed appearing into the last 45 ms of each disruption. Even if the achieved results in terms of correct predictions were good, it has to be highlighted that the choice of such a fixed temporal window might have limited the prediction performance. In fact, it generates confusing information in cases of disruptions with disruptive phase different from 45 ms. The assessment of a specific disruptive phase for each disruptive discharge represents a relevant issue in understanding the disruptive events. In this paper, the Mahalanobis distance is applied to define a specific disruptive phase for each disruption, and a logistic regressor has been trained as disruption predictor. The results show that enhancements on the achieved performance on disruption prediction are possible by defining a specific disruptive phase for each disruption.

Improvements in disruption prediction at ASDEX Upgrade

International Nuclear Information System (INIS)

Aledda, R.; Cannas, B.; Fanni, A.; Pau, A.; Sias, G.

2015-01-01

Highlights: • A disruption prediction system for AUG, based on a logistic model, is designed. • The length of the disruptive phase is set for each disruption in the training set. • The model is tested on dataset different from that used during the training phase. • The generalization capability and the aging of the model have been tested. • The predictor performance is compared with the locked mode detector. - Abstract: In large-scale tokamaks disruptions have the potential to create serious damage to the facility. Hence disruptions must be avoided, but, when a disruption is unavoidable, minimizing its severity is mandatory. A reliable detection of a disruptive event is required to trigger proper mitigation actions. To this purpose machine learning methods have been widely studied to design disruption prediction systems at ASDEX Upgrade. The training phase of the proposed approaches is based on the availability of disrupted and non-disrupted discharges. In literature disruptive configurations were assumed appearing into the last 45 ms of each disruption. Even if the achieved results in terms of correct predictions were good, it has to be highlighted that the choice of such a fixed temporal window might have limited the prediction performance. In fact, it generates confusing information in cases of disruptions with disruptive phase different from 45 ms. The assessment of a specific disruptive phase for each disruptive discharge represents a relevant issue in understanding the disruptive events. In this paper, the Mahalanobis distance is applied to define a specific disruptive phase for each disruption, and a logistic regressor has been trained as disruption predictor. The results show that enhancements on the achieved performance on disruption prediction are possible by defining a specific disruptive phase for each disruption.

Internal transport barrier triggering by rational magnetic flux surfaces in tokamaks

International Nuclear Information System (INIS)

Joffrin, E.; Challis, C.D.; Conway, G.D.

2003-01-01

The formation of internal transport barriers (ITBs) has been experimentally associated with the presence of rational q surfaces in both JET and ASDEX Upgrade. The triggering mechanisms are related to the occurrence of magneto-hydrodynamic (MHD) instabilities such as mode coupling and fishbone activity. These events could locally modify the poloidal velocity and increase transiently the shearing rate to values comparable with the linear growth rate of ion temperature gradient modes. For JET reversed magnetic shear scenarios, ITB emergence occurs preferentially when the minimum q reaches an integral value. In this case, transport effects localized in the vicinity of zero magnetic shear and close to rational q values may be at the origin of ITB formation. The role of rational q surfaces in ITB triggering stresses the importance of q profile control for an advanced tokamak scenario and could assist in substantially lowering the access power to these scenarios in next step facilities. (author)

Correlation of the tokamak H-mode density limit with ballooning stability at the separatrix

Science.gov (United States)

Eich, T.; Goldston, R. J.; Kallenbach, A.; Sieglin, B.; Sun, H. J.; ASDEX Upgrade Team; Contributors, JET

2018-03-01

We show for JET and ASDEX Upgrade, based on Thomson-scattering measurements, a clear correlation of the density limit of the tokamak H-mode high-confinement regime with the approach to the ideal ballooning instability threshold at the periphery of the plasma. It is shown that the MHD ballooning parameter at the separatrix position α_sep increases about linearly with the separatrix density normalized to Greenwald density, n_e, sep/n_GW for a wide range of discharge parameters in both devices. The observed operational space is found to reach at maximum n_e, sep/n_GW≈ 0.4 -0.5 at values for α_sep≈ 2 -2.5, in the range of theoretical predictions for ballooning instability. This work supports the hypothesis that the H-mode density limit may be set by ballooning stability at the separatrix.

MHD phenomena in advanced scenarios on ASDEX upgrade and the influence of localised electron heating and current drive

International Nuclear Information System (INIS)

Guenter, S.; Gude, A.; Hobirk, J.; Maraschek, M.; Peeters, A.G.; Pinches, S.D.; Schade, S.; Wolf, R.C.; Saarelma, S.

2001-01-01

MHD instabilities in advanced tokamak scenarios on the one hand are favourable as they can contribute to the stationarity of the current profiles and act as a trigger for the formation of internal transport barriers. In particular fishbone oscillations driven by fast particles arising from neutral beam injection (NBI) are shown to trigger internal transport barriers in low and reversed magnetic shear discharges. During the whistling down period of the fishbone oscillation the transport is reduced around the corresponding rational surface, leading to an increased pressure gradient. This behaviour is explained by the redistribution of the resonant fast particles resulting in a sheared plasma rotation due to the return current in the bulk plasma, which is equivalent to a radial electric field. On the other hand MHD instabilities limit the accessible operating regime. Ideal and resistive MHD modes such as double tearing modes, infernal modes and external kinks degrade the confinement or even lead to disruptions in ASDEX Upgrade reversed shear discharges. Localized electron cyclotron heating and current drive is shown to significantly affect the MHD stability of this type of discharges. (author)

MHD phenomena in advanced scenarios on ASDEX Upgrade and the influence of localized electron heating and current drive

International Nuclear Information System (INIS)

Guenter, S.; Gude, A.; Hobirk, J.; Maraschek, M.; Schade, S.; Wolf, R.C.; Saarelma, S.

2001-01-01

On the one hand, MHD instabilities in advanced tokamak scenarios are favourable as they can contribute to the stationarity of the current profiles and act as a trigger for the formation of internal transport barriers (ITBs). In particular, fishbone oscillations driven by fast particles arising from NBI are shown to trigger ITBs in low and reversed magnetic shear discharges. During the whistling down period of the fishbone oscillation the transport is reduced around the corresponding rational surface, leading to an increased pressure gradient. This behaviour could be explained by the redistribution of the resonant fast particles resulting in a sheared plasma rotation due to the return current in the bulk plasma, which is equivalent to a radial electric field. On the other hand, MHD instabilities limit the accessible operating regime. Ideal and resistive MHD modes such as double tearing modes, infernal modes and external kinks degrade the confinement or even lead to disruptions in ASDEX Upgrade reversed shear discharges. Localized electron cyclotron heating and current drive are shown to significantly affect the MHD stability of this type of discharge. (author)

Studies of tungsten erosion at the inner and outer main chamber wall of the ASDEX Upgrade tokamak

Science.gov (United States)

Tabasso, A.; Maier, H.; Roth, J.; Krieger, K.; ASDEX Upgrade Team

2001-03-01

A critical issue for the choice of main chamber first wall materials in future fusion devices such as ITER is the erosion rate due to bombardment by charge-exchange (CX) neutrals. Due to the relatively small flux density of impacting particles, respective measurements are only possible using long term samples (LTS) exposed for a full experimental campaign. In ASDEX Upgrade, CX erosion has been studied extensively for tungsten on the inner heat shield by placing four W coated tiles at different poloidal positions in one toroidal sector. During the same campaign, several LTS were placed at different poloidal and toroidal positions of the outer wall. 13C and Cu coated graphite probes were also used in order to test and compare W low and medium Z alternatives. The erosion results from the probes are compared with the calculated erosion [W. Eckstein, C. Garcia-Rosales, J. Roth, W. Ottenberger IPP Report, IPP 9/82]; [H. Verbeek, J. Stober, D. Coster, W. Eckstein, R. Schneider Nucl. Fus. 38 (1998) 12] and a figure of merit (F. of M.) between several materials is proposed which also takes into account the plasma isotope effect in CX erosion.

Studies of tungsten erosion at the inner and outer main chamber wall of the ASDEX Upgrade tokamak

International Nuclear Information System (INIS)

Tabasso, A.; Maier, H.; Roth, J.; Krieger, K.

2001-01-01

A critical issue for the choice of main chamber first wall materials in future fusion devices such as ITER is the erosion rate due to bombardment by charge-exchange (CX) neutrals. Due to the relatively small flux density of impacting particles, respective measurements are only possible using long term samples (LTS) exposed for a full experimental campaign. In ASDEX Upgrade, CX erosion has been studied extensively for tungsten on the inner heat shield by placing four W coated tiles at different poloidal positions in one toroidal sector. During the same campaign, several LTS were placed at different poloidal and toroidal positions of the outer wall. 13 C and Cu coated graphite probes were also used in order to test and compare W low and medium Z alternatives. The erosion results from the probes are compared with the calculated erosion [W. Eckstein, C. Garcia-Rosales, J. Roth, W. Ottenberger IPP Report, IPP 9/82]; [H. Verbeek, J. Stober, D. Coster, W. Eckstein, R. Schneider Nucl. Fus. 38 (1998) 12] and a figure of merit (F. of M.) between several materials is proposed which also takes into account the plasma isotope effect in CX erosion

The European Integrated Tokamak Modelling (ITM) effort: achievements and first physics results

International Nuclear Information System (INIS)

Falchetto, G.L.; Nardon, E.; Artaud, J.F.; Basiuk, V.; Huynh, Ph.; Imbeaux, F.; Coster, D.; Scott, B.D.; Coelho, R.; Alves, L.L.; Bizarro, João P.S.; Ferreira, J.; Figueiredo, A.; Figini, L.; Nowak, S.; Farina, D.; Kalupin, D.; Boulbe, C.; Faugeras, B.; Dinklage, A.

2014-01-01

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 edge MHD stability 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. (paper)

Electromagnetic and structural global model of the TF magnet system in ASDEX Upgrade

Energy Technology Data Exchange (ETDEWEB)

Zammuto, I., E-mail: irene.zammuto@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, EURATOM Association, D-85740 Garching (Germany); Streibl, B.; Giannone, L.; Herrmann, A.; Kallenbach, A.; Mertens, V. [Max-Planck-Institut für Plasmaphysik, EURATOM Association, D-85740 Garching (Germany)

2013-10-15

Highlights: ► An electromagnetic and structural FE 3D model is set up for ASDEX Upgrade. ► The model is benchmarked against the old design results, present displacement measurements. ► The benchmarked model is applied to the present plasma configurations, which have a different poloidal field distribution with respect to the design case. ► The different poloidal field influences the out-of-plane force distribution, thus requiring an update of the TF safety system. -- Abstract: The enhancements carried out in the tokamak ASDEX Upgrade (AUG) are oriented toward the preparation of the future physics-related activities of ITER and DEMO. To address the main ITER issues, plasma configurations with a wider operational limit (e.g. higher triangularity) are planned for the future experimental campaigns in AUG. To evaluate the mechanical impact on the toroidal field (TF) magnet system a combined electromagnetic and structural finite element model was set up. At first extensive benchmarks of the models are carried out against the AUG reference design configurations with respect to stress [1–3], lateral displacement measurements and poloidal flux pattern. The numerical model was then applied to a set of actual high triangularity (HT) configurations generated by a more favorable poloidal field (PF) current distribution made possible by an extension of the power supply system. The resulting change of the poloidal flux pattern and the lateral force distribution has consequences for the coil shear stress and vault stability. Both aspects are monitored by a safety system measuring the PF flux placed on top and bottom of the outer surface of two TF coils (TFCs) between vault and the TFC supporting structure, so called Turn Over Structure (TOS). The range of the new HT configurations has induced a modification of the flux pattern, so that an adaptation of safety system is required to protect the TFCs system. Following the same criteria of the old safety system [4,5], a new

The ASDEX upgrade digital video processing system for real-time machine protection

Energy Technology Data Exchange (ETDEWEB)

Drube, Reinhard, E-mail: reinhard.drube@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, EURATOM Association, Boltzmannstr. 2, 85748 Garching (Germany); Neu, Gregor [Max-Planck-Institut für Plasmaphysik, EURATOM Association, Boltzmannstr. 2, 85748 Garching (Germany); Cole, Richard H.; Lüddecke, Klaus [Unlimited Computer Systems GmbH, Seeshaupterstr. 15, 82393 Iffeldorf (Germany); Lunt, Tilmann; Herrmann, Albrecht [Max-Planck-Institut für Plasmaphysik, EURATOM Association, Boltzmannstr. 2, 85748 Garching (Germany)

2013-11-15

Highlights: • We present the Real-Time Video diagnostic system of ASDEX Upgrade. • We show the implemented image processing algorithms for machine protection. • The way to achieve a robust operating multi-threading Real-Time system is described. -- Abstract: This paper describes the design, implementation, and operation of the Video Real-Time (VRT) diagnostic system of the ASDEX Upgrade plasma experiment and its integration with the ASDEX Upgrade Discharge Control System (DCS). Hot spots produced by heating systems erroneously or accidentally hitting the vessel walls, or from objects in the vessel reaching into the plasma outer border, show up as bright areas in the videos during and after the reaction. A system to prevent damage to the machine by allowing for intervention in a running discharge of the experiment was proposed and implemented. The VRT was implemented on a multi-core real-time Linux system. Up to 16 analog video channels (color and b/w) are acquired and multiple regions of interest (ROI) are processed on each video frame. Detected critical states can be used to initiate appropriate reactions – e.g. gracefully terminate the discharge. The system has been in routine operation since 2007.

Edge density X-mode reflectometry of RF-heated plasmas on ASDEX

International Nuclear Information System (INIS)

Schubert, R.

1991-09-01

In the present work microwave reflectometry is extended to the outermost part of tokamak plasmas (n e ≅ 10 11 to 1.5x10 13 cm -3 ), which is subject to strong electron density fluctuations. The perturbations of electron density profile measurements by these fluctuations, which lead to strong modulations in intensity and phase of the reflected signal is analysed in detail. By increasing the frequency of the interference fringes to values between 800 kHz and 2.4 MHz it is possible to make reliable profile measurements even in the region of very strong fluctuations. Measurements in the low density region are only possible with reasonable errors in the X-mode (Eperpendicular toB), as only the cut-off frequency of this mode, in contrast to that of the O-mode (EparallelB), takes a finite value (f ce ) for n e ->O. Taking advantage of this property, a method is presented to calibrate the measurements on the first reflection, which occurs directly in front of the microwave antennas (1-4 mm from the opening) thus giving a high precision even in the outermost part of the plasma close to the microwave antennas. For the calculation of the electron density profile a new and numerically stable algorithm has been developed. Measurements in connection with Lower Hybrid have been made with a set of 2 reflectometer antennas installed in ASDEX. (orig./AH)

Advanced statistics for tokamak transport colinearity and tokamak to tokamak variation

International Nuclear Information System (INIS)

Riedel, K.S.

1989-03-01

This is a compendium of three separate articles on the statistical analysis of tokamak transport. The first article is an expository introduction to advanced statistics and scaling laws. The second analyzes two important problems of tokamak data---colinearity and tokamak to tokamak variation in detail. The third article generalizes the Swamy random coefficient model to the case of degenerate matrices. Three papers have been processed separately

Operation of ASDEX Upgrade with tungsten coated walls

International Nuclear Information System (INIS)

Rohde, V.

2002-01-01

An alternative for low-Z materials in the main chamber of a future fusion device are high-Z materials, but the maximal tolerable concentration in the plasma core is restricted. A step by step approach to employ tungsten at the central column of ASDEX Upgrade was started in 1999. Meanwhile almost the whole central column is covered with tiles, which were coated by PVD with tungsten. Up to now 9000 s of plasma discharge covering all relevant scenarios were performed. Routine operation of ASDEX Upgrade was not affected by the tungsten. Typical concentrations below 10 -5 were found. The tungsten concentration is mostly connected to the transport into the core plasma, not to the tungsten erosion. It can be demonstrated, that additional central heating can eliminate the tungsten accumulation. These experiments demonstrate the compatibility of fusion plasmas with W plasma facing components under reactor relevant conditions. The erosion pattern found by post mortem analysis indicates that the main effect is ion sputtering. The main erosion of tungsten seems to occur during plasma ramp-up and ramp-down. (author)

Full Tokamak discharge simulation and kinetic plasma profile control for ITER

International Nuclear Information System (INIS)

Hee Kim, S.

2009-10-01

Understanding non-linearly coupled physics between plasma transport and free-boundary equilibrium evolution is essential to operating future tokamak devices, such as ITER and DEMO, in the advanced tokamak operation regimes. To study the non-linearly coupled physics, we need a simulation tool which can self-consistently calculate all the main plasma physics, taking the operational constraints into account. As the main part of this thesis work, we have developed a full tokamak discharge simulator by combining a non-linear free-boundary plasma equilibrium evolution code, DINA-CH, and an advanced transport modelling code, CRONOS. This tokamak discharge simulator has been used to study the feasibility of ITER operation scenarios and several specific issues related to ITER operation. In parallel, DINA-CH has been used to study free-boundary physics questions, such as the magnetic triggering of edge localized modes (ELMs) and plasma dynamic response to disturbances. One of the very challenging tasks in ITER, the active control of kinetic plasma profiles, has also been studied. In the part devoted to free-boundary tokamak discharge simulations, we have studied dynamic responses of the free-boundary plasma equilibrium to either external voltage perturbations or internal plasma disturbances using DINA-CH. Firstly, the opposite plasma behaviour observed in the magnetic triggering of ELMs between TCV and ASDEX Upgrade has been investigated. Both plasmas experience similar local flux surface expansions near the upper G-coil set and passive stabilization loop (PSL) when the ELMs are triggered, due to the presence of the PSLs located inside the vacuum vessel of ASDEX Upgrade. Secondly, plasma dynamic responses to strong disturbances anticipated in ITER are examined to study the capability of the feedback control system in rejecting the disturbances. Specified uncontrolled ELMs were controllable with the feedback control systems. However, the specifications for fast H-L mode

ECRH experiments in ASDEX Upgrade

International Nuclear Information System (INIS)

Leuterer, F.; Guenter, S.; Hobirk, J.; Kirov, K.; Ryter, F.; Wolf, R.; Zohm, H.; Gantenbein, G.

2001-01-01

In ASDEX Upgrade ECRH and ECCD with a power of up to 2 MW has been used. With counter ECCD in discharges with an internal transport barrier we achieved an electron temperature of ≅13 keV. In low density ohmic plasmas we obtained an rf driven current of ≅80% both in co- and counter directions. Neoclassical tearing modes have been completly stabilised by driving a current on the resonant q-surface. Electron heat transport has been studied in standard L- and H-mode plasmas and can be described by a dependence on a critical electron temperature gradient

Commissioning and initial operation experience with ASDEX Upgrade's new real-time control and data acquisition

International Nuclear Information System (INIS)

Raupp, G.; Behler, K.; Cole, R.; Engelhardt, K.; Lohs, A.; Lueddecke, K.; Treutterer, W.; Neu, G.; Vijverberg, T.; Zasche, D.; Zehetbauer, Th.

2006-01-01

ASDEX Upgrade was equipped with a distributed real-time (RT) control and diagnostic system. Application processes can be freely mapped onto controllers. They communicate through RT data exchanged via a shared memory network. Processes run self-organized with a data-driven scheme, i.e. a process executes when all required data has become available and produced data drive other processes waiting for these. The process chain starts periodically through a time-synchronous cycle master process to deterministically execute closed-loop control. Generic processes were implemented for feedback of plasma position and shape, and for performance control with fuelling and heating systems, for evaluation and monitoring of plasma quantities and Tokamak components, and generation of reference values in RT. Upon commissioning the system was speed-optimized to run a 1.6 ms cycle. Methods to exchange data and time information in RT and operate distributed data-driven processes work efficiently and reliably. The ability to freely map processes to computing nodes and RT data to generic processes provides outstanding configurational flexibility for optimizing system performance or supporting in situ replay and simulation runs. In the computation of reference values in RT acting on underlying feedback controllers, demonstrated for soft-landing, we see great potential to counteract instabilities or optimize pulses

Poloidal structure of the plasma response to n = 1 Resonant Magnetic Perturbations in ASDEX Upgrade

Science.gov (United States)

Marrelli, L.; Bettini, P.; Piovesan, P.; Terranova, D.; Giannone, L.; Igochine, V.; Maraschek, M.; Suttrop, W.; Teschke, M.; Liu, Y. Q.; Ryan, D.; Eurofusion Mst1 Team; ASDEX Upgrade Team

2017-10-01

The hybrid scenario, a candidate for high-beta steady-state tokamak operations, becomes highly sensitive to 3D magnetic field near the no-wall limit. A predictive understanding of the plasma response to 3D fields near ideal MHD limits in terms of validated MHD stability codes is therefore important in order to safely operate future devices. Slowly rotating (5 - 10 Hz) n = 1 external magnetic fields have been applied in hybrid discharges in ASDEX Upgrade for an experimental characterization: the global n = 1 kink response has been measured by means of SXR and complete poloidal arrays of bθ probes located at different toroidal angles and compared to predictions of MHD codes such as MARS-F and V3FIT-VMEC. A Least-Squares Spectral Analysis approach has been developed together with a Monte Carlo technique to extract the small plasma response and its confidence interval from the noisy magnetic signals. MARS-F correctly reproduces the poloidal structure of the n = 1 measurements: for example, the dependence of the dominant poloidal mode number at the plasma edge from q95 is the same as in the experiment. Similar comparisons with V3FIT-VMEC and will be presented. See author list of ``H. Meyer et al. 2017 Nucl. Fusion 57 102014''.

New drive converter and digital control for the pulsed power supply system of ASDEX Upgrade

Energy Technology Data Exchange (ETDEWEB)

Käsemann, Claus-Peter, E-mail: c.p.kaesemann@ipp.mpg.de [Max Planck Institute for Plasma Physics, Boltzmannstraße 2, 85748 Garching (Germany); Jacob, Christian; Nguyen, Hong Ha; Stobbe, Ferdinand; Mayer, Alois [Max Planck Institute for Plasma Physics, Boltzmannstraße 2, 85748 Garching (Germany); Sachs, Edgar; Klein, Reiner [Siemens AG, Industrial Automation Systems, Gleiwitzer Straße 555, 90475 Nürnberg (Germany)

2015-10-15

Highlights: • IGBT converter system with integrated control. • Proven technology reduces time and budget. • Flexibility to be integrated into a 35 years old installation. • Stable control algorithms for static and dynamic speed control. • Possibilities for active and reactive power management. - Abstract: Safety and reliability are major issues for the ASDEX Upgrade (AUG) pulsed power supply systems. To avoid long downtimes during an experimental campaign, fault-prone components have to be identified and treated early. This becomes even more important due to the AUG participation in the EUROfusion Medium Sized Tokamak (MST) program. Operating equipment which is up to 40 years old adds additional complications. This contribution describes one such example where a 35 year old flywheel generator at AUG was identified as fault-prone and pre-emptively upgraded with a new drive converter with integrated control. Most challenging was to adapt a modern converter, originally designed for wind turbines, toward a drive system for a flywheel-motor-generator system. To identify the layout of the controller and the control parameters, accurate modeling and comprehensive simulations were performed. This effort paid off during commissioning and measuring results verified the calculated design values. Finally, the system shows good performance during AUG plasma experiments.

Sensitivity study of a proposed polarimetry diagnostic on ASDEX upgrade

International Nuclear Information System (INIS)

Callaghan, H.P.; McCarthy, P.J.

1994-09-01

ASDEX-Upgrade currently uses FIR interferometry (DCN, 195 μm) as a technique for measuring line integrated electron density along eight chords of the plasma cross-section. A polarimetry diagnostic based on Faraday rotation using the existing setup would yield ∫ n e B.dl along the same chords which, in combination with the ∫ n e dl measurements, would provide additional information about the poloidal magnetic field. This would be helpful for reconstructing the q(ψ) profile, which is difficult to recover from external magnetic measurements alone. A sensitivity study to determine the effectiveness of adding polarimetry to ASDEX Upgrade is carried out using function parameterization on a simulated equilibrium database, together with a database of randomly chosen density profiles with four degrees of freedom. The robustness of the recovery in the presence of measurement noise and the effects of plasma birefringence are taken into account. (orig.)

Runaway electron mitigation by 3D fields in the ASDEX-Upgrade experiment

Science.gov (United States)

Gobbin, M.; Li, L.; Liu, Y. Q.; Marrelli, L.; Nocente, M.; Papp, G.; Pautasso, G.; Piovesan, P.; Valisa, M.; Carnevale, D.; Esposito, B.; Giacomelli, L.; Gospodarczyk, M.; McCarthy, P. J.; Martin, P.; Suttrop, W.; Tardocchi, M.; Teschke, M.; the ASDEX Upgrade Team; the EUROfusion MST1 Team

2018-01-01

Disruption-generated runaway electron (RE) beams represent a severe threat for tokamak plasma-facing components in high current devices like ITER, thus motivating the search of mitigation techniques. The application of 3D fields might aid this purpose and recently was investigated also in the ASDEX Upgrade experiment by using the internal active saddle coils (termed B-coils). Resonant magnetic perturbations (RMPs) with dominant toroidal mode number n = 1 have been applied by the B-coils, in a RE specific scenario, before and during disruptions, which are deliberately created via massive gas injection. The application of RMPs affects the electron temperature profile and seemingly changes the dynamics of the disruption; this results in a significantly reduced current and lifetime of the generated RE beam. A similar effect is observed also in the hard-x-ray (HXR) spectrum, associated to RE emission, characterized by a partial decrease of the energy content below 1 MeV when RMPs are applied. The strength of the observed effects strongly depends on the upper-to-lower B-coil phasing, i.e. on the poloidal spectrum of the applied RMPs, which has been reconstructed including the plasma response by the code MARS-F. A crude vacuum approximation fails in the interpretation of the experimental findings: despite the relatively low β (< 0.5 % ) of these discharges, a modest amplification (factor of 2) of the edge kink response occurs, which has to be considered to explain the observed suppression effects.

Tokamak

International Nuclear Information System (INIS)

Wesson, John.

1996-01-01

This book is the first compiled collection about tokamak. At first chapter tokamak is represented from fusion point of view and also the necessary conditions for producing power. The following chapters are represent plasma physics, the specifications of tokamak, plasma heating procedures and problems related to it, equilibrium, confinement, magnetohydrodynamic stability, instabilities, plasma material interaction, plasma measurement and experiments regarding to tokamak; an addendum is also given at the end of the book

An Asdex-type divertor for ITER

International Nuclear Information System (INIS)

Fowler, T.K.

1989-01-01

An Asdex-type local divertor is proposed for ITER consisting of a copper poloidal field coil adjacent to the plasma. Estimates indicate that the power consumption is acceptable. Advantages would be a much reduced heat load not very sensitive to magnetic perturbations. A disadvantage is the finite lifetime under neutron bombardment that would require periodic replacement of the divertor coils in a reactor, but probably not in ITER because of its limited fluence. Another disadvantage would be poorer blanket coverage unless the divertor coil itself incorporates breeding material. 3 figs

Recent ECRH results in ASDEX upgrade

International Nuclear Information System (INIS)

Leuterer, F.; Dux, R.; Gantenbein, G.

2003-01-01

This paper provides an overview on recent experimental results obtained in ASDEX Upgrade using electron cyclotron heating and current drive. The following topics are covered: determination of the power deposition profile, modulated power deposition, studies of the electron heat transport via power balance and heat wave analysis and a comparison with turbulent transport theory, generation of an internal transport barrier for the electron heat flux, impact of electron cyclotron resonance heating (ECRH) on particle and impurity transport, and studies related to neoclassical tearing modes and to sawteeth. (author)

Plasma rotation and ion temperature measurements by collective Thomson scattering at ASDEX Upgrade

DEFF Research Database (Denmark)

Stejner Pedersen, Morten; Nielsen, Stefan Kragh; Jacobsen, Asger Schou

2015-01-01

We present the first deuterium ion temperature and rotation measurements by collective Thomson scattering at ASDEX Upgrade. The results are in general agreement with boron-based charge exchange recombination spectroscopy measurements and consistent with neoclassical simulations for the plasma sce...... scenario studied here. This demonstration opens the prospect for direct non-perturbative measurements of the properties of the main ion species in the plasma core with applications in plasma transport and confinement studies.......We present the first deuterium ion temperature and rotation measurements by collective Thomson scattering at ASDEX Upgrade. The results are in general agreement with boron-based charge exchange recombination spectroscopy measurements and consistent with neoclassical simulations for the plasma...

2D heat flux pattern in ASDEX upgrade L-mode with magnetic perturbation

Energy Technology Data Exchange (ETDEWEB)

Faitsch, Michael; Sieglin, Bernhard; Eich, Thomas; Herrmann, Albrecht; Suttrop, Wolfgang [Max-Planck-Institute for Plasma Physics, Boltzmannstr. 2, D-85748 Garching (Germany); Collaboration: the ASDEDX Upgrade Team

2016-07-01

A future fusion reactor is likely to operate in high confinement mode (H-mode). This mode is associated with a periodic instability at the plasma edge that expels particles and energy. This instability is called edge localized mode (ELM). External magnetic perturbation (MP) is one technique that is thought to be able to mitigate or even suppress large ELMs in next step fusion devices such as ITER, where the ELM induced heat load for unmitigated ELMs might limit the lifetime of the divertor. Applying an external magnetic perturbation breaks the axisymmetry and leads to a 2D steady state heat flux pattern at the divertor. The ASDEX Upgrade tokamak is equipped with 16 perturbation coils, 8 above (upper row) and 8 below (lower row) the outer mid plane, toroidal equally distributed. A high resolution infra red system is measuring the heat flux at the outer target at a fixed toroidal position with a resolution of around 0.6 mm. In order to measure the 2D structure a slow rotation of the MP field was applied (1 Hz) with a toroidal mode number n=2. The differential phase between the upper and lower row was changed to investigate the effect of the alignment with the field lines at the edge. The density was varied to study the density dependence of the heat transport with applied external MP and compare it to the axisymmetric scenario.

Update on the ASDEX Upgrade data acquisition and data management environment

Energy Technology Data Exchange (ETDEWEB)

Behler, K., E-mail: karl.behler@ipp.mpg.de; Blank, H.; Buhler, A.; Drube, R.; Eixenberger, H.; Engelhardt, K.; Lohs, A.; Merkel, R.; Raupp, G.; Treutterer, W.

2014-05-15

Highlights: • An exponential growth of data amount was managed over more than twenty years of experiment operation. • Continuous adaptation of the diagnostic software and configuration keeps track with actual experiment demands. • A great number of distributed, varying diagnostics is centrally managed. - Abstract: It has been a while since it had been reported on the status of ASDEX Upgrade data acquisition (DAQ) and data management environment. An update on changes, expansions, and enhancements applied in the last years will be given. The acquired amount of data per shot increased from 4 GiB to 40 GiB in eight years. Network, storage, and archive challenges have been managed by stepwise improvements. New DAQ techniques have been introduced to replace outdated technologies. Real-time diagnostics speed-up data provisioning and contribute to feedback control. Information technology applied to ASDEX Upgrade is under permanent change. Recent and future steps are outlined.

Operational experience with reactive power control methods optimized for tokamak power supplies

International Nuclear Information System (INIS)

Sihler, C.; Huart, M.; Kaesemann, C.-P.; Streibl, B.

2003-01-01

The power and energy of the ASDEX Upgrade (AUG) tokamak are provided by two separate 10.5 kV, 110-85 Hz networks based on the flywheel generators EZ3-EZ4 in addition to the generator EZ2 dedicated to the toroidal field coil. The 10.5 kV networks supply the thyristor converters allowing fast control of the DC currents in the AUG poloidal field coils. Two methods for improving the load power factor in the present experimental campaign of AUG have been investigated, namely the control of the phase-to-neutral voltage in thyristor converters fitted with neutral thyristors, such as the new 145 MVA modular thyristor converter system (Group 6), and reactive power control achieved by means of static VAr compensators (SVC). The paper shows that reliable compensation up to 90 MVAr was regularly achieved and that electrical transients in SVC modules can be kept at an acceptable level. The paper will discuss the results from the reactive power reduction by SVC and neutral thyristor control and draw a comparative conclusion

Determination of the stochastic layer properties induced by magnetic perturbations via heat pulse experiments at ASDEX upgrade

Directory of Open Access Journals (Sweden)

D. Brida

2017-08-01

Full Text Available A heat pulse experiment was carried out in the tokamak ASDEX Upgrade to estimate the stochastic layer width of a deuterium L-mode discharge with externally applied Magnetic Perturbations. The method relies on the deposition of ECRH pulses in the plasma edge while measuring the divertor target heat flux with high temporal resolution IR thermography and Langmuir probes. The experimental results were compared to simulations of the time dependent heat pulse propagation on a constant plasma background with the EMC3-EIRENE code package, using an ad-hoc screening model. If no screening was taken into account in the simulations a decrease in the characteristic heat pulse propagation time was observed, which shows that the heat transport is enhanced compared to the screened cases. No such enhancement was found in the experiment, indicating strong screening. In further simulations the effect of screening on the target fluxes was investigated for varying densities. For low densities it was found that screening reduces the strike line splitting strongly, while for higher densities no strong strike line splitting was found, independent of the screening degree. For strongly detached L-mode conditions with MPs experiments at AUG indicate that the lobe structures vanish completely.

Tungsten erosion and redeposition in the all-tungsten divertor of ASDEX Upgrade

Energy Technology Data Exchange (ETDEWEB)

Mayer, M; Krieger, K; Matern, G; Neu, R; Rasinski, M; Rohde, V; Sugiyama, K; Wiltner, A [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Boltzmannstrasse 2, 85748 Garching (Germany); Andrzejczuk, M; Fortuna-Zalesna, E; Kurzydlowski, K J; Zielinski, W [Faculty of Materials Science and Engineering, Warsaw University of Technology, Association EURATOM-IPPLM, 02-507 Warsaw (Poland); Hakola, A; Koivuranta, S; Likonen, J [VTT Materials for Power Engineering, EURATOM Association, PO Box 1000, FI-02044 VTT (Finland); Ramos, G [CICATA-Qro, Instituto Politecnico Nacional, Queretaro (Mexico); Dux, R, E-mail: matej.mayer@ipp.mpg.de

2009-12-15

Net erosion and deposition of tungsten (W) in the ASDEX Upgrade divertor were determined after the 2007 campaign by using thin W marker stripes. ASDEX Upgrade had full-W plasma-facing components during this campaign. The inner divertor and the roof baffle were net W deposition areas with a maximum deposition of about 1x10{sup 18} W-atoms cm{sup -2} in the private flux region below the inner strike point. Net erosion of W was observed in the whole outer divertor, with the largest erosion close to the outer strike point. Only a small fraction of the W eroded in the main chamber and in the outer divertor was found in redeposits in the inner divertor, while a large fraction was either redeposited at unidentified places in the main chamber or has formed dust.

Energy and particle control characteristics of the ASDEX Upgrade 'LYRA' divertor

International Nuclear Information System (INIS)

Kaufmann, M.; Bosch, H.S.; Herrmann, A.

1999-01-01

In 1997 the new 'LYRA' divertor went into operation at ASDEX Upgrade and the neutral beam heating power was increased to 20 MW by installation of a second injector. This leads to the relatively high value of P/R of 12 MW/m. It has been shown that the ASDEX Upgrade LYRA divertor is capable of handling such high heating powers. Measurements presented in this paper reveal a reduction of the maximum heat flux in the LYRA divertor by more than a factor of two compared to the open Divertor I. This reduction is caused by radiative losses inside the divertor region. Carbon radiation cools the divertor plasma down to a few eV where hydrogen radiation losses become significant. They are increased due to an effective reflection of neutrals into the hot separatrix region. B2-Eirene modelling of the performed experiments supports the experimental findings and refines the understanding of loss processes in the divertor region. (author)

Energy and particle control characteristics of the ASDEX Upgrade 'LYRA' divertor

International Nuclear Information System (INIS)

Kaufmann, M.; Bosch, H.-S.; Herrmann, A.

2001-01-01

In 1997 the new 'LYRA' divertor went into operation at ASDEX Upgrade and the neutral beam heating power was increased to 20 MW by installation of a second injector. This leads to the relatively high value of P/R of 12 MW/m. It has been shown that the ASDEX Upgrade LYRA divertor is capable of handling such high heating powers. Measurements presented in this paper reveal a reduction of the maximum heat flux in the LYRA divertor by more than a factor of two compared to the open Divertor I. This reduction is caused by radiative losses inside the divertor region. Carbon radiation cools the divertor plasma down to a few eV where hydrogen radiation losses become significant. They are increased due to an effective reflection of neutrals into the hot separatrix region. B2-Eirene modelling of the performed experiments supports the experimental findings and refines the understanding of loss processes in the divertor region. (and others)

Stability and propagation of the high field side high density front in the fluctuating state of detachment in ASDEX Upgrade

Directory of Open Access Journals (Sweden)

P. Manz

2017-08-01

Full Text Available During detachment a structure of strongly enhanced density develops close to the inner target. Its dynamics is approximated by those of radiative fluctuations appearing at a similar position and studied by means of a time-delay-estimation technique in the ASDEX Upgrade tokamak. Compared to theory the dynamics can be described as follows: at increasing density the ionization front moves upstream to reduce ionization radiation in order to balance the increased recombination radiation. The recombination zone stays close to the target strike point. The parallel motion of the ionization front is determined by the perpendicular neutral motion. The divertor nose constitutes an obstacle for the perpendicular neutral flux from the target to the region above the X-point. Passing into this shadow the neutral flux above the X-point is strongly reduced, the ionization front fades away and the heat flux from upstream can increase the temperature in the recombination region, subsequently reducing recombination and reforming an ionization front below the X-point. A cyclic reformation of the ionization front propagating from below to above the X-point occurs leading to a fluctuation as observed in the experiment.

ASDEX Upgrade-JT-60U comparison and ECRH power requirements for NTM stabilization in ITER

International Nuclear Information System (INIS)

Urso, L.; Zohm, H.; Maraschek, M.; Poli, E.; Isayama, A.

2010-01-01

Neoclassical tearing modes (NTMs) are experimentally controlled with local electron cyclotron current drive (ECCD) and the island width decay during NTM stabilization is modelled using the so-called modified Rutherford equation (MRE). In this paper, a modelling of the MRE is carried out and simulations of the island width decay are compared with the experimentally observed ones in order to fit the two free machine-independent parameters present in the equation. A systematic study on a database of NTM stabilization discharges from ASDEX Upgrade and JT-60U is done for extrapolating the ECCD power requirements for ITER. The extrapolation to ITER of the NTM stabilization results from ASDEX Upgrade and JT-60U shows that 10 MW of ECCD power are enough to stabilize large NTMs. The 10 MW power estimate for ITER is based on the assumption that the free parameters in the MRE are machine independent. Indeed, this assumption is verified in this paper for ASDEX Upgrade and JT-60U. An interesting consequence of the relatively modest power requirement for ITER is that the installed 20 MW will suffice for simultaneous 2/1 and 3/2 NTM stabilization.

Linear stability of microtearing modes in ASDEX

International Nuclear Information System (INIS)

Giannone, L.

1987-12-01

The linear stability of microtearing modes in typical ASDEX discharges have been calculated. In the case of Ohmic discharges it was found that unstable modes are predicted to be located towards the centre of the plasma. For L and H discharges the zone of instability shifts towards the plasma edge. The interpretation of an increase or decrease in the amplitude of broadband magnetic fluctuations during L and H discharges must be interpreted with caution, since the amplitude observed is strongly dependent on the radial position of the instability. (orig./GG)

Experimental studies and modelling of high radiation and high density plasmas in the ASDEX upgrade tokamak

Energy Technology Data Exchange (ETDEWEB)

Casali, Livia

2015-11-24

Fusion plasmas contain impurities, either intrinsic originating from the wall, or injected willfully with the aim of reducing power loads on machine components by converting heat flux into radiation. The understanding and the prediction of the effects of these impurities and their radiation on plasma performances is crucial in order to retain good confinement. In addition, it is important to understand the impact of pellet injection on plasma performance since this technique allows higher core densities which are required to maximise the fusion power. This thesis contributes to these efforts through both experimental investigations and modelling. Experiments were conducted at ASDEX Upgrade which has a full-W wall. Impurity seeding was applied to H-modes by injecting nitrogen and also medium-Z impurities such as Kr and Ar to assess the impact of both edge and central radiation on confinement. A database of about 25 discharges has been collected and analysed. A wide range of plasma parameters was achieved up to ITER relevant values such as high Greenwald and high radiation fractions. Transport analyses taking into account the radiation distribution reveal that edge localised radiation losses do not significantly impact confinement as long as the H-mode pedestal is sustained. N seeding induces higher pedestal pressure which is propagated to the core via profile stiffness. Central radiation must be limited and controlled to avoid confinement degradation. This requires reliable control of the impurity concentration but also possibilities to act on the ELM frequency which must be kept high enough to avoid an irreversible impurity accumulation in the centre and the consequent radiation collapse. The key role of the f{sub ELM} is confirmed also by the analysis of N+He discharges. Non-coronal effects affect the radiation of low-Z impurities at the plasma edge. Due to the radial transport, the steep temperature gradients and the ELM flush out, a local equilibrium cannot be

Continuous tokamaks

International Nuclear Information System (INIS)

Peng, Y.K.M.

1978-04-01

A tokamak configuration is proposed that permits the rapid replacement of a plasma discharge in a ''burn'' chamber by another one in a time scale much shorter than the elementary thermal time constant of the chamber first wall. With respect to the chamber, the effective duty cycle factor can thus be made arbitrarily close to unity minimizing the cyclic thermal stress in the first wall. At least one plasma discharge always exists in the new tokamak configuration, hence, a continuous tokamak. By incorporating adiabatic toroidal compression, configurations of continuous tokamak compressors are introduced. To operate continuous tokamaks, it is necessary to introduce the concept of mixed poloidal field coils, which spatially groups all the poloidal field coils into three sets, all contributing simultaneously to inducing the plasma current and maintaining the proper plasma shape and position. Preliminary numerical calculations of axisymmetric MHD equilibria in continuous tokamaks indicate the feasibility of their continued plasma operation. Advanced concepts of continuous tokamaks to reduce the topological complexity and to allow the burn plasma aspect ratio to decrease for increased beta are then suggested

MHD equilibrium identification on ASDEX-Upgrade

International Nuclear Information System (INIS)

McCarthy, P.J.; Schneider, W.; Lakner, K.; Zehrfeld, H.P.; Buechl, K.; Gernhardt, J.; Gruber, O.; Kallenbach, A.; Lieder, G.; Wunderlich, R.

1992-01-01

A central activity accompanying the ASDEX-Upgrade experiment is the analysis of MHD equilibria. There are two different numerical methods available, both using magnetic measurements which reflect equilibrium states of the plasma. The first method proceeds via a function parameterization (FP) technique, which uses in-vessel magnetic measurements to calculate up to 66 equilibrium parameters. The second method applies an interpretative equilibrium code (DIVA) for a best fit to a different set of magnetic measurements. Cross-checks with the measured particle influxes from the inner heat shield and the divertor region and with visible camera images of the scrape-off layer are made. (author) 3 refs., 3 figs

Attempt to model the edge turbulence of a tokamak as a random superposition of eddies

Energy Technology Data Exchange (ETDEWEB)

Endler, M; Theimer, G; Weinlich, M; Carlson, A; Giannone, L.; Niedermeyer, H; Rudyj, A [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

1993-12-31

Turbulence is considered to be the most likely origin of the anomalous transport in tokamaks. Although the main interest is focussed on the bulk plasma, transport in the scrape-off layer is very important for reactor design. For this reason extensive experimental investigations of the edge turbulence were performed on the ASDEX divertor tokamak. Langmuir probe arrays were used in the floating potential mode and in the ion saturation mode to measure the poloidal distribution of density and plasma potential fluctuations neglecting temperature fluctuations. Density fluctuations integrated radially over the boundary layer were derived from H{sub {alpha}}-measurements. Data from up to 16 channels were sampled with a frequency of 1 MHz during time windows of 1 s. Often one parameter like the plasma density or the radial probe position were scanned during this interval. It is impossible to derive physical mechanisms directly from these statistical observations. We draw general conclusions about the physics involved from the entity of observations and propose a set of basic effects to include in a theoretical model. Being still unable to solve the complex nonlinear problem of the fully developed turbulence exactly we attempt to describe the turbulence with a simple non-self-consistent statistical model. This allows to derive plausible physical interpretations of several features of the statistical functions and may be used as a guide-line for the development of a manageable theoretical model. (author) 6 refs., 3 figs.

Fast-ion transport studies using FIDA spectroscopy at the ASDEX Upgrade tokamak

International Nuclear Information System (INIS)

Geiger, Benedikt

2013-01-01

A good confinement of fast-ions, i.e. ions with energies above the thermal energy, is essential for the success of fusion devices as it determines, amongst others, the plasma performance and the heating and current drive efficiencies. In case of a turbulent or magneto-hydrodynamic (MHD) active background plasma, various mechanisms have to be considered in order to estimate the spatial distribution of the fast-ions: the slowing down and radial diffusion by Coulomb collisions on electrons and ions, the effect of potential fluctuations and the effect of perturbations of the magnetic field structure. These can lead to a broadening of the fast-ion distribution function which is not yet completely understood. At the fusion experiment ASDEX Upgrade, the fast-ions are generated by heating sources such as neutral beam injection (NBI). Their transport properties can be studied by a fast-ion D-alpha (FIDA) spectroscopy diagnostic which has been built in the framework of this thesis. Through charge exchange reactions with neutrals, fast-ions can receive a bound electron and emit Balmer alpha line radiation. This so-called FIDA radiation can be measured with large Doppler shifts and is localized along the NBI path where a high density of neutrals is present. The FIDA diagnostic uses radially distributed lines of sight that intersect, in the horizontal and in the vertical plane, the path of a 2.5 MW NBI heating source. Thereby different parts of the fast-ion phase space above 25 keV can be analyzed. To interpret the FIDA radiation quantitatively, a forward modelling code has been implemented, tested and further developed. The code calculates, based on theoretical fast-ion distribution functions, synthetic FIDA spectra that can be compared to the measurement. In MHD-quiescent plasmas, the possible effect of turbulence on the fast-ion transport has been investigated with the FIDA diagnostic. The measurements obtained under different experimental conditions, such as during on- and

Simultaneous Measurements of Electrostatic and Magnetic Fluctuations in ASDEX Upgrade Edge Plasma

DEFF Research Database (Denmark)

Ionita, Codrina; Vianello, Nicola; Müller, H.W.

2009-01-01

In ASDEX Upgrade (AUG) electrostatic and magnetic fluctuations in the edge plasma region were measured simultaneously during ELMy H-mode (high confinement) plasmas and L-mode (low confinement) plasmas and during a transition between the two modes. A special probe was used containing six Langmuir...

Time and space-resolved energy flux measurements in the divertor of the ASDEX tokamak by computerized infrared thermography

International Nuclear Information System (INIS)

Mueller, E.R.; Steinmetz, K.; Bein, B.K.

1984-06-01

A new, fully computerized and automatic thermographic system has been developed. Its two central components are an AGA THV 780 infrared camera and a PDP-11/34 computer. A combined analytical-numerical method of solving the 1-dimensional heat diffusion equation for a solid of finite thickness bounded by two parallel planes was developed. In high-density (anti nsub(e) = 8 x 10 13 cm -3 ) neutral-beam-heated (L-mode) divertor discharges in ASDEX, the power deposition on the neutralizer plates is reduced to about 10-15% of the total heating power, owing to the inelastic scattering of the divertor plasma from a neutral gas target. Between 30% and 40% of the power is missing in the global balance. The power flow inside the divertor chambers is restricted to an approximately 1-cm-thick plasma scrape-off layer. This width depends only weakly on the density and heating power. During H-phases free of Edge Localized Mode (ELM) activity the energy flow into the divertor is blocked. During H-phases with ELM activity the energy is expelled into the divertor in very short intense pulses (several MW for about one hundred μs). Sawtooth events are able to transport significant amounts of energy from the plasma core to the peripheral zones and the scrape-off layer, and they are frequently correlated with transitions from the L to the H mode. (orig./AH)

Heat and density pulse propagation in ASDEX

International Nuclear Information System (INIS)

Giannone, L.; Riedel, K.; Stroth, U.; Eberhagen, A.; Gruber, O.; Mertens, V.

1990-01-01

Experimental measurements of the electron thermal conductivity, derived from the radial propagation of the heat pulse generated by a sawtooth crash, have consistently yielded larger values than those obtained by power balance. It has been proposed that this discrepancy could be the result of the coupling of density and temperature perturbations. Numerical modelling of heat and density pulse propagation on ASDEX has been used to address this question. In addition, measurements at various electron densities and in hydrogen and deuterium were undertaken, with the aim of providing a broad base of experimental measurements for testing the various transport models proposed. (orig.)

Recent results of reflectometry on ASDEX-upgrade

International Nuclear Information System (INIS)

Manso, M.; Serra, F.; Numes, I.; Cupido, L.; Grossmann, V.; Meneses, L.; Santos, J.; Silva, A.; Silva, F.; Varela, P.; Vergamota, S.; Maraschek, M.

1999-01-01

Reflectometry is well known to be very sensitive to plasma density fluctuations. The study of plasma response in broadband frequency operation is concentrated on the obtention of the main peak and many techniques have been developed to filter the unwanted components. In comparison little work has been done to understand the remaining part of the signal. This paper presents some recent results about plasma fluctuations obtained with FM-reflectometry on ASDEX-Upgrade. They demonstrate the rich content information of both the fixed frequency and broadband signals and suggest that they can be used in a complementary way. (A.L.B.)

Time accuracy requirements for fusion experiments: A case study at ASDEX Upgrade

International Nuclear Information System (INIS)

Raupp, Gerhard; Behler, Karl; Eixenberger, Horst; Fitzek, Michael; Kollotzek, Horst; Lohs, Andreas; Lueddecke, Klaus; Mueller, Peter; Merkel, Roland; Neu, Gregor; Schacht, Joerg; Schramm, Gerold; Treutterer, Wolfgang; Zasche, Dieter; Zehetbauer, Thomas

2010-01-01

To manage and operate a fusion device and measure meaningful data an accurate and stable time is needed. As a benchmark, we suggest to consider time accuracy as sufficient if it is better than typical data errors or process timescales. This allows to distinguish application domains and chose appropriate time distribution methods. For ASDEX Upgrade a standard NTP method provides Unix time for project and operation management tasks, and a dedicated time system generates and distributes a precise experiment time for physics applications. Applying the benchmark to ASDEX Upgrade shows that physics measurements tagged with experiment time meet the requirements, while correlation of NTP tagged operation data with physics data tagged with experiment time remains problematic. Closer coupling of the two initially free running time systems with daily re-sets was an efficient and satisfactory improvement. For ultimate accuracy and seamless integration, however, continuous adjustment of the experiment time clock frequency to NTP is needed, within frequency variation limits given by the benchmark.

Management of complex data flows in the ASDEX Upgrade plasma control system

International Nuclear Information System (INIS)

Treutterer, Wolfgang; Neu, Gregor; Raupp, Gerhard; Zasche, Dieter; Zehetbauer, Thomas; Cole, Richard; Lüddecke, Klaus

2012-01-01

Highlights: ► Control system architectures with data-driven workflows are efficient, flexible and maintainable. ► Signal groups provide coherence of interrelated signals and increase the efficiency of process synchronisation. ► Sample tags indicating sample quality form the fundament of a local event handling strategy. ► A self-organising workflow benefits from sample tags consisting of time stamp and stream activity. - Abstract: Establishing adequate technical and physical boundary conditions for a sustained nuclear fusion reaction is a challenging task. Phased feedback control and monitoring for heating, fuelling and magnetic shaping is mandatory, especially for fusion devices aiming at high performance plasmas. Technical and physical interrelations require close collaboration of many components in sequential as well as in parallel processing flows. Moreover, handling of asynchronous, off-normal events has become a key element of modern plasma performance optimisation and machine protection recipes. The manifoldness of plasma states and events, the variety of plant system operation states and the diversity in diagnostic data sampling rates can hardly be mastered with a rigid control scheme. Rather, an adaptive system topology in combination with sophisticated synchronisation and process scheduling mechanisms is suited for such an environment. Moreover, the system is subject to real-time control constraints: response times must be deterministic and adequately short. Therefore, the experimental tokamak device ASDEX Upgrade employs a discharge control system DCS, whose core has been designed to meet these requirements. In the paper we will compare the scheduling schemes for the parallelised realisation of a control workflow and show the advantage of a data-driven workflow over a managed workflow. The data-driven workflow as used in DCS is based on signals connecting process outputs and inputs. These are implemented as real-time streams of data samples

Tokamak Systems Code

International Nuclear Information System (INIS)

Reid, R.L.; Barrett, R.J.; Brown, T.G.

1985-03-01

The FEDC Tokamak Systems Code calculates tokamak performance, cost, and configuration as a function of plasma engineering parameters. This version of the code models experimental tokamaks. It does not currently consider tokamak configurations that generate electrical power or incorporate breeding blankets. The code has a modular (or subroutine) structure to allow independent modeling for each major tokamak component or system. A primary benefit of modularization is that a component module may be updated without disturbing the remainder of the systems code as long as the imput to or output from the module remains unchanged

Experiments on transient melting of tungsten by ELMs in ASDEX Upgrade

Science.gov (United States)

Krieger, K.; Balden, M.; Coenen, J. W.; Laggner, F.; Matthews, G. F.; Nille, D.; Rohde, V.; Sieglin, B.; Giannone, L.; Göths, B.; Herrmann, A.; de Marne, P.; Pitts, R. A.; Potzel, S.; Vondracek, P.; ASDEX-Upgrade Team; EUROfusion MST1 Team

2018-02-01

Repetitive melting of tungsten by power transients originating from edge localized modes (ELMs) has been studied in ASDEX Upgrade. Tungsten samples were exposed to H-mode discharges at the outer divertor target plate using the divertor manipulator II (DIM-II) system (Herrmann et al 2015 Fusion Eng. Des. 98-9 1496-9). Designed as near replicas of the geometries used also in separate experiments on the JET tokamak (Coenen et al 2015 J. Nucl. Mater. 463 78-84 Coenen et al 2015 Nucl. Fusion 55 023010; Matthews et al 2016 Phys. Scr. T167 7), the samples featured a misaligned leading edge and a sloped ridge respectively. Both structures protrude above the default target plate surface thus receiving an increased fraction of the parallel power flux. Transient melting by ELMs was induced by moving the outer strike point to the sample location. The temporal evolution of the measured current flow from the samples to vessel potential confirmed transient melting. Current magnitude and dependency from surface temperature provided strong evidence for thermionic electron emission as main origin of the replacement current driving the melt motion. The different melt patterns observed after exposures at the two sample geometries support the thermionic electron emission model used in the MEMOS melt motion code, which assumes a strong decrease of the thermionic net current at shallow magnetic field to surface angles (Pitts et al 2017 Nucl. Mater. Energy 12 60-74). Post exposure ex situ analysis of the retrieved samples show recrystallization of tungsten at the exposed surface areas to a depth of up to several mm. The melt layer transport to less exposed surface areas leads to ratcheting pile up of re-solidified debris with zonal growth extending from the already enlarged grains at the surface.

H-mode and confinement studies in ASDEX Upgrade

International Nuclear Information System (INIS)

Suttrop, W.; Ryter, F.; Mertens, V.; Gruber, O.; Murmann, H.; Salzmann, H.; Schweinzer, J.

2001-01-01

H-mode operational boundaries and H-mode confinement are investigated on ASDEX Upgrade. The local edge parameter threshold for H-mode holds independent of divertor geometry and changes little with ion mass. The deviation of the H-mode power threshold at densities near the Greenwald limit can be understood as a consequence of a confinement deterioration, caused by 'stiff' temperature profiles and lack of core density gradients in gas puff fuelled discharges. Ion and electron temperature profiles can be described by a lower limit of gradient length L T =T/T'. (author)

ICRF Mode Conversion Current Drive for Plasma Stability Control in Tokamaks

International Nuclear Information System (INIS)

Grekov, D.; Kock, R.; Lyssoivan, A.; Noterdaeme, J. M.; Ongena, J.

2007-01-01

There is a substantial incentive for the International Thermonuclear Experimental Reactor (ITER) to operate at the highest attainable beta (plasma pressure normalized to magnetic pressure), a point emphasized by requirements of attractive economics in a reactor. Recent experiments aiming at stationary high beta discharges in tokamak plasmas have shown that maximum achievable beta value is often limited by the onset of instabilities at rational magnetic surfaces (neoclassical tearing modes). So, methods of effective control of these instabilities have to be developed. One possible way for neoclassical tearing modes control is an external current drive in the island to locally replace the missing bootstrap current and thus to suppress the instability. Also, a significant control of the sawtooth behaviour was demonstrated when the magnetic shear was modified by driven current at the magnetic surface where safety factor equals to 1. In the ion cyclotron range of frequencies (ICRF), the mode conversion regime can be used to drive the local external current near the position of the fast-to-slow wave conversion layer, thus providing an efficient means of plasma stability control. The slow wave energy is effectively absorbed in the vicinity of mode conversion layer by electrons with such parallel to confining magnetic field velocities that the Landau resonance condition is satisfied. For parameters of present day tokamaks and for ITER parameters the slow wave phase velocity is rather low, so the large ratio of momentum to energy content would yield high current drive efficiency. In order to achieve noticeable current drive effect, it is necessary to create asymmetry in the ICRF power absorption between top and bottom parts of the plasma minor cross-section. Such asymmetric electron heating may be realized using: - shifted from the torus midplane ICRF antenna in TEXTOR tokamak; - plasma displacement in vertical direction that is feasible in ASDEX-Upgrade; - the

A large divertor manipulator for ASDEX Upgrade

Energy Technology Data Exchange (ETDEWEB)

Herrmann, Albrecht, E-mail: albrecht.herrmann@ipp.mpg.de; Jaksic, Nikola; Leitenstern, Peter; Greuner, Henri; Krieger, Karl; Marné, Pascal de; Oberkofler, Martin; Rohde, Volker; Schall, Gerd

2015-10-15

Highlights: • A large divertor manipulator for ASDEX Upgrade is developed and tested. • It allows replacing a relevant part of the divertor by dedicated targets and probes. • Modified solid standard targets. • Electrical and mechanical probes for dedicated investigations. • Test of actively cooled component. - Abstract: In 2013 a new bulk tungsten divertor, Div-III, was installed in ASDEX Upgrade (AUG). During the concept and design phase of Div-III the option of adaptable divertor instrumentation and divertor modification as contribution for divertor investigations in preparation of ITER was given a high priority. To gain flexibility for the test of divertor modifications without affecting the operational space of AUG, the large divertor manipulator, DIM-II, was designed and installed. DIM-II allows to retract 2 out of 128 outer divertor target tiles including the water cooled support structure into a target exchange box and to replace these targets without breaking the vacuum of the AUG vessel. DIM-II is based on a carriage-rail system with a driving rod pushing a front-end with the target module into the divertor position for plasma operation. Three types of front-ends are foreseen for physics investigations: (i) modified standard targets clamped to the standard cooling structure, (ii) dedicated front-ends making use of the whole available volume of about 230 × 160 × 80 mm{sup 3} and (iii) actively cooled/heated targets for cooling water temperatures up to 230 °C. This paper presents the DIM-II design including the FEM calculations for the modified divertor support structure and the front-end options, as well as the test procedure and operation mode.

ASDEX papers at the 13th European conference on controlled fusion and plasma physics

International Nuclear Information System (INIS)

1986-05-01

This report provides 29 ASDEX papers concerning pellet refuelling, confinement, high-beta plasma and MHD-equilibrium, heating by ICR, lower hybrid and current-drive, impurity studies and plasma diagnostics. All of these papers have been indexed separately. (GG)

Heat and density pulse propagation in ASDEX

International Nuclear Information System (INIS)

Giannone, L.; Riedl, K.; Stroth, U.; Eberhagen, A.; Gruber, O.; Mertens, V.

1990-01-01

Experimental measurements of the electron thermal conductivity, derived from the radial propagation of the heat pulse generated by a sawtooth crash, have consistently yielded larger values than those obtained by power balance. It has been proposed that this discrepancy could be the result of the coupling of density and temperature perturbations. Numerical modelling of heat and density pulse propagation on ASDEX has been used to address this question. In addition, measurements at various electron densities and in hydrogen and deuterium were undertaken, with the aim of providing a broad base of experimental measurements for testing the various transport models proposed. (author) 9 refs., 1 fig

Tokamak experiments

International Nuclear Information System (INIS)

Robinson, D.C.

1987-01-01

With the advent of the new large tokamaks JET, JT-60 and TFTR important advances in magnetic confinement have been made. These include the exploitation of radio frequency and neutral beam heating on a much larger scale than previously, the demonstration of regimes of improved confinement and the demonstration of current drive at the Megamp level. A number of small and medium sized tokamaks have also come into operation recently such as WT-3 in Japan with an emphasis on radio frequency current drive and HL-1 a medium sized tokamak in China. Each of these new tokamaks is addressing specific problems which remain for the future development of the system. Of these particular problems: β, density and q limits remain important issues for the future development of the tokamak. β limits are being addressed on the DIII-D device in the USA. The anomalous confinement that the tokamak displays is being explored in detail on the TEXT device in the USA. Two other problems are impurity control and current drive. There is significant emphasis on divertor configurations at the present time with their enhanced confinement in the so called H mode. Due to improved discharge cleaning techniques and the ability to repetitively refuel using pellets, purer plasmas can be obtained even without divertors. Current drive remains a crucial issue for quasi of near steady state operation of the tokamak in the future and many current drive schemes are being investigated. (author) [pt

Observation with the low energy neutral analyser (LENA) on ASDEX. Pt. 1

International Nuclear Information System (INIS)

Verbeek, H.

1991-02-01

This report is a compilation of the observation with the Low Energy Neutral Particle Analyzers (LENA) at ASDEX during Ohmic discharges. The dependence of the energy distributions, the integrated fluxes, and their mean energies on various plasma parameters is documented. Connections and correlations with other edge and divertor diagnostics are discussed. (orig.)

PPPL tokamak program

International Nuclear Information System (INIS)

Furth, H.P.

1984-10-01

The economic prospects of the tokamak are reviewed briefly and found to be favorable - if the size of ignited tokamak plasmas can be kept small and appropriate auxiliary systems can be developed. The main objectives of the Princeton Plasma Physics Laboratory tokamak program are: (1) exploration of the physics of high-temperature toroidal confinement, in TFTR; (2) maximization of the tokamak beta value, in PBX; (3) development of reactor-relevant rf techniques, in PLT

Control of MHD instabilities by ECCD: ASDEX Upgrade results and implications for ITER

International Nuclear Information System (INIS)

Zohm, H.; Gantenbein, G.; Leuterer, F.; Manini, A.; Maraschek, M.; Yu, Q.

2007-01-01

The requirements for control of MHD instabilities by electron cyclotron current drive (ECCD) are reviewed. It is shown that a localized current drive is needed for control of both sawteeth and neoclassical tearing modes (NTMs). In the case of NTMs, the deposition width should be smaller than the island width for efficient control. At island widths smaller than the deposition width, as is predicted to occur in ITER, theory suggests that efficient control is possible only by modulating the ECCD power in phase with the island. These predictions are experimentally confirmed in ASDEX Upgrade for NTM control. Narrow deposition has also been used to extend the operational range of NTM stabilization in ASDEX Upgrade to lower q 95 and in the improved H-mode scenario. Our results suggest that, for the ITER ECCD system, good localization of the driven current profile as well as the capability to modulate the ECCD in phase with rotating modes will be needed for efficient MHD control by ECCD

Commissioning of inline ECE system within waveguide based ECRH transmission systems on ASDEX upgrade

NARCIS (Netherlands)

Bongers, W.A.; Kasparek, W.; Doelman, N. J.; Braber, R. van den; Brand, H. van den; Meo, F.; Baar, M.R. de; Amerongen, F.J.; Donné, A.J.H.; Elzendoorn, B.S.Q.; Erckmann, V.; Goede, A.P.H.; Giannone, L.; Grünwald, G.; Hollman, F.; Kaas, G.; Krijger, B.; Michel, G.; Lubyako, L.; Monaco, F.; Noke, F.; Petelin, M.; Plaum, B.; Purps, F.; Pierik, J.G.W. ten; Schüller, C.; Slob, J.W.; Stober, J.K.; Schütz, H.; Wagner, D.; Westerhof, E.; Ronden, D.M.S.

2012-01-01

A CW capable inline electron cyclotron emission (ECE) separation system for feedback control, featuring oversized corrugated waveguides, is commissioned on ASDEX upgrade (AUG). The system is based on a combination of a polarization independent, non-resonant, Mach-Zehnder diplexer equipped with

Enhancement of the FIDA diagnostic at ASDEX Upgrade for velocity space tomography

DEFF Research Database (Denmark)

Weiland, M.; Geiger, B.; Jacobsen, Asger Schou

2016-01-01

Recent upgrades to the FIDA (fast-ion D-alpha) diagnostic at ASDEX Upgrade are discussed. The diagnostic has been extended from three to five line of sight arrays with different angles to the magnetic field, and a spectrometer redesign allows the simultaneous measurement of red- and blue-shifted ......Recent upgrades to the FIDA (fast-ion D-alpha) diagnostic at ASDEX Upgrade are discussed. The diagnostic has been extended from three to five line of sight arrays with different angles to the magnetic field, and a spectrometer redesign allows the simultaneous measurement of red- and blue......-shifted parts of the Doppler spectrum. These improvements make it possible to reconstruct the 2D fast-ion velocity distribution from the FIDA measurements by tomographic inversion under a wide range of plasma parameters. Two applications of the tomography are presented: a comparison between the distributions...... resulting from 60 keV and 93 keV neutral beam injection and a velocity-space resolved study of fast-ion redistribution induced by a sawtooth crash inside and outside the sawtooth inversion radius....

The ASDEX Upgrade discharge schedule

International Nuclear Information System (INIS)

Neu, G.; Engelhardt, K.; Raupp, G.; Treutterer, W.; Zasche, D.; Zehetbauer, T.

2007-01-01

ASDEX Upgrade's recently commissioned discharge control system (DCS) marks the transition from a traditional programmed system to a highly flexible 'data driven' one. The allocation of application processes (APs) to controllers, the interconnection of APs through uniquely named signals, and AP control parameter values are all defined as data, and can easily be adapted to the requirements of a particular discharge. The data is laid down in a set of XML documents which APs request via HTTP from a configuration server before a discharge. The use of XML allows for easy parsing, and structural validation through (XSD) schemas. The central input to the configuration process is the discharge schedule (DS), which embodies the dynamic behaviour of a planned discharge as reference trajectories grouped in segments, concatenated through transition conditions. Editing, generation and validation tools, and version control through CVS allow for efficient management of DSs

Tokamak engineering mechanics

International Nuclear Information System (INIS)

Song, Yuntao; Wu, Weiyue; Du, Shijun

2014-01-01

Provides a systematic introduction to tokamaks in engineering mechanics. Includes design guides based on full mechanical analysis, which makes it possible to accurately predict load capacity and temperature increases. Presents comprehensive information on important design factors involving materials. Covers the latest advances in and up-to-date references on tokamak devices. Numerous examples reinforce the understanding of concepts and provide procedures for design. Tokamak Engineering Mechanics offers concise and thorough coverage of engineering mechanics theory and application for tokamaks, and the material is reinforced by numerous examples. Chapter topics include general principles, static mechanics, dynamic mechanics, thermal fluid mechanics and multiphysics structural mechanics of tokamak structure analysis. The theoretical principle of the design and the methods of the analysis for various components and load conditions are presented, while the latest engineering technologies are also introduced. The book will provide readers involved in the study of mechanical/fusion engineering with a general understanding of tokamak engineering mechanics.

Commissioning of inline ECE system within waveguide based ECRH transmission systems on ASDEX upgrade

DEFF Research Database (Denmark)

Bongers, W. A.; Kasparek, W.; Doelman, N.

2012-01-01

A CW capable inline electron cyclotron emission (ECE) separation system for feedback control, featuring oversized corrugated waveguides, is commissioned on ASDEX upgrade (AUG). The system is based on a combination of a polarization independent, non-resonant, Mach-Zehnder diplexer equipped with di...

Comparison of fast ion collective Thomson scattering measurements at ASDEX Upgrade with numerical simulations

DEFF Research Database (Denmark)

Salewski, Mirko; Meo, Fernando; Stejner Pedersen, Morten

2010-01-01

Collective Thomson scattering (CTS) experiments were carried out at ASDEX Upgrade to measure the one-dimensional velocity distribution functions of fast ion populations. These measurements are compared with simulations using the codes TRANSP/NUBEAM and ASCOT for two different neutral beam injecti...

ASDEX Upgrade Discharge Control System—A real-time plasma control framework

International Nuclear Information System (INIS)

Treutterer, W.; Cole, R.; Lüddecke, K.; Neu, G.; Rapson, C.; Raupp, G.; Zasche, D.; Zehetbauer, T.

2014-01-01

Highlights: • The ASDEX Upgrade Discharge Control System (DCS) is a comprehensive control system to conduct fusion experiments. • DCS supports real-time diagnostic integration, adaptable feedback schemes, actuator management and exception handling. • DCS offers workflow management, logging and archiving, self-monitoring and inter-process communication. • DCS is based on a distributed, modular software framework architecture designed for real-time operation. • DCS is composed of re-usable generic but highly customisable components. - Abstract: ASDEX Upgrade is a fusion experiment with a size and complexity to allow extrapolation of technical and physical conditions and requirements to devices like ITER and even beyond. In addressing advanced physics topics it makes extensive use of sophisticated real-time control methods. It comprises real-time diagnostic integration, dynamically adaptable multivariable feedback schemes, actuator management including load distribution schemes and a powerful monitoring and pulse supervision concept based on segment scheduling and exception handling. The Discharge Control System (DCS) supplies all this functionality on base of a modular software framework architecture designed for real-time operation. It provides system-wide services like workflow management, logging and archiving, self-monitoring and inter-process communication on Linux, VxWorks and Solaris operating systems. By default DCS supports distributed computing, and a communication layer allows multi-directional signal transfer and data-driven process synchronisation over shared memory as well as over a number of real-time networks. The entire system is built following the same common design concept combining a rich set of re-usable generic but highly customisable components with a configuration-driven component deployment method. We will give an overview on the architectural concepts as well as on the outstanding capabilities of DCS in the domains of inter

ASDEX Upgrade Discharge Control System—A real-time plasma control framework

Energy Technology Data Exchange (ETDEWEB)

Treutterer, W., E-mail: Wolfgang.Treutterer@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, EURATOM Association, Boltzmannstraße 2, 85748 Garching (Germany); Cole, R.; Lüddecke, K. [Unlimited Computer Systems GmbH, Iffeldorf (Germany); Neu, G.; Rapson, C.; Raupp, G.; Zasche, D.; Zehetbauer, T. [Max-Planck-Institut für Plasmaphysik, EURATOM Association, Boltzmannstraße 2, 85748 Garching (Germany)

2014-03-15

Highlights: • The ASDEX Upgrade Discharge Control System (DCS) is a comprehensive control system to conduct fusion experiments. • DCS supports real-time diagnostic integration, adaptable feedback schemes, actuator management and exception handling. • DCS offers workflow management, logging and archiving, self-monitoring and inter-process communication. • DCS is based on a distributed, modular software framework architecture designed for real-time operation. • DCS is composed of re-usable generic but highly customisable components. - Abstract: ASDEX Upgrade is a fusion experiment with a size and complexity to allow extrapolation of technical and physical conditions and requirements to devices like ITER and even beyond. In addressing advanced physics topics it makes extensive use of sophisticated real-time control methods. It comprises real-time diagnostic integration, dynamically adaptable multivariable feedback schemes, actuator management including load distribution schemes and a powerful monitoring and pulse supervision concept based on segment scheduling and exception handling. The Discharge Control System (DCS) supplies all this functionality on base of a modular software framework architecture designed for real-time operation. It provides system-wide services like workflow management, logging and archiving, self-monitoring and inter-process communication on Linux, VxWorks and Solaris operating systems. By default DCS supports distributed computing, and a communication layer allows multi-directional signal transfer and data-driven process synchronisation over shared memory as well as over a number of real-time networks. The entire system is built following the same common design concept combining a rich set of re-usable generic but highly customisable components with a configuration-driven component deployment method. We will give an overview on the architectural concepts as well as on the outstanding capabilities of DCS in the domains of inter

Tokamak concept innovations

International Nuclear Information System (INIS)

1986-04-01

This document contains the results of the IAEA Specialists' Meeting on Tokamak Concept Innovations held 13-17 January 1986 in Vienna. Although it is the most advanced fusion reactor concept the tokamak is not without its problems. Most of these problems should be solved within the ongoing R and D studies for the next generation of tokamaks. Emphasis for this meeting was placed on innovations that would lead to substantial improvements in a tokamak reactor, even if they involved a radical departure from present thinking

Experimental investigations of turbulent temperature fluctuations and phase angles in ASDEX Upgrade

Science.gov (United States)

Freethy, Simon

2017-10-01

A complete experimental understanding of the turbulent fluctuations in tokamak plasmas is essential for providing confidence in the extrapolation of heat transport models to future experimental devices and reactors. Guided by ``predict first'' nonlinear gyrokinetic simulations with the GENE code, two new turbulence diagnostics were designed and have been installed on ASDEX Upgrade (AUG) to probe the fundamentals of ion-scale turbulent electron heat transport. The first, a 30-channel correlation ECE (CECE) radiometer, measures radial profiles (0.5 a levels are in the range 0.3 - 0.8%. The second is formed by the addition of a reflectometer on the same line of sight to enable measurements of the phase angle between turbulent density and temperature fluctuations. Design predictions are followed by a more traditional ``post-diction'' validation study with GENE. Using a cutting edge synthetic diagnostic GENE shows a factor 1.6 - 2 over-prediction of the fluctuation amplitude, while matching both ion and electron heat fluxes within experimental error. Detailed sensitivity scans are underway to understand the robustness of this disagreement and a detailed assessment of the experimental errors has been carried out. The discrepancy opens questions about the role of multi-scale turbulence physics, but also indicates the need for the comparison of more experimental turbulence properties to have a more complete validation hierarchy. In an effort to understand the discrepancy, predictions of the nT-phase and the radial correlation length have been made along with an assessment of their sensitivity to experimental errors. Comparison to experimental measurements will be discussed. This work is supported in part by the US DOE under Grants DE-SC0006419 and DE-SC0017381. This work has also received funding from the European Union's Horizon 2020 research and innovation programme under Grant agreement number 633053.

Quantitative AMS depth profiling of the hydrogen isotopes collected in graphite divertor and wall tiles of the tokamak ASDEX-Upgrade

International Nuclear Information System (INIS)

Sun, G.Y.; Friedrich, M.; Groetzschel, R.; Buerger, W.; Behrisch, R.; Garcia-Rosales, C.

1997-01-01

The accelerator mass spectrometry (AMS) facility at the 3 MV Tandetron in Rossendorf has been applied for quantitative depth profiling of deuterium and tritium in samples cut from graphite protection tiles at the vessel walls of the fusion experiment ASDEX-Upgrade at the Max-Planck-Institut fuer Plasmaphysik in Garching. The tritium originates from D(d,p)T fusion reactions in the plasma and it is implanted in the vessel walls together with deuterium atoms and ions from the plasma. The T concentrations in the surface layers down to the analyzing depth of about 25 μm are in the range of 10 11 to 5 x 10 15 T-atoms/cm 3 corresponding to a tritium retention of 3 x 10 10 to 3.5 x 10 12 T-atoms/cm 2 . The much higher deuterium concentrations in the samples were simultaneously measured by calibrated conventional SIMS. In the surface layers down to the analyzing depth of about 25 μm the deuterium concentrations are between 3 x 10 18 and 8 x 10 21 atoms/cm 3 , corresponding to a deuterium retention of 2.5 x 10 16 to 2.5 x 10 18 atoms/cm 2 The estimated total amount of tritium in the vessel walls is of the same order of magnitude as the total number of neutrons produced in D(d,n) 3 He reactions. (orig.)

Real-time disruption handling at ASDEX upgrade

International Nuclear Information System (INIS)

Zehetbauer, Th.; Pautasso, G.; Tichmann, C.; Egorov, S.; Lorenz, A.; Mertens, V.; Neu, G.; Raupp, G.; Treutterer, W.; Zasche, D.

2001-01-01

A neural network for prediction of disruptions has been developed at ASDEX Upgrade with the goal to mitigate or avoid these. The novel idea is to compute the remaining time-to-disruption to indicate the stability level of the discharge. The neural network has been specified, trained and then implemented within the real-time plasma control system. The current version of the system terminates the discharge with an impurity pellet when the computed time-to-disruption falls below a threshold of 80 ms. Routine operation shows that disruptions are recognized reliably. Vessel currents and forces are considerably reduced. The system will be enhanced to avoid disruptions with a soft landing initiated in time

Joint research using small tokamaks

International Nuclear Information System (INIS)

Gryaznevich, M.P.; Bosco, E. Del; Malaquias, A.; Mank, G.; Oost, G. van; He, Yexi; Hegazy, H.; Hirose, A.; Hron, M.; Kuteev, B.; Ludwig, G.O.; Nascimento, I.C.; Silva, C.; Vorobyev, G.M.

2005-01-01

Small tokamaks have an important role in fusion research. More than 40 small tokamaks are operational. Research on small tokamaks has created a scientific basis for the scaling-up to larger tokamaks. Well-known scientific and engineering schools, which are now determining the main directions of fusion science and technology, have been established through research on small tokamaks. Combined efforts within a network of small and medium size tokamaks will further enhance the contribution of small tokamaks. A new concept of interactive coordinated research using small tokamaks in the mainstream fusion science areas, in testing of new diagnostics, materials and technologies as well as in education, training and broadening of the geography of fusion research in the scope of the IAEA Coordinated Research Project, is presented

Management of complex data flows in the ASDEX Upgrade plasma control system

Energy Technology Data Exchange (ETDEWEB)

Treutterer, Wolfgang, E-mail: Wolfgang.Treutterer@ipp.mpg.de [Max-Planck Institut fuer Plasmaphysik, EURATOM Association, Garching (Germany); Neu, Gregor; Raupp, Gerhard; Zasche, Dieter; Zehetbauer, Thomas [Max-Planck Institut fuer Plasmaphysik, EURATOM Association, Garching (Germany); Cole, Richard; Lueddecke, Klaus [Unlimited Computer Systems, Iffeldorf (Germany)

2012-12-15

Highlights: Black-Right-Pointing-Pointer Control system architectures with data-driven workflows are efficient, flexible and maintainable. Black-Right-Pointing-Pointer Signal groups provide coherence of interrelated signals and increase the efficiency of process synchronisation. Black-Right-Pointing-Pointer Sample tags indicating sample quality form the fundament of a local event handling strategy. Black-Right-Pointing-Pointer A self-organising workflow benefits from sample tags consisting of time stamp and stream activity. - Abstract: Establishing adequate technical and physical boundary conditions for a sustained nuclear fusion reaction is a challenging task. Phased feedback control and monitoring for heating, fuelling and magnetic shaping is mandatory, especially for fusion devices aiming at high performance plasmas. Technical and physical interrelations require close collaboration of many components in sequential as well as in parallel processing flows. Moreover, handling of asynchronous, off-normal events has become a key element of modern plasma performance optimisation and machine protection recipes. The manifoldness of plasma states and events, the variety of plant system operation states and the diversity in diagnostic data sampling rates can hardly be mastered with a rigid control scheme. Rather, an adaptive system topology in combination with sophisticated synchronisation and process scheduling mechanisms is suited for such an environment. Moreover, the system is subject to real-time control constraints: response times must be deterministic and adequately short. Therefore, the experimental tokamak device ASDEX Upgrade employs a discharge control system DCS, whose core has been designed to meet these requirements. In the paper we will compare the scheduling schemes for the parallelised realisation of a control workflow and show the advantage of a data-driven workflow over a managed workflow. The data-driven workflow as used in DCS is based on signals

The ASDEX 100 keV neutral lithium beam diagnostic gun

International Nuclear Information System (INIS)

McCormick, K.; Kick, M.

1983-04-01

The neutral lithium beam gun intended for measurement of the poloidal magnetic field and of the density gradient in the scrape-off layer of ASDEX is described, and test results over a beam energy range of 27-100 keV are presented. In the gun, lithium ions are extracted from a solid emitter (#betta#-Eurcryptite) in a Pierce-type configuration, accelerated and focused in a two-tube immersion lens, and neutralized in a charge-exchange cell using sodium. The beam can be pulsed from less than one to several seconds, depending on experimental needs. At a distance of 165 cm from the gun the neutral beam equivalent current is typically greater than 1 mA (0.16 mA) for a beam energy of 100 keV (27 keV), the beam FWHM being about 8-9 mm. It is found that to produce a particular beam with a certain ratio must be maintained between the extraction and total beam voltages, this relationship depending in turn on the emitter-extractor separation. The principal features which distinguish the ASDEX gun from that employed on W7a are the greater compactness - all the active elements, i.e. emitter, extractor, lens, deflection plates and neutralizer, are contained with 57 cm - and the vacuum vessel, which simultaneously serves as the magnetic shielding. (orig.)

Turbulence in the SOL of ASDEX and W7-AS

Energy Technology Data Exchange (ETDEWEB)

Endler, M [Association Euratom-Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Giannone, L. [Association Euratom-Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); McCormick, K [Association Euratom-Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Niedermeyer, H [Association Euratom-Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Rudyj, A [Association Euratom-Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Theimer, G [Association Euratom-Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Tsois, N [Association Euratom-Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Zoletnik, S [Hungarian Academy of Sciences, Budapest (Hungary). Central Research Inst. for Physics; ASDEX Team; W7-AS Team

1995-05-01

Electrostatic fluctuations have been measured in the scrape-off layer of ASDEX and W7-AS by Langmuir probes and by observation of H{sub {alpha}} light and light emitted from a fast Li atom beam with high spatial and temporal resolution. It was demonstrated that these fluctuations contribute a significant, if not dominant, fraction of the anomalous radial particle transport. The basic properties of the fluctuations are the same in both experiments. A model for an instability mechanism specific to the SOL is presented including density, temperature and electric potential fluctuations. From this model mixing length estimates for the radial transport and resulting density and pressure gradients in the SOL are derived and compared to measurements in the mid-plane and in the divertor of ASDEX. In spite of several simplifications in the model a quantitative agreement up to factors of 1-3 and a qualitative agreement for variations of discharge parameters is achieved between the model predictions and the measurements. Data from poloidal multi-pin probe arrays are decomposed into a sum of spatial-temporal ``events`` by means of a fitting procedure. Centres of selected events serve as reference points for the analysis of the dynamics in their surroundings. Averaging shows that positive and negative potential events appear mostly in pairs with the E x B drift in between directed radially outward. (orig.).

Joint research using small tokamaks

International Nuclear Information System (INIS)

Gryaznevich, M.P.; Del Bosco, E.; Malaquias, A.; Mank, G.; Oost, G. van

2005-01-01

Small tokamaks have an important role in fusion research. More than 40 small tokamaks are operational. Research on small tokamaks has created a scientific basis for the scaling-up to larger tokamaks. Well-known scientific and engineering schools, which are now determining the main directions of fusion science and technology, have been established through research on small tokamaks. Combined efforts within a network of small and medium size tokamaks will further enhance the contribution of small tokamaks. A new concept of interactive co-ordinated research using small tokamaks in the mainstream fusion science areas, in testing of new diagnostics, materials and technologies as well as in education, training and broadening of the geography of fusion research in the scope of the IAEA Co-ordinated Research Project is presented. (author)

L-mode radiative plasma edge studies for model validation in ASDEX Upgrade and JET

Energy Technology Data Exchange (ETDEWEB)

Aho-Mantila, L., E-mail: leena.aho-mantila@vtt.fi [VTT Technical Research Centre of Finland, FI-02044 VTT (Finland); Bernert, M. [Max-Planck Institut für Plasmaphysik, D-85748 Garching (Germany); Coenen, J.W. [Energie- und Klimaforschung IEK-4, FZJ, EURATOM Association, TEC, 52425 Jülich (Germany); Fischer, R. [Max-Planck Institut für Plasmaphysik, D-85748 Garching (Germany); Lehnen, M. [Energie- und Klimaforschung IEK-4, FZJ, EURATOM Association, TEC, 52425 Jülich (Germany); Lowry, C. [EFDA JET CSU, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Marsen, S. [Max-Planck-Institut für Plasmaphysik, Teilinsitut Greifswald, D-17491 Greifswald (Germany); McCormick, K.; Müller, H.W.; Sieglin, B. [Max-Planck Institut für Plasmaphysik, D-85748 Garching (Germany); Stamp, M.F. [Culham Centre for Fusion Energy, EURATOM-CCFE Association, Abingdon (United Kingdom); Wischmeier, M. [Max-Planck Institut für Plasmaphysik, D-85748 Garching (Germany); Bonnin, X. [LSPM, CNRS, Université Paris 13, F-93430 Villetaneuse (France); Coster, D.P. [Max-Planck Institut für Plasmaphysik, D-85748 Garching (Germany); Reiter, D.; Brezinsek, S. [Energie- und Klimaforschung IEK-4, FZJ, EURATOM Association, TEC, 52425 Jülich (Germany)

2013-07-15

The presently favoured option for reactor power handling combines metallic plasma-facing components and impurity seeding to achieve highly radiative scrape-off layer and divertor plasmas. It is uncertain whether tolerable divertor power loads will be obtained in this scenario, necessitating the development of predictive modelling tools. L-mode experiments with N{sub 2} seeding have been conducted at both ASDEX Upgrade and JET for benchmarking the critically important impurity radiation models in edge fluid codes. In both machines, particle and power loads are observed to first reduce at the inner target, and only then at the outer target. The outer divertor cools down with increasing N seeding rate, evolving from low-recycling conditions to a regime with peak temperature of 8–10 eV in both devices. First SOLPS5.0 simulations of N{sub 2} seeding in ASDEX Upgrade geometry show a similar in–out asymmetry in the effect of impurity radiation when drifts are activated in the simulations.

Tokamaks. 2. ed.

International Nuclear Information System (INIS)

Wesson, John; Campbell, D.J.; Connor, J.W.

1997-01-01

It is interesting to recall the state of tokamak research when the first edition of this book was written. My judgement of the level of real understanding at that time is indicated by the virtual absence of comparisons of experiment with theory in that edition. The need then was for a 'handbook' which collected in a single volume the concepts and models which form the basis of everyday tokamak research. The experimental and theoretical endeavours of the subsequent decade have left almost all of this intact, but have brought a massive development of the subject. Firstly, there are now several areas where the experimental behaviour is described in terms of accepted theory. This is particularly true of currents parallel to the magnetic field, and of the stability limitations on the plasma pressure. Next there has been the research on large tokamaks, hardly started at the writing of the first edition. Now our thinking is largely based on the results from these tokamaks and this work has led to the long awaited achievement of significant amounts of fusion power. Finally, the success of tokamak research has brought us face to face with the problems involved in designing and building a tokamak reactor. The present edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes an account of the advances outlined above. (Author)

Tokamak devices: towards controlled fusion

International Nuclear Information System (INIS)

Trocheris, M.

1975-01-01

The Tokamak family is from Soviet Union. These devices were exclusively studied at the Kurchatov Institute in Moscow for more than ten years. The first occidental Tokamak started in 1970 at Princeton. The TFR (Tokamak Fontenay-aux-Roses) was built to be superior to the Russian T4. Tokamak future is now represented by the JET (Joint European Tokamak) [fr

Imaging motional Stark effect measurements at ASDEX Upgrade

Energy Technology Data Exchange (ETDEWEB)

Ford, O. P.; Burckhart, A.; McDermott, R.; Pütterich, T.; Wolf, R. C. [Max-Planck Institut für Plasmaphysik, Greifswald/Garching (Germany)

2016-11-15

This paper presents an overview of results from the Imaging Motional Stark Effect (IMSE) diagnostic obtained during its first measurement campaign at ASDEX Upgrade since installation as a permanent diagnostic. A brief overview of the IMSE technique is given, followed by measurements of a standard H-mode discharge, which are compared to equilibrium reconstructions showing good agreement where expected. The development of special discharges for the calibration of pitch angle is reported and safety factor profile changes during sawteeth crashes are shown, which can be resolved to a few percent due to the high sensitivity at good time resolution of the new IMSE system.

A solid tungsten divertor for ASDEX Upgrade

International Nuclear Information System (INIS)

Herrmann, A; Greuner, H; Jaksic, N; Böswirth, B; Maier, H; Neu, R; Vorbrugg, S

2011-01-01

The conceptual design of a solid tungsten divertor for ASDEX Upgrade (AUG) is presented. The Div-III design is compatible with the existing divertor structure. It re-establishes the energy and heat receiving capability of a graphite divertor and overcomes the limitations of tungsten coatings. In addition, a solid tungsten divertor allows us to investigate erosion and bulk deuterium retention as well as test castellation and target tilting. The design criteria as well as calculations of forces due to halo and eddy currents are presented. The thermal properties of the proposed sandwich structure are calculated with finite element method models. After extensive testing of a target tile in the high heat flux test facility GLADIS, two solid tungsten tiles were installed in AUG for in-situ testing.

The ASDEX integrated data analysis system AIDA

International Nuclear Information System (INIS)

Grassie, K.; Gruber, O.; Kardaun, O.; Kaufmann, M.; Lackner, K.; Martin, P.; Mast, K.F.; McCarthy, P.J.; Mertens, V.; Pohl, D.; Rang, U.; Wunderlich, R.

1989-11-01

Since about two years, the ASDEX integrated data analysis system (AIDA), which combines the database (DABA) and the statistical analysis system (SAS), is successfully in operation. Besides a considerable, but meaningful, reduction of the 'raw' shot data, it offers the advantage of carefully selected and precisely defined datasets, which are easily accessible for informative tabular data overviews (DABA), and multi-shot analysis (SAS). Even rather complicated, statistical analyses can be performed efficiently within this system. In this report, we want to summarise AIDA's main features, give some details on its set-up and on the physical models which have been used for the derivation of the processed data. We also give short introduction how to use DABA and SAS. (orig.)

Real-time control of the plasma density profile on ASDEX upgrade

International Nuclear Information System (INIS)

Mlynek, Alexander

2010-01-01

The tokamak concept currently is the most promising approach to future power generation by controlled thermonuclear fusion. The spatial distribution of the particle density in the toroidally confined fusion plasma is of particular importance. This thesis work therefore focuses on the question as to what extent the shape of the density profile can be actively controlled by a feedback loop in the fusion experiment ASDEX Upgrade. There are basically two essential requirements for such feedback control of the density profile, which has been experimentally demonstrated within the scope of this thesis work: On the one hand, for this purpose the density profile must be continuously calculated under real-time constraints during a plasma discharge. The calculation of the density profile is based on the measurements of a sub-millimeter interferometer, which provides the line-integrated electron density along 5 chords through the plasma. Interferometric density measurements can suffer from counting errors by integer multiples of 2π when detecting the phase difference between a probing and a reference beam. As such measurement errors have severe impact on the reconstructed density profile, one major part of this work consists in the development of new readout electronics for the interferometer, which allows for detection of such measurement errors in real-time with high reliability. A further part of this work is the design of a computer algorithm which reconstructs the spatial distribution of the plasma density from the line-integrated measurements. This algorithm has to be implemented on a computer which communicates the measured data to other computers in real-time, especially to the tokamak control system. On the other hand, a second fundamental requirement for the successful implementation of a feedback controller is the identification of at least one actuator which enables a modification of the density profile. Here, electron cyclotron resonance heating (ECRH) has been

Commissioning activities and first results from the collective Thomson scattering diagnostic on ASDEX Upgrade (invited)

DEFF Research Database (Denmark)

Meo, Fernando; Bindslev, Henrik; Korsholm, Søren Bang

2008-01-01

of the system. First results in near perpendicular of scattered spectra in a neutral beam injection (NBI) and ion cyclotron resonance heating (ICRH) plasma (minority hydrogen) on ASDEX Upgrade have shown evidence of ICRH heating phase of hydrogen. ©2008 American Institute of Physics...

The spatial structure of type-I ELMs at the mid-plane in ASDEX Upgrade and a comparison with data from MAST

International Nuclear Information System (INIS)

Kirk, A; Eich, T; Herrmann, A; Muller, H W; Horton, L D; Counsell, G F; Price, M; Rohde, V; Bobkov, V; Kurzan, B; Neuhauser, J; Wilson, H

2005-01-01

The radial extent and spatial structure of type-I edge localized modes (ELMs) in ASDEX Upgrade are investigated using data from a mid-plane manipulator equipped with Langmuir probes and a fast visible imaging camera and are compared to data from MAST. Plasmas with a range of toroidal magnetic fields have been studied. The radial extent of the ELM efflux is found to be largest at the smaller toroidal magnetic field. A study of a series of shots on ASDEX Upgrade with different plasma edge to wall separation suggests that the closeness of the wall does not have a stabilizing effect on the radial extent of the ELM. The data from the mid-plane manipulator and from visible imaging are consistent with non-linear ballooning mode theory, which predicts that the ELM has a filament like structure. On both devices these structures have a poloidal extent of 5-10 cm and a typical toroidal mode number of ∼15 and are found to accelerate away from the plasma edge. The acceleration is ∼3 times larger on MAST than on ASDEX Upgrade

Improvements for real-time magnetic equilibrium reconstruction on ASDEX Upgrade

International Nuclear Information System (INIS)

Giannone, L.; Fischer, R.; McCarthy, P.J.; Odstrcil, T.; Zammuto, I.; Bock, A.; Conway, G.; Fuchs, J.C.; Gude, A.; Igochine, V.; Kallenbach, A.; Lackner, K.; Maraschek, M.; Rapson, C.; Ruan, Q.; Schuhbeck, K.H.; Suttrop, W.; Wenzel, L.

2015-01-01

Highlights: • Spline basis current functions with second-order linear regularisation. • Perturbations of magnetic probe measurements due to ferromagnetic tiles on the inner wall and from oscillations in the fast position coil current are corrected. • A constraint of the safety factor on the magnetic axis is introduced. Soft X-ray tomography is used to assess the quality of the real-time magnetic equilibrium reconstruction. • External loop voltage measurements and magnetic probe pairs inside and outside the vessel wall were used to measure the vacuum vessel wall resistivity. - Abstract: Real-time magnetic equilibria are needed for NTM stabilization and disruption avoidance experiments on ASDEX Upgrade. Five improvements to real-time magnetic equilibrium reconstruction on ASDEX Upgrade have been investigated. The aim is to include as many features of the offline magnetic equilibrium reconstruction code in the real-time equilibrium reconstruction code. Firstly, spline current density basis functions with regularization are used in the offline equilibrium reconstruction code, CLISTE [1]. It is now possible to have the same number of spline basis functions in the real-time code. Secondly, in the presence of edge localized modes, (ELM's), it is found to be necessary to include the low pass filter effect of the vacuum vessel on the fast position control coil currents to correctly compensate the magnetic probes for current oscillations in these coils. Thirdly, the introduction of ferromagnetic tiles in ASDEX Upgrade means that a real-time algorithm for including the perturbations of the magnetic equilibrium generated by these tiles is required. A methodology based on tile surface currents is described. Fourthly, during current ramps it was seen that the difference between fitted and measured magnetic measurements in the equilibrium reconstruction were larger than in the constant current phase. External loop voltage measurements and magnetic probe pairs inside and

Improvements for real-time magnetic equilibrium reconstruction on ASDEX Upgrade

Energy Technology Data Exchange (ETDEWEB)

Giannone, L.; Fischer, R. [Max Planck Institute for Plasma Physics, 85748 Garching (Germany); McCarthy, P.J. [Department of Physics, University College Cork, Cork (Ireland); Odstrcil, T.; Zammuto, I.; Bock, A.; Conway, G.; Fuchs, J.C.; Gude, A.; Igochine, V.; Kallenbach, A.; Lackner, K.; Maraschek, M.; Rapson, C. [Max Planck Institute for Plasma Physics, 85748 Garching (Germany); Ruan, Q. [National Instruments, Austin, TX 78759-3504 (United States); Schuhbeck, K.H.; Suttrop, W. [Max Planck Institute for Plasma Physics, 85748 Garching (Germany); Wenzel, L. [National Instruments, Austin, TX 78759-3504 (United States)

2015-11-15

Highlights: • Spline basis current functions with second-order linear regularisation. • Perturbations of magnetic probe measurements due to ferromagnetic tiles on the inner wall and from oscillations in the fast position coil current are corrected. • A constraint of the safety factor on the magnetic axis is introduced. Soft X-ray tomography is used to assess the quality of the real-time magnetic equilibrium reconstruction. • External loop voltage measurements and magnetic probe pairs inside and outside the vessel wall were used to measure the vacuum vessel wall resistivity. - Abstract: Real-time magnetic equilibria are needed for NTM stabilization and disruption avoidance experiments on ASDEX Upgrade. Five improvements to real-time magnetic equilibrium reconstruction on ASDEX Upgrade have been investigated. The aim is to include as many features of the offline magnetic equilibrium reconstruction code in the real-time equilibrium reconstruction code. Firstly, spline current density basis functions with regularization are used in the offline equilibrium reconstruction code, CLISTE [1]. It is now possible to have the same number of spline basis functions in the real-time code. Secondly, in the presence of edge localized modes, (ELM's), it is found to be necessary to include the low pass filter effect of the vacuum vessel on the fast position control coil currents to correctly compensate the magnetic probes for current oscillations in these coils. Thirdly, the introduction of ferromagnetic tiles in ASDEX Upgrade means that a real-time algorithm for including the perturbations of the magnetic equilibrium generated by these tiles is required. A methodology based on tile surface currents is described. Fourthly, during current ramps it was seen that the difference between fitted and measured magnetic measurements in the equilibrium reconstruction were larger than in the constant current phase. External loop voltage measurements and magnetic probe pairs inside

Path-oriented early reaction to approaching disruptions in ASDEX Upgrade and TCV in view of the future needs for ITER and DEMO

Science.gov (United States)

Maraschek, M.; Gude, A.; Igochine, V.; Zohm, H.; Alessi, E.; Bernert, M.; Cianfarani, C.; Coda, S.; Duval, B.; Esposito, B.; Fietz, S.; Fontana, M.; Galperti, C.; Giannone, L.; Goodman, T.; Granucci, G.; Marelli, L.; Novak, S.; Paccagnella, R.; Pautasso, G.; Piovesan, P.; Porte, L.; Potzel, S.; Rapson, C.; Reich, M.; Sauter, O.; Sheikh, U.; Sozzi, C.; Spizzo, G.; Stober, J.; Treutterer, W.; ZancaP; ASDEX Upgrade Team; TCV Team; the EUROfusion MST1 Team

2018-01-01

Routine reaction to approaching disruptions in tokamaks is currently largely limited to machine protection by mitigating an ongoing disruption, which remains a basic requirement for ITER and DEMO [1]. Nevertheless, a mitigated disruption still generates stress to the device. Additionally, in future fusion devices, high-performance discharge time itself will be very valuable. Instead of reacting only on generic features, occurring shortly before the disruption, the ultimate goal is to actively avoid approaching disruptions at an early stage, sustain the discharges whenever possible and restrict mitigated disruptions to major failures. Knowledge of the most relevant root causes and the corresponding chain of events leading to disruption, the disruption path, is a prerequisite. For each disruption path, physics-based sensors and adequate actuators must be defined and their limitations considered. Early reaction facilitates the efficiency of the actuators and enhances the probability of a full recovery. Thus, sensors that detect potential disruptions in time are to be identified. Once the entrance into a disruption path is detected, we propose a hierarchy of actions consisting of (I) recovery of the discharge to full performance or at least continuation with a less disruption-prone backup scenario, (II) complete avoidance of disruption to sustain the discharge or at least delay it for a controlled termination and, (III), only as last resort, a disruption mitigation. Based on the understanding of disruption paths, a hierarchical and path-specific handling strategy must be developed. Such schemes, testable in present devices, could serve as guidelines for ITER and DEMO operation. For some disruption paths, experiments have been performed at ASDEX Upgrade and TCV. Disruptions were provoked in TCV by impurity injection into ELMy H-mode discharges and in ASDEX Upgrade by forcing a density limit in H-mode discharges. The new approach proposed in this paper is discussed for

Status of the tokamak program

Science.gov (United States)

Sheffield, J.

1981-08-01

For a specific configuration of magnetic field and plasma to be economically attractive as a commercial source of energy, it must contain a high-pressure plasma in a stable fashion while thermally isolating the plasma from the walls of the containment vessel. The tokamak magnetic configuration is presently the most successful in terms of reaching the considered goals. Tokamaks were developed in the USSR in a program initiated in the mid-1950s. By the early 1970s tokamaks were operating not only in the USSR but also in the U.S., Australia, Europe, and Japan. The advanced state of the tokamak program is indicated by the fact that it is used as a testbed for generic fusion development - for auxiliary heating, diagnostics, materials - as well as for specific tokamak advancement. This has occurred because it is the most economic source of a large, reproducible, hot, dense plasma. The basic tokamak is considered along with tokamak improvements, impurity control, additional heating, particle and power balance in a tokamak, aspects of microscopic transport, and macroscopic stability.

The halo current in ASDEX Upgrade

International Nuclear Information System (INIS)

Pautasso, G.; Giannone, L.; Gruber, O.; Herrmann, A.; Maraschek, M.; Schuhbeck, K.H.

2011-01-01

Due to the complexity of the phenomena involved, a self-consistent physical model for the prediction of the halo current is not available. Therefore the ITER specifications of the spatial distribution and evolution of the halo current rely on empirical assumptions. This paper presents the results of an extensive analysis of the halo current measured in ASDEX Upgrade with particular emphasis on the evolution of the halo region, on the magnitude and time history of the halo current, and on the structure and duration of its toroidal and poloidal asymmetries. The effective length of the poloidal path of the halo current in the vessel is found to be rather insensitive to plasma parameters. Large values of the toroidally averaged halo current are observed in both vertical displacement events and centred disruptions but last a small fraction of the current quench; they coincide typically with a large but short-lived MHD event.

The halo current in ASDEX Upgrade

Science.gov (United States)

Pautasso, G.; Giannone, L.; Gruber, O.; Herrmann, A.; Maraschek, M.; Schuhbeck, K. H.; ASDEX Upgrade Team

2011-04-01

Due to the complexity of the phenomena involved, a self-consistent physical model for the prediction of the halo current is not available. Therefore the ITER specifications of the spatial distribution and evolution of the halo current rely on empirical assumptions. This paper presents the results of an extensive analysis of the halo current measured in ASDEX Upgrade with particular emphasis on the evolution of the halo region, on the magnitude and time history of the halo current, and on the structure and duration of its toroidal and poloidal asymmetries. The effective length of the poloidal path of the halo current in the vessel is found to be rather insensitive to plasma parameters. Large values of the toroidally averaged halo current are observed in both vertical displacement events and centred disruptions but last a small fraction of the current quench; they coincide typically with a large but short-lived MHD event.

Scattering effects of small-scale density fluctuations on reflectometric measurements in a tokamak plasma

International Nuclear Information System (INIS)

Garcia, J.P.; Manso, M.E.; Serra, F.M.; Mendonca, J.T.

1989-01-01

When a wave propagates in a non homogeneous fluctuating plasma part of the incident energy is scattered out to the nonlinear interaction between the wave and the oscillating modes perturbing the plasma. The possibility of enhanced scattering at the cutoff layer, where reflection of the incident wave occurs, has been recently suggested as the basis of a reflectometric experiment to determine the spatial location of small scale fluctuations in a fusion plasma. Here we report on the development of a theoretical model to evaluate the flux of energy scattered by fluctuations, in order to give insight about the interpretation of measurements using a microwave reflectometry diagnostic in a tokamak. The scattered field is obtained through the resolution of a (non-homogeneous) wave propagation equation where the source term is related with the nonlinear current due to the interaction between the incident wave and local fluctuations. We use a slab model for the plasma, and an ordinary (0) wave propagation along the density gradient is considered. The amplitude of the scattered wave at the border of the plasma is estimated. In order to know the contributions to the energy scattered both from the propagation region and the reflecting layer, an approach was used where perturbations are modelled by spatial step functions at several layers. The main contribution to the scattered power comes from the cutoff region, where the electric field amplitude swells as compared with the incident value. Considering the reflectometric system recently installed on the ASDEX tokamak, and using typical density profiles, expected values of the 'swelling factor' have been numerically evaluated. The role of incoherent scattering due to drift wave activity is discussed as well as the coherent scattering due to fluctuations induced by lower hybrid (LH) waves. (author) 2 refs., 4 figs

IPP annual report 1981

International Nuclear Information System (INIS)

1982-01-01

In part A of this annual report the tokamak and stellarator projects at the IPP are reported: ASDEX, ASDEX upgrade, JET collaboration, NET collaboration, Wendelstein VII-7, Wendelstein VII-AS, Wendelstein VII-X and stellarator reactor system studies. In part B the departments and research groups give a brief, but detailed report of the results in the field of research and development. In part C a review is presented of the IPP organisation. Part D includes the papers and conference reports published in 1981. Finally a brief description of the IPP projects at German universities is presented. (GG) [de

Plasma surface interaction with tungsten in ASDEX Upgrade

International Nuclear Information System (INIS)

Dux, R.; Herrmann, A.; Kallenbach, A.; Neu, R.; Neuhauser, J.; Maier, H.; Pugno, R.; Puetterich, T.; Rohde, V.

2005-01-01

ASDEX Upgrade pursues the progressive increase of W coated plasma facing components. At present, the central column, the upper passive stabiliser loop, the complete upper divertor, the baffles at the lower divertor, as well as six tiles of one guard limiter at the low field side are W coated, representing about 65% of the total surface area. W erosion at these guard limiter tiles exceeds the erosion found at other main chamber components by more than one order of magnitude, and spectroscopically determined erosion yields indicate a strong contribution from fast particles. Upper single null discharges do not show an obviously increased W content compared to discharges run in the lower C based divertor

Present status of Tokamak research

International Nuclear Information System (INIS)

Basu, Jayanta

1991-01-01

The scenario of thermonuclear fusion research is presented, and the tokamak which is the most promising candidate as a fusion reactor is introduced. A brief survey is given of the most noteworthy tokamaks in the global context, and fusion programmes relating to Next Step devices are outlined. Supplementary heating of tokamak plasma by different methods is briefly reviewed; the latest achievements in heating to fusion temperatures are also reported. The progress towards the high value of the fusion product necessary for ignition is described. The improvement in plasma confinement brought about especially by the H-mode, is discussed. The latest situation in pushing up Β for increasing the efficiency of a tokamak is elucidated. Mention is made of the different types of wall treatment of the tokamak vessel for impurity control, which has led to a significant improvement in tokamak performance. Different methods of current drive for steady state tokamak operation are reviewed, and the issue of current drive efficiency is addressed. A short resume is given of the various diagnostic methods which are employed on a routine basis in the major tokamak centres. A few diagnostics recently developed or proposed in the context of the advanced tokamaks as well as the Next Step devices are indicated. The important role of the interplay between theory, experiment and simulation is noted, and the areas of investigation requiring concerted effort for further progress in tokamak research are identified. (author). 17 refs

Tokamak engineering mechanics

CERN Document Server

Song, Yuntao; Du, Shijun

2013-01-01

Tokamak Engineering Mechanics offers concise and thorough coverage of engineering mechanics theory and application for tokamaks, and the material is reinforced by numerous examples. Chapter topics include general principles, static mechanics, dynamic mechanics, thermal fluid mechanics and multiphysics structural mechanics of tokamak structure analysis. The theoretical principle of the design and the methods of the analysis for various components and load conditions are presented, while the latest engineering technologies are also introduced. The book will provide readers involved in the study

The tokamak as a neutron source

International Nuclear Information System (INIS)

Hendel, H.W.; Jassby, D.L.

1989-11-01

This paper describes the tokamak in its role as a neutron source, with emphasis on experimental results for D-D neutron production. The sections summarize tokamak operation, sources of fusion and non-fusion neutrons, principal neutron detection methods and their calibration, neutron energy spectra and fluxes outside the tokamak plasma chamber, history of neutron production in tokamaks, neutron emission and fusion power gain from JET and TFTR (the largest present-day tokamaks), and D-T neutron production from burnup of D-D tritons. This paper also discusses the prospects for future tokamak neutron production and potential applications of tokamak neutron sources. 100 refs., 16 figs., 4 tabs

Relevance, Realization and stability of a cold layer at the plasma edge for fusion reactors

International Nuclear Information System (INIS)

1990-09-01

The workshop was dedicated to the realization and stability of a cold layer at the plasma edge for fusion reactors. The subjects of the communications presented were: impurity transport, and control, plasma boundary layers, power balance, radiation control and modifications, limiter discharges, tokamak density limit, Asdex divertor discharges, thermal stability of a radiating diverted plasma, plasma stability, auxiliary heating in Textor, detached plasma in Tore Supra, poloidal divertor tokamak, radiation cooling, neutral-particle transport, plasma scrape-off layer, edge turbulence

Doppler coherence imaging of ion dynamics in VINETA.II and ASDEX-upgrade

Energy Technology Data Exchange (ETDEWEB)

Gradic, Dorothea; Ford, Oliver; Wolf, Robert [Max-Planck-Institut fuer Plasmaphysik, Greifswald (Germany); Lunt, Tilmann [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

2016-07-01

In magnetically confining plasma experiments, diagnosis of ion flows is of great importance to measure the plasma response to the magnetic field or the exhaust particle flows in the divertor areas. Doppler coherence imaging spectroscopy (CIS) is a relatively new technique for the observation of plasma bulk ion dynamics. It is a passive optical diagnostic enabling line-integrated measurements to obtain 2D images of the ion flow and ion temperature. The general principle is similar to traditional Doppler spectroscopy, however CIS uses an imaging interferometer to perform narrow-bandwidth Fourier spectroscopy. A major advantage of the coherence imaging technique is the large amount of spatial information recovered. This allows tomographic inversion of the line-integrated measurements. With existing CIS setups, scrape-off-layer and high field side edge impurity flows could be observed in the MAST, core and edge poloidal He II flows in the WEGA stellarator and divertor impurity flows in DIII-D. The main objective of this study is the research of ion dynamics in the small linear plasma experiment VINETA.II and ASDEX-Upgrade. First Doppler CIS measurements from Ar-II plasma discharges in VINETA.II and He-II, C-III divertor flows in ASDEX-Upgrade and their preliminary interpretation will be presented.

Particle influx measurements with the ASDEX-upgrade multichord visible spectroscopy system

International Nuclear Information System (INIS)

Kallenbach, A.; Mayer, H.M.; Schneider, W.

1992-01-01

This report describes the hardware and software components of the ASDEX-Upgrade multichord visible spectroscopy system. Main emphasis is laid on a detailed description of the detector, a free programmable charge-coupled device intensified by a microchannel plate. As an experimental application, flux measurements of different impurity species from the inner heat shield are presented. Poloidal profiles of the released impurity amount obtained for various experimental situations are used to check the plasma position which is derived by the function parametrization analysis. (orig.)

Characterization of ion heat conduction in JET and ASDEX Upgrade plasmas with and without internal transport barriers

Energy Technology Data Exchange (ETDEWEB)

Wolf, R C [Institut fuer Plasmaphysik, Forschungszentrum Juelich, Association EURATOM/FZJ, Trilateral Euregio Cluster, D-52425 Juelich (Germany); Baranov, Y [UKAEA/EURATOM Fusion Association, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Garbet, X [Association EURATOM-CEA sur la fusion, CEA Cadarache, F-13108 St Paul lez Durance (France); Hawkes, N [UKAEA/EURATOM Fusion Association, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Peeters, A G [Max-Planck-Institut fuer Plasmaphysik, EURATOM-Assoziation, D-85748 Garching (Germany); Challis, C [UKAEA/EURATOM Fusion Association, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Baar, M de [FOM Instituut voor Plasmafyisica Rijnhuizen, Association EURATO-FOM, Trilateral Euregio Cluster, PO Box 1207, 3430 BE Nieuwegein (Netherlands); Giroud, C [FOM Instituut voor Plasmafyisica Rijnhuizen, Association EURATO-FOM, Trilateral Euregio Cluster, PO Box 1207, 3430 BE Nieuwegein (Netherlands); Joffrin, E [Association EURATOM-CEA sur la fusion, CEA Cadarache, F-13108 St Paul lez Durance (France); Mantsinen, M [Helsinki University of Technology, Association-EURATOM Tekes, FIN-02015 HUT (Finland); Mazon, D [Association EURATOM-CEA sur la fusion, CEA Cadarache, F-13108 St Paul lez Durance (France); Meister, H [Max-Planck-Institut fuer Plasmaphysik, EURATOM-Assoziation, D-85748 Garching (Germany); Suttrop, W [Max-Planck-Institut fuer Plasmaphysik, EURATOM-Assoziation, D-85748 Garching (Germany); Zastrow, K-D [UKAEA/EURATOM Fusion Association, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom)

2003-09-01

In ASDEX Upgrade and JET, the ion temperature profiles can be described by R/L{sub Ti} which exhibits only little variations, both locally, when comparing different discharges, and radially over a wide range of the poloidal cross-section. Considering a change of the local ion heat flux of more than a factor of two, this behaviour indicates some degree of profile stiffness. In JET, covering a large ion temperature range from 1 to 25 keV, the normalized ion temperature gradient, R/L{sub Ti}, shows a dependence on the electron to ion temperature ratio or toroidal rotational shear. In particular, in hot ion plasmas, produced predominantly by neutral beam heating at low densities, in which large T{sub i}/T{sub e} is coupled to strong toroidal rotation, the effect of the two quantities cannot be distinguished. Both in ASDEX Upgrade and JET, plasmas with internal transport barriers (ITBs), including the PEP mode in JET, are characterized by a significant increase of R/L{sub Ti} above the value of L- and H-mode plasmas. In agreement with previous ASDEX Upgrade results, no increase of the ion heat transport in reversed magnetic shear ITB plasmas is found in JET when raising the electron heating. Evidence is presented that magnetic shear directly influences R/L{sub Ti}, namely decreasing the ion heat transport when going from weakly positive to negative magnetic shear.

Status of tokamak research

International Nuclear Information System (INIS)

Rawls, J.M.

1979-10-01

An overall review of the tokamak program is given with particular emphasis upon developments over the past five years in the theoretical and experimental elements of the program. A summary of the key operating parameters for the principal tokamaks throughout the world is given. Also discussed are key issues in plasma confinement, plasma heating, and tokamak design

Tokamak ARC damage

International Nuclear Information System (INIS)

Murray, J.G.; Gorker, G.E.

1985-01-01

Tokamak fusion reactors will have large plasma currents of approximately 10 MA with hundreds of megajoules stored in the magnetic fields. When a major plasma instability occurs, the disruption of the plasma current induces voltage in the adjacent conducting structures, giving rise to large transient currents. The induced voltages may be sufficiently high to cause arcing across sector gaps or from one protruding component to another. This report reviews a tokamak arcing scenario and provides guidelines for designing tokamaks to minimize the possibility of arc damage

Tokamak ARC damage

Energy Technology Data Exchange (ETDEWEB)

Murray, J.G.; Gorker, G.E.

1985-01-01

Tokamak fusion reactors will have large plasma currents of approximately 10 MA with hundreds of megajoules stored in the magnetic fields. When a major plasma instability occurs, the disruption of the plasma current induces voltage in the adjacent conducting structures, giving rise to large transient currents. The induced voltages may be sufficiently high to cause arcing across sector gaps or from one protruding component to another. This report reviews a tokamak arcing scenario and provides guidelines for designing tokamaks to minimize the possibility of arc damage.

Observation of precursor magnetic oscillations to the H-mode transition of ASDEX

International Nuclear Information System (INIS)

Toi, K.; Gernhardt, J.; Klueber, O.; Kornherr, M.

1988-05-01

Precursor oscillations to the H-mode transition are identified in magnetic fluctuations of the ASDEX H-mode discharges initiated without a sawtooth. This precursor is m=4/n=1 mode, rotating with f ≅ 10 kHz in the opposite direction to co-injected neutral beams. Time behaviour of the amplitude suggests that the H-mode transition is caused, not by the edge electron temperature, but by the edge current density. (orig.)

Kalman filters for real-time magnetic island phase tracking

NARCIS (Netherlands)

Borgers, D. P.; Lauret, M.; M.R. de Baar,

2013-01-01

For control of neoclassical tearing modes (NTMs) and the resulting rotating magnetic islands in tokamak plasmas, the frequency and phase of the magnetic islands need to be accurately tracked in real-time. In previous experiments on TEXTOR, this was achieved using a phase-locked loop (PLL). For ASDEX

Upgrades and Real Time Ntm Control Application of the Ece Radiometer on Asdex Upgrade

Science.gov (United States)

Hicks, N. K.; Suttrop, W.; Behler, K.; Giannone, L.; Manini, A.; Maraschek, M.; Raupp, G.; Reich, M.; Sips, A. C. C.; Stober, J.; Treutterer, W.; ASDEX Upgrade Team; Cirant, S.

2009-04-01

The 60-channel electron cyclotron emission (ECE) radiometer diagnostic on the ASDEX Upgrade tokamak is presently being upgraded to include a 1 MHz sampling rate data acquisition system. This expanded capability allows electron temperature measurements up to 500 kHz (anti-aliasing filter cut-off) with spatial resolution ~1 cm, and will thus provide measurement of plasma phenomena on the MHD timescale, such as neoclassical tearing modes (NTMs). The upgraded and existing systems may be run in parallel for comparison, and some of the first plasma measurements using the two systems together are presented. A particular planned application of the upgraded radiometer is integration into a real-time NTM stabilization loop using targeted deposition of electron cyclotron resonance heating (ECRH). For this loop, it is necessary to determine the locations of the NTM and ECRH deposition using ECE measurements. As the magnetic island of the NTM repeatedly rotates through the ECE line of sight, electron temperature fluctuations at the NTM frequency are observed. The magnetic perturbation caused by the NTM is independently measured using Mirnov coils, and a correlation profile between these magnetic measurements and the ECE data is constructed. The phase difference between ECE oscillations on opposite sides of the island manifests as a zero-crossing of the correlation profile, which determines the NTM location in ECE channel space. To determine the location of ECRH power deposition, the power from a given gyrotron may be modulated at a particular frequency. Correlation analysis of this modulated signal and the ECE data identifies a particular ECE channel associated with the deposition of that gyrotron. Real time equilibrium reconstruction allows the ECE channels to be translated into flux surface and spatial coordinates for use in the feedback loop.

Tokamak reactor studies

International Nuclear Information System (INIS)

Baker, C.C.

1981-01-01

This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, non-divertor particle control, and improved reactor safety features

Disassembly of JT-60 tokamak device and ancillary facilities for JT-60 tokamak

International Nuclear Information System (INIS)

Okano, Fuminori; Ichige, Hisashi; Miyo, Yasuhiko; Kaminaga, Atsushi; Sasajima, Tadayuki; Nishiyama, Tomokazu; Yagyu, Jun-ichi; Ishige, Youichi; Suzuki, Hiroaki; Komuro, Kenichi; Sakasai, Akira; Ikeda, Yoshitaka

2014-03-01

The disassembly of JT-60 tokamak device and its peripheral equipments, where the total weight was about 5400 tons, started in 2009 and accomplished in October 2012. This disassembly was required process for JT-60SA project, which is the Satellite Tokamak project under Japan-EU international corroboration to modify the JT-60 to the superconducting tokamak. This work was the first experience of disassembling a large radioactive fusion device based on Radiation Hazard Prevention Act in Japan. The cutting was one of the main problems in this disassembly, such as to cut the welded parts together with toroidal field coils, and to cut the vacuum vessel into two. After solving these problems, the disassembly completed without disaster and accident. This report presents the outline of the JT-60 disassembly, especially tokamak device and ancillary facilities for tokamak device. (author)

Tokamak confinement scaling laws

International Nuclear Information System (INIS)

Connor, J.

1998-01-01

The scaling of energy confinement with engineering parameters, such as plasma current and major radius, is important for establishing the size of an ignited fusion device. Tokamaks exhibit a variety of modes of operation with different confinement properties. At present there is no adequate first principles theory to predict tokamak energy confinement and the empirical scaling method is the preferred approach to designing next step tokamaks. This paper reviews a number of robust theoretical concepts, such as dimensional analysis and stability boundaries, which provide a framework for characterising and understanding tokamak confinement and, therefore, generate more confidence in using empirical laws for extrapolation to future devices. (author)

Spatial radiation profiles in the ASDEX Upgrade divertor for detached plasmas

International Nuclear Information System (INIS)

Wenzel, U.; Thoma, A.; Dux, R.; Fuchs, C.; Herrmann, A.; Hirsch, S.; Kallenbach, A.; Kastelewicz, H.; Laux, M.; Mast, F.; Napiontek, B.

1997-01-01

In this paper we describe impurity line emission measurements in the divertor of ASDEX Upgrade during high power neutral beam heated discharges. We focus on detached conditions where the dominating part of the radiation comes from the X-point region. Spatially resolved line emission in the VUV and visible spectral region of the intrinsic carbon and additionally puffed impurities (neon and nitrogen) is presented. A simple interpretation of the line emission profiles is given and they are also compared to the results of bolometry. (orig.)

Characterization of Scrape-Off layer turbulence changes induced by a non-axisymmetric magnetic perturbation in an ASDEX upgrade low density L-mode

International Nuclear Information System (INIS)

Mueller, H.W.; Carralero, D.; Birkenmeier, G.; Conway, G.D.; Fischer, R.; Happel, T.; Manz, P.; Suttrop, W.; Wolfrum, E.

2014-01-01

In the tokamak ASDEX Upgrade the influence of a non-axisymmetric n = 2 error field on the turbulence in the far scrape-off layer of a low density L-mode discharge has been studied. There is no density pump-out with the non-axisymmetric perturbation but an increase of the scrape-off layer density at the outer midplane. While the relative ion saturation current fluctuation level in the far scrape-off layer is decreasing, the skewness rises and especially the excess kurtosis grows by a factor of 1.5-3. The frequency of intermittent events (blobs) is increasing by 50 %. Also the poloidal velocity grows with the magnetic perturbation while the typical turbulent structure size becomes smaller by a factor 5-10 about 20-25 mm outside the separatrix. The local spectral density has been calculated from a two-point measurement of the ion saturation current. It is used to derive a dispersion relation. Two poloidal propagation velocities depending on the wave number have been found. One is an upper limit for the bulk E x B velocity and the second one the lower limit of the phase velocity. There is a significant contribution of the phase velocity to the propagation speed in the far scrape-off layer. (copyright 2014 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

Far-reaching Impact of Intermittent Transport across the Scrape-off Layer: Latest Results from ASDEX Upgrade

Energy Technology Data Exchange (ETDEWEB)

Kocan, M.; Muller, W.; Conway, G.; De Marne, P.; Eich, T.; Fischer, R.; Fuchs, C.; Herrmann, A.; Ionita, C.; Kallenbach, A.; Lunt, T.; Maraschek, M.; Muller, S.; Nold, B.; Ribeiro, T.; Rohde, V.; Scott, B.; Stroth, U.; Suttrop, W.; Wolfrum, E., E-mail: martin.kocan@ipp.mpg.de [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Garching (Germany); Adamek, J.; Horacek, J.; Komm, M. [Association EURATOM-IPP CR, Prague (Czech Republic); Gennrich, F.; Maszl, C.; Mehlmann, F.; Schrittwieser, R. [Institute for Ion Physics and Applied Physics, Association Euratom-OAW (Austria); Huang, Z. [Institut fuer Plasmaforschung, Universitat Stuttgart, Stuttgart (Germany)

2012-09-15

Full text: Latest research of intermittent transport in the scrape-off layer (SOL) of the ASDEX Upgrade tokamak is presented. Near the separatrix the fluctuations of the plasma and the floating potentials, measured by various Langmuir probes (LPs), are found to be anti-correlated due to fluctuations of the electron temperature. This indicates that, in contrast to a widely used experimental practice, a free exchange of both potentials is unjustified and can lead to significant error. Measurements of ion energies in turbulent L-mode and ELM filaments were carried out using a retarding field analyzer. In L-mode plasma, the filament ion temperature measured at 2 cm outside the separatrix is 80 - 110 eV, i.e., 3 - 4x the background ion temperature. Turbulent filaments also convect plasma to the wall with larger density than the background plasma density. Qualitatively similar observations were obtained during inter-ELM periods. Such enhanced particle and energy fluxes can potentially raise the erosion of the first wall in ITER. The ion temperature averaged over an ELM measured 35 - 60 mm outside the separatrix is in the range of 20 - 200 eV (5 - 50% of the pedestal top ion temperature). This demonstrates that ELM filaments carry hot ions over large radial distances in the SOL, which, in turn, can lead to enhanced sputtering from the first wall in future tokamaks. Lowest ion energies are observed during ELMs mitigated by in-vessel magnetic perturbations (MPs). The ELM ion temperature in the far SOL is found to increase with the ELM energy, indicating that on average the filaments in large ELMs propagate faster radially. The filamentary structure of the ion current density measured by LPs at the outboard mid-plane during mitigated ELMs is qualitatively similar to that observed during type I ELMs. The amplitude of the ion current density decreases only slightly when ELMs are mitigated, but, in contrast to type I ELMs, bursts of the ion current are observed throughout the

Large Aspect Ratio Tokamak Study

International Nuclear Information System (INIS)

Reid, R.L.; Holmes, J.A.; Houlberg, W.A.; Peng, Y.K.M.; Strickler, D.J.; Brown, T.G.; Wiseman, G.W.

1980-06-01

The Large Aspect Ratio Tokamak Study (LARTS) at Oak Ridge National Laboratory (ORNL) investigated the potential for producing a viable longburn tokamak reactor by enhancing the volt-second capability of the ohmic heating transformer through the use of high aspect ratio designs. The plasma physics, engineering, and economic implications of high aspect ratio tokamaks were assessed in the context of extended burn operation. Using a one-dimensional transport code plasma startup and burn parameters were addressed. The pulsed electrical power requirements for the poloidal field system, which have a major impact on reactor economics, were minimized by optimizing the startup and shutdown portions of the tokamak cycle. A representative large aspect ratio tokamak with an aspect ratio of 8 was found to achieve a burn time of 3.5 h at capital cost only approx. 25% greater than that of a moderate aspect ratio design tokamak

Kalman filters for real-time magnetic island phase tracking

International Nuclear Information System (INIS)

Borgers, D.P.; Lauret, M.; Baar, M.R. de

2013-01-01

Highlights: • We propose two Kalman filters for tracking of NTMs on ASDEX Upgrade. • The Kalman filters can track NTMs in a much larger frequency range than PLLs. • The filters are tested on synthetic and experimental data from TEXTOR and TCV. • We conclude that the unscented Kalman filter can be useful for NTM control. -- Abstract: For control of neoclassical tearing modes (NTMs) and the resulting rotating magnetic islands in tokamak plasmas, the frequency and phase of the magnetic islands need to be accurately tracked in real-time. In previous experiments on TEXTOR, this was achieved using a phase-locked loop (PLL). For ASDEX Upgrade however, the desired frequency range in which the islands are to be tracked (100 Hz–10 kHz) is much larger than is possible with a PLL. In this contribution, an extended Kalman filter (EKF) and an unscented Kalman filter (UKF) are proposed for real-time frequency, phase and amplitude tracking of sinusoidal signals, based on noisy measurements. Compared to PLLs, the EKF and UKF are able to track sinusoidal signals in a much larger frequency range. The filters are applied on synthetic data and on experimental data from the TEXTOR and TCV tokamaks, from which we conclude that the UKF can be useful for real-time control of magnetic islands on ASDEX Upgrade

Kalman filters for real-time magnetic island phase tracking

Energy Technology Data Exchange (ETDEWEB)

Borgers, D.P. [Hybrid and Networked Systems, Department of Mechanical Engineering – Eindhoven University of Technology, P.O. Box 513, 5600 MB Eindhoven (Netherlands); Lauret, M., E-mail: M.Lauret@tue.nl [FOM Institute DIFFER – Dutch Institute for Fundamental Energy Research, Association EURATOM-FOM, Trilateral Euregio Cluster, P.O. Box 1207, Nieuwegein (Netherlands); Control Systems Technology, Department of Mechanical Engineering – Eindhoven University of Technology, P.O. Box 513, 5600 MB Eindhoven (Netherlands); Baar, M.R. de [FOM Institute DIFFER – Dutch Institute for Fundamental Energy Research, Association EURATOM-FOM, Trilateral Euregio Cluster, P.O. Box 1207, Nieuwegein (Netherlands); Control Systems Technology, Department of Mechanical Engineering – Eindhoven University of Technology, P.O. Box 513, 5600 MB Eindhoven (Netherlands)

2013-11-15

Highlights: • We propose two Kalman filters for tracking of NTMs on ASDEX Upgrade. • The Kalman filters can track NTMs in a much larger frequency range than PLLs. • The filters are tested on synthetic and experimental data from TEXTOR and TCV. • We conclude that the unscented Kalman filter can be useful for NTM control. -- Abstract: For control of neoclassical tearing modes (NTMs) and the resulting rotating magnetic islands in tokamak plasmas, the frequency and phase of the magnetic islands need to be accurately tracked in real-time. In previous experiments on TEXTOR, this was achieved using a phase-locked loop (PLL). For ASDEX Upgrade however, the desired frequency range in which the islands are to be tracked (100 Hz–10 kHz) is much larger than is possible with a PLL. In this contribution, an extended Kalman filter (EKF) and an unscented Kalman filter (UKF) are proposed for real-time frequency, phase and amplitude tracking of sinusoidal signals, based on noisy measurements. Compared to PLLs, the EKF and UKF are able to track sinusoidal signals in a much larger frequency range. The filters are applied on synthetic data and on experimental data from the TEXTOR and TCV tokamaks, from which we conclude that the UKF can be useful for real-time control of magnetic islands on ASDEX Upgrade.

Micro-NRA and micro-3HIXE with 3He microbeam on samples exposed in ASDEX Upgrade and Pilot-PSI machines

Science.gov (United States)

Kelemen, Mitja; Založnik, Anže; Vavpetič, Primož; Pečovnik, Matic; Pelicon, Primož; Hakola, Antti; Lahtinen, Aki; Karhunen, Juuso; Piip, Kaarel; Paris, Peeter; Laan, Matti; Krieger, Karl; Oberkofler, Martin; van der Meiden, Hennie; Markelj, Sabina

2017-08-01

Micro nuclear reaction analysis (micro-NRA) exploiting the nuclear reaction D(3He,p)4He was used for post-mortem analyses of special marker samples, exposed to deuterium plasma inside ASDEX Upgrade (AUG) tokamak and to the deuterium plasma jet in the Pilot-PSI linear plasma gun. Lateral concentration profiles of deuterium and erosion/deposition profiles of the marker materials were obtained by a combination of micro-NRA and particle induced X-ray emission by 3He beam (3HIXE). In the case of AUG samples, where 25 nm thick W marker layers had been deposited on unpolished and polished graphite substrates, the effect of surface roughness on local erosion and deposition was also investigated. The lateral distribution of W concentration showed that erosion is much more distinct in the case of polished samples and the resulting surface shows a ;leopard; skin pattern of W accumulated on carbon aggregates left on the surface from polishing. The Pilot-PSI samples indicated preferential accumulation of deuterium a few mm off from the centre of the region affected by the plasma beam. This is connected with the largest surface modifications while the thick deposited layers at the centre do not favour deuterium retention per se. The results were cross correlated with those obtained using laser-induced breakdown spectroscopy (LIBS). With its quantitative abilities, micro-NRA provided essential calibration data for in situ LIBS operation, as well as for complementary post mortem Secondary Ion Mass Spectroscopy (SIMS).

Extension of electron cyclotron heating at ASDEX Upgrade with respect to high density operation

Directory of Open Access Journals (Sweden)

Schubert Martin

2017-01-01

Full Text Available The ASDEX Upgrade electron cyclotron resonance heating operates at 105 GHz and 140 GHz with flexible launching geometry and polarization. In 2016 four Gyrotrons with 10 sec pulse length and output power close to 1 MW per unit were available. The system is presently being extended to eight similar units in total. High heating power and high plasma density operation will be a part of the future ASDEX Upgrade experiment program. For the electron cyclotron resonance heating, an O-2 mode scheme is proposed, which is compatible with the expected high plasma densities. It may, however, suffer from incomplete single-pass absorption. The situation can be improved significantly by installing holographic mirrors on the inner column, which allow for a second pass of the unabsorbed fraction of the millimetre wave beam. Since the beam path in the plasma is subject to refraction, the beam position on the holographic mirror has to be controlled. Thermocouples built into the mirror surface are used for this purpose. As a protective measure, the tiles of the heat shield on the inner column were modified in order to increase the shielding against unabsorbed millimetre wave power.

Nonlinear evolution of the mode structure of ELMs in realistic ASDEX Upgrade geometry

Energy Technology Data Exchange (ETDEWEB)

Krebs, Isabel; Hoelzl, Matthias; Lackner, Karl; Guenter, Sibylle [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Boltzmannstr. 2, Garching (Germany); Collaboration: The ASDEX Upgrade Team

2013-07-01

Edge-localized modes (ELMs) are edge instabilities in H-mode plasmas, which eject particles and energy. The suitability of the H-mode for future fusion reactors depends crucially on the exact ELM dynamics as they can damage plasma facing components if too large. We have simulated ELMs in ASDEX Upgrade geometry using the nonlinear MHD code JOREK. Emphasis was put on the mode structure evolution in the early ELM phase which is characterized by the exponential growth of the unstable toroidal Fourier harmonics followed by a phase of saturation. In the linear phase, toroidal harmonics grow independently, whereas at larger amplitudes, the nonlinear interaction between the toroidal harmonics influences their growth and structure. Prior to mode saturation, the evolution of the mode structure can be reproduced well by a simple quadratic mode-interaction model, which yields a possible explanation for the strong n=1 component of type-I ELMs observed in ASDEX Upgrade. In the linear phase of the simulations, intermediate toroidal mode numbers (n 6-14) are most unstable as predicted by the peeling-ballooning model. But non-linearly, the n=1 component becomes important due to an energy transfer from pairs of linearly dominant toroidal harmonics with neighboring mode numbers to the n=1. The latter thereby changes its spatial structure.

Fusion potential for spherical and compact tokamaks

International Nuclear Information System (INIS)

Sandzelius, Mikael

2003-02-01

The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high β-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect

Fusion potential for spherical and compact tokamaks

Energy Technology Data Exchange (ETDEWEB)

Sandzelius, Mikael

2003-02-01

The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high {beta}-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect.

Microwave Tokamak Experiment

International Nuclear Information System (INIS)

Anon.

1988-01-01

The Microwave Tokamak Experiment, now under construction at the Laboratory, will use microwave heating from a free-electron laser. The intense microwave pulses will be injected into the tokamak to realize several goals, including a demonstration of the effects of localized heat deposition within magnetically confined plasma, a better understanding of energy confinement in tokamaks, and use of the new free-electron laser technology for plasma heating. The experiment, soon to be operational, provides an opportunity to study dense plasmas heated by powers unprecedented in the electron-cyclotron frequency range required by the especially high magnetic fields used with the MTX and needed for reactors. 1 references, 5 figures, 3 tables

Advanced Tokamak Stability Theory

Science.gov (United States)

Zheng, Linjin

2015-03-01

The intention of this book is to introduce advanced tokamak stability theory. We start with the derivation of the Grad-Shafranov equation and the construction of various toroidal flux coordinates. An analytical tokamak equilibrium theory is presented to demonstrate the Shafranov shift and how the toroidal hoop force can be balanced by the application of a vertical magnetic field in tokamaks. In addition to advanced theories, this book also discusses the intuitive physics pictures for various experimentally observed phenomena.

Simulation of the ASDEX divertor performance after hardening

International Nuclear Information System (INIS)

Schneider, W.; Lackner, K.; Neuhauser, J.; Wunderlich, R.

1985-05-01

Two combined computer models - a fluid description of the plasma scrape-off layer (SOLID) and a Monte-Carlo code for the neutral gas dynamics (DEGAS) - are used to assess changes in the divertor performance expected from the modifications in geometry needed for hardening the ASDEX divertor chamber for long-pulse, high-power heating. Stand-alone DEGAS calculations with assumed fixed scrape-off plasma parameters predict a doubling of the neutral escape probability, which, however, still remains so low, that achievement of the high divertor recycling regime can be expected over roughly the same operational regime as before modifications. This conclusion is also supported by fully self-consistent calculations with the combined model. Due to the reduced divertor, a significant reduction is predicted in the divertor time constant, which is expected to affect transient phenomena. (orig.)

Expected performance properties of the ASDEX upgrade toroidal field magnet derived from calculations and materials investigations

International Nuclear Information System (INIS)

Streibl, B.; Mukherjee, S.

1989-11-01

This is a summary of the TF-magnet calculation results for the 1984 phase-II proposal including supplements (also considering disturbances) of the performance of ASDEX Upgrade. Calculation results are as reliable as the assumptions incorporated, so that investigations of materials and design components were always used to complete the calculations. (orig.) [de

Burning plasma simulation and environmental assessment of tokamak, spherical tokamak and helical reactors

International Nuclear Information System (INIS)

Yamazaki, K.; Uemura, S.; Oishi, T.; Arimoto, H.; Shoji, T.; Garcia, J.

2009-01-01

Reference 1-GWe DT reactors (tokamak TR-1, spherical tokamak ST-1 and helical HR-1 reactors) are designed using physics, engineering and cost (PEC) code, and their plasma behaviours with internal transport barrier operations are analysed using toroidal transport analysis linkage (TOTAL) code, which clarifies the requirement of deep penetration of pellet fuelling to realize steady-state advanced burning operation. In addition, economical and environmental assessments were performed using extended PEC code, which shows the advantage of high beta tokamak reactors in the cost of electricity (COE) and the advantage of compact spherical tokamak in life-cycle CO 2 emission reduction. Comparing with other electric power generation systems, the COE of the fusion reactor is higher than that of the fission reactor, but on the same level as the oil thermal power system. CO 2 reduction can be achieved in fusion reactors the same as in the fission reactor. The energy payback ratio of the high-beta tokamak reactor TR-1 could be higher than that of other systems including the fission reactor.

Parametric dependencies of the experimental tungsten transport coefficients in ICRH and ECRH assisted ASDEX Upgrade H-modes

Science.gov (United States)

Sertoli, M.; Angioni, C.; Odstrcil, T.; ASDEX Upgrade Team; Eurofusion MST1 Team

2017-11-01

The profiles of the W transport coefficients have been experimentally calculated for a large database of identical ASDEX Upgrade H-mode discharges where only the radio-frequency (RF) power characteristics have been varied [Angioni et al., Nucl. Fusion 57, 056015 (2017)]. Central ion cyclotron resonance heating (ICRH) in the minority heating scheme has been compared with central and off-axis electron cyclotron resonance heating (ECRH), using both localized and broad heat deposition profiles. The transport coefficients have been calculated applying the gradient-flux relation to the evolution of the intrinsic W density in-between sawtooth cycles as measured using the soft X-ray diagnostic. For both ICRH and ECRH, the major player in reducing the central W density peaking is found to be the reduction of inward pinch and, in the case of ECRH, the rise of an outward convection. The impurity convection increases, from negative to positive, almost linearly with RF-power, while no appreciable changes are observed in the diffusion coefficient, which remains roughly at neoclassical levels independent of RF power or background plasma conditions. The ratio vW/DW is consistent with the equilibrium ∇ n W / n W prior to the sawtooth crash, corroborating the separate estimates of diffusion and convection. These experimental findings are slightly different from previous results obtained analysing the evolution of impurity injections over many sawtooth cycles. Modelling performed using the drift-kinetic code NEO and the gyro-kinetic code GKW (assuming axisymmetry) overestimates the diffusion coefficient and underestimates the experimental positive convection. This is a further indication that magneto-hydrodynamic/neoclassical models accounting for 3D effects may be needed to characterize impurity transport in sawtoothing tokamak plasmas.

Tokamak control simulator

International Nuclear Information System (INIS)

Edelbaum, T.N.; Serben, S.; Var, R.E.

1976-01-01

A computer model of a tokamak experimental power reactor and its control system is being constructed. This simulator will allow the exploration of various open loop and closed loop strategies for reactor control. This paper provides a brief description of the simulator and some of the potential control problems associated with this class of tokamaks

Behaviour of the ASDEX pressure gauge at high neutral gas pressure and applications for ITER

International Nuclear Information System (INIS)

Scarabosio, A.; Haas, G.

2008-01-01

The ASDEX Pressure Gauge is, at present, the main candidate for in-vessel neutral pressure measurement in ITER. Although the APG output is found to saturate at around 15 Pa, below the ITER requirement of 20 Pa. We show, here, that with small modifications of the gauge geometry and potentials settings we can achieve satisfactory behaviour up to 30 Pa at 6 T

Survey of Tokamak experiments

International Nuclear Information System (INIS)

Bickerton, R.J.

1977-01-01

The survey covers the following topics:- Introduction and history of tokamak research; review of tokamak apparatus, existing and planned; remarks on measurement techniques and their limitations; main results in terms of electron and ion temperatures, plasma density, containment times, etc. Empirical scaling; range of operating densities; impurities, origin, behaviour and control (including divertors); data on fluctuations and instabilities in tokamak plasmas; data on disruptive instabilities; experiments on shaped cross-sections; present experimental evidence on β limits; auxiliary heating; experimental and theoretical problems for the future. (author)

Comparison of Advanced Machine Learning Tools for Disruption Prediction and Disruption Studies

Czech Academy of Sciences Publication Activity Database

Odstrčil, Michal; Murari, A.; Mlynář, Jan

2013-01-01

Roč. 41, č. 7 (2013), s. 1751-1759 ISSN 0093-3813 R&D Projects: GA ČR GAP205/10/2055 Institutional support: RVO:61389021 Keywords : Learning Machines * Support Vector Machines * Neural Network * ASDEX Upgrade * JET * Disruption mitigation * Tokamaks * ITER Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.950, year: 2013

Expression for the thermal H-mode energy confinement time under ELM-free conditions

International Nuclear Information System (INIS)

Ryter, F.; Gruber, O.; Kardaun, O.J.W.F.; Menzler, H.P.; Wagner, F.; Schissel, D.P.; DeBoo, J.C.; Kaye, S.M.

1992-07-01

The design of future tokamaks, which are supposed to reach ignition with the H-mode, requires a reliable scaling expression for the H-mode energy confinement time. In the present work, an H-mode scaling expression for the thermal plasma energy confinement time has been developed by combining data from four existing divertor tokamaks, ASDEX, DIII-D, JET and PBX-M. The plasma conditions, which were as similar as possible to ensure a coherent set of data, were ELM-free deuterium discharges heated by deuterium neutral beam injection. By combining four tokamaks, the parametric dependence of the thermal energy confinement on the main plasma parameters, including the three main geometrical variables, was determined. (orig./WL)

Tokamak building-design considerations for a large tokamak device

International Nuclear Information System (INIS)

Barrett, R.J.; Thomson, S.L.

1981-01-01

Design and construction of a satisfactory tokamak building to support FED appears feasible. Further, a pressure vessel building does not appear necessary to meet the plant safety requirements. Some of the building functions will require safety class systems to assure reliable and safe operation. A rectangular tokamak building has been selected for FED preconceptual design which will be part of the confinement system relying on ventilation and other design features to reduce the consequences and probability of radioactivity release

Modelling of Ohmic discharges in ADITYA tokamak using the Tokamak Simulation Code

International Nuclear Information System (INIS)

Bandyopadhyay, I; Ahmed, S M; Atrey, P K; Bhatt, S B; Bhattacharya, R; Chaudhury, M B; Deshpande, S P; Gupta, C N; Jha, R; Joisa, Y Shankar; Kumar, Vinay; Manchanda, R; Raju, D; Rao, C V S; Vasu, P

2004-01-01

Several Ohmic discharges of the ADITYA tokamak are simulated using the Tokamak Simulation Code (TSC), similar to that done earlier for the TFTR tokamak. Unlike TFTR, the dominant radiation process in ADITYA is through impurity line radiation. TSC can follow the experimental plasma current and position to very good accuracy. The thermal transport model of TSC including impurity line radiation gives a good match of the simulated results with experimental data for the Ohmic flux consumption, electron temperature and Z eff . Even the simulated magnetic probe signals are in reasonably good agreement with the experimental values

Modelling of Ohmic discharges in ADITYA tokamak using the Tokamak Simulation Code

Energy Technology Data Exchange (ETDEWEB)

Bandyopadhyay, I; Ahmed, S M; Atrey, P K; Bhatt, S B; Bhattacharya, R; Chaudhury, M B; Deshpande, S P; Gupta, C N; Jha, R; Joisa, Y Shankar; Kumar, Vinay; Manchanda, R; Raju, D; Rao, C V S; Vasu, P [Institute for Plasma Research, Bhat, Gandhinagar 382428 (India)

2004-09-01

Several Ohmic discharges of the ADITYA tokamak are simulated using the Tokamak Simulation Code (TSC), similar to that done earlier for the TFTR tokamak. Unlike TFTR, the dominant radiation process in ADITYA is through impurity line radiation. TSC can follow the experimental plasma current and position to very good accuracy. The thermal transport model of TSC including impurity line radiation gives a good match of the simulated results with experimental data for the Ohmic flux consumption, electron temperature and Z{sub eff}. Even the simulated magnetic probe signals are in reasonably good agreement with the experimental values.

Advanced commercial tokamak study

International Nuclear Information System (INIS)

Thomson, S.L.; Dabiri, A.E.; Keeton, D.C.; Brown, T.G.; Bussell, G.T.

1985-12-01

Advanced commercial tokamak studies were performed by the Fusion Engineering Design Center (FEDC) as a participant in the Tokamak Power Systems Studies (TPSS) project coordinated by the Office of Fusion Energy. The FEDC studies addressed the issues of tokamak reactor cost, size, and complexity. A scoping study model was developed to determine the effect of beta on tokamak economics, and it was found that a competitive cost of electricity could be achieved at a beta of 10 to 15%. The implications of operating at a beta of up to 25% were also addressed. It was found that the economics of fusion, like those of fission, improve as unit size increases. However, small units were found to be competitive as elements of a multiplex plant, provided that unit cost and maintenance time reductions are realized for the small units. The modular tokamak configuration combined several new approaches to develop a less complex and lower cost reactor. The modular design combines the toroidal field coil with the reactor structure, locates the primary vacuum boundary at the reactor cell wall, and uses a vertical assembly and maintenance approach. 12 refs., 19 figs

D-D tokamak reactor assessment

International Nuclear Information System (INIS)

Baxter, D.C.; Dabiri, A.E.

1983-01-01

A quantitative comparison of the physics and technology requirements, and the cost and safety performance of a d-d tokamak relative to a d-t tokamak has been performed. The first wall/blanket and energy recovery cycle for the d-d tokamak is simpler, and has a higher efficiency than the d-t tokamak. In most other technology areas (such as magnets, RF, vacuum, etc.) d-d requirements are more severe and the systems are more complex, expensive and may involve higher technical risk than d-t tokamak systems. Tritium technology for processing the plasma exhaust, and tritium refueling technology are required for d-d reactors, but no tritium containment around the blanket or heat transport system is needed. Cost studies show that for high plasma beta and high magnetic field the cost of electricity from d-d and d-t tokamaks is comparable. Safety analysis shows less radioactivity in a d-d reactor but larger amounts of stored energy and thus higher potential for energy release. Consequences of all postulated d-d accidents are significantly smaller than those from d-t reactor tritium releases

Scrape-off layer radiation and heat load to the ASDEX Upgrade LYRA divertor

International Nuclear Information System (INIS)

Kallenbach, A.; Kaufmann, M.; Coster, D.P.

1999-01-01

In 1997 the new 'LYRA' divertor went into operation at ASDEX Upgrade and, in parallel, the neutral beam heating power was increased to 20 MW by installation of a second injector leading to a P/R value of 12 MW/m. Experiments have shown that the ASDEX Upgrade LYRA divertor is capable of handling such high heating powers. There is an overall reduction of the maximum heat flux in the LYRA divertor by about a factor of 2 compared with the previous open divertor Div I. This reduction is mainly due to increased radiative losses inside the divertor region, which are caused by an effective reflection of hydrogen neutrals into the hot separatrix region. The main channel of radiative loss is carbon radiation, which cools the divertor plasma down to a few electronvolts, where hydrogen radiation losses become significant. The radiative losses preferentially reduce the power flux at the separatrix, leading to early detachment around the strike point position. With increasing density, the detached region extends upwards on the vertical target. The power fraction radiated in the LYRA divertor is around 45% and nearly independent of the heating power. This value is a factor of 2 higher than the typical radiation fraction in Div I. B2-EIRENE modelling of the performed experiments supports the experimental finding and refines the understanding of loss processes in the divertor region. (author)

Summary discussion: An integrated advanced tokamak reactor

International Nuclear Information System (INIS)

Sauthoff, N.R.

1994-01-01

The tokamak concept improvement workshop addressed a wide range of issues involved in the development of a more attractive tokamak. The agenda for the workshop progressed from a general discussion of the long-range energy context (with the objective being the identification of a set of criteria and ''figures of merit'' for measuring the attractiveness of a tokamak concept) to particular opportunities for the improvement of the tokamak concept. The discussions concluded with a compilation of research program elements leading to an improved tokamak concept

Energy flux to the ASDEX-upgrade diverter plates determined by thermography and calorimetry

International Nuclear Information System (INIS)

Herrmann, A.; Junker, W.; Guenther, K.

1995-01-01

A new thermography system with high time resolution was put into operation at ASDEX-Upgrade and is routinely used to determine the energy flux onto the lower diverter plates. The measurements allow the power deposition to be characterized during dynamic events such as ELMs and disruptions, as well as the asymmetry of the inboard/outboard power load. A power balance is set up even during single discharges and the losses are found to be fairly equal to the power input. (author)

IPP: Annual report 1982

International Nuclear Information System (INIS)

1982-01-01

This report presents a detailed outline of the work accomplished within the framework of the Tokamak projects ASDEX and ASDEX Upgrade and the JET and NET projects, the stellarator projects VIII-A, VII-AS and VII-X, and of stellarator reactor system studies. The various scientific research departments, such as Experimental Plasma Physics 1, 2 and 3, Theoretical Plasma Physics 1 and 2, Plasma-Wall. Interactions, and Information Sciences give an account of the results of their research and development work. Further reports on the 1982 activities deal with a) the organisational structure of the IPP, b) administrative and business items, c) scientific publications and conference proceedings, and with d) IPP-supported projects at universities. (GG) [de

Radial transport in the far scrape-off layer of ASDEX upgrade during L-mode and ELMy H-mode

DEFF Research Database (Denmark)

Ionita, C.; Naulin, Volker; Mehlmann, F.

2013-01-01

The radial turbulent particle flux and the Reynolds stress in the scrape-off layer (SOL) of ASDEX Upgrade were investigated for two limited L-mode (low confinement) and one ELMy H-mode (high confinement) discharge. A fast reciprocating probe was used with a probe head containing five Langmuir...

Modelling and experiments on NTM stabilisation at ASDEX upgrade

Energy Technology Data Exchange (ETDEWEB)

Urso, Laura

2009-07-27

In the next fusion device ITER the so-called neoclassical tearing modes (NTMs) are foreseen as being extremely detrimental to plasma confinement. This type of resistive instability is related to the presence in the plasma of magnetic islands. These are experimentally controlled with local electron cyclotron current drive (ECCD) and the island width decay during NTM stabilisation is modelled using the so-called Modified Rutherford equation. In this thesis, a modelling of the Modified Rutherford equation is carried out and simulations of the island width decay are compared with the experimentally observed ones in order to fit the two free machine-independent parameters present in the equation. A systematic study on a database of NTM stabilisation discharges from ASDEX Upgrade and JT-60U is done within the context of a multi-machine benchmark for extrapolating the ECCD power requirements for ITER. The experimental measurements in both devices are discussed by means of consistency checks and sensitivity analysis and used to evaluate the two fitting parameters present in the Modified Rutherford equation. The influence of the asymmetry of the magnetic island on stabilisation is for the first time included in the model and the effect of ECCD on the marginal island after which the mode naturally decays is quantified. The effect of radial misalignment and over-stabilisation during the experiment are found to be the key quantities affecting the NTM stabilisation. As a main result of this thesis, the extrapolation to ITER of the NTM stabilisation results from ASDEX Upgrade and JT-60U shows that 10MW of ECCD power are enough to stabilise large NTMs as long as the O-point of the island and the ECCD beam are perfectly aligned. In fact, the high ratio between the island size at saturation and the deposition width of the ECCD beam foreseen for ITER is found to imply a maximum allowable radial misalignment of 2-3 cm and little difference in terms of gained performance between

Modelling and experiments on NTM stabilisation at ASDEX upgrade

International Nuclear Information System (INIS)

Urso, Laura

2009-01-01

In the next fusion device ITER the so-called neoclassical tearing modes (NTMs) are foreseen as being extremely detrimental to plasma confinement. This type of resistive instability is related to the presence in the plasma of magnetic islands. These are experimentally controlled with local electron cyclotron current drive (ECCD) and the island width decay during NTM stabilisation is modelled using the so-called Modified Rutherford equation. In this thesis, a modelling of the Modified Rutherford equation is carried out and simulations of the island width decay are compared with the experimentally observed ones in order to fit the two free machine-independent parameters present in the equation. A systematic study on a database of NTM stabilisation discharges from ASDEX Upgrade and JT-60U is done within the context of a multi-machine benchmark for extrapolating the ECCD power requirements for ITER. The experimental measurements in both devices are discussed by means of consistency checks and sensitivity analysis and used to evaluate the two fitting parameters present in the Modified Rutherford equation. The influence of the asymmetry of the magnetic island on stabilisation is for the first time included in the model and the effect of ECCD on the marginal island after which the mode naturally decays is quantified. The effect of radial misalignment and over-stabilisation during the experiment are found to be the key quantities affecting the NTM stabilisation. As a main result of this thesis, the extrapolation to ITER of the NTM stabilisation results from ASDEX Upgrade and JT-60U shows that 10MW of ECCD power are enough to stabilise large NTMs as long as the O-point of the island and the ECCD beam are perfectly aligned. In fact, the high ratio between the island size at saturation and the deposition width of the ECCD beam foreseen for ITER is found to imply a maximum allowable radial misalignment of 2-3 cm and little difference in terms of gained performance between

Progress of the ECH·ECCD experiments. Research progress of the ECH·ECCD experiments in tokamaks and spherical tokamaks

International Nuclear Information System (INIS)

Isayama, Akihiko; Tanaka, Hitoshi

2009-01-01

Recent progress in the ECH·ECCD study in tokamak and spherical tokamak devices is described. As for the tokamak study, results on the control of neoclassical tearing modes and sawtooth oscillations, the current profile, the internal transport barrier, the plasma start-up and the discharge cleaning are given. As for the spherical tokamak study, the plasma start-up by ECH·ECCD and the electron-Bernstein-wave heating and the current drive are described. (T.I.)

Preliminary Design of Alborz Tokamak

Science.gov (United States)

Mardani, M.; Amrollahi, R.; Saramad, S.

2012-04-01

The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. The most important part of the tokamak design is the design of TF coils. In this paper a refined design of the TF coil system for the Alborz tokamak is presented. This design is based on cooper cable conductor with 5 cm width and 6 mm thickness. The TF coil system is consist of 16 rectangular shape coils, that makes the magnetic field of 0.7 T at the plasma center. The stored energy in total is 160 kJ, and the power supply used in this system is a capacitor bank with capacity of C = 1.32 mF and V max = 14 kV.

Moving Divertor Plates in a Tokamak

International Nuclear Information System (INIS)

Zweben, S.J.; Zhang, H.

2009-01-01

Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions

Moving Divertor Plates in a Tokamak

Energy Technology Data Exchange (ETDEWEB)

S.J. Zweben, H. Zhang

2009-02-12

Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions.

Radial transport of poloidal momentum in ASDEX Upgrade in L-mode and H-mode

DEFF Research Database (Denmark)

Schrittwieser, R.; Mehlmann, F.; Naulin, Volker

2012-01-01

A reciprocating probe was used for localized measurements of the radial transport of poloidal momentum in the scrape-off layer (SOL) of ASDEX Upgrade (AUG). The probe measured poloidal and radial electric field components and density. We concentrate on three components of the momentum transport: ......: Reynolds stress, convective momentum flux and triple product of the fluctuating components of density, radial and poloidal electric field. For the evaluation we draw mainly on the probability density functions (PDFs)....

The ARIES tokamak reactor study

International Nuclear Information System (INIS)

1989-10-01

The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D 3 He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions

Determination of the stochastic layer properties induced by magnetic perturbations via heat pulse experiments at ASDEX upgrade

Czech Academy of Sciences Publication Activity Database

Brida, D.; Lunt, T.; Wischmeier, M.; Birkenmeier, G.; Cahyna, Pavel; Carralero, D.; Faitsch, M.; Feng, Y.; Kurzan, B.; Schubert, M.; Sieglin, B.; Suttrop, W.; Wolfrum, E.

2017-01-01

Roč. 12, August (2017), s. 831-837 ISSN 2352-1791 EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : ASDEX upgrade * Magnetic perturbations * Divertor heat flux Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) http://www.sciencedirect.com/science/article/pii/S2352179116302150

Summary report on tokamak confinement experiments

International Nuclear Information System (INIS)

1982-03-01

There are currently five major US tokamaks being operated and one being constructed under the auspices of the Division of Toroidal Confinement Systems. The currently operating tokamaks include: Alcator C at the Massachusetts Institute of Technology, Doublet III at the General Atomic Company, the Impurity Studies Experiment (ISX-B) at the Oak Ridge National Laboratory, and the Princeton Large Torus (PLT) and the Poloidal Divertor Experiment (PDX) at the Princeton Plasma Physics Laboratory. The Tokamak Fusion Test Reactor (TFTR) is under construction at Princeton and should be completed by December 1982. There is one major tokamak being funded by the Division of Applied Plasma Physics. The Texas Experimental Tokamak (TEXT) is being operated as a user facility by the University of Texas. The TEXT facility includes a complete set of standard diagnostics and a data acquisition system available to all users

First operations with the new Collective Thomson Scattering diagnostic on the Frascati Tokamak Upgrade device

Science.gov (United States)

Bin, W.; Bruschi, A.; D'Arcangelo, O.; Castaldo, C.; De Angeli, M.; Figini, L.; Galperti, C.; Garavaglia, S.; Granucci, G.; Grosso, G.; Korsholm, S. B.; Lontano, M.; Mellera, V.; Minelli, D.; Moro, A.; Nardone, A.; Nielsen, S. K.; Rasmussen, J.; Simonetto, A.; Stejner, M.; Tartari, U.

2015-10-01

Anomalous emissions were found over the last few years in spectra of Collective Thomson Scattering (CTS) diagnostics in tokamak devices such as TEXTOR, ASDEX and FTU, in addition to real CTS signals. The signal frequency, down-shifted with respect to the probing one, suggested a possible origin in Parametric Decay Instability (PDI) processes correlated with the presence of magnetic islands and occurring for pumping wave power levels well below the threshold predicted by conventional models. A threshold below or close to the Electron Cyclotron Resonance Heating (ECRH) power levels could limit, under certain circumstances, the use of the ECRH in fusion devices. An accurate characterization of the conditions for the occurrence of this phenomenon and of its consequences is thus of primary importance. Exploiting the front-steering configuration available with the real-time launcher, the implementation of a new CTS setup now allows studying these anomalous emission phenomena in FTU under conditions of density and wave injection geometry that are more similar to those envisaged for CTS in ITER. The upgrades of the diagnostic are presented as well as a few preliminary spectra detected with the new system during the very first operations in 2014. The present work has been carried out under an EUROfusion Enabling Research project. A shorter version of this contribution is due to be published in PoS at: 1st EPS conference on Plasma Diagnostics

A forward model for the helium plume effect and the interpretation of helium charge exchange measurements at ASDEX Upgrade

Science.gov (United States)

Kappatou, A.; McDermott, R. M.; Pütterich, T.; Dux, R.; Geiger, B.; Jaspers, R. J. E.; Donné, A. J. H.; Viezzer, E.; Cavedon, M.; the ASDEX Upgrade Team

2018-05-01

The analysis of the charge exchange measurements of helium is hindered by an additional emission contributing to the spectra, the helium ‘plume’ emission (Fonck et al 1984 Phys. Rev. A 29 3288), which complicates the interpretation of the measurements. The plume emission is indistinguishable from the active charge exchange signal when standard analysis of the spectra is applied and its intensity is of comparable magnitude for ASDEX Upgrade conditions, leading to a significant overestimation of the He2+ densities if not properly treated. Furthermore, the spectral line shape of the plume emission is non-Gaussian and leads to wrong ion temperature and flow measurements when not taken into account. A kinetic model for the helium plume emission has been developed for ASDEX Upgrade. The model is benchmarked against experimental measurements and is shown to capture the underlying physics mechanisms of the plume effect, as it can reproduce the experimental spectra and provides consistent values for the ion temperature, plasma rotation, and He2+ density.

Spatiotemporal Oscillations in Tokamak Edge Layer and their Generation by Various Mechanisms

Energy Technology Data Exchange (ETDEWEB)

Daybelge, U.; Yarim, C., E-mail: daybelge@itu.edu.tr [Istanbul Technical University, Istanbul (Turkey); Nicolai, A. [Forschungszentrum Juelich, Juelich (Germany)

2012-09-15

Full text: Toroidal and poloidal rotations of plasma at the edge region of tokamak devices have long been known to play an important role, such as enhancing the confinement properties by suppressing turbulent behaviour, improving tolerance to error fields and increasing stability to neoclassical tearing modes. Hence, understanding of creation and evolution of rotation is important, since external momentum would not be enough or could not even be realized especially for future large fusion devices. In addition to the externally applied momentum, several mechanisms have been suggested to explain the reasons for spontaneous toroidal rotation of plasmas. For a tokamak edge region as found, for example, within the operational boundaries of the ASDEX upgrade, relevance of the collisional neoclassical theory was recently emphasized. In this regime gyrostresses play a considerable role in modifying the coupled flux surface averaged continuity, energy and momentum equations. Examination of the terms in these equations that are responsible for diffusion or reaction and acting as sources, can show the share of the neoclassical mechanisms to terms like intrinsic rotation, etc. Using similarities of our equations to the nonlinear reaction-diffusion equations with a susceptibility to the Turing instability and applying some robust numerical methods, we present here an approach based on the spatiotemporal simulation of the oscillations in plasma temperature, density, toroidal and poloidal rotation velocities under various perturbative effects. Present study considers a subsonic, collisional plasma in front of the magnetic separatrix. Study indicates a nonlinear, three-time-scales-coupling between the evolutions of the density, temperature and poloidal and toroidal rotation velocities. Numerical solutions of the coupled system for the vector W = [T,N,U{sub {phi}}, U{sub {theta}}] were studied under various given sources such as a periodic pellet injection or loop voltage variation

Research using small tokamaks

International Nuclear Information System (INIS)

1991-01-01

The technical reports contained in this collection of papers on research using small tokamaks fall into four main categories, i.e., (i) experimental work (heating, stability, plasma radial profiles, fluctuations and transport, confinement, ultra-low-q tokamaks, wall physics, a.o.), (ii) diagnostics (beam probes, laser scattering, X-ray tomography, laser interferometry, electron-cyclotron absorption and emission systems), (iii) theory (strong turbulence, effects of heating on stability, plasma beta limits, wave absorption, macrostability, low-q tokamak configurations and bootstrap currents, turbulent heating, stability of vortex flows, nonlinear islands growth, plasma-drift-induced anomalous transport, ergodic divertor design, a.o.), and (iv) new technical facilities (varistors applied to establish constant current and loop voltage in HT-6M), lower-hybrid-current-drive systems for HT-6B and HT-6M, radio-frequency systems for HT-6M ICR heating experimentation, and applications of fiber optics for visible and vacuum ultraviolet radiation detection as applied to tokamaks and reversed-field pinches. A total number of 51 papers are included in the collection. Refs, figs and tabs

Three-wave interaction during electron cyclotron resonance heating and current drive

DEFF Research Database (Denmark)

Nielsen, Stefan Kragh; Jacobsen, Asger Schou; Hansen, Søren Kjer

2016-01-01

Non-linear wave-wave interactions in fusion plasmas, such as the parametric decay instability (PDI) of gyrotron radiation, can potentially hamper the use of microwave diagnostics. Here we report on anomalous scattering in the ASDEX Upgrade tokamak during electron cyclotron resonance heating...... experiments. The observations can be linked to parametric decay of the gyrotron radiation at the second harmonic upper hybrid resonance layer....

Accelerator technology in tokamaks

International Nuclear Information System (INIS)

Kustom, R.L.

1977-01-01

This article presents the similarities in the technology required for high energy accelerators and tokamak fusion devices. The tokamak devices and R and D programs described in the text represent only a fraction of the total fusion program. The technological barriers to producing successful, economical tokamak fusion power plants are as many as the plasma physics problems to be overcome. With the present emphasis on energy problems in this country and elsewhere, it is very likely that fusion technology related R and D programs will vigorously continue; and since high energy accelerator technology has so much in common with fusion technology, more scientists from the accelerator community are likely to be attracted to fusion problems

Texas Experimental Tokamak

International Nuclear Information System (INIS)

Wootton, A.J.

1993-04-01

This progress report covers the period from November 1, 1990 to April 30, 1993. During that period, TEXT was operated as a circular tokamak with a material limiter. It was devoted to the study of basic plasma physics, in particular to study of fluctuations, turbulence, and transport. The purpose is to operate and maintain TEXT Upgrade as a complete facility for applied tokamak physics, specifically to conduct a research program under the following main headings: (1) to elucidate the mechanisms of working gas, impurity, and thermal transport in tokamaks, in particular to understand the role of turbulence; (2) to study physics of the edge plasma, in particular the turbulence; (3) to study the physics or resonant magnetic fields (ergodic magnetic divertors, intra island pumping); and (4) to study the physics of electron cyclotron heating (ECRH). Results of studies in each of these areas are reported

START: the creation of a spherical tokamak

International Nuclear Information System (INIS)

Sykes, Alan

1992-01-01

The START (Small Tight Aspect Ratio Tokamak) plasma fusion experiment is now operational at AEA Fusion's Culham Laboratory. It is the world's first experiment to explore an extreme limit of the tokamak - the Spherical Tokamak - which theoretical studies predict may have substantial advantages in the search for economic fusion power. The Head of the START project, describes the concept, some of the initial experimental results and the possibility of developing a spherical tokamak power reactor. (author)

Tokamaks (Second Edition)

Energy Technology Data Exchange (ETDEWEB)

Stott, Peter [JET, UK (United Kingdom)

1998-10-01

The first edition of John Wesson's book on tokamaks, published in 1987, established itself as essential reading for researchers in the field of magnetic confinement fusion: it was an excellent introduction for students to tokamak physics and also a valuable reference work for the more experienced. The second edition, published in 1997, has been completely rewritten and substantially enlarged (680 pages compared with 300). The new edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes discussion of the substantial advances in fusion research during the past decade. The new book, like its predecessor, is well written and commendable for its clarity and accuracy. In fact many of the chapters are written by a series of co-authors bringing the benefits of a wide range of expertise but, by careful editing, Wesson has maintained a uniformity of style and presentation. The chapter headings and coverage for the most part remain the same - but are expanded considerably and brought up to date. The most substantial change is that the single concluding chapter in the first edition on 'Experiments' has been replaced by three chapters: 'Tokamak experiments' which deals with some of the earlier key experiments plus a selection of recent small and medium-sized devices, 'Large experiments' which gives an excellent summary of the main results from the four large tokamaks - TFTR, JET, JT60/JT60U and DIII-D, and 'The future' which gives a very short (possibly too short in my opinion) account of reactors and ITER. This is an excellent book, which I strongly recommend should have a place - on the desk rather than in the bookshelf - of researchers in magnetic confinement fusion. (book review)

Tokamaks (Second Edition)

International Nuclear Information System (INIS)

Stott, Peter

1998-01-01

The first edition of John Wesson's book on tokamaks, published in 1987, established itself as essential reading for researchers in the field of magnetic confinement fusion: it was an excellent introduction for students to tokamak physics and also a valuable reference work for the more experienced. The second edition, published in 1997, has been completely rewritten and substantially enlarged (680 pages compared with 300). The new edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes discussion of the substantial advances in fusion research during the past decade. The new book, like its predecessor, is well written and commendable for its clarity and accuracy. In fact many of the chapters are written by a series of co-authors bringing the benefits of a wide range of expertise but, by careful editing, Wesson has maintained a uniformity of style and presentation. The chapter headings and coverage for the most part remain the same - but are expanded considerably and brought up to date. The most substantial change is that the single concluding chapter in the first edition on 'Experiments' has been replaced by three chapters: 'Tokamak experiments' which deals with some of the earlier key experiments plus a selection of recent small and medium-sized devices, 'Large experiments' which gives an excellent summary of the main results from the four large tokamaks - TFTR, JET, JT60/JT60U and DIII-D, and 'The future' which gives a very short (possibly too short in my opinion) account of reactors and ITER. This is an excellent book, which I strongly recommend should have a place - on the desk rather than in the bookshelf - of researchers in magnetic confinement fusion. (book review)

Edge Localized Modes: resent experimental findings and related issues

International Nuclear Information System (INIS)

Kamiya, K.

2007-01-01

Edge Localized Mode (ELM) measurements in the tokamaks, including JT-60U, DIII-D, ASDEX-U and JET, are reviewed. An ELMy H-mode operation having Type-I ELMs is nominated as the reference inductive operational scenario for ITER (Q DT =10), which is normally observed for the best performing H-mode in many tokamaks,. However, the ELMs produce pulsed heat and particle fluxes that can lead to a rapid erosion of the divertor plate. It is estimated that the peak heat flux to the divertor would reduce the lifetime of the divertor to several hundred shots in ITER (e.g. an acceptable divertor lifetime could be realized only by an upper limit of ELM energy loss normalized by pedestal stored energy, ΔDW ELM /W ped ∼ 5-6%). Approaches to control the Type-I ELMs, such as '' Ergodization '' on DIII-D, '' Pace making by a shallow pellet injection '' on ASDEX-U, '' Vertical motion '' on TCV, have been successfully demonstrated in many tokamaks. On the other hand, finding alternative scenarios to Type-I ELMy H-mode operation are also a key area of research for current tokamaks. Specifically, '' Quiescent H-mode (QH-mode) '' on DIII-D, ASDEX-U and JT-60U, and '' Grassy ELMs '' on JT-60U demonstrated a high confinement (being comparable to that of Type-I ELMy H-mode plasmas at similar parameters) in the absence of large, ELM induced, transient heat/particle fluxes to the divertor targets. ELM dynamics measurements in the SOL at the midplane show large, rapid variations of the SOL parameters. Recent data from a fast resolved measurements, such as scanning probe, radial interferometer chord, BES and tangentially viewing fast-gated camera at the midplane, suggest a filamentary structure of the perturbation with fast radial propagation in later phases and parallel propagation of the ELM pulse at around the sound speed of pedestal ions. The results are qualitatively consistent with nonlinear ballooning theory, although a more quantitative physics understanding, including detailed

Modelling dust transport in tokamaks

International Nuclear Information System (INIS)

Martin, J.D.; Martin, J.D.; Bacharis, M.; Coppins, M.; Counsell, G.F.; Allen, J.E.; Counsell, G.F.

2008-01-01

The DTOKS code, which models dust transport through tokamak plasmas, is described. The floating potential and charge of a dust grain in a plasma and the fluxes of energy to and from it are calculated. From this model, the temperature of the dust grain can be estimated. A plasma background is supplied by a standard tokamak edge modelling code (B2SOLPS5.0), and dust transport through MAST (the Mega-Amp Spherical Tokamak) and ITER plasmas is presented. We conclude that micron-radius tungsten dust can reach the separatrix in ITER. (authors)

Compact tokamak reactors

International Nuclear Information System (INIS)

Wootton, A.J.; Wiley, J.C.; Edmonds, P.H.; Ross, D.W.

1997-01-01

The possible use of tokamaks for thermonuclear power plants is discussed, in particular tokamaks with low aspect ratio and copper toroidal field coils. Three approaches are presented. First, the existing literature is reviewed and summarized. Second, using simple analytic estimates, the size of the smallest tokamak to produce an ignited plasma is derived. This steady state energy balance analysis is then extended to determine the smallest tokamaks power plant, by including the power required to drive the toroidal field and by considering two extremes of plasma current drive efficiency. Third, the analytic results are augmented by a numerical calculation that permits arbitrary plasma current drive efficiency and different confinement scaling relationships. Throughout, the importance of various restrictions is emphasized, in particular plasma current drive efficiency, plasma confinement, plasma safety factor, plasma elongation, plasma beta, neutron wall loading, blanket availability and recirculation of electric power. The latest published reactor studies show little advantage in using low aspect ratios to obtain a more compact device (and a low cost of electricity) unless either remarkably high efficiency plasma current drive and low safety factor are combined, or unless confinement (the H factor), the permissible elongation and the permissible neutron wall loading increase as the aspect ratio is reduced. These results are reproduced with the analytic model. (author). 22 refs, 3 figs

Magnetic confinement experiment -- 1: Tokamaks

International Nuclear Information System (INIS)

Goldston, R.J.

1994-01-01

This report reviews presentations made at the 15th IAEA Conference on Plasma Physics and Controlled Nuclear Fusion on experimental tokamak physics, particularly on advances in core plasma physics, divertor and edge physics, heating and current drive, and tokamak concept optimization

Disruption prediction with adaptive neural networks for ASDEX Upgrade

International Nuclear Information System (INIS)

Cannas, B.; Fanni, A.; Pautasso, G.; Sias, G.

2011-01-01

In this paper, an adaptive neural system has been built to predict the risk of disruption at ASDEX Upgrade. The system contains a Self Organizing Map, which determines the 'novelty' of the input of a Multi Layer Perceptron predictor module. The answer of the MLP predictor will be inhibited whenever a novel sample is detected. Furthermore, it is possible that the predictor produces a wrong answer although it is fed with known samples. In this case, a retraining procedure will be performed to update the MLP predictor in an incremental fashion using data coming from both the novelty detection, and from wrong predictions. In particular, a new update is performed whenever a missed alarm is triggered by the predictor. The performance of the adaptive predictor during the more recent experimental campaigns until November 2009 has been evaluated.

Turbulence in high-beta ASDEX upgrade advanced scenarios

Science.gov (United States)

Doerk, H.; Bock, A.; Di Siena, A.; Fable, E.; Görler, T.; Jenko, F.; Stober, J.; The ASDEX Upgrade Team

2018-01-01

Recent experiments at ASDEX Upgrade achieve non-inductive operation in full tungsten wall conditions by applying electron cyclotron and neutral beam current drive. These discharges are characterised by a well-measured safety factor profile, which does not drop below one, and a good energy confinement. By reproducing the experimental heat fluxes, nonlinear gyrokinetic simulations suggest that the observed strong peaking of the ion temperature in the core is caused by the stabilising impact of a significant beam ion content, as well as strong electromagnetic effects on turbulent transport. Quasilinear transport models are not yet applicable in this interesting and reactor relevant parameter regime, but available simulation data may serve as a testbed for improvements. As the present plasma is close to the kinetic ballooning (KBM) threshold, elevating the safety factor profile under otherwise identical conditions is proposed to clarify, whether profiles are ultimately limited by KBM turbulence, or by global stability constraints.

Real time capable infrared thermography for ASDEX Upgrade

Energy Technology Data Exchange (ETDEWEB)

Sieglin, B., E-mail: Bernhard.Sieglin@ipp.mpg.de; Faitsch, M.; Herrmann, A.; Brucker, B.; Eich, T.; Kammerloher, L.; Martinov, S. [Max-Planck Institute for Plasma Physics, Boltzmannstr. 2, D-85748 Garching (Germany)

2015-11-15

Infrared (IR) thermography is widely used in fusion research to study power exhaust and incident heat load onto the plasma facing components. Due to the short pulse duration of today’s fusion experiments, IR systems have mostly been designed for off-line data analysis. For future long pulse devices (e.g., Wendelstein 7-X, ITER), a real time evaluation of the target temperature and heat flux is mandatory. This paper shows the development of a real time capable IR system for ASDEX Upgrade. A compact IR camera has been designed incorporating the necessary magnetic and electric shielding for the detector, cooler assembly. The camera communication is based on the Camera Link industry standard. The data acquisition hardware is based on National Instruments hardware, consisting of a PXIe chassis inside and a fibre optical connected industry computer outside the torus hall. Image processing and data evaluation are performed using real time LabVIEW.

Microwave Tokamak Experiment: Overview and status

International Nuclear Information System (INIS)

1990-05-01

The Microwave Tokamak Experiment, now under construction at the Laboratory, will use microwave heating from a free-electron laser. The intense microwave pulses will be injected into the tokamak to realize several goals, including a demonstration of the effects of localized heat deposition within magnetically confined plasma, a better understanding of energy confinement in tokamaks, and use of the new free-electron laser technology for plasma heating. 3 figs., 3 tabs

Advanced statistics for tokamak transport colinearity and tokamak to tokamak variation

International Nuclear Information System (INIS)

Riedel, K.S.

1989-01-01

This paper is an expository introduction to advanced statistics and scaling laws and their application to tokamak devices. Topics of discussion are as follows: implicit assumptions in the standard analysis; advanced regression techniques; specialized tools in statistics and their applications in fusion physics; and improved datasets for transport studies

Magnetic ''islandography'' in tokamaks

International Nuclear Information System (INIS)

Callen, J.D.; Waddell, B.V.; Hicks, H.R.

1978-09-01

Tearing modes are shown to be responsible for most of the experimentally observed macroscopic behavior of tokamak discharges. The effects of these collective magnetic perturbations on magnetic topology and plasma transport in tokamaks are shown to provide plausible explanations for: internal disruptions (m/n = 1); Mirnov oscillations (m/n = 2,3...); and major disruptions (coupling of 2/1-3/2 modes). The nonlinear evolution of the tearing modes is followed with fully three-dimensional computer codes. The effects on plasma confinement of the magnetic islands or stochastic field lines induced by the macroscopic tearing modes are discussed and compared with experiment. Finally, microscopic magnetic perturbations are shown to provide a natural model for the microscopic anomalous transport processes in tokamaks

Dimensionally similar studies of confinement and H-mode transition in ASDEX Upgrade and JET

International Nuclear Information System (INIS)

Ryter, F.; Stober, J.; Suttrop, W.

2001-01-01

Joint experiments on confinement and L-H transition were performed in ASDEX Upgrade and JET. The confinement experiments suggest that the invariance principle is not always fulfilled at high density. For the L-H transition studies, the dimensionless variables taken at the plasma edge can be in general only made identical per pair, due to the condition imposed by the L-H transition. This new approach to investigate the L-H physics suggests a weak dependence of the L-H transition mechanism on collisionality. (author)

Dimensionally similar studies of confinement and H-mode transition in ASDEX Upgrade and JET

International Nuclear Information System (INIS)

Ryter, F.; Stober, J.; Suttrop, W.

1999-01-01

Joint experiments on confinement and L-H transition were performed in ASDEX Upgrade and JET. The confinement experiments suggest that the invariance principle is not always fulfilled at high density. For the L-H transition studies, the dimensionless variables taken at the plasma edge can be in general only made identical per pair, due to the condition imposed by the L-H transition. This new approach to investigate the L-H physics suggests a weak dependence of the L-H transition mechanism on collisionality. (author)

Current drive by spheromak injection into a tokamak

International Nuclear Information System (INIS)

Brown, M.R.; Bellan, P.M.

1990-01-01

The authors report the first observation of current drive by spheromak injection into a tokamak due to the process of helicity injection. Current drive is observed in Caltech's ENCORE tokamak (30% increase, ΔI > 1 kA) only when both the tokamak and injected spheromak have the same sign of helicity (where helicity is defined as positive if current flows parallel to magnetic field lines and negative if anti-parallel). The initial increase (decrease) in current is accompanied by a sharp decrease (increase) in loop voltage and the increase in tokamak helicity is consistent with the helicity content of the injected spheromak. In addition, the injection of the spheromak raises the tokamak central density by a factor of six. The introduction of cold spheromak plasma causes sudden cooling of the tokamak discharge from 12 eV to 4 eV which results in a gradual decline in tokamak plasma current by a factor of three. In a second experiment, the authors inject spheromaks into the magnetized toroidal vacuum vessel (with no tokamak plasma). An m = 1 magnetic structure forms in the vessel after the spheromak undergoes a double tilt; once in the cylindrical entrance between gun and tokamak, then again in the tokamak vessel. A horizontal shift of the spheromak equilibrium is observed in the direction opposite that of the static toroidal field. In the absence of net toroidal flux, the structure develops a helical pitch as predicted by theory. Experiments with a number of refractory metal coatings have shown that tungsten and chrome coatings provide some improvement in spheromak parameters. They have also designed and will soon construct a larger, higher current spheromak gun with a new accelerator section for injection experiments on the Phaedrus-T tokamak

A need for non-tokamak approaches to magnetic fusion energy

International Nuclear Information System (INIS)

Bathke, C.G.; Krakowski, R.A.; Miller, R.L.

1992-01-01

Focusing exclusively on conventional tokamak physics in the quest for commercial fusion power is premature, and the options for both advanced-tokamak and non-tokamak concepts need continued investigation. The basis for this claim is developed, and promising advanced-tokamak and non-tokamak options are suggested

Development of laser induced breakdown spectroscopy for studying erosion, deposition, and fuel retention in ASDEX Upgrade

Energy Technology Data Exchange (ETDEWEB)

Paris, Peeter; Piip, Kaarel [Institute of Physics, University of Tartu, Tartu (Estonia); Hakola, Antti [VTT Technical Research Centre of Finland, Espoo (Finland); Laan, Matti, E-mail: matti.laan@ut.ee [Institute of Physics, University of Tartu, Tartu (Estonia); Aints, Märt [Institute of Physics, University of Tartu, Tartu (Estonia); Koivuranta, Seppo; Likonen, Jari [VTT Technical Research Centre of Finland, Espoo (Finland); Lissovski, Aleksandr [Institute of Physics, University of Tartu, Tartu (Estonia); Mayer, Matej [Max-Planck-Institut für Plasmaphysik, Garching (Germany); Neu, Rudolf [Max-Planck-Institut für Plasmaphysik, Garching (Germany); Technische Universität München, Fachgebt Plasma-Material-Wechelwirkung, Garching (Germany); Rohde, Volker; Sugiyama, Kazuyoshi [Max-Planck-Institut für Plasmaphysik, Garching (Germany)

2015-10-15

Highlights: • LIBS development for in situ monitoring of first walls of fusion reactors. • Testing of samples extracted from the divertor tiles of ASDEX Upgrade. • Reliable detection of deuterium depth profiles. • A method of LIBS data processing which allows to find the elemental depth profiles. • Comparison of LIBS results with those of other surface characterization methods. - Abstract: The paper deals with the development of laser induced breakdown spectroscopy (LIBS) into an in situ method for studying erosion/deposition processes at the first walls of fusion reactors. To this end, samples extracted from the divertor tiles of ASDEX Upgrade after the 2009 plasma operations were analyzed using LIBS for their composition and the results were compared with other post mortem deposition data. Quantitative depth profiles for the elemental concentrations were extracted from LIBS spectra by applying a novel data processing method. In addition, both multiline and multispot averaging procedures were applied to reduce fluctuations in the data. The LIBS concentration profiles matched qualitatively with those given by secondary ion mass spectrometry and quantitatively with the ion-beam data. The deuterium content of the samples could be reliably determined if the surface densities were >10{sup 17} at/cm{sup 2}.

Submillimeter wave propagation in tokamak plasmas

International Nuclear Information System (INIS)

Ma, C.H.; Hutchinson, D.P.; Staats, P.A.; Vander Sluis, K.L.; Mansfield, D.K.; Park, H.; Johnson, L.C.

1985-01-01

The propagation of submillimeter-waves (smm) in tokamak plasmas has been investigated both theoretically and experimentally to ensure successful measurements of electron density and plasma current distributions in tokamak devices. Theoretical analyses have been carried out to study the polarization of the smm waves in TFTR and ISX-B tokamaks. A multichord smm wave interferometer/polarimeter system has been employed to simultaneously measure the line electron density and poloidal field-induced Faraday rotation in the ISX-B tokamak. The experimental study on TFTR is under way. Computer codes have been developed and have been used to study the wave propagation and to reconstruct the distributions of plasma current and density from the measured data. The results are compared with other measurements

Submillimeter wave propagation in tokamak plasmas

International Nuclear Information System (INIS)

Ma, C.H.; Hutchinson, D.P.; Staats, P.A.; Vander Sluis, K.L.; Mansfield, D.K.; Park, H.; Johnson, L.C.

1986-01-01

Propagation of submillimeter waves (smm) in tokamak plasma was investigated both theoretically and experimentally to ensure successful measurements of electron density and plasma current distributions in tokamak devices. Theoretical analyses were carried out to study the polarization of the smm waves in TFTR and ISX-B tokamaks. A multichord smm wave interferometer/polarimeter system was employed to simultaneously measure the line electron density and poloidal field-induced Faraday rotation in the ISX-B tokamak. The experimental study on TFTR is under way. Computer codes were developed and have been used to study the wave propagation and to reconstruct the distributions of plasma current and density from the measured data. The results are compared with other measurements. 5 references, 2 figures

The role of the spherical tokamak in clarifying tokamak physics

International Nuclear Information System (INIS)

Morris, A.W.; Akers, R.J.; Connor, J.W.; Counsell, G.F.; Gryaznevich, M.P.; Hender, T.C.; Maddison, G.P.; Martin, T.J.; McClements, K.G.; Roach, C.M.; Robinson, D.C.; Sykes, A.; Valovic, M.; Wilson, H.R.; Fonck, R.J.; Gusev, V.; Kaye, S.M.; Majeski, R.; Peng, Y.-K.M.; Medvedev, S.; Sharapov, S.; Walsh, M.J.

1999-01-01

The spherical tokamak (ST) provides a unique environment in which to perform complementary and exacting tests of the tokamak physics required for a burning plasma experiment of any aspect ratio, while also having the potential for long-term fusion applications in its own right. New experiments are coming on-line in the UK (MAST), USA (NSTX, Pegasus), Russia (Globus-M), Brazil (ETE) and elsewhere, and the status of these devices will be reported, along with newly-analysed data from START. Those physics issues where the ST provides an opportunity to remove degeneracy in the databases or clarify one's understanding will be emphasized. (author)

Real-time feedback control of the plasma density profile on ASDEX Upgrade

International Nuclear Information System (INIS)

Mlynek, A.; Reich, M.; Giannone, L.; Treutterer, W.; Behler, K.; Blank, H.; Buhler, A.; Cole, R.; Eixenberger, H.; Fischer, R.; Lohs, A.; Lueddecke, K.; Merkel, R.; Neu, G.; Ryter, F.; Zasche, D.

2011-01-01

The spatial distribution of density in a fusion experiment is of significant importance as it enters in numerous analyses and contributes to the fusion performance. The reconstruction of the density profile is therefore commonly done in offline data analysis. In this paper, we present an algorithm which allows for density profile reconstruction from the data of the submillimetre interferometer and the magnetic equilibrium in real-time. We compare the obtained results to the profiles yielded by a numerically more complex offline algorithm. Furthermore, we present recent ASDEX Upgrade experiments in which we used the real-time density profile for active feedback control of the shape of the density profile.

Concepts for improving the accuracy of gas balance measurement at ASDEX Upgrade

Energy Technology Data Exchange (ETDEWEB)

Härtl, T., E-mail: thomas.haertl@ipp.mpg.de; Rohde, V.; Mertens, V.

2013-10-15

The ITER fusion reactor which is under construction will use a deuterium–tritium gas mixture for operation. A fraction of this fusion fuel remains inside of the machine due to various mechanisms. The evaluation of this retention in present fusion experiments is of crucial importance to estimate the expected tritium inventory in ITER which shall be limited due to safety considerations. At ASDEX Upgrade (AUG) sufficiently time-resolved measurements should take place to extrapolate from current 10 s discharges to the at least intended 400 s ones of ITER. To achieve this, a new measurement system has been designed that enables accuracy of better than one per cent.

Chapter 7: High-Density H-Mode Operation in ASDEX Upgrade

International Nuclear Information System (INIS)

Stober, Joerg Karl; Lang, Peter Thomas; Mertens, Vitus

2003-01-01

Recent results are reported on the maximum achievable H-mode density and the behavior of pedestal density and central density peaking as this limit is approached. The maximum achievable H-mode density roughly scales as the Greenwald density, though a dependence on B t is clearly observed. In contrast to the stiff temperature profiles, the density profiles seem to allow more shape variation and especially with high-field-side pellet-injection, strongly peaked profiles with good confinement have been achieved. Also, spontaneous density peaking at high densities is observed in ASDEX Upgrade, which is related to the generally observed large time constants for the density profile equilibration. The equilibrated density profile shapes depend strongly on the heat-flux profile in the sense that central heating leads to significantly flatter profiles

Advanced tokamak burning plasma experiment

International Nuclear Information System (INIS)

Porkolab, M.; Bonoli, P.T.; Ramos, J.; Schultz, J.; Nevins, W.N.

2001-01-01

A new reduced size ITER-RC superconducting tokamak concept is proposed with the goals of studying burn physics either in an inductively driven standard tokamak (ST) mode of operation, or in a quasi-steady state advanced tokamak (AT) mode sustained by non-inductive means. This is achieved by reducing the radiation shield thickness protecting the superconducting magnet by 0.34 m relative to ITER and limiting the burn mode of operation to pulse lengths as allowed by the TF coil warming up to the current sharing temperature. High gain (Q≅10) burn physics studies in a reversed shear equilibrium, sustained by RF and NB current drive techniques, may be obtained. (author)

Computational studies of tokamak plasmas

International Nuclear Information System (INIS)

Takizuka, Tomonori; Tsunematsu, Toshihide; Tokuda, Shinji

1981-02-01

Computational studies of tokamak plasmas are extensively advanced. Many computational codes have been developed by using several kinds of models, i.e., the finite element formulation of MHD equations, the time dependent multidimensional fluid model, and the particle model with the Monte-Carlo method. These codes are applied to the analyses of the equilibrium of an axisymmetric toroidal plasma (SELENE), the time evolution of the high-beta tokamak plasma (APOLLO), the low-n MHD stability (ERATO-J) and high-n ballooning mode stability (BOREAS) in the INTOR tokamak, the nonlinear MHD stability, such as the positional instability (AEOLUS-P), resistive internal mode (AEOLUS-I) etc., and the divertor functions. (author)

Combined confinement system applied to tokamaks

International Nuclear Information System (INIS)

Ohkawa, Tihiro

1986-01-01

From particle orbit point of view, a tokamak is a combined confinement configuration where a closed toroidal volume is surrounded by an open confinement system like a magnetic mirror. By eliminating a cold halo plasma, the energy loss from the plasma becomes convective. The H-mode in diverted tokamaks is an example. Because of the favorable scaling of the energy confinement time with temperature, the performance of the tokamak may be significantly improved by taking advantage of this effect. (author)

Commissioning of inline ECE system within waveguide based ECRH transmission systems on ASDEX upgrade

Directory of Open Access Journals (Sweden)

Donné A.J.H.

2012-09-01

Full Text Available A CW capable inline electron cyclotron emission (ECE separation system for feedback control, featuring oversized corrugated waveguides, is commissioned on ASDEX upgrade (AUG. The system is based on a combination of a polarization independent, non-resonant, Mach-Zehnder diplexer equipped with dielectric plate beam splitters [2, 3] employed as corrugated oversized waveguide filter, and a resonant Fast Directional Switch, FADIS [4, 5, 6, 7] as ECE/ECCD separation system. This paper presents an overview of the system, the low power characterisation tests and first high power commissioning on AUG.

ASDEX contributions to the 17th European conference on controlled fusion and plasma heating

International Nuclear Information System (INIS)

1990-09-01

The 'ASDEX contributions to the 17th European conference on controlled fusion and plasma heating' (Amsterdam, June 25-29, 1990) hold one invited paper (Physics of enhanced confinement with peaked and board density profiles) and 12 chapters containing 44 contributed papers dealing with the following topics: Lower hybrid current drive and heating; Ion cyclotron heating; General confinement studies; Fluctuation studies; Direct measurement of transport coefficients; H-mode studies; Pellet studies; Divertor and SOL-studies; Impurity and impurity transport studies; Density limit studies; MHD studies; Diagnostic development. (orig./AH)

Commissioning of inline ECE system within waveguide based ECRH transmission systems on ASDEX upgrade

Science.gov (United States)

Bongers, W. A.; Kasparek, W.; Doelman, N.; van den Braber, R.; van den Brand, H.; Meo, F.; de Baar, M. R.; Amerongen, F. J.; Donné, A. J. H.; Elzendoorn, B. S. Q.; Erckmann, V.; Goede, A. P. H.; Giannone, L.; Grünwald, G.; Hollman, F.; Kaas, G.; Krijger, B.; Michel, G.; Lubyako, L.; Monaco, F.; Noke, F.; Petelin, M.; Plaum, B.; Purps, F.; ten Pierik, J. G. W.; Schüller, C.; Slob, J. W.; Stober, J. K.; Schütz, H.; Wagner, D.; Westerhof, E.; Ronden, D. M. S.

2012-09-01

A CW capable inline electron cyclotron emission (ECE) separation system for feedback control, featuring oversized corrugated waveguides, is commissioned on ASDEX upgrade (AUG). The system is based on a combination of a polarization independent, non-resonant, Mach-Zehnder diplexer equipped with dielectric plate beam splitters [2, 3] employed as corrugated oversized waveguide filter, and a resonant Fast Directional Switch, FADIS [4, 5, 6, 7] as ECE/ECCD separation system. This paper presents an overview of the system, the low power characterisation tests and first high power commissioning on AUG.

Research into controlled fusion in tokamaks

International Nuclear Information System (INIS)

Zacek, F.

1992-01-01

During the thirty years of tokamak research, physicists have been approaching step by step the reactor breakeven condition defined by the Lawson criterion. JET, the European Community tokamak is probably the first candidate among the world largest tokamaks to reach the ignition threshold and thus to demonstrate the physical feasibility of thermonuclear reaction. The record plasma parameters achieved in JET at H plasma modes due to powerful additional plasma heating and due to substantial reduction of plasma impurities, opened the door to the first experiment with a deuterium-tritium plasma. In the paper, the conditions and results of these tritium experiments are described in detail. The prospects of the world tokamak research and of the participation of Czechoslovak physicists are also discussed. (J.U.) 3 figs., 6 refs

The density limit in Tokamaks

International Nuclear Information System (INIS)

Alladio, F.

1985-01-01

A short summary of the present status of experimental observations, theoretical ideas and understanding of the density limit in tokamaks is presented. It is the result of the discussion that was held on this topic at the 4th European Tokamak Workshop in Copenhagen (December 4th to 6th, 1985). 610 refs

Interpretation of low ionized impurity distributions in the ASDEX Upgrade divertor

International Nuclear Information System (INIS)

Lieder, G.; Napiontek, B.; Radtke, R.; Field, A.; Fussmann, G.; Kallenbach, A.; Kiemer, K.; Mayer, H.M.

1993-01-01

Design studies for reactor-like devices, like ITER, have particularly emphasized the importance of erosion and transport of material from the divertor target plates. In this context experimental measurements which can lead to a better understanding of the underlying physics are highly desirable. We discuss the spatial profiles of line emission from impurities measured in the divertor of ASDEX Upgrade with a recently developed multi-chord divertor spectrometer system. These profiles are obtained from observations in the ultra-violet/visible spectral range. The divertor spectrometer system was developed particularly to measure the erosion of the divertor plates and to study transport of the impurities and the ionization and recombination processes in the divertor region. (author) 6 refs., 3 figs., 2 tabs

Interpretation of low ionized impurity distributions in the ASDEX Upgrade divertor

Energy Technology Data Exchange (ETDEWEB)

Lieder, G; Napiontek, B; Radtke, R; Field, A; Fussmann, G; Kallenbach, A; Kiemer, K; Mayer, H M [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

1994-12-31

Design studies for reactor-like devices, like ITER, have particularly emphasized the importance of erosion and transport of material from the divertor target plates. In this context experimental measurements which can lead to a better understanding of the underlying physics are highly desirable. We discuss the spatial profiles of line emission from impurities measured in the divertor of ASDEX Upgrade with a recently developed multi-chord divertor spectrometer system. These profiles are obtained from observations in the ultra-violet/visible spectral range. The divertor spectrometer system was developed particularly to measure the erosion of the divertor plates and to study transport of the impurities and the ionization and recombination processes in the divertor region. (author) 6 refs., 3 figs., 2 tabs.

Development of Operation Scenario for Spherical Tokamak at SNU

International Nuclear Information System (INIS)

Sung, C. K.; Park, Y. S.; Lee, H. Y.; Kang, J.; Hwang, Y. S.

2009-01-01

Several concepts for nuclear fusion plant exist. In these concepts, tokamak is the most promising one to realize nuclear fusion plant. Though tokamak has leading concept, and this has world record in fusion heating power, tokamak has the critical drawback: low heating efficiency. That is the reason why we need another alternative concept which compensates tokamak's disadvantage. Spherical Torus(ST) is one of these kinds of concepts. ST is a kind of tokamak which has low aspect ratio. This feature gives ST advantages compared to conventional tokamak: high efficiency, compactness, low cost. However, ST lacks central region for solenoid that is needed to start-up and sustain. Since it is the most efficient that initializing and sustaining by using solenoid, this is ST's intrinsic limitation. To overcome this, a new device which can start-up and sustain ST plasmas by means of continuous tokamak plasma injection has been designed

Economic considerations of commercial tokamak options

International Nuclear Information System (INIS)

Dabiri, A.E.

1986-05-01

Systems studies have been performed to assess commercial tokamak options. Superconducting, as well as normal, magnet coils in either first or second stability regimes have been considered. A spherical torus (ST), as well as an elongated tokamak (ET), is included in the study. The cost of electricity (COE) is selected as the figure of merit, and beta and first-wall neutron wall loads are selected to represent the physics and technology characteristics of various options. The results indicate that an economical optimum for tokamaks is predicted to require a beta of around 10%, as predicted to be achieved in the second stability regime, and a wall load of about 5 MW/m 2 , which is assumed to be optimum technologically. This tokamak is expected to be competitive with fission plants if efficient, noninductive current drive is developed. However, if this regime cannot be attained, all other tokamaks operating in the first stability regime, including spherical torus and elongated tokamak and assuming a limiting wall load of 5 MW/m 2 , will compete with one another with a COE of about 50 mill/kWh. This 40% higher than the COE for the optimum reactor in the second stability regime with fast-wave current drive. The above conclusions pertain to a 1200-MW(e) net electric power plant. A comparison was also made between ST, ET, and superconducting magnets in the second stability regime with fast-wave current drive at 600 MW(e)

Large aspect ratio tokamak study

International Nuclear Information System (INIS)

Reid, R.L.; Holmes, J.A.; Houlberg, W.A.; Peng, Y.K.M.; Strickler, D.J.; Brown, T.G.; Sardella, C.; Wiseman, G.W.

1979-01-01

The Large Aspect Ratio Tokamak Study (LARTS) investigated the potential for producing a viable long burn tokamak reactor through enhanced volt-second capability of the ohmic heating transformer by employing high aspect ratio designs. The plasma physics, engineering, and economic implications of high aspect ratio tokamaks were accessed in the context of extended burn operation. Plasma startup and burn parameters were addressed using a one-dimensional transport code. The pulsed electrical power requirements for the poloidal field system, which have a major impact on reactor economics, were minimized by optimizing the field in the ohmic heating coil and the wave shape of the ohmic heating discharge. A high aspect ratio reference reactor was chosen and configured

Plasma boundary phenomena in tokamaks

International Nuclear Information System (INIS)

Stangeby, P.C.

1989-06-01

The focus of this review is on processes occurring at the edge, and on the connection between boundary plasma - the scrape-off layer (SOL) and the radiating layer - and central plasma processes. Techniques used for edge diagnosis are reviewed and basic experimental information (n e and T e ) is summarized. Simple models of the SOL are summarized, and the most important effects of the boundary plasma - the influence on the fuel particles, impurities, and energy - on tokamak operation dealt with. Methods of manipulating and controlling edge conditions in tokamaks and the experimental data base for the edge during auxiliary heating of tokamaks are reviewed. Fluctuations and asymmetries at the edge are also covered. (9 tabs., 134 figs., 879 refs.)

Nuclear fusion research at Tokamak Energy Ltd

International Nuclear Information System (INIS)

Windridge, Melanie J.; Gryaznevich, Mikhail; Kingham, David

2017-01-01

Tokamak Energy's approach is close to the mainstream of nuclear fusion, and chooses a spherical tokamak, which is an economically developed form of Tokamak reactor design, as research subjects together with a high-temperature superconducting magnet. In the theoretical prediction, it is said that spherical tokamak can make tokamak reactor's scale compact compared with ITER or DEMO. The dependence of fusion energy multiplication factor on reactor size is small. According to model studies, it has been found that the center coil can be protected from heat and radiation damage even if the neutron shielding is optimized to 35 cm instead of 1 m. As a small tokamak with a high-temperature superconducting magnet, ST25 HTS, it demonstrated in 2015 continuous operation for more than 24 hours as a world record. Currently, this company is constructing a slightly larger ST40 type, and it is scheduled to start operation in 2017. ST40 is designed to demonstrate that it can realize a high magnetic field with a compact size and aims at attaining 8-10 keV (reaching the nuclear fusion reaction temperature at about 100 million degrees). This company will verify the startup and heating technology by the coalescence of spherical tokamak expected to have plasma current of 2 MA, and will also use 2 MW of neutral particle beam heating. In parallel with ST40, it is promoting a development program for high-temperature superconducting magnet. (A.O.)

An improved routine for the fast estimate of ion cyclotron heating efficiency in tokamak plasmas

International Nuclear Information System (INIS)

Brambilla, M.

1992-02-01

The subroutine ICEVAL for the rapid simulation of Ion Cyclotron Heating in tokamak plasmas is based on analytic estimates of the wave behaviour near resonances, and on drastic but reasonable simplifications of the real geometry. The subroutine has been rewritten to improve the model and to facilitate its use as input in transport codes. In the new version the influence of quasilinear minority heating on the damping efficiency is taken into account using the well-known Stix analytic approximation. Among other improvements are: a) the possibility of considering plasmas with more than two ion species; b) inclusion of Landau, Transit Time and collisional damping on the electrons non localised at resonances; c) better models for the antenna spectrum and for the construction of the power deposition profiles. The results of ICEVAL are compared in detail with those of the full-wave code FELICE for the case of Hydrogen minority heating in a Deuterium plasma; except for details which depend on the excitation of global eigenmodes, agreement is excellent. ICEVAL is also used to investigate the enhancement of the absorption efficiency due to quasilinear heating of the minority ions. The effect is a strongly non-linear function of the available power, and decreases rapidly with increasing concentration. For parameters typical of Asdex Upgrade plasmas, about 4 MW are required to produce a significant increase of the single-pass absorption at concentrations between 10 and 20%. (orig.)

Advances in the density profile evaluation from broadband reflectometry on ASDEX upgrade

International Nuclear Information System (INIS)

Varela, P.; Manso, M.; Conway, G.

2001-01-01

The high temporal and spatial resolutions provided by broadband microwave reflectometry make it an attractive diagnostic technique to measure the density profile in fusion plasmas. However, great problems have been encountered due to the plasma turbulence that difficult, and sometimes prevent, the routine evaluation of density profiles. Advanced broadband systems employ ultra-fast sweeping in an attempt to perform the profile measurement in a time window smaller than the temporal scale of the main plasma fluctuations but this is not sufficient. Indeed, abrupt plasma movements and/or spatial turbulence always affect the reflectometry signals, as shown by numerical studies (with both one- and two-dimensional codes), for the case of ultra-fast sweeping and pulse radar systems. For this reason not only the system performance is important but the software tools also play a crucial role for reflectometry to become a standard density profile diagnostic. Here we present the recent advances towards automatic evaluation of density profiles from broadband reflectometry on ASDEX Upgrade. For regimes with moderate levels of plasma turbulence, density profiles are obtained from single reflectometry samples (temporal resolution of 20 μs), and for higher turbulence levels average profiles are obtained from bursts of ultra-fast (20 μs), closely spaced (10 μs) sweeps. This method improved the accuracy and reliability of density profiles, which can now be obtained automatically from the edge to the bulk plasma - using reflectometry alone - in most plasma regimes of ASDEX Upgrade. New data processing capability has been implemented that allows the profiles to be available to the end-users 10-12 minutes after each discharge. These developments were possible due to the flexibility and high performance of the control and data acquisition systems and to the large number of measurements that can be performed with the diagnostic during each discharge (720 profiles both on the low- and

Tokamak simulation code manual

International Nuclear Information System (INIS)

Chung, Moon Kyoo; Oh, Byung Hoon; Hong, Bong Keun; Lee, Kwang Won

1995-01-01

The method to use TSC (Tokamak Simulation Code) developed by Princeton plasma physics laboratory is illustrated. In KT-2 tokamak, time dependent simulation of axisymmetric toroidal plasma and vertical stability have to be taken into account in design phase using TSC. In this report physical modelling of TSC are described and examples of application in JAERI and SERI are illustrated, which will be useful when TSC is installed KAERI computer system. (Author) 15 refs., 6 figs., 3 tabs

Mercier criterion for high-β tokamaks

International Nuclear Information System (INIS)

Galvao, R.M.O.

1984-01-01

An expression, for the application of the Mercier criterion to numerical studies of diffuse high-β tokamaks (β approximatelly Σ,q approximatelly 1), which contains only leading order contributions in the high-β tokamak approximation is derived. (L.C.) [pt

Replacement strategy for ASDEX upgrade's new control and data acquisition

International Nuclear Information System (INIS)

Raupp, G.; Behler, K.; Cole, R.; Engelhardt, K.; Lohs, A.; Lueddecke, K.; Neu, G.; Treutterer, W.; Vijverberg, Th.; Zasche, D.; Zehetbauer, Th.

2004-01-01

ASDEX Upgrade is being equipped with a new real-time plasma control and data acquisition system and a novel time system. Major components were implemented and installed. While much work for performance optimisation and application programming remains to be done, commissioning of the new system parallel to experiment operation is being prepared. Commissioning of the new system will be done step-by-step. To facilitate testing the old and new control systems share all input signals. Switching between old and new system can be performed within 60 min: 23 fibre optics for output of actuator commands and input triggers must be connected to the active system and minor modifications done to interface the machine protection. Commissioning phases include background listening, technical discharges and full plasma operation. With the strategy chosen we minimize risk to the machine and reduce interference with ongoing experiment campaigns

Measurement of neoclassically predicted edge current density at ASDEX Upgrade

Science.gov (United States)

Dunne, M. G.; McCarthy, P. J.; Wolfrum, E.; Fischer, R.; Giannone, L.; Burckhart, A.; the ASDEX Upgrade Team

2012-12-01

Experimental confirmation of neoclassically predicted edge current density in an ELMy H-mode plasma is presented. Current density analysis using the CLISTE equilibrium code is outlined and the rationale for accuracy of the reconstructions is explained. Sample profiles and time traces from analysis of data at ASDEX Upgrade are presented. A high time resolution is possible due to the use of an ELM-synchronization technique. Additionally, the flux-surface-averaged current density is calculated using a neoclassical approach. Results from these two separate methods are then compared and are found to validate the theoretical formula. Finally, several discharges are compared as part of a fuelling study, showing that the size and width of the edge current density peak at the low-field side can be explained by the electron density and temperature drives and their respective collisionality modifications.

Measurement of neoclassically predicted edge current density at ASDEX Upgrade

International Nuclear Information System (INIS)

Dunne, M.G.; McCarthy, P.J.; Wolfrum, E.; Fischer, R.; Giannone, L.; Burckhart, A.

2012-01-01

Experimental confirmation of neoclassically predicted edge current density in an ELMy H-mode plasma is presented. Current density analysis using the CLISTE equilibrium code is outlined and the rationale for accuracy of the reconstructions is explained. Sample profiles and time traces from analysis of data at ASDEX Upgrade are presented. A high time resolution is possible due to the use of an ELM-synchronization technique. Additionally, the flux-surface-averaged current density is calculated using a neoclassical approach. Results from these two separate methods are then compared and are found to validate the theoretical formula. Finally, several discharges are compared as part of a fuelling study, showing that the size and width of the edge current density peak at the low-field side can be explained by the electron density and temperature drives and their respective collisionality modifications. (paper)

Tokamak Physics Experiment (TPX) design

International Nuclear Information System (INIS)

Schmidt, J.A.

1995-01-01

TPX is a national project involving a large number of US fusion laboratories, universities, and industries. The element of the TPX requirements that is a primary driver for the hardware design is the fact that TPX tokamak hardware is being designed to accommodate steady state operation if the external systems are upgraded from the 1,000 second initial operation. TPX not only incorporates new physics, but also pioneers new technologies to be used in ITER and other future reactors. TPX will be the first tokamak with fully superconducting magnetic field coils using advanced conductors, will have internal nuclear shielding, will use robotics for machine maintenance, and will remove the continuous, concentrated heat flow from the plasma with new dispersal techniques and with special materials that are actively cooled. The Conceptual Design for TPX was completed during Fiscal Year 1993. The Preliminary Design formally began at the beginning of Fiscal Year 1994. Industrial contracts have been awarded for the design, with options for fabrication, of the primary tokamak hardware. A large fraction of the design and R and D effort during FY94 was focused on the tokamak and in turn on the tokamak magnets. The reason for this emphasis is because the magnets require a large design and R and D effort, and are critical to the project schedule. The magnet development is focused on conductor development, quench protection, and manufacturing R and D. The Preliminary Design Review for the Magnets is planned for fall, 1995

Test of the predictive capability of B2-Eirene on ASDEX-Upgrade

International Nuclear Information System (INIS)

Schneider, R.; Coster, D.P.; Kallenbach, A.

2001-01-01

Based on validated B2-Eirene results for the previous divertor of ASDEX Upgrade, the modelling predictions for the new divertor are compared with the actual experimental results. For the same experimental scenarios (L-mode) in both divertors the predictions are robust and in agreement with experimental results. For a full quantitative agreement in H-mode both the carbon chemical sputtering yield and the radial transport had to be adjusted. The new divertor has a reduced power load due to larger radiation losses. These are caused by larger hydrogen losses, enhancement of carbon radiation due to radial transport and convective energy transport into the radiation zone, and larger radial energy transport in the divertor. (author)

Steady State Advanced Tokamak (SSAT): The mission and the machine

International Nuclear Information System (INIS)

Thomassen, K.; Goldston, R.; Nevins, B.; Neilson, H.; Shannon, T.; Montgomery, B.

1992-03-01

Extending the tokamak concept to the steady state regime and pursuing advances in tokamak physics are important and complementary steps for the magnetic fusion energy program. The required transition away from inductive current drive will provide exciting opportunities for advances in tokamak physics, as well as important impetus to drive advances in fusion technology. Recognizing this, the Fusion Policy Advisory Committee and the US National Energy Strategy identified the development of steady state tokamak physics and technology, and improvements in the tokamak concept, as vital elements in the magnetic fusion energy development plan. Both called for the construction of a steady state tokamak facility to address these plan elements. Advances in physics that produce better confinement and higher pressure limits are required for a similar unit size reactor. Regimes with largely self-driven plasma current are required to permit a steady-state tokamak reactor with acceptable recirculating power. Reliable techniques of disruption control will be needed to achieve the availability goals of an economic reactor. Thus the central role of this new tokamak facility is to point the way to a more attractive demonstration reactor (DEMO) than the present data base would support. To meet the challenges, we propose a new ''Steady State Advanced Tokamak'' (SSAT) facility that would develop and demonstrate optimized steady state tokamak operating mode. While other tokamaks in the world program employ superconducting toroidal field coils, SSAT would be the first major tokamak to operate with a fully superconducting coil set in the elongated, divertor geometry planned for ITER and DEMO

Heavy ion transport in the core of ASDEX upgrade

Energy Technology Data Exchange (ETDEWEB)

Odstrcil, Tomas [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstrasse 2, D-85748 Garching (Germany); Physik-Department E28, Technische Universitaet Muenchen, 85747 Garching (Germany); Puetterich, Thomas; Angioni, Clemente; Bilato, Roberto; Gude, Anja; Vezinet, Didier [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstrasse 2, D-85748 Garching (Germany); Mazon, Didier [CEA, IRFM F-13108 Saint Paul-lez-Durance (France); Collaboration: ASDEX Upgrade Team

2016-07-01

High impurity concentration in the core of the future fusion reactors can lead to the serious degradation of the achievable fusion gain. Therefore, a better understanding of the underlying impurity transport processes is necessary for higher performance, more efficient power exhaust and avoidance of impurity accumulation. Radial impurity transport is mainly driven by neoclassical and turbulent particle fluxes. Both these components show substantial variation depending on the poloidal angle. Consequently, an asymmetry in the poloidal distribution of impurities leads to significant changes in the radial impurity flow and the total content of the plasma core. The aim of this contribution is to experimentally verify a model describing the poloidal asymmetry of heavy impurities using measurements from ASDEX Upgrade. The observed asymmetries are caused mainly by the centrifugal force and poloidal electric force created by the fast particles produced by intensive ion-cyclotron heating. Finally, a change in the radial transport of the tungsten ions will be presented in the case of large inboard and outboard impurity accumulation.

A new thermal He-beam diagnostic for electron density and temperature measurements in the scrape-off layer of ASDEX Upgrade

Energy Technology Data Exchange (ETDEWEB)

Griener, Michael; Wolfrum, Elisabeth; Eich, Thomas; Herrmann, Albrecht; Rohde, Volker [Max Planck Institute for Plasma Physics, Garching (Germany); Schmitz, Oliver [Engineering Physics Department, University of Wisconsin-Madison (United States); Stroth, Ulrich [Max Planck Institute for Plasma Physics, Garching (Germany); Physik Department E28, Technische Universitaet Muenchen, Garching (Germany); Collaboration: the ASDEX Upgrade Team

2015-05-01

In a nuclear fusion device power is exhausted across the last closed flux surface into the so-called 'scrape-off layer', SOL. In order to study the transport dynamics to (a) the divertor via parallel heat flux and (b) to the wall via filaments, a diagnostic for the determination of n{sub e} and T{sub e} with high spatial and temporal resolution is required. Although the diagnostic capabilities of the ASDEX Upgrade edge plasma are excellent, there is a lack of spatially and temporally highly resolved electron temperature measurements in the SOL. Therefore a piezo valve will be installed in ASDEX Upgrade in April 2015. It allows fast chopping of a thermal He-beam which is part of the new diagnostic. In the first campaign, existing lines of sight of the CXRS diagnostic will be used to measure various He I transitions to confirm the collisional radiative model for He. The principle of the thermal He-diagnostic as well as calculations of the achievable spatial resolution of the initial set-up are presented.

Summary of the 1982 small tokamak users meeting

International Nuclear Information System (INIS)

Sprott, J.C.

1982-11-01

On November 1, 1982, the sixth in a series of approximately annual meetings of the users of small tokamaks was held in conjunction with the APS Division of Plasma Physics meeting at New Orleans. The meeting lasted three hours, with 34 people attending. The interest was on strengthening the ties between the small tokamaks and the large tokamaks. Accordingly, the latest meeting was dedicated to this theme, and in contrast to previous meetings, a few representatives from the large tokamaks were invited to attend and make presentations. Summaries of the various talks are included

Magnetic confinement by Tokamak: physical aspects

International Nuclear Information System (INIS)

Tachon, J.

1980-01-01

After describing the Tokamak configuration concept, the author provides an analysis of the principal physical aspects of this type of installation and concludes by estimating that the Tokamak concept is a 'plausible candidate' as a means of producing controlled thermonuclear fusion [fr

Tokamak research in the Soviet Union

International Nuclear Information System (INIS)

Strelkov, V.S.

1981-01-01

Important milestones on the way to the tokamak fusion reactor are recapitulated. Soviet tokamak research concentrated at the I.V. Kurchatov Institute in Moscow, the A.F. Ioffe Institute in Leningrad and the Physical-Technical Institute in Sukhumi successfully provides necessary scientific and technological data for reactor design. Achievments include, the successful operation of the first tokamak with superconducting windings (T-7) and the gyrotron set for microwave plasma heating in the T-10 tokamak. The following problems have intensively been studied: Various methods of additional plasma heating, heat and particle transport, and impurity control. The efficiency of electron-cyclotron resonance heating was demonstrated. In the Joule heating regime, both the heat conduction and diffusion rates are anomalously high, but the electron heat conduction rate decreases with increasing plasma density. Progress in impurity control makes it possible to obtain a plasma with effective charge approaching unity. (J.U.)

Neutral beam in ALVAND IIC tokamak

International Nuclear Information System (INIS)

Ghrannevisse, M.; Moradshahi, M.; Avakian, M.

1992-01-01

Neutral beams have a wide application in tokamak experiments. It used to heat; fuel; adjust electric potentials in plasmas and diagnose particles densities and momentum distributions. It may be used to sustain currents in tokamaks to extend the pulse length. A 5 KV; 500 mA ion source has been constructed by plasma physics group, AEOI and it used to produce plasma and study the plasma parameters. Recently this ion source has been neutralized and it adapted to a neutral beam source; and it used to heat a cylindrical DC plasma and the plasma of ALVAND IIC Tokamak which is a small research tokamak with a minor radius of 12.6 cm, and a major radius of 45.5 cm. In this paper we report the neutralization of the ion beam and the results obtained by injection of this neutral beam into plasmas. (author) 2 refs., 4 figs

Multiprocessor systems for real-time data acquisition on the Asdex upgrade and future plasma experiments

International Nuclear Information System (INIS)

Zilker, M.; Hallatschek, K.; Heimann, P.; Hertweck, F.

1999-01-01

In this paper we present our transputer-based multitop multiprocessor systems for data acquisition, which are currently used on the Asdex upgrade experiment. The bandwidth of these systems goes from low-speed like the calorimetry diagnostic up to highspeed and large data volume systems like the soft-X-ray and Mirnov diagnostics, which collect several hundreds of megabytes of data during a plasma discharge of ∼8 s. Further, we present the multitop-MX, a newly developed system based on transputers and powerPCs, which provides real-time facilities for analysing the acquired data, to generate necessary information for the dynamic adaptation of sample rates, and to deliver triggers when certain events in the plasma are detected. The algorithm running on the powerPCs performs a wavelet like time-frequency transform. In the last part we give an outlook how to build the next generation of data acquisition systems to be used on the future plasma experiments W7-X and ITER, but also on Asdex upgrade. The hardware of these new distributed systems should be mainly based on established industry standards like the VME-bus, PCI-bus and FiberChannel, but also emerging technologies like SCI (scalable coherent interconnect) should be considered. The systems software should be well designed with object oriented methods to simplify the maintenance process and to enable further expansions and adaptations to new problems in an easy way. (orig.)

The particle fluxes in the edge plasma during discharges with improved ohmic confinement in ASDEX

International Nuclear Information System (INIS)

Verbeek, H.; Poschenrieder, W.; Fu, J.K.; Soeldner, F.X.

1989-01-01

In the recent experimental period of ASDEX a new regime of improved ohmic confinement (IOC) was discovered. So far the energy confinement time τ E increased linearly with increasing line averaged density n e up to n e = 3·10 13 cm -3 saturated, however, at higher densities. In the new IOC regime τ E increases further with increasing n e up to ∼5·10 13 cm -3 . The IOC regime is achieved for D 2 discharges only since the last modification of the ASDEX divertor which substantially increased the recycling from the divertor through the divertor slits. It also led to a reduction in gas consumption for a discharge by a factor of about 2. As it appears, the high fuelling rate required during a fast ramp-up of the plasma density leads to a transition into the Saturated Ohmic Confinememt (SOC) regime. Vice versa, the strong reduction in the external gas feed when the preprogrammed density plateau is reached seems to be essential for establishing the IOC. It is characterized by a pronounced peaking of the density profile. During the transition from the SOC to the IOC regime large variations in the signals of all edge and divertor related diagnostics are observed. In this paper we concentrate on the results of the Low Energy Neutral Particle Analyser (LENA), the sniffer probe, on the mass spectrometers measuring the divertor exhaust pressure. (author) 7 refs., 2 figs

Diagnostics for the Rijnhuizen Tokamak Project

NARCIS (Netherlands)

Donne, A. J. H.

1994-01-01

The research program of the Rijnhuizen Tokamak Project is concentrated on the study of plasma transport processes. The RTP tokamak is therefore equipped with an extensive set of multichannel diagnostics, including a 19-channel FIR interferometer, a 20-channel heterodyne ECE system, an 80-channel

Plasma position control in TCABR Tokamak

International Nuclear Information System (INIS)

Galvao, R.M.O.; Kuznetsov, Yu. K.; Nascimento, I.C.; Fonseca, A.M.M.; Silva, R.P. da; Ruchko, L.F.; Tuszel, A.G.; Reis, A.P. dos; Sanada, E.K.

1998-01-01

The plasma control position in the TCABR tokamak is described. The TCA tokamak was transferred from the Centre de Recherches en Physique des Plasmas, Lausanne, to the Institute of Physics of University of Sao Paulo, renamed TCABR (α=0.18 m, R = 0.62 m, B = 1 T,I p = 100 kA). The control system was reconstructed using mainly components obtained from the TCA tokamak. A new method of plasma position determination is used in TCABR to improve its accuracy. A more detailed theoretical analysis of the feed forward and feedback control is performed as compared with. (author)

Global gyrokinetic simulation of tokamak transport

International Nuclear Information System (INIS)

Furnish, G.; Horton, W.; Kishimoto, Y.; LeBrun, M.J.; Tajima, T.

1998-10-01

A kinetic simulation code based on the gyrokinetic ion dynamics in global general metric (including a tokamak with circular or noncircular cross-section) has been developed. This gyrokinetic simulation is capable of examining the global and semi-global driftwave structures and their associated transport in a tokamak plasma. The authors investigate the property of the ion temperature gradient (ITG) or η i (η i ≡ ∂ ell nT i /∂ ell n n i ) driven drift waves in a tokamak plasma. The emergent semi-global drift wave modes give rise to thermal transport characterized by the Bohm scaling

Fast IR diodes thermometer for tokamak

International Nuclear Information System (INIS)

Chen Xiangbo

2001-01-01

A 30 channel fast IR pyrometry array has been constructed for tokamak, which has 0.5 μs time response, 10 mm diameter spatial resolution and 5 degree C temperature resolution. The temperature measuring range is from 250 degree C to 1200 degree C. The two dimensional temperature profiles of the first wall during both major and minor disruptions can be measured with an accuracy of about 1% measuring temperature, which is adequate for tokamak experiments. This gives a very useful tool for the disruption study, especially for the divertor physics and edge heat flux research on tokamak and other magnetic confinement devices

The ASDEX Upgrade Parameter Server

Energy Technology Data Exchange (ETDEWEB)

Neu, Gregor, E-mail: gregor.neu@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Cole, Richard [Unlimited Computer Systems, Seeshaupter Str. 15, 82393 Iffeldorf (Germany); Gräter, Alex [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Lüddecke, Klaus [Unlimited Computer Systems, Seeshaupter Str. 15, 82393 Iffeldorf (Germany); Rapson, Christopher J.; Raupp, Gerhard; Treutterer, Wolfgang; Zasche, Dietrich; Zehetbauer, Thomas [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany)

2015-10-15

Highlights: • We describe our main tool in the plasma control configuration process. • Parameter access and computation are configurable with XML files. • Simple implementation of in situ tests by rerouting requests to test data. • Pulse specific overriding of parameters. - Abstract: Concepts for the configuration of plant systems and plasma control of modern devices such as ITER and W7-X are based on global data structures, or “pulse schedules” or “experiment programs”, which specify all physics characteristics (waveforms for controlled actuators and plasma quantities) and all technical characteristics of the plant systems (diagnostics and actuators operation settings) for a planned pulse. At ASDEX Upgrade we use different approach. We observed that the physics characteristics driving the discharge control system (DCS) are frequently modified on a pulse-to-pulse basis. Plant system operation, however, relies on technical standard settings, or “basic configurations” to provide guaranteed resources or services, which evolve according to longer term session or campaign operation schedules. This is why AUG manages technical configuration items separately from physics items. Consistent computation of the DCS configuration requires access to all this physics and technical data, which include the discharge programme (DP), settings of actuator systems and real-time diagnostics, the current system state and a database of static parameters. A Parameter Server provides a unified view on all these parameter sets and acts as the central point of access. We describe the functionality and architecture of the Parameter Server and its embedding into the control environment.

Prospects for Tokamak Fusion Reactors

International Nuclear Information System (INIS)

Sheffield, J.; Galambos, J.

1995-01-01

This paper first reviews briefly the status and plans for research in magnetic fusion energy and discusses the prospects for the tokamak magnetic configuration to be the basis for a fusion power plant. Good progress has been made in achieving fusion reactor-level, deuterium-tritium (D-T) plasmas with the production of significant fusion power in the Joint European Torus (up to 2 MW) and the Tokamak Fusion Test Reactor (up to 10 MW) tokamaks. Advances on the technologies of heating, fueling, diagnostics, and materials supported these achievements. The successes have led to the initiation of the design phases of two tokamaks, the International Thermonuclear Experimental Reactor (ITER) and the US Toroidal Physics Experiment (TPX). ITER will demonstrate the controlled ignition and extended bum of D-T plasmas with steady state as an ultimate goal. ITER will further demonstrate technologies essential to a power plant in an integrated system and perform integrated testing of the high heat flux and nuclear components required to use fusion energy for practical purposes. TPX will complement ITER by testing advanced modes of steady-state plasma operation that, coupled with the developments in ITER, will lead to an optimized demonstration power plant

Calculation of voltages and currents induced in the vacuum vessel of ASDEX by plasma disruptions

International Nuclear Information System (INIS)

Preis, H.

1978-01-01

An approximation method is used to analyze the electromagnetic diffusion process induced in the walls of the ASDEX vacuum vessel by plasma disruptions. For this purpose the rotational-symmetric vessel is regarded as N = 82 circular conductors connected in parallel and inductively coupled with one another and with the plasma. The transient currents and voltages occurring in this circuit are calculated with computer programs. From the calculated currents it is possible to determine the time behavior of the distributions of the current density and magnetic force density in the vessel walls. (orig.) [de

Presheath profiles in simulated tokamak edge plasmas

International Nuclear Information System (INIS)

LaBombard, B.; Conn, R.W.; Hirooka, Y.; Lehmer, R.; Leung, W.K.; Nygren, R.E.; Ra, Y.; Tynan, G.

1988-04-01

The PISCES plasma surface interaction facility at UCLA generates plasmas with characteristics similar to those found in the edge plasmas of tokamaks. Steady state magnetized plasmas produced by this device are used to study plasma-wall interaction phenomena which are relevant to tokamak devices. We report here progress on some detailed investigations of the presheath region that extends from a wall surface into these /open quotes/simulated tokamak/close quotes/ edge plasma discharges along magnetic field lines

Robust Sliding Mode Control for Tokamaks

Directory of Open Access Journals (Sweden)

I. Garrido

2012-01-01

Full Text Available Nuclear fusion has arisen as an alternative energy to avoid carbon dioxide emissions, being the tokamak a promising nuclear fusion reactor that uses a magnetic field to confine plasma in the shape of a torus. However, different kinds of magnetohydrodynamic instabilities may affect tokamak plasma equilibrium, causing severe reduction of particle confinement and leading to plasma disruptions. In this sense, numerous efforts and resources have been devoted to seeking solutions for the different plasma control problems so as to avoid energy confinement time decrements in these devices. In particular, since the growth rate of the vertical instability increases with the internal inductance, lowering the internal inductance is a fundamental issue to address for the elongated plasmas employed within the advanced tokamaks currently under development. In this sense, this paper introduces a lumped parameter numerical model of the tokamak in order to design a novel robust sliding mode controller for the internal inductance using the transformer primary coil as actuator.

Main-ion temperature and plasma rotation measurements based on scattering of electron cyclotron heating waves in ASDEX Upgrade

DEFF Research Database (Denmark)

Pedersen, Morten Stejner; Rasmussen, Jesper; Nielsen, Stefan Kragh

2017-01-01

We demonstrate measurements of spectra of O-mode electron cyclotron resonance heating (ECRH) waves scattered collectively from microscopic plasma fluctuations in ASDEX Upgrade discharges with an ITER-like ECRH scenario. The measured spectra are shown to allow determination of the main ion...... temperature and plasma rotation velocity. This demonstrates that ECRH systems can be exploited for diagnostic purposes alongside their primary heating purpose in a reactor relevant scenario....

Three novel tokamak plasma regimes in TFTR

International Nuclear Information System (INIS)

Furth, H.P.

1985-10-01

Aside from extending ''standard'' ohmic and neutral beam heating studies to advanced plasma parameters, TFTR has encountered a number of special plasma regimes that have the potential to shed new light on the physics of tokamak confinement and the optimal design of future D-T facilities: (1) High-powered, neutral beam heating at low plasma densities can maintain a highly reactive hot-ion population (with quasi-steady-state beam fueling and current drive) in a tokamak configuration of modest bulk-plasma confinement requirements. (2) Plasma displacement away from limiter contact lends itself to clarification of the role of edge-plasma recycling and radiation cooling within the overall pattern of tokamak heat flow. (3) Noncentral auxiliary heating (with a ''hollow'' power-deposition profile) should serve to raise the central tokamak plasma temperature without deterioration of central region confinement, thus facilitating the study of alpha-heating effects in TFTR. The experimental results of regime (3) support the theory that tokamak profile consistency is related to resistive kink stability and that the global energy confinement time is determined by transport properties of the plasma edge region

Internal disruption in tokamaks

International Nuclear Information System (INIS)

Kuvshinov, B.N.; Savrukhin, P.V.

1990-01-01

A review of results of experimental and theoretical investigations of internal disruption in tokamaks is given. Specific features of various types of saw-tooth oscillations are described and their classification is performed. Theoretical models of the process of development of internal disruption instability are discussed. Effect of internal disruption on parameters of plasma, confined in tokamak, is considered. Scalings of period and amplitude of saw-tooth oscillations, as well as version radius are presented. Different methods for stabilizing instability of internal disruption are described

Internal disruptions in tokamaks

International Nuclear Information System (INIS)

Kuvshinov, B.N.; Savrukhin, P.V.

1990-01-01

Experimental and theoretical studies of the phenomenon of internal disruptions in tokamaks are reviewed. A classification scheme is introduced and the features of different types of sawtooth oscillations are described. A theoretical model for the development of the internal disruption instability is discussed. The effect of internal disruptions on the parameters of plasma confined in tokamaks is discussed. Scaling laws for the period and amplitude of sawtooth oscillations, as well as for the inversion radius, are presented. Different methods of stabilizing the internal disruption instability are described

Possibilities for breakeven and ignition of D-3He fusion fuel in a near term tokamak

International Nuclear Information System (INIS)

Emmert, G.A.; El-Guebaly, L.; Kulcinski, G.L.; Santarius, J.F.; Scharer, J.E.; Sviatoslavsky, I.N.; Walstrom, P.L.; Klinghoefer, R.; Wittenberg, J.L.

1988-09-01

The recent realization that the moon contains a large amount of the isotope 3 He has rekindled interest in the D- 3 He fuel cycle. In this study we consider the feasibility of investigating D- 3 He reactor plasma conditions in a tokamak of the NET/INTOR class. We have found that, depending on the energy confinement scaling law, energy breakeven may be achieved without significant modification to the NET design. The best results are for the more optimistic ASDEX H-mode scaling law. Kaye-Goldston scaling with a modest improvement due to the H-mode is more pessimistic and makes achieving breakeven more difficult. Significant improvement in Q (ratio of the fusion power to the injected power), or the ignition margin, can be achieved by taking advantage of the much reduced neutron production of the D- 3 He fuel cycle. Removal of the tritium producing blanket and replacing the inboard neutron shield by a thinner shield optimized for the neutron spectrum in D- 3 He allows the plasma to be increased without changing the magnetic field at the toroidal field magnet. This allows the plasma to achieve higher beta and Q values up to about 3. The implications of D- 3 He operation for fast ion loss, neutron shielding, heat loads on the first wall and divertor, plasma refuelling, changes to the poloidal field coil system, and pumping of the helium from the vacuum chamber are considered in the report. (orig.)

ANTWKB: a code for the simulation of ion cyclotron antennas in tokamaks

International Nuclear Information System (INIS)

Brambilla, M.

1995-04-01

We have developed a code which evaluates the complex input impedance, the loading, and the spectral distribution of the launched power, of metallic antennas for ion cyclotron heating of large tokamak plasmas. The current distribution along the conductors is obtained selfconsistently from a variational method. The plasma response is evaluated assuming that the WKB approximation can be used already at the plasma edge, thereby avoiding the lengthy integration of the wave equations in the plasma. This makes possible systematic scans over frequency or other parameters, while retaining a sufficiently large number of Fourier components in the radiated fields to ensure convergence of both the resistive and reactive part of the power. Optionally, the code can evaluate the antenna response in vacuum or with a dummy load, for comparison with test bank measurements. We have applied the code to a few antennas of practical interest. The code reproduces accurately the expected transmission-line-like behaviour of a simple feeder-to-short antenna, and reasonably well the measured properties of the folded antenna of the ASDEX Upgrade ICRF experiment. This antenna is found to have particularly favourable properties, since its outer conductors present to the plasma a relatively uniform current over a broad range of frequencies, which, moreover, is always larger than in the return conductors. The loading of the ''violin antenna'' recently proposed for use in ITER is found to be satisfactory in the vicinity of antenna resonance, although rather poor at other frequencies. In the case of simple strap antennas replacing the short by an adjustable capacity, as in TORE SUPRA, is confirmed to be a good way of optimizing the loading. (orig.)

Project and analysis of the toroidal magnetic field production circuits and the plasma formation of the ETE (Spherical Tokamak Experiment) tokamak; Projeto e analise dos circuitos de producao de campo magnetico toroidal e de formacao do plasma do Tokamak ETE (Experimento Tokamak Esferico)

Energy Technology Data Exchange (ETDEWEB)

Barbosa, Luis Filipe F.P.W.; Bosco, Edson del

1994-12-31

This report presents the project and analysis of the circuit for production of the toroidal magnetic field in the Tokamak ETE (Spherical Tokamak Experiment). The ETE is a Tokamak with a small-aspect-ratio parameter to be used for studying the plasma physics for the research on thermonuclear fusion. This machine is being constructed at the Laboratorio Associado de Plasma (LAP) of the Instituto Nacional de Pesquisas Espaciais (INPE) in Sao Jose dos Campos, SP, Brazil. (author). 20 refs., 39 figs., 4 tabs.

Soft X-Ray measurements and analysis on Tokamaks in view of real-time control

International Nuclear Information System (INIS)

Vezinet, Didier

2013-01-01

This thesis focuses on measuring and interpreting the Soft X-Ray (SXR) radiation (approximately [1 keV; 15 keV]) in Tokamaks. As explained in Chapter 2, this radiation conveys information about the plasma density, temperature, magnetic equilibrium and impurity content. However, the measured data is spectrally and spatially-integrated and results from several physical phenomena affecting every ion species. Tore Supra's SXR diagnostics is based on semiconductor diodes presented in Chapter 3, along with a new gas detector successfully tested in laboratory and on Tore Supra. A new methodology for absolute spectral characterisation of photo detectors using a portable SXR tube is presented. Tomographic inversion algorithms, that grant access to reconstructions of the SXR emissivity field in a poloidal cross-section, are presented in Chapter 4. Improvements implemented on one particular algorithm are detailed with examples of application. A comparison between the position of the SXR emissivity maximum and the magnetic axis reconstructed by an equilibrium code is presented in Chapter 5. Chapter 6 presents an approach used to derive an impurity density from its SXR emissivity using the robustness of its SXR cooling factor with respect to impurity transport. The physics accounting for this robustness is studied and a first map of the domain of validity of this method is provided. Chapter 7 addresses poloidal asymmetries of the SXR emissivity field. Two types of asymmetries are presented as well as experiments conducted on ASDEX-U to verify their parametric dependences. A new type of SXR asymmetry, observed on Tore Supra is introduced. (author) [fr

Tokamak pump limiters

International Nuclear Information System (INIS)

Conn, R.W.

1984-05-01

Recent experiments with a scoop limiter without active internal pumping have been carried out in the PDX tokamak with up to 6MW of auxiliary neutral beam heating. Experiments have also been done with a rotating head pump limiter in the PLT tokamak in conjunction with RF plasma heating. Extensive experiments have been done in the ISX-B tokamak and first experiments have been completed with the ALT-I limiter in TEXTOR. The pump limiter modules in these latter two machines have internal getter pumping. Experiments in ISX-B are with ohmic and auxiliary neutral beam heating. The results in ISX-B and TEXTOR show that active density control and particle removal is achieved with pump limiters. In ISX-B, the boundary layer (or scape-off layer) plasma partially screens the core plasma from gas injection. In both ISX-B and TEXTOR, the pressure internal to the module scales linearly with plasma density but in ISX-B, with neutral beam injection, a nonlinear increase is observed at the highest densities studied. Plasma plugging is the suspected cause. Results from PDX suggest that a region may exist in which core plasma energy confinement improves using a pump limiter during neutral beam injection. Asymmetric radial profiles and an increased edge electron temperature are observed in discharges with improved confinement. The injection of small amounts of neon into ISX-B has more clearly shown an improved electron core energy confinement during neutral beam injection. While carried out with a regular limiter, this Z-mode of operation is ideal for use with pump limiters and should be a way to achieve energy confinement times similar to values for H-mode tokamak plasmas. The implication of all these results for the design of a reactor pump limiter is described

Tokamak pump limiters

International Nuclear Information System (INIS)

Conn, R.W.; California Univ., Los Angeles

1984-01-01

Recent experiments with a scoop limiter without active internal pumping have been carried out in the PDX tokamak with up to 6 MW of auxiliary neutral beam heating. Experiments have also been performed with a rotating head pump limiter in the PLT tokamak in conjunction with RF plasma heating. Extensive experiments have been done in the ISX-B tokamak and first experiments have been completed with the ALT-I limiter in TEXTOR. The pump limiter modules in these latter two machines have internal getter pumping. Experiments in ISX-B are with ohmic and auxiliary neutral beam heating. The results in ISX-B and TEXTOR show that active density control and particle removal is achieved with pump limiters. In ISX-B, the boundary layer (or scrape-off layer) plasma partially screens the core plasma from gas injection. In both ISX-B and TEXTOR, the pressure internal to the module scales linearly with plasma density but in ISX-B, with neutral beam injection, a nonlinear increase is observed at the highest densities studied. Plasma plugging is the suspected cause. Results from PDX suggest that a regime may exist in which core plasma energy confinement improves using a pump limiter during neutral beam injection. Asymmetric radial profiles and an increased edge electron temperature are observed in discharges with improved confinement. The injection of small amounts of neon into ISX-B has more clearly shown an improved electron core energy confinement during neutral beam injection. While carried out with a regular limiter, this 'Z-mode' of operation is ideal for use with pump limiters and should be a way to achieve energy confinement times similar to values for H-mode tokamak plasmas. The implication of all these results for the design of a reactor pump limiter is described. (orig.)

Estimation of Zeff in Novillo Tokamak

International Nuclear Information System (INIS)

Valencia, R.; Olayo, G.; Cruz, G.; Lopez, R.; Chavez, E.; Melendez, L.; Flores, A.; Gaytan, E.

1996-01-01

We estimated the Z eff in the Novillo Tokamak after having applied a HeGDC process through two different methods: anomaly factor and mass spectrometry. The first one gave a Z eff value of 2.07 for a tokamak discharge of 4350 A plasma current and 3 V of loop voltage. By mass spectrometry 30 s after the discharge had finished a Z eff of 4.19 was obtained for the same discharge. By mass spectrometry we observed that the Z eff value is a time function. Furthermore this method is helpful for evaluating the level of impurities after many discharges in Novillo Tokamak. (orig.)

The enhanced ASDEX Upgrade pellet centrifuge launcher

International Nuclear Information System (INIS)

Plöckl, B.; Lang, P. T.

2013-01-01

Pellets played an important role in the program of ASDEX Upgrade serving both for investigations on efficient particle fuelling and high density scenarios but also for pioneering work on Edge Localised Mode (ELM) pacing and mitigation. Initially designed for launching fuelling pellets from the magnetic low field side, the system was converted already some time ago to inject pellets from the magnetic high field side as much higher fuelling efficiency was found using this configuration. In operation for more than 20 years, the pellet launching system had to undergo a major revision and upgrading, in particular of its control system. Furthermore, the control system installed adjacent to the launcher had to be transferred to a more distant location enforcing a complete galvanic separation from torus potential and a fully remote control solution. Changing from a hybrid system consisting of PLC S5/S7 and some hard wired relay control to a state of the art PLC system allowed the introduction of several new operational options enabling more flexibility in the pellet experiments. This article describes the new system architecture of control hardware and software, the operating procedure, and the extended operational window. First successful applications for ELM pacing and triggering studies are presented as well as utilization for the development of high density scenarios

The enhanced ASDEX Upgrade pellet centrifuge launcher

Energy Technology Data Exchange (ETDEWEB)

Plöckl, B.; Lang, P. T. [Max Planck Institute for Plasma Physics, EURATOM Association, Boltzmannstr. 2, 85748 Garching (Germany)

2013-10-15

Pellets played an important role in the program of ASDEX Upgrade serving both for investigations on efficient particle fuelling and high density scenarios but also for pioneering work on Edge Localised Mode (ELM) pacing and mitigation. Initially designed for launching fuelling pellets from the magnetic low field side, the system was converted already some time ago to inject pellets from the magnetic high field side as much higher fuelling efficiency was found using this configuration. In operation for more than 20 years, the pellet launching system had to undergo a major revision and upgrading, in particular of its control system. Furthermore, the control system installed adjacent to the launcher had to be transferred to a more distant location enforcing a complete galvanic separation from torus potential and a fully remote control solution. Changing from a hybrid system consisting of PLC S5/S7 and some hard wired relay control to a state of the art PLC system allowed the introduction of several new operational options enabling more flexibility in the pellet experiments. This article describes the new system architecture of control hardware and software, the operating procedure, and the extended operational window. First successful applications for ELM pacing and triggering studies are presented as well as utilization for the development of high density scenarios.

Results of high heat flux tests and structural analysis of the new solid tungsten divertor tile for ASDEX Upgrade

Energy Technology Data Exchange (ETDEWEB)

Jaksic, Nikola, E-mail: nikola.jaksic@ipp.mpg.de; Greuner, Henri; Herrmann, Albrecht; Böswirth, Bernd; Vorbrugg, Stefan

2015-10-15

Highlights: • The main motivation for the HHF investigation of tungsten tiles was an untypical deformation of some specimens under thermal loading, observed during the previous tests in GLADIS test facility. • A nonlinear finite element (FE) model for simulations of the GLADIS tests has been built. • The unexpected plastic deformations are mainly caused by internal stresses due to the manufacturing process. The small discrepancies among the FEA investigated and measured plastic deformations are most likely caused, beside of the practical difficulties by measuring of low items, also by tile internal stresses. • The influences of the residual stresses caused by special production processes have to be taken into account by design of the structural part made of solid tungsten. - Abstract: Tungsten as plasma-facing material for fusion devices is currently the most favorable candidate. In general solid tungsten is used for shielding the plasma chamber interior against the high heat generated from the plasma. For the purposes of implementation at ASDEX Upgrade and as a contribution to ITER the thermal performance of tungsten tiles has been extensively tested in the high heat flux test facility GLADIS during the development phase and beyond. These tests have been performed on full scale tungsten tile prototypes including their clamping and cooling structure. Simulating the adiabatically thermal loading due to plasma operation in ASDEX Upgrade, the tungsten tiles have been subjected to a thermal load with central heat flux of 10–24 MW/m{sup 2} and absorbed energy between 370 and 680 kJ. This loading results in maximum surface temperatures between 1300 °C and 2800 °C. The tests in GLADIS have been accompanied by intensive numerical investigations using FEA methods. For this purpose a multiple nonlinear finite element model has been set up. This paper discusses the main results of the high heat flux final tests and their numerical simulation. Moreover, first

Results of high heat flux tests and structural analysis of the new solid tungsten divertor tile for ASDEX Upgrade

International Nuclear Information System (INIS)

Jaksic, Nikola; Greuner, Henri; Herrmann, Albrecht; Böswirth, Bernd; Vorbrugg, Stefan

2015-01-01

Highlights: • The main motivation for the HHF investigation of tungsten tiles was an untypical deformation of some specimens under thermal loading, observed during the previous tests in GLADIS test facility. • A nonlinear finite element (FE) model for simulations of the GLADIS tests has been built. • The unexpected plastic deformations are mainly caused by internal stresses due to the manufacturing process. The small discrepancies among the FEA investigated and measured plastic deformations are most likely caused, beside of the practical difficulties by measuring of low items, also by tile internal stresses. • The influences of the residual stresses caused by special production processes have to be taken into account by design of the structural part made of solid tungsten. - Abstract: Tungsten as plasma-facing material for fusion devices is currently the most favorable candidate. In general solid tungsten is used for shielding the plasma chamber interior against the high heat generated from the plasma. For the purposes of implementation at ASDEX Upgrade and as a contribution to ITER the thermal performance of tungsten tiles has been extensively tested in the high heat flux test facility GLADIS during the development phase and beyond. These tests have been performed on full scale tungsten tile prototypes including their clamping and cooling structure. Simulating the adiabatically thermal loading due to plasma operation in ASDEX Upgrade, the tungsten tiles have been subjected to a thermal load with central heat flux of 10–24 MW/m"2 and absorbed energy between 370 and 680 kJ. This loading results in maximum surface temperatures between 1300 °C and 2800 °C. The tests in GLADIS have been accompanied by intensive numerical investigations using FEA methods. For this purpose a multiple nonlinear finite element model has been set up. This paper discusses the main results of the high heat flux final tests and their numerical simulation. Moreover, first results

Numerical Tokamak Project code comparison

International Nuclear Information System (INIS)

Waltz, R.E.; Cohen, B.I.; Beer, M.A.

1994-01-01

The Numerical Tokamak Project undertook a code comparison using a set of TFTR tokamak parameters. Local radial annulus codes of both gyrokinetic and gyrofluid types were compared for both slab and toroidal case limits assuming ion temperature gradient mode turbulence in a pure plasma with adiabatic electrons. The heat diffusivities were found to be in good internal agreement within ± 50% of the group average over five codes

Confinement and diffusion in tokamaks

International Nuclear Information System (INIS)

McWilliams, R.

1988-01-01

The effect of electric field fluctuations on confinement and diffusion in tokamak is discussed. Based on the experimentally determined cross-field turbolent diffusion coefficient, D∼3.7*cT e /eB(δn i /n i ) rms which is also derived by a simple theory, the cross-field diffusion time, tp=a 2 /D, is calculated and compared to experimental results from 51 tokamak for standard Ohmic operation

Thyristor crowbar system for the high current power supplies of ASDEX upgrade

International Nuclear Information System (INIS)

Kaesemann, C.-P.; Lieshout, L. van; Huart, M.; Sihler, C.

2005-01-01

The ohmic heating system and the poloidal field coils of ASDEX upgrade are supplied by 15 thyristor converter units with an installed apparent power of 600 MVA. To protect the thyristor converters against dc overvoltage arising from abnormal operations and resulting damages caused by the large energy stored in the AUG magnet coils an overvoltage protection system was required. The paper describes the motivation for-and the design and testing of the thyristor crowbar system representing the thyristor converter overvoltage protection system. It will present the layout, analyse the results of measurements obtained during commissioning, compare them to the calculated (design) values and report on the first experience of operation on the AUG coils improving the safety of the equipment

Edge transport and its interconnection with main chamber recycling in ASDEX upgrade

International Nuclear Information System (INIS)

Kallenbach, A.; Dux, R.; Gafert, J.

2003-01-01

Edge profiles of electron temperature and density are measured in ASDEX Upgrade with high spatial resolution of 2-3 mm with Thomson scattering. In the region of the edge transport barrier in ELMy H-mode, the gradient lengths of T e and n e are found closely coupled, with the temperature profile twice as steep as the density profile corresponding to η e ∼ 2. The edge density in the region of the barrier foot is closely coupled to the main chamber recycling, with no strong dependence on other parameters. In contrast the density rise from the outer barrier foot to the pedestal exhibits pronounced dependence on plasma current and shaping, indicating quite different mechanisms determining the absolute density and its gradient. (author)

Numerical studies of edge localized instabilities in tokamaks

International Nuclear Information System (INIS)

Wilson, H.R.; Snyder, P.B.; Huysmans, G.T.A.; Miller, R.L.

2002-01-01

A new computational tool, edge localized instabilities in tokamaks equilibria (ELITE), has been developed to help our understanding of short wavelength instabilities close to the edge of tokamak plasmas. Such instabilities may be responsible for the edge localized modes observed in high confinement H-mode regimes, which are a serious concern for next step tokamaks because of the high transient power loads which they can impose on divertor target plates. ELITE uses physical insight gained from analytic studies of peeling and ballooning modes to provide an efficient way of calculating the edge ideal magnetohydrodynamic stability properties of tokamaks. This paper describes the theoretical formalism which forms the basis for the code

Tokamak COMPASS

Czech Academy of Sciences Publication Activity Database

Řípa, Milan; Křenek, Petr

2011-01-01

Roč. 17, č. 1 (2011), s. 32-34 ISSN 1210-4612 Institutional research plan: CEZ:AV0Z20430508 Keywords : fusion * tokamak * Compass * Golem * Institute of Plasma Physics AVCR v.v * NBI * diagnostics Subject RIV: BL - Plasma and Gas Discharge Physics

An enhanced tokamak startup model

Science.gov (United States)

Goswami, Rajiv; Artaud, Jean-François

2017-01-01

The startup of tokamaks has been examined in the past in varying degree of detail. This phase typically involves the burnthrough of impurities and the subsequent rampup of plasma current. A zero-dimensional (0D) model is most widely used where the time evolution of volume averaged quantities determines the detailed balance between the input and loss of particle and power. But, being a 0D setup, these studies do not take into consideration the co-evolution of plasma size and shape, and instead assume an unchanging minor and major radius. However, it is known that the plasma position and its minor radius can change appreciably as the plasma evolves in time to fill in the entire available volume. In this paper, an enhanced model for the tokamak startup is introduced, which for the first time takes into account the evolution of plasma geometry during this brief but highly dynamic period by including realistic one-dimensional (1D) effects within the broad 0D framework. In addition the effect of runaway electrons (REs) has also been incorporated. The paper demonstrates that the inclusion of plasma cross section evolution in conjunction with REs plays an important role in the formation and development of tokamak startup. The model is benchmarked against experimental results from ADITYA tokamak.

Spherical tokamak development in Brazil

International Nuclear Information System (INIS)

Ludwig, G.O.; Del Bosco, E.; Ferreira, J.G.; Berni, L.A.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Ueda, M.; Barroso, J.J.; Castro, P.J.; Barbosa, L.F.W.; Patire Junior, H.; The high-power microwave sources group

2003-01-01

This paper describes the general characteristics of spherical tokamaks, or spherical tori, with a brief overview of work in this area already performed or in progress at several institutions worldwide. The paper presents also the steps in the development of the ETE (Experimento Tokamak Esferico) project, its research program, technical characteristics and operating conditions as of December, 2002 at the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

Enhancement of confinement in tokamaks

International Nuclear Information System (INIS)

Furth, H.P.

1986-01-01

The analysis begins by identifying a hypothetical model of tokamak confinement that is designed to take into account the conflict between Tsub(e)(r)-profile shapes arising from microscopic transport and J(r)-profile shapes required for gross stability. On the basis of this model, a number of hypothetical lines of advance are developed. Some TFTR experiments that may point the way to a particularly attractive type of tokamak reactor regime are discussed. (author)

Spherical tokamak development in Brazil

Energy Technology Data Exchange (ETDEWEB)

Ludwig, Gerson Otto; Bosco, Edson Del; Ferreira, Julio Guimaraes [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma] (and others)

2003-07-01

The general characteristics of spherical tokamaks, or spherical tori, with a brief view of work in this area already performed or in progress at several institutions worldwide are described. The paper presents also the steps in the development of the ETE (Experiment Tokamak spheric) project, its research program, technical characteristics and operating conditions as of December, 2002 a the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

Spherical tokamak development in Brazil

Energy Technology Data Exchange (ETDEWEB)

Ludwig, G.O.; Del Bosco, E.; Ferreira, J.G.; Berni, L.A.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Ueda, M.; Barroso, J.J.; Castro, P.J. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma; Barbosa, L.F.W. [Universidade do Vale do Paraiba (UNIVAP), Sao Jose dos Campos, SP (Brazil). Faculdade de Engenharia, Arquitetura e Urbanismo; Patire Junior, H. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Div. de Mecanica Espacial e Controle; The high-power microwave sources group

2003-12-01

This paper describes the general characteristics of spherical tokamaks, or spherical tori, with a brief overview of work in this area already performed or in progress at several institutions worldwide. The paper presents also the steps in the development of the ETE (Experimento Tokamak Esferico) project, its research program, technical characteristics and operating conditions as of December, 2002 at the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

Spherical tokamak development in Brazil

International Nuclear Information System (INIS)

Ludwig, Gerson Otto; Bosco, Edson Del; Ferreira, Julio Guimaraes

2003-01-01

The general characteristics of spherical tokamaks, or spherical tori, with a brief view of work in this area already performed or in progress at several institutions worldwide are described. The paper presents also the steps in the development of the ETE (Experiment Tokamak spheric) project, its research program, technical characteristics and operating conditions as of December, 2002 a the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

Bibliography of fusion product physics in tokamaks

International Nuclear Information System (INIS)

Hively, L.M.; Sigmar, D.J.

1989-09-01

Almost 700 citations have been compiled as the first step in reviewing the recent research on tokamak fusion product effects in tokamaks. The publications are listed alphabetically by the last name of the first author and by subject category

Divertor power and particle fluxes between and during type-I ELMs in the ASDEX Upgrade

Science.gov (United States)

Kallenbach, A.; Dux, R.; Eich, T.; Fischer, R.; Giannone, L.; Harhausen, J.; Herrmann, A.; Müller, H. W.; Pautasso, G.; Wischmeier, M.; ASDEX Upgrade Team

2008-08-01

Particle, electric charge and power fluxes for type-I ELMy H-modes are measured in the divertor of the ASDEX Upgrade tokamak by triple Langmuir probes, shunts, infrared (IR) thermography and spectroscopy. The discharges are in the medium to high density range, resulting in predominantly convective edge localized modes (ELMs) with moderate fractional stored energy losses of 2% or below. Time resolved data over ELM cycles are obtained by coherent averaging of typically one hundred similar ELMs, spatial profiles from the flush-mounted Langmuir probes are obtained by strike point sweeps. The application of simple physics models is used to compare different diagnostics and to make consistency checks, e.g. the standard sheath model applied to the Langmuir probes yields power fluxes which are compared with the thermographic measurements. In between ELMs, Langmuir probe and thermography power loads appear consistent in the outer divertor, taking into account additional load due to radiation and charge exchange neutrals measured by thermography. The inner divertor is completely detached and no significant power flow by charged particles is measured. During ELMs, quite similar power flux profiles are found in the outer divertor by thermography and probes, albeit larger uncertainties in Langmuir probe evaluation during ELMs have to be taken into account. In the inner divertor, ELM power fluxes from thermography are a factor 10 larger than those derived from probes using the standard sheath model. This deviation is too large to be caused by deficiencies of probe analysis. The total ELM energy deposition from IR is about a factor 2 higher in the inner divertor compared with the outer divertor. Spectroscopic measurements suggest a quite moderate contribution of radiation to the target power load. Shunt measurements reveal a significant positive charge flow into the inner target during ELMs. The net number of elementary charges correlates well with the total core particle loss

Tokamak startup: problems and scenarios related to the transient phases of ignited tokamak operations

International Nuclear Information System (INIS)

Sheffield, J.

1985-01-01

During recent years improvements have been made to tokamak startup procedures, which are important to the optimization of ignited tokamaks. The use of rf-assisted startup and noninductive current drive has led to substantial reduction and even complete elimination of the volt-seconds used during startup, relaxing constraints on poloidal coil, vacuum vessel, and structure design. This paper reviews these and other improvements and discusses the various bulk heating techniques that may be used to ignite a D-T plasma

Pellet injection into ASDEX upgrade plasmas

International Nuclear Information System (INIS)

Lang, P.T.; Zohm, H.; Buechl, K.; Fuchs, J.C.; Gehre, O.; Gruber, O.; Lang, R.S.; Mertens, V.; Neuhauser, J.; Salzmann, H.

1996-04-01

This work comprises results obtained using the new centrifuge injection system for the two first years of pellet injection experiments at Asdex Upgrade until the end of the 1995 experimental campaign. The main aim of the pellet injection investigation is to develop scenarios allowing for a more flexible plasma density control means of injection of cryogenic solid hydrogen pellets. Efforts have been made to develop scenarios allowing more flexible plasma density control by injecting cryogenic solid hydrogen pellets. While the injection of pellets during ohmic discharges was found to be most efficient and also improves the plasma performance, increasing the auxiliary heating power causes a detoriation of the pellet fuelling efficiency. A further strong reduction of the pellet fuelling efficiency by an additional process was observed for the more reactor-relevant conditions of shallow particle deposition during H-mode phases. With injection during type I ELMy H-mode phases, each pellet was found to trigger the release of an ELM and therefore cause particle losses mainly from the edge region. In the type I ELMy H-mode, only sufficient pellet penetration allowed noticeable, persistent particle deposition in the plasma by the pellets. Applying adequate pellet injection conditions and favourable scenarios using combined pellet/gas puff refuelling, significant density ramp-up to densities exceeding the empirical Greenwald limit by up to a factor of two was achieved even for strongly heated H-mode plasmas. (orig.)

MAST: a Mega Amp Spherical Tokamak

International Nuclear Information System (INIS)

Darke, A.C.; Harbar, J.R.; Hay, J.H.; Hicks, J.B.; Hill, J.W.; McKenzie, J.S.; Morris, A.W.; Nightingale, M.P.S.; Todd, T.N.; Voss, G.M.; Watkins, J.R.

1995-01-01

The highly successful tight aspect ratio tokamak research pioneered on the START machine at Culham, together with the attractive possibilities of the concept, suggest a larger device should be considered. The design of a Mega Amp Spherical Tokamak is described, operating at much higher currents and over longer pulses than START and compatible with strong additional heating. (orig.)

The ARIES-I tokamak reactor study

International Nuclear Information System (INIS)

1991-01-01

This report contains an overview of the Aries-I tokamak reactor study. The following topics are discussed on this tokamak: Systems studies; equilibrium, stability, and transport; summary and conclusions; current drive; impurity control system; tritium systems; magnet engineering; fusion-power-core engineering; power conversion; Aries-I safety design and analysis; design layout and maintenance; and start-up and operations

Economic evaluation of tokamak power plants

International Nuclear Information System (INIS)

Reid, R.L.; Steiner, D.

1977-01-01

This study reports the impact of plasma operating characteristics, engineering options, and technology on the capital cost trends of tokamak power plants. Tokamak power systems are compared to other advanced energy systems and found to be economically competitive. A three-phase strategy for demonstrating commercial feasibility of fusion power, based on a common-site multiple-unit concept, is presented

Tokamaks: from A D Sakharov to the present (the 60-year history of tokamaks)

International Nuclear Information System (INIS)

Azizov, E A

2012-01-01

The paper is prepared on the basis of the report presented at the session of the Physical Sciences Division of the Russian Academy of Sciences (RAS) at the Lebedev Physical Institute, RAS on 25 May 2011, devoted to the 90-year jubilee of Academician Andrei D Sakharov - the initiator of controlled nuclear fusion research in the USSR. The 60-year history of plasma research work in toroidal devices with a longitudinal magnetic field suggested by Andrei D Sakharov and Igor E Tamm in 1950 for the confinement of fusion plasma and known at present as tokamaks is described in brief. The recent (2006) agreement among Russia, the EU, the USA, Japan, China, the Republic of Korea, and India on the joint construction of the international thermonuclear experimental reactor (ITER) in France based on the tokamak concept is discussed. Prospects for using the tokamak as a thermonuclear (14 MeV) neutron source are examined. (conferences and symposia)

Overview of Tokamak Results

International Nuclear Information System (INIS)

Unterberg, Bernhard; Samm, Ulrich

2004-01-01

An overview is given of recent results obtained in tokamak devices. We introduce basic confinement scenarios as L-mode, H-mode and plasmas with an internal transport barrier and discuss methods for profile control. Important findings in DT-experiments at JET as α-particle heating are described. Methods for power exhaust like plasma regimes with a radiating mantle and radiative divertor scenarios are discussed. The overall impact of plasma edge conditions on the general plasma performance in tokamaks is illustrated by describing the impact of wall conditions on confinement and the edge operational diagram of H-mode plasmas

Plea for stellarator funding raps tokamaks

International Nuclear Information System (INIS)

Blake, M.

1992-01-01

The funding crunch in magnetic confinement fusion development has moved the editor of a largely technical publication to speak out on a policy issue. James A. Rome, who edits Stellarator News from the Fusion Energy Division at Oak Ridge National Laboratory, wrote an editorial that appeared on the front page of the May 1992 issue. It was titled open-quotes The US Stellarator Program: A Time for Renewal,close quotes and while it focused chiefly on that subject (and lamented the lack of funding for the operation of the existing ATF stellarator at Oak Ridge), it also cited some of the problems inherent in the mainline MCF approach--the tokamak--and stated that if the money can be found for further tokamak design upgrades, it should also be found for stellarators. Rome wrote, open-quotes There is growing recognition in the US, and elsewhere, that the conventional tokamak does not extrapolate to a commercially competitive energy source except with very high field coils ( 1000 MWe).close quotes He pointed up open-quotes the difficulty of simultaneously satisfying conflicting tokamak requirements for efficient current drive, high bootstrap-current fraction, complete avoidance of disruptions, adequate beta limits, and edge-plasma properties compatible with improved (H-mode) confinement and acceptable erosion of divertor plates.close quotes He then called for support for the stellarator as open-quotes the only concept that has performance comparable to that achieved in tokamaks without the plasma-current-related limitations listed above.close quotes

Electron thermal transport in tokamak plasmas

Energy Technology Data Exchange (ETDEWEB)

Konings, J A

1994-11-30

The process of fusion of small nuclei thereby releasing energy, as it occurs continuously in the sun, is essential for the existence of mankind. The same process applied in a controlled way on earth would provide a clean and an abundant energy source, and be the long term solution of the energy problem. Nuclear fusion requires an extremely hot (10{sup 8} K) ionized gas, a plasma, that can only be maintained if it is kept insulated from any material wall. In the so called `tokamak` this is achieved by using magnetic fields. The termal insulation, which is essential if one wants to keep the plasma at the high `fusion` temperature, can be predicted using basic plasma therory. A comparison with experiments in tokamaks, however, showed that the electron enery losses are ten to hundred times larger than this theory predicts. This `anomalous transport` of thermal energy implies that, to reach the condition for nuclear fusion, a fusion reactor must have very large dimensions. This may put the economic feasibility of fusion power in jeopardy. Therefore, in a worldwide collaboration, physicists study tokamak plasmas in an attempt to understand and control the energy losses. From a scientific point of view, the mechanisms driving anomalous transport are one of the challenges in fudamental plasma physics. In Nieuwegein, a tokamak experiment (the Rijnhuizen Tokamak Project, RTP) is dedicated to the study of anomalous transport, in an international collaboration with other laboratories. (orig./WL).

Digital controlled pulsed electric system of the ETE tokamak. First report; Sistema eletrico pulsado com controle digital do Tokamak ETE (experimento Tokamak esferico). Primeiro relatorio

Energy Technology Data Exchange (ETDEWEB)

Barbosa, Luis Felipe de F.P.W.; Del Bosco, Edson

1997-12-31

This reports presents a summary on the thermonuclear fusion and application for energy supply purposes. The tokamak device operation and the magnetic field production systems are described. The ETE tokamak is a small aspect ratio device designed for plasma physics and thermonuclear fusion studies, which presently is under construction at the Laboratorio Associado de Plasma (LAP), Instituto Nacional de Pesquisas Espaciais (INPE) - S.J. dos Campos - S. Paulo. (author) 55 refs., 40 figs.

Digital controlled pulsed electric system of the ETE tokamak. First report; Sistema eletrico pulsado com controle digital do Tokamak ETE (experimento Tokamak esferico). Primeiro relatorio

Energy Technology Data Exchange (ETDEWEB)

Barbosa, Luis Felipe de F.P.W.; Del Bosco, Edson

1998-12-31

This reports presents a summary on the thermonuclear fusion and application for energy supply purposes. The tokamak device operation and the magnetic field production systems are described. The ETE tokamak is a small aspect ratio device designed for plasma physics and thermonuclear fusion studies, which presently is under construction at the Laboratorio Associado de Plasma (LAP), Instituto Nacional de Pesquisas Espaciais (INPE) - S.J. dos Campos - S. Paulo. (author) 55 refs., 40 figs.

Magnetic diagnostic of SOL-filaments generated by type I ELMs on JET and ASDEX Upgrade

DEFF Research Database (Denmark)

Naulin, Volker; Vianello, N.; Schrittwieser, R.

2011-01-01

to a simple model, motivated by observations. A new diagnostic in the form of a reciprocating probe with three magnetic pickup loops was developed for ASDEX Upgrade (AUG). Measurements during the passage of type-I ELM filaments determine the filaments to be in the scrape off layer (SOL) and to carry currents......This contribution is focused on the magnetic signatures of type I ELM filaments. On JET a limited number of high time resolution magnetic coils were used to derive essential ELM filament parameters. The method uses forward modelling and simultaneous fitting of magnetic pickup coil signals...

Comparison of the filament behaviour observed during type I ELMs in ASDEX upgrade and MAST

International Nuclear Information System (INIS)

Kirk, A; Ayed, B; Counsell, G F; Lisgo, S; Price, M; Tallents, S; Herrmann, A; Eich, T; Muller, H W; Schmid, A; Wilson, H

2008-01-01

A study of the evolution of the filaments observed during Type I ELMs on ASDEX Upgrade and MAST is presented. The filaments start off rotating toroidally/poloidally with velocities close to that of the pedestal. This velocity then decreases as the filaments propagate radially. On both devices the ion saturation current e-folding lengths of the filaments show a weak, if any, dependence on the size of the ELM (δW ELM /W ped ). On MAST the measured radial velocities of the filaments also show at most a weak dependence on δW ELM /W ped

Analysis of tokamak plasma confinement modes using the fast

Indian Academy of Sciences (India)

The Fourier analysis is a satisfactory technique for detecting plasma confinement modes in tokamaks. The confinement mode of tokamak plasma was analysed using the fast Fourier transformation (FFT). For this purpose, we used the data of Mirnov coils that is one of the identifying tools in the IR-T1 tokamak, with and ...

Investigation of fusion proton and triton emission in ASDEX

International Nuclear Information System (INIS)

Leinberger, U.

1991-01-01

A diagnostic method of measuring the fusion rate profile was developed on ASDEX. The collimated protons and tritons from d-d fusion reaction are simultaneously detected by a semiconductor counter at a single position in the vacuum vessel for different viewing directions. The detection efficiency profiles for these viewing directions are numerically calculated from the measured currents in the coils and assumed plasma current distributions. Folding the detection efficiency profile with a fusion rate profile yields the proton and triton fluxes to the detector. Comparison with measured fluxes allows one to find a fusion rate profile in agreement with the experimental data. In certain cases the detection efficiency profile strongly on the plasma current density profile, and information on the current distribution in the plasma can thus be achieved. It was proved that the spectra from rotating plasmas are in accordance with the theory of a rotating thermal plasma. Deviations can only be found in the case of strong vignetting of the detection efficiency by structures in the vacuum vessel. (orig.)

Empirical scaling for present Ohmically heated tokamaks

International Nuclear Information System (INIS)

Daughney, C.

1975-01-01

Experimental results from the Adiabatic Toroidal Compressor (ATC) tokamak are used to obtain empirical scaling laws for the average electron temperature and electron energy confinement time as functions of the average electron density, the effective ion charge, and the plasma current. These scaling laws are extended to include dependence upon minor and major plasma radius and toroidal field strength through a comparison of the various tokamaks described in the literature. Electron thermal conductivity is the dominant loss process for the ATC tokamak. The parametric dependences of the observed electron thermal conductivity are not explained by present theoretical considerations. The electron temperature obtained with Ohmic heating is shown to be a function of current density - which will not be increased in the next generation of large tokamaks. However, the temperature dependence of the electron energy confinement time suggests that significant improvement in confinement time will be obtained with supplementary electron heating. (author)

The conceptual design for the modification of HL-2A tokamak

International Nuclear Information System (INIS)

Dequan Liu

2006-01-01

The medium-sized tokamak HL-2A, based on the former ASDEX's main components has approached its rating operational parameters with I p = 450 kA and B T =2.8 T so far. The HL-2A was originally designed to operate under an axisymmetrical double-null configuration by two triplets of shaping coils located in the vacuum vessel. Because the shaping coils were set very close to the plasma column, the nulls are normally fixed and the plasma has a nearly circular cross-section. It is difficult to increase the parameters further and obtain a preferable plasma shape with certain values of triangularity and elongation. Due to the defects on upper shaping coils' electrical insulation, the HL-2A is presently just operated with lower single null divertor configuration. Additionally, the experiment has to occasionally terminate due to incidental damages of the weak parts. With the experimental progress on HL-2A, the plasma stored energy and auxiliary heating power is being increased, a more efficient, compact and tight divertor is needed. Based on present status of HL-2A tokamak and for satisfying the requirements of the experiment, the modification of HL-2A tokamak device will be carried out. The modification design of HL-2A intents to obtain optimized plasma parameters, e.g., aspect ratio, I p , b, plasma volume, certain plasma shaping with preferable elongation and triangularity, a stable vacuum vessel easy to be maintained and an advanced divertor. The primary consideration of the modification is to take all of the shaping coils out of the vessel to enlarge the plasma volume. The elongated and triangular cross section with double null of plasma is to be obtained by composite distribution of eight sets of poloidal coils. Because all poloidal coils will be located between the TF coil and the vessel, the vacuum vessel will be remade with a smaller size to make rooms for the poloidal coil system. With an optimized operation mode, the volt-second of the plasma may be increased a

Analysis of a global energy confinement database for JET ohmic plasmas

International Nuclear Information System (INIS)

Bracco, G.; Thomsen, K.

1997-01-01

A database containing global energy confinement data for JET ohmic plasmas in the campaigns from 1984 to 1992 has been established. An analysis is presented of this database and the results are compared with data from other tokamaks, such as the Axially Symmetric Divertor Experiment (ASDEX), Frascati Tokamak Upgrade (FTU) and Tore Supra. The trends of JET ohmic confinement appear to be similar to those observed on other tokamaks: a linear dependence of the global energy confinement time on density is observed up to a density value where a saturation is attained; this density value defines the border between the linear and the saturated ohmic confinement regimes; this border is shifted towards higher density values if the q value of the discharge is decreased; the global confinement time in the saturated ohmic regime increases less than linearly with the value of the magnetic field. (author). 20 refs, 13 figs, 4 tabs

Simulation of a major tokamak disruption

International Nuclear Information System (INIS)

White, R.B.; Monticello, D.A.; Rosenbluth, M.N.

1977-08-01

It is known that the internal tokamak disruption leads to a current profile which is flattened inside the surface where the safety factor equals unity. It is shown that such a profile can lead to m = 2 magnetic islands which grow to fill a substantial part of the tokamak cross section in a time consistent with the observations of the major disruption

Edge plasma diagnostics on Tore Supra tokamak

International Nuclear Information System (INIS)

Fujita, Junji

1991-01-01

From 1988 to 1991, the international scientific research 'Diagnosis of peripheral plasma in Tore Supra tokamak' was carried out as a three-year plan receiving the support of the scientific research expense of the Ministry of Education. This is to apply the method of measuring electron density distribution by neutral lithium beam probe spectroscopy to the measurement of the electron density distribution in the peripheral plasma in Tore Supra Tokamak in France. Among many tokamaks in operation doing respective characteristics researches, the Tore Supra generates the toroidal magnetic field by using superconducting coils, and aims at the long time discharge for 30 sec. for the time being, and for 300 sec. in future. In the plasma generators for long time discharge like this, the technology of particle control is a large problem. For this purpose, a divertor was added to the Tore Supra. In order to advance the research on particle control, it is necessary to examine the behavior of plasma in the peripheral part in detail. The measurement of peripheral plasma in tokamaks, beam probe spectroscopy, the Tore Supra tokamak, the progress of the joint research, the problems in the joint research and the perspective of hereafter are reported. (K.I.)

Integral torque balance in tokamaks

International Nuclear Information System (INIS)

Pustovitov, V.D.

2011-01-01

The study is aimed at clarifying the balance between the sinks and sources in the problem of intrinsic plasma rotation in tokamaks reviewed recently by deGrassie (2009 Plasma Phys. Control. Fusion 51 124047). The integral torque on the toroidal plasma is calculated analytically using the most general magnetohydrodynamic (MHD) plasma model taking account of plasma anisotropy and viscosity. The contributions due to several mechanisms are separated and compared. It is shown that some of them, though, possibly, important in establishing the rotation velocity profile in the plasma, may give small input into the integral torque, but an important contribution can come from the magnetic field breaking the axial symmetry of the configuration. In tokamaks, this can be the error field, the toroidal field ripple or the magnetic perturbation created by the correction coils in the dedicated experiments. The estimates for the error-field-induced electromagnetic torque show that the amplitude of this torque is comparable to the typical values of torques introduced into the plasma by neutral beam injection. The obtained relations allow us to quantify the effect that can be produced by the existing correction coils in tokamaks on the plasma rotation, which can be used in experiments to study the origin and physics of intrinsic rotation in tokamaks. Several problems are proposed for theoretical studies and experimental tests.

Tokamak and RFP ignition requirements

International Nuclear Information System (INIS)

Werley, K.A.

1991-01-01

A plasma model is applied to calculate numerically transport- confinement (nτ E ) requirements and steady-state operation tokamak. The CIT tokamak and RFP ignition conditions are examined. Physics differences between RFP and tokamaks, and their consequences for a DT ignition machine, are discussed. The ignition RFP, compared to a tokamak, has many physics advantages, including ohmic heating to ignition (no need for auxiliary heating systems), higher beta, low ignition current, less sensitivity of ignition requirements to impurity effects, no hard disruptions (associated with beta or density limits), and successful operation with high radiation fractions (f RAD ∼ 0.95). These physics advantages, coupled with important engineering advantages associated with lower external magnetic fields, larger aspect ratios, and smaller plasma cross sections translate into significant cost reductions for both ignition and power reactor. The primary drawback of the RFP is the uncertainty that the present confinement scaling will extrapolate to reactor regimes. The 4-MA ZTH was expected to extend the nτ E transport scaling data three order of magnitude above ZT-40M results, and if the present scaling held, to achieve a DT-equivalent scientific energy breakeven, Q=1. A basecase RFP ignition point is identified with a plasma current of 8.1 MA and no auxiliary heating. 16 refs., 4 figs., 1 tab

Different ELM regimes at ASDEX upgrade and their linear stability analysis

International Nuclear Information System (INIS)

Burckhart, Andreas O.

2013-01-01

Edge localised modes (ELMs) are magnetohydrodynamic (MHD) instabilities that occur at the edge of magnetically confined fusion plasmas. They periodically expel particles and energy from the confined region. In addition to limiting the confinement, they cause high heat fluxes to the walls of the tokamak which may not be manageable in larger, next-generation devices. However, the exact nature of the instabilities that drive ELMs is still unknown. The most commonly invoked theory to explain the occurrence of ELMs is the peeling-ballooning model which posits a critical edge pressure gradient and current density. In this thesis, this model is tested against experimental data gathered at the ASDEX Upgrade (AUG) tokamak. For the first time, a broad selection of ELM scenarios is analysed with respect to ideal MHD stability using the same methodology. The comparison of experiment and theory is performed using a stability analysis chain, which consists of combining kinetic and magnetic measurements to generate self-consistent plasma equilibria with the Grad-Shafranov solver CLISTE, refining this equilibrium with the HELENA code, and, finally, determining its stability using ILSA, a linear MHD stability code. In theory the peeling ballooning model should apply to all type-I ELM scenarios. Therefore, the stability of several different type-I ELMy H-mode plasmas is analysed with respect to peeling ballooning modes. While some of them are consistent with the model, in others ELMs occur well below or above the ideal MHD stability limit. The standard type-I ELMy H-mode regime exhibits considerable variations with equilibria both well below and at the stability limit depending on the discharge. In addition, a nitrogen-seeded case in which the edge pressure gradient greatly exceeds the stability limit is identified. In another discharge, the edge pressure gradient and current density, which are on the threshold for marginal stability, relax when edge heating is applied. Contrary to

Different ELM regimes at ASDEX upgrade and their linear stability analysis

Energy Technology Data Exchange (ETDEWEB)

Burckhart, Andreas O.

2013-12-16

Edge localised modes (ELMs) are magnetohydrodynamic (MHD) instabilities that occur at the edge of magnetically confined fusion plasmas. They periodically expel particles and energy from the confined region. In addition to limiting the confinement, they cause high heat fluxes to the walls of the tokamak which may not be manageable in larger, next-generation devices. However, the exact nature of the instabilities that drive ELMs is still unknown. The most commonly invoked theory to explain the occurrence of ELMs is the peeling-ballooning model which posits a critical edge pressure gradient and current density. In this thesis, this model is tested against experimental data gathered at the ASDEX Upgrade (AUG) tokamak. For the first time, a broad selection of ELM scenarios is analysed with respect to ideal MHD stability using the same methodology. The comparison of experiment and theory is performed using a stability analysis chain, which consists of combining kinetic and magnetic measurements to generate self-consistent plasma equilibria with the Grad-Shafranov solver CLISTE, refining this equilibrium with the HELENA code, and, finally, determining its stability using ILSA, a linear MHD stability code. In theory the peeling ballooning model should apply to all type-I ELM scenarios. Therefore, the stability of several different type-I ELMy H-mode plasmas is analysed with respect to peeling ballooning modes. While some of them are consistent with the model, in others ELMs occur well below or above the ideal MHD stability limit. The standard type-I ELMy H-mode regime exhibits considerable variations with equilibria both well below and at the stability limit depending on the discharge. In addition, a nitrogen-seeded case in which the edge pressure gradient greatly exceeds the stability limit is identified. In another discharge, the edge pressure gradient and current density, which are on the threshold for marginal stability, relax when edge heating is applied. Contrary to