Thermal pattern measurements during disruptions on TEXTOR-94

Ciotti, Marco; Denner, T.; Maruccia, G.; Finken, K. H.; Hobirk, J.; Kremer-Flecken, A.; Maddaluno, G.; Mank, G.; Pasqua, P. F.; Schüller, F. C.; Zanino, Roberto

doi:10.1016/s0022-3115(98)00559-5

Cited by 4 publications

(4 citation statements)

References 5 publications

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“…The surface of the interacting zone (S) is not known precisely. However, from previous results obtained in TEXTOR, it appears that in a limiter machine no broadening of thermal pattern is appearing during disruption [5]. S is comparable to the one observed during normal operation, S $ 1 m 2 .…”

Section: Erosion During Transient Processes As Disruptionsupporting

confidence: 70%

From eroded material to dust: An experimental evaluation of the mobilised dust production in Tore Supra

Grisolia

Rosanvallon

Loarte

et al. 2009

Journal of Nuclear Materials

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Section: Erosion During Transient Processes As Disruptionsupporting

confidence: 70%

From eroded material to dust: An experimental evaluation of the mobilised dust production in Tore Supra

Grisolia

Rosanvallon

Loarte

et al. 2009

Journal of Nuclear Materials

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“…plasma-facing surface in MAST [313], JET [314] and ASDEX Upgrade [315], that show evidence of disruption power fluxes outside the main divertor region. TEXTOR [316,317] is equipped with an IR camera monitoring the limiter. The evolution of the temperature and density profiles in the plasma SOL or, later on in the disruption, in the halo region, is usually not diagnosed directly (e.g.…”

Section: Plasma Energy Magnitudes and Thermal Energies At Disruptionmentioning

confidence: 99%

“…Routine IR monitoring of the firstwall (other than the portions in direct view of the divertor camera) is less common though there are recent developments in the IR measurements of an appreciable fraction the whole plasma-facing surface in MAST [313], JET [314] and ASDEX Upgrade [315], that show evidence of disruption power fluxes outside the main divertor region. TEXTOR [316,317] is equipped with an IR camera monitoring the limiter. The evolution of the temperature and density profiles in the plasma SOL or, later on in the disruption, in the halo region, is usually not diagnosed directly (e.g.…”

Section: Plasma Energy Magnitudes and Thermal Energies At Disruptionmentioning

confidence: 99%

Chapter 3: MHD stability, operational limits and disruptions

Hender¹,

Wesley²,

Bialek³

et al. 2007

Nucl. Fusion

1,000

947

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Progress in the area of MHD stability and disruptions, since the publication of the 1999 ITER Physics Basis document Nucl. Fusion 39 2137-2664, is reviewed. Recent theoretical and experimental research has made important advances in both understanding and control of MHD stability in tokamak plasmas. Sawteeth are anticipated in the ITER baseline ELMy H-mode scenario, but the tools exist to avoid or control them through localized current drive or fast ion generation. Active control of other MHD instabilities will most likely be also required in ITER. Extrapolation from existing experiments indicates that stabilization of neoclassical tearing modes by highly localized feedback-controlled current drive should be possible in ITER. Resistive wall modes are a key issue for S128 Chapter 3: MHD stability, operational limits and disruptions advanced scenarios, but again, existing experiments indicate that these modes can be stabilized by a combination of plasma rotation and direct feedback control with non-axisymmetric coils. Reduction of error fields is a requirement for avoiding non-rotating magnetic island formation and for maintaining plasma rotation to help stabilize resistive wall modes. Recent experiments have shown the feasibility of reducing error fields to an acceptable level by means of non-axisymmetric coils, possibly controlled by feedback. The MHD stability limits associated with advanced scenarios are becoming well understood theoretically, and can be extended by tailoring of the pressure and current density profiles as well as by other techniques mentioned here. There have been significant advances also in the control of disruptions, most notably by injection of massive quantities of gas, leading to reduced halo current fractions and a larger fraction of the total thermal and magnetic energy dissipated by radiation. These advances in disruption control are supported by the development of means to predict impending disruption, most notably using neural networks. In addition to these advances in means to control or ameliorate the consequences of MHD instabilities, there has been significant progress in improving physics understanding and modelling. This progress has been in areas including the mechanisms governing NTM growth and seeding, in understanding the damping controlling RWM stability and in modelling RWM feedback schemes. For disruptions there has been continued progress on the instability mechanisms that underlie various classes of disruption, on the detailed modelling of halo currents and forces and in refining predictions of quench rates and disruption power loads. Overall the studies reviewed in this chapter demonstrate that MHD instabilities can be controlled, avoided or ameliorated to the extent that they should not compromise ITER operation, though they will necessarily impose a range of constraints.

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“…The reduction of the convective heat load to the main limiter with increasing γ has been measured with the help of a fast thermographic camera, showing both a decrease of the averaged heat flux onto the limiter blades and of the peak load at the leading edges of the limiter. The heat flux pattern becomes more uniform [17]. The change in edge properties described so far leads to two further effects.…”

Section: B195mentioning

confidence: 82%

The influence of plasma-edge properties on high confinement discharges with a radiating plasma mantle at the tokamak TEXTOR-94

Unterberg

Messiaen

Ongena

et al. 1997

Plasma Phys. Control. Fusion

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The radiative improved mode obtained on the limiter tokamak TEXTOR-94 combines the possibility of power exhaust by a radiating plasma boundary (with a fraction of the radiated power with respect to the total input power up to 90% with neon or argon cooling) with improved energy confinement (as good as in the ELM-free H-mode in divertor tokamaks) at high plasma densities (line-averaged central-electron density equal to or even above the Greenwald density limit ) in quasi-stationary discharges. An overview is given of the substantial changes in plasma-edge properties occurring at high radiated power levels . These changes are characterized by a reduction of the plasma-edge density and temperature, a reduction of particle transport out of the confined plasma volume and an increase of the penetration depth of deuterium and impurity atoms. As a consequence, the particle confinement time increases and the electron-density profiles steepen. The transition to improved confinement takes place as soon as the density peaking reaches a critical threshold. An internal transport barrier is observed in the bulk of RI-mode plasmas (at ) characterized by an increase of the pressure gradient and of the shear of the toroidal velocity compared to discharges without additional impurity seeding. The dilution at the plasma boundary is strongly increased by the seeded impurities whereas the central dilution is only weakly affected.

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Thermal pattern measurements during disruptions on TEXTOR-94

Cited by 4 publications

References 5 publications

From eroded material to dust: An experimental evaluation of the mobilised dust production in Tore Supra

From eroded material to dust: An experimental evaluation of the mobilised dust production in Tore Supra

Chapter 3: MHD stability, operational limits and disruptions

The influence of plasma-edge properties on high confinement discharges with a radiating plasma mantle at the tokamak TEXTOR-94

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