Disruption mitigation with high-pressure noble gas injection

Whyte, D.G.; Jernigan, T.C.; Humphreys, D. A.; Hyatt, A.W.; Lasnier, C.J.; Parks, P. B.; Evans, T. E.; Taylor, P.L.; Kellman, A.G.; Gray, D.S.; Hollmann, E. M.

doi:10.1016/s0022-3115(02)01525-8

Cited by 82 publications

(132 citation statements)

References 11 publications

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“…In some cases, that structure has been observed to rotation toroidally [28]. Given these results, it is important to note that the reduction in halo currents with "Massive Gas Injection" (MGI) [50][51][52][53][54][55][56][57][58][59] has been documented in ALCATOR C-Mod [53], DIII-D [51], JET [58], and ASDEX-Upgrade [56].…”

Section: : Introduction and Backgroundmentioning

confidence: 99%

Characterization of disruption halo currents in the National Spherical Torus Experiment

Gerhardt¹,

Ménard²,

Sabbagh³

et al. 2012

Nucl. Fusion

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This paper describes the general characteristics of disruptions halo currents in the National Spherical Torus Experiment [M. Ono, et al. Nuclear Fusion 40, 557 (2000)].The commonly observed types of vertical motion and resulting halo current patterns are described, and it is shown that plasma discharges developing between components can facilitate halo current flow. The halo current fractions and toroidal peaking factors at various locations in the device are presented. The maximum product of these two metrics for localized halo current measurements is always significantly less than the worst-case expectations from conventional aspect ratio tokamaks (which are typically written in terms of the total halo current). The halo current fraction and impulse is often largest in cases with the fastest plasma current quenches and highest quench rates. The effective duration of the halo current pulse is comparable to or shorter than the plasma current quench time. The largest halo currents have tended to occur in lower β and lower elongation plasmas. The sign of the poloidal halo current is reversed when the toroidal field direction is reversed. PACS: 52.55. Ez, 52.55.Tn

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Section: : Introduction and Backgroundmentioning

confidence: 99%

Characterization of disruption halo currents in the National Spherical Torus Experiment

Gerhardt¹,

Ménard²,

Sabbagh³

et al. 2012

Nucl. Fusion

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“…In this paper we discuss suppression of REs by massive gas injection (MGI) in TEXTOR. MGI has an advantage that it can also mitigate heat loads and halo currents [10,11,12]. First we briefly summarize established results of the theory of runaway electrons.…”

Section: Introductionmentioning

confidence: 99%

Generation and suppression of runaway electrons in disruption mitigation experiments in TEXTOR

Bozhenkov¹,

Lehnen²,

Finken³

et al. 2008

Plasma Phys. Control. Fusion

122

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Abstract. Runaway electrons represent a serious problem for the reliable operation of the future experimental tokamak ITER. Due to the multiplication factor of exp(50) in the avalanche even a few seed runaway electrons will result in a beam of high energetic electrons that is able to damage the machine. Thus suppression of runaway electrons is a task of high importance, for which reason we present here a systematic study of runaway electrons following massive gas injection in TEXTOR. Argon injection can cause generation of runaways carrying up to 30% of the initial plasma current, while disruptions triggered by injection of helium or of mixtures of argon (5, 10, 20%) with deuterium are runaway free. Disruptions caused by argon injection finally become runaway free for very large amounts of injected atoms. The appearance/absence of runaway electrons is related to the fraction of atoms delivered to the plasma center. This so called mixing efficiency is deduced from a 0D model of the current quench. The estimated mixing efficiency is: 3% for argon, 15% for an argon/deuterium mixture and about 40% for helium. A low mixing efficiency of high-Z impurities can have a strong implication for the design of the disruption mitigation system for ITER. However, a quantitative prediction requires a better understanding of the mixing mechanism.

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“…Experiments on DIII-D have demonstrated that the impact of disruptions on first wall or in-vessel components can be greatly reduced by the injection of noble gas using a high pressure gas jet [21][22][23]. Modelling indicates that this technique extrapolates well to future burning plasmas experiments.…”

Section: Disruption Mitigationmentioning

confidence: 99%

“…Our work on the quiescent H-mode (QH-mode) demonstrates a solution to the pulsed divertor heat load in future burning plasma devices caused by edge localized modes (ELM) [17,19,20]. In addition, we have demonstrated disruption mitigation using a massive gas puff of either neon or argon [21][22][23]. This substantially reduces thermal and mechanical vessel loading while simultaneously suppressing runaway electrons in the current quench.…”

Section: Introductionmentioning

confidence: 99%

Overview of recent experimental results from the DIII-D advanced tokamak program

Team

2003

Nucl. Fusion

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The DIII-D research program is developing the scientific basis for advanced tokamak (AT) modes of operation in order to enhance the attractiveness of the tokamak as an energy producing system. Since the last international atomic energy agency (IAEA) meeting, we have made significant progress in developing the building blocks needed for AT operation: (1) we have doubled the magnetohydrodynamic (MHD) stable tokamak operating space through rotational stabilization of the resistive wall mode; (2) using this rotational stabilization, we have achieved β N H 89 10 for 4τ E limited by the neoclassical tearing mode (NTM); (3) using real-time feedback of the electron cyclotron current drive (ECCD) location, we have stabilized the (m, n) = (3, 2) NTM and then increased β T by 60%; (4) we have produced ECCD stabilization of the (2, 1) NTM in initial experiments; (5) we have made the first integrated AT demonstration discharges with current profile control using ECCD; (6) ECCD and electron cyclotron heating (ECH) have been used to control the pressure profile in high performance plasmas; and (7) we have demonstrated stationary tokamak operation for 6.5 s (36τ E ) at the same fusion gain parameter of β N H 89 /q 2 95 ∼ = 0.4 as ITER but at much higher q 95 = 4.2. We have developed general improvements applicable to conventional and AT operating modes: (1) we have an existence proof of a mode of tokamak operation, quiescent H-mode, which has no pulsed, edge localized modes (ELM) heat load to the divertor and which can run for long periods of time (3.8 s or 25τ E ) with constant density and constant radiated power; (2) we have demonstrated real-time disruption detection and mitigation for vertical disruption events using high pressure gas jet injection of noble gases; (3) we have found that the heat and particle fluxes to the inner strike points of balanced, double-null divertors are much smaller than to the outer strike points. We have made detailed investigations of the edge pedestal and scrape-off layer (SOL): (1) atomic physics and plasma physics both play significant roles in setting the width of the edge density barrier in H-mode; (2) ELM heat flux conducted to the divertor decreases as density increases; (3) intermittent, bursty transport contributes to cross field particle transport in the SOL of H-mode and, especially, L-mode plasmas.

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Disruption mitigation with high-pressure noble gas injection

Cited by 82 publications

References 11 publications

Characterization of disruption halo currents in the National Spherical Torus Experiment

Characterization of disruption halo currents in the National Spherical Torus Experiment

Generation and suppression of runaway electrons in disruption mitigation experiments in TEXTOR

Overview of recent experimental results from the DIII-D advanced tokamak program

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