Triggered Confinement Enhancement and Pedestal Expansion in High-Confinement-Mode Discharges in the National Spherical Torus Experiment

Maingi, R.; Bell, R.E.; Canik, J.; Gerhardt, S. P.; Kaye, S. M.; LeBlanc, B.; Osborne, T.H.; Bell, M.G.; Fredrickson, E. D.; Lee, K. C.; Ménard, J.; Park, J.-K.; Sabbagh, S. A.

doi:10.1103/physrevlett.105.135004

Cited by 44 publications

(50 citation statements)

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“…Further, such scenarios require H 98 =1.2-1.4 and would be limited to 1.5s duration by NBI ion dump operating limits. H 98 =1.3-1.4 has been obtained transiently in NSTX, but sustaining H 98 =1.15-1.2 is only now beginning to be achieved with Li conditioning [6] in ELM-free conditions in NSTX with a goal of extending this enhanced confinement to small-ELM regimes. With the addition of the 2nd NBI of the Upgrade, Figure 10c shows that higher NBI power (10MW vs. 7.5MW) can reduce the required confinement to H 98 =1.2 for 100% non-inductive scenarios and also enables control of q min with ∆q min = 0.6 by varying the NBI source mix at fixed density.…”

Section: B 2nd Neutral Beam Injectormentioning

confidence: 99%

Overview of the physics and engineering design of NSTX upgrade

et al. 2012

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Abstract-The spherical tokamak (ST) is a leading candidate for a fusion nuclear science facility (FNSF) due to its compact size and modular configuration. The National Spherical Torus eXperiment (NSTX) is a MA-class ST facility in the U.S. actively developing the physics basis for an ST-based FNSF. In plasma transport research, ST experiments exhibit a strong (nearly inverse) scaling of normalized confinement with collisionality, and if this trend holds at low collisionality, high fusion neutron fluences could be achievable in very compact ST devices. A major motivation for the NSTX Upgrade (NSTX-U) is to span the next factor of 3-6 reduction in collisionality. To achieve this collisionality reduction with equilibrated profiles, NSTX-U will double the toroidal field, plasma current, and NBI heating power and increase the pulse length from 1-1.5s to 5s. In the area of stability and advanced scenarios, plasmas with higher aspect ratio and elongation, high βN , and broad current profiles approaching those of an ST-based FNSF have been produced in NSTX using active control of the plasma β and advanced resistive wall mode control. High non-inductive current fractions of 70% have been sustained for many current diffusion times, and the more tangential injection of the 2nd NBI of the Upgrade is projected to increase the NBI current drive by up to a factor of 2 and support 100% non-inductive operation. More tangential NBI injection is also projected to provide non-solenoidal current ramp-up (from IP = 0.4MA up to 0.8-1MA) as needed for an ST-based FNSF. In boundary physics, NSTX and higher-A tokamaks measure an inverse relationship between the scrape-off layer heat-flux width and plasma current that could unfavorably impact nextstep devices. Recently, NSTX has successfully demonstrated very high flux expansion and substantial heat-flux reduction using a snowflake divertor configuration, and this type of divertor is incorporated in the NSTX-U design. The physics and engineering design supporting NSTX Upgrade are described.

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Section: B 2nd Neutral Beam Injectormentioning

confidence: 99%

Overview of the physics and engineering design of NSTX upgrade

et al. 2012

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“…For ITER, predictive modeling shows that fusion Q scales roughly with square of the pedestal pressure so that to fulfill its mission of high fusion gain (Q = 10), it is essential to achieve an H-mode with sufficiently high pedestal pressure. In NSTX, a very high confinement regime called the enhanced pedestal H-mode with confinement enhanced relative to the predictions of the ITER 1997 H-mode scaling up to H H98y,2 ∼ 1.7 has been observed following lithium application [18]. This level of confinement is sufficient for future ST devices such as a Fusion Nuclear Science Facility (FNSF) [19] or an ST Pilot Plant [20].…”

Section: H-mode Power Thresholdmentioning

confidence: 97%

Recent progress of NSTX lithium program and opportunities for magnetic fusion research

Ono

Bell

Kaita

et al. 2012

Fusion Engineering and Design

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“…In this section we discuss progress on research on the I-mode [19,[104][105][106][107][108][109], the quiescent H-mode [15,83,[110][111][112][113][114], and the Enhanced Pedestal (EP) H-mode [115][116][117]. Another candidate regime, Enhanced D α (EDA) H-mode [21,118], is discussed briefly in section 6 as studies [ 119] have shown that small ELMs appear in EDA H-mode at sufficiently high β.…”

Section: Naturally Occurring Quiescent Regimesmentioning

confidence: 99%

“…In this scenario, an ELM typically provides a trigger for a confinement bifurcation in which the pedestal T e and T i increase substantially, while the rate of line-average density ramping slows substantially or is arrested. The temporal evolution of a relatively long-lived EP H-mode [116] is shown in figure 30. The ELM trigger preceding the confinement transition is marked with a dashed vertical arrow.…”

Section: Ep H-modementioning

confidence: 99%

“…The local drag on toroidal rotation results in an increased rotational shear, which would reduce any residual H-mode turbulence. It is likely that the E × B shear is the critical quantity, but analysis [116] has shown that the changes in the E × B shear were dominated by the changes in the toroidal rotation profile, i.e. the changes in the poloidal rotation made a smaller contribution to the E × B shearing rate.…”

Section: Ep H-modementioning

confidence: 99%

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Eroshenko¹

2013

Phys.-Usp.

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Large edge localized modes (ELMs) typically accompany good H-mode confinement in fusion devices, but can present problems for plasma facing components because of high transient heat loads. Here the range of techniques for ELM control deployed in fusion devices is reviewed. Two strategies in the ITER baseline design are emphasized: rapid ELM triggering and peak heat flux control via pellet injection, and the use of magnetic perturbations to suppress or mitigate ELMs. While both of these techniques are moderately well developed, with reasonable physical bases for projecting to ITER, differing observations between multiple devices are also discussed to highlight the needed community R&D. In addition, recent progress in ELM-free regimes, namely quiescent H-mode, I-mode, and enhanced pedestal H-mode is reviewed, and open questions for extrapolability are discussed. Finally progress and outstanding issues in alternate ELM control techniques are reviewed: supersonic molecular beam injection, edge electron cyclotron heating, lower hybrid heating and/or current drive, controlled periodic jogs of the vertical centroid position, ELM pace-making via periodic magnetic perturbations, ELM elimination with lithium wall conditioning, and naturally occurring small ELM regimes.

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Triggered Confinement Enhancement and Pedestal Expansion in High-Confinement-Mode Discharges in the National Spherical Torus Experiment

Cited by 44 publications

References 20 publications

Overview of the physics and engineering design of NSTX upgrade

Overview of the physics and engineering design of NSTX upgrade

Recent progress of NSTX lithium program and opportunities for magnetic fusion research

Physics news on the Internet (based on electronic preprints)

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