Tokamak Power Exhaust with the Snowflake Divertor: Present Results and Outstanding Issues

Soukhanovskii, V.; Xu, X. Q.

doi:10.1007/s10894-015-9999-z

Cited by 3 publications

(5 citation statements)

References 31 publications

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“…The divertor heat flux width strongly decreased as I p increased in NSTX, DIII-D and C-Mod [41]. The snowflake divertor experiment in NSTX with P NBI = 4 MW, P SOL = 3 MW has demonstrated significant reduction in divertor heat flux (from 3-7 to 0.5-1 MW m −2 ) [27]. In TCV, the inverted AXUV radiation imaging has clearly shown the power redistribution to secondary strike points as the shape transitioned from a single-null configuration to a snowflake configuration.…”

Section: Reduction Of Heat Loadmentioning

confidence: 93%

“…The snowflake divertor for mitigation of the divertor heat flux has been reported from NSTX [27] and TCV [28]. Both experiments have succeeded in the redistribution of divertor heat flux and demonstrated significant reduction in divertor heat load.…”

Section: Reduction Of Heat Loadmentioning

confidence: 98%

“…Innovative divertor concepts: super-X and snowflake have been investigated in MAST [15], NSTX [27], TCV [28] and DIII-D [29]. The initial experiment of the closed helical divertor has been reported from LHD [30].…”

Section: Reduction Of Heat Loadmentioning

confidence: 99%

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Magnetic confinement experiments: plasma–material interactions, divertors, limiters, scrape-off layer (EX/D), stability (EX/S), wave–plasma interactions, current drive, heating, energetic particles (EX/W)

Yamada¹

2013

Nucl. Fusion

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This article summarizes the results presented at the 24th IAEA Fusion Energy Conference in the categories of plasma-material interactions, divertors, limiters, scrape-off layer (EX/D), stability (EX/S), wave-plasma interactions, current drive, heating, energetic particles (EX/W) in magnetic confinement experiments. In total, 149 papers including post-deadline papers have contributed to these categories. Several closely related papers, which are actually categorized in confinement (EX/C), have also been included. The understanding of experimental results has progressed remarkably, in particular, in the topics of Resonant Magnetic Perturbation and ITER-Like Wall, which are the highlight of this conference. At the same time, identification of the bridging mechanism between the actuator and the consequence still requires further dedicated efforts so as to provide more accurate and reliable extrapolations to ITER and DEMO.

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Section: Reduction Of Heat Loadmentioning

confidence: 93%

Section: Reduction Of Heat Loadmentioning

confidence: 98%

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Magnetic confinement experiments: plasma–material interactions, divertors, limiters, scrape-off layer (EX/D), stability (EX/S), wave–plasma interactions, current drive, heating, energetic particles (EX/W)

Yamada¹

2013

Nucl. Fusion

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“…In NSTX-U, the projected peak divertor heat fluxes can reach 20-40 MW m −2 with I p 2 MA, and P NBI 12 MW, with pulse length up to 5 s. NSTX-U will explore novel solutions to the boundary physics power exhaust challenge by testing the so-called 'snowflake' (SF) divertor configuration, and liquid metal PFCs to mitigate erosion and melting problems. In NSTX-U, two (upper and lower) sets of four divertor coils will be used to test up-downsymmetric snowflake (SF) divertors as shown in figure 20 [31]. In NSTX, a single lower SF divertor was investigated as seen in figure 6 [9].…”

Section: Novel Power Exhaust Solutionsmentioning

confidence: 99%

Progress toward commissioning and plasma operation in NSTX-U

et al. 2015

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The National Spherical Torus Experiment-Upgrade (NSTX-U) is the most powerful spherical torus facility at PPPL, Princeton USA. The major mission of NSTX-U is to develop the physics basis for an ST-based Fusion Nuclear Science Facility (FNSF). The ST-based FNSF has the promise of achieving the high neutron fluence needed for reactor component testing with relatively modest tritium consumption. At the same time, the unique operating regimes of NSTX-U can contribute to several important issues in the physics of burning plasmas to optimize the performance of ITER. NSTX-U further aims to determine the attractiveness of the compact ST for addressing key research needs on the path toward a fusion demonstration power plant (DEMO). The upgrade will nearly double the toroidal magnetic field BT to 1 T at a major radius of R0 = 0.93 m, plasma current Ip to 2 MA and neutral beam injection (NBI) heating power to 14 MW. The anticipated plasma performance enhancement is a quadrupling of the plasma stored energy and near doubling of the plasma confinement time, which would result in a 5–10 fold increase in the fusion performance parameter nτ T. A much more tangential 2nd NBI system, with 2–3 times higher current drive efficiency compared to the 1st NBI system, is installed to attain the 100% non-inductive operation needed for a compact FNSF design. With higher fields and heating powers, the NSTX-U plasma collisionality will be reduced by a factor of 3–6 to help explore the favourable trend in transport towards the low collisionality FNSF regime. The NSTX-U first plasma is planned for the Summer of 2015, at which time the transition to plasma operations will occur.

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“…Kotschenreuther et al [7] and Soukhanovskii and Xu [8] review efforts to understand and optimize the divertor configuration in order to reduce target heat loads while maintaining high confinement of the fusion core. Raman et al [9] describe deep particle fueling and momentum injection using compact tori (CT) and using off-axis current drive from electron Bernstein waves (EBW) in order to optimize the performance of the steadystate advanced tokamak and spherical tokamak.…”

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confidence: 99%

Preface to the Special Issue: Strategic Opportunities for Fusion Energy

et al. 2016

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The Journal of Fusion Energy provides a forum for discussion of broader policy and planning issues that play a crucial role in energy fusion programs. In keeping with this purpose and in response to several recent strategic planning efforts worldwide, this Special Issue on Strategic Opportunities was launched with the goal to invite fusion scientists and engineers to record viewpoints of the scientific opportunities and policy issues that can drive continued advancements in fusion energy research.

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Tokamak Power Exhaust with the Snowflake Divertor: Present Results and Outstanding Issues

Cited by 3 publications

References 31 publications

Magnetic confinement experiments: plasma–material interactions, divertors, limiters, scrape-off layer (EX/D), stability (EX/S), wave–plasma interactions, current drive, heating, energetic particles (EX/W)

Magnetic confinement experiments: plasma–material interactions, divertors, limiters, scrape-off layer (EX/D), stability (EX/S), wave–plasma interactions, current drive, heating, energetic particles (EX/W)

Progress toward commissioning and plasma operation in NSTX-U

Preface to the Special Issue: Strategic Opportunities for Fusion Energy

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