The ARIES Advanced and Conservative Tokamak Power Plant Study

Kessel, C.; Tillack, M. S.; Najmabadi, F.; Poli, F. M.; Ghantous, K.; Gorelenkov, N. N.; Wang, X. R.; Navaei, D.; Toudeshki, H. H.; Koehly, C.; El-Guebaly, L.; Blanchard, James P.; Martin, Carl J.; Mynsburge, L.; Humrickhouse, Paul W.; Rensink, M.E.; Rognlien, T. D.; Yoda, Minami; Abdel‐Khalik, S. I.; Hageman, Mitchell D.; Mills, Brantley; Rader, Jordan D.; Sadowski, D. L.; Snyder, P. B.; John, H. St.; Turnbull, A. D.; Waganer, Lester M.; Malang, S.; Rowcliffe, A.F.

doi:10.13182/fst14-794

Cited by 50 publications

(31 citation statements)

References 2 publications

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“…, significantly exceed those envisioned in recent ARIES-ACT steady-state fusion reactor studies (β N ~ β P ~ 2.5, β T~1 .5%) [34]. The large values of β N and energy confinement quality achieved in these experiments compensate for the low plasma current relative to the magnetic field, i.e.…”

Section: Discussionmentioning

confidence: 52%

Compatibility of internal transport barrier with steady-state operation in the high bootstrap fraction regime on DIII-D

Garofalo¹,

Gong²,

Grierson³

et al. 2015

Nucl. Fusion

117

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Recent EAST/DIII-D joint experiments on the high poloidal beta tokamak regime in DIII-D have demonstrated fully noninductive operation with an internal transport barrier (ITB) at large minor radius, at normalized fusion performance increased by ⩾30% relative to earlier work (Politzer et al 2005 Nucl. Fusion 45 417). The advancement was enabled by improved understanding of the 'relaxation oscillations', previously attributed to repetitive ITB collapses, and of the fast ion behavior in this regime. It was found that the 'relaxation oscillations' are coupled core-edge modes amenable to wall-stabilization, and that fast ion losses which previously dictated a large plasma-wall separation to avoid wall over-heating, can be reduced to classical levels with sufficient plasma density. By using optimized waveforms of the plasma-wall separation and plasma density, fully noninductive plasmas have been sustained for long durations with excellent energy confinement quality, bootstrap fraction ⩾80%, β ⩽ 4 N , β ⩾ 3 P , and β ⩾ % 2 T . These results bolster the applicability of the high poloidal beta tokamak regime toward the realization of a steady-state fusion reactor.

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Section: Discussionmentioning

confidence: 52%

Compatibility of internal transport barrier with steady-state operation in the high bootstrap fraction regime on DIII-D

Garofalo¹,

Gong²,

Grierson³

et al. 2015

Nucl. Fusion

117

View full text Add to dashboard Cite

show abstract

“…Japan proposes a more advanced SlimCS device [18], while South Korea targets its KSTAR program on the K-DEMO device [19,20]. In the United States, the original ARIES design [9] has been updated with various 'Advanced and Conservative Tokamak' (ACT) versions [21], while proposals are also made for more compact lower power designs with ARC [22] and ST Pilot Plant [23] facilities. In addition, the AT forms the basis for various proposed fusion nuclear science testing facilities such as the Fusion Development Facility (FDF) [24] and Fusion Nuclear Science Facility (FNSF) [25][26][27] in the United States, and the China Fusion Engineering Test Reactor (CFETR) [28].…”

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confidence: 99%

DIII-D Research to Prepare for Steady State Advanced Tokamak Power Plants

et al. 2018

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We review progress made on the advanced tokamak path to fusion energy by the DIII-D National Fusion Facility (Luxon et al. in Nucl Fusion 42:614, 2002). The advanced tokamak represents a highly attractive approach for a future steady state fusion power plant. In this concept, there is a natural alignment between high pressure operation, favorable stability and transport properties, and a highly self-driven ('bootstrap') plasma current to sustain operation efficiently and without disruptions. Research on DIII-D has identified several promising plasma configurations for fully non-inductive operation with potential applications to a range of future devices, from ITER to nuclear science facilities, to compact or large scale fusion power plants. Significant progress has been made toward realizing these scenarios, with the demonstration of high b access, off-axis current drive techniques, model based profile control, and stability and ELM control in reactor relevant physics regimes. Radiative techniques have also been pioneered to develop improved compatibility with divertor requirements, and simultaneous access to high performance pedestals. Research has also developed major advances in physics understanding, validating concepts of kinetic damping of ideal MHD instabilities that enable high b operation, identifying how current profile and b influence plasma turbulence in order to validate and improve turbulent transport models, and understanding the physics of energetic particle redistribution due to Alfvénic and other instabilities. These advances have been partnered with development of a rigorous integrated modeling framework used to interpret and validate individual physics models of the various aspects of plasma behavior, and to guide development of improved regimes and upgrades. These tools are also being used to develop and validate concepts for future reactors directly. Having established these foundations, DIII-D is now undergoing a substantial upgrade to raise power, current drive, electron heating and 3-D field capabilities in order to validate this physics and test conceptual solutions in reactor-relevant physics regimes, with a goal to resolve the key scientific and technology questions to enable a decision on a future steady state fusion power plant.

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“…However, a HFS antenna integrated into the blanket module has less significant LOS given that it is in a low neutron flux region. The thickness of a HFS blanket module could increase slightly (5-10 cm) to compensate for void of waveguide between the plasma facing armor and the breeding material, although some blanket designs such as those for the recent ARIES-ACT1 and ACT2 studies [27] already include a similarly sized void between the plasma facing surface and the breeding material. The ARIES-ACT1 blanket design includes inner and outer interlocking "C" shaped blanket modules with a radial gap between the blankets for divertor pumping.…”

Section: Conceptual Engineering Assessmentmentioning

confidence: 99%