High non-inductive fraction H-mode discharges generated by high-harmonic fast wave heating and current drive in the National Spherical Torus Experiment

Taylor, G.; Hosea, J.; Kessel, C. E.; LeBlanc, B.; Mueller, D.; Phillips, C. K.; Valeo, E. J.; Wilson, J. R.; Ryan, P. M.; Bonoli, P. T.; Wright, J. C.; Harvey, R. W.

doi:10.1063/1.3699364

Cited by 24 publications

(11 citation statements)

References 25 publications

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“…Further, edge losses in the scrape-off-layer to the divertor [84][85][86][87] can degrade HHFW core coupling efficiency, and parasitic absorption by NBI fast-ions [88][89][90] can compete with thermal electron heating and current drive. For these reasons, HHFW appears most applicable to heating and driving current in low-current high-bootstrap-fraction plasmas [91] serving as targets for subsequent non-inductive current rampup through other means. Electron Bernstein waves (EBW) are another potentially very attractive wave for heating and driving current in over-dense plasma conditions [92][93][94], but the necessity for precise tailoring of the edge density gradient [95] to maximize the double-mode-conversion efficiency combined with other loss mechanisms in the plasma edge [96,97] have thus far made efficiently coupling to the EBW operationally challenging.…”

Section: Energy Confinementmentioning

confidence: 99%

Fusion nuclear science facilities and pilot plants based on the spherical tokamak

et al. 2016

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Section: Energy Confinementmentioning

confidence: 99%

Fusion nuclear science facilities and pilot plants based on the spherical tokamak

et al. 2016

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“…Present NSTX research is pursuing noninductive formation of plasma current using CHI [140] to form a closed-flux plasma of 0.2-0.3 MA to be heated and sustained by high-harmonic fast-waves in a high bootstrapcurrent-fraction H-mode plasma. Promising HHFW heating and current drive results have recently been obtained [141] in low-current (I P = 300 kA) ohmic target plasmas being developed as plasma current ramp-up targets for NBI. In these plasmas, an HHFW-induced H-mode with a central electron temperature of up to 3 keV was achieved with 65% noninductive current drive (f BS = 43%, f RFCD = 22%) with 1.4 MW of injected HHFW power.…”

Section: Overview Of Non-inductive Current Formation and Ramp-upmentioning

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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“…The computed bootstrap current varies from 100 to 230 kA. The current generated directly by HHFW power was generated inside a normalized minor radius ∼0.2, and 75% of the non-inductive current was generated inside a normalized minor radius ∼0.4 [135]. Over the entire range of NBI heated plasmas, up to 65% NICF was experimentally reached (computed by TRANSP), peaking at plasma current value of I p = 0.7 MA (figure 31).…”

Section: Non-inductive Current By Nbi and Rfmentioning

confidence: 99%

Overview of physics results from the conclusive operation of the National Spherical Torus Experiment

Sabbagh¹,

Ahn²,

Allain³

et al. 2013

Nucl. Fusion

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Research on the National Spherical Torus Experiment, NSTX, targets physics understanding needed for extrapolation to a steady-state ST Fusion Nuclear Science Facility, pilot plant, or DEMO. The unique ST operational space is leveraged to test physics theories for next-step tokamak operation, including ITER. Present research also examines implications for the coming device upgrade, NSTX-U. An energy confinement time, τ E , scaling unified for varied wall conditions exhibits a strong improvement of B T τ E with decreased electron collisionality, accentuated by lithium (Li) wall conditioning. This result is consistent with nonlinear microtearing simulations that match the experimental electron diffusivity quantitatively and predict reduced electron heat transport at lower collisionality. Beam-emission spectroscopy measurements in the steep gradient region of the pedestal indicate the poloidal correlation length of turbulence of about ten ion gyroradii increases at higher electron density gradient and lower T i gradient, consistent with turbulence caused by trapped electron instabilities. Density fluctuations in the pedestal top region indicate ion-scale microturbulence compatible with ion temperature gradient and/or kinetic ballooning mode instabilities. Plasma characteristics change nearly continuously with increasing Li evaporation and edge localized modes (ELMs) stabilize due to edge density gradient alteration. Global mode stability studies show stabilizing resonant kinetic effects are enhanced at lower collisionality, but in stark contrast have almost no dependence on collisionality when the plasma is off-resonance. Combined resistive wall mode radial and poloidal field sensor feedback was used to control n = 1 perturbations and improve stability. The disruption probability due to unstable resistive wall modes (RWMs) was surprisingly reduced at very high β N /l i > 10 consistent with low frequency magnetohydrodynamic spectroscopy measurements of mode stability. Greater instability seen at intermediate β N is consistent with decreased kinetic RWM stabilization. A model-based RWM state-space controller produced long-pulse discharges exceeding β N = 6.4 and β N /l i = 13. Precursor analysis shows 96.3% of disruptions can be predicted with 10 ms warning and a false positive rate of only 2.8%. Disruption halo currents rotate toroidally and can have significant toroidal asymmetry. of this phenomenon in designing future RF systems. The snowflake divertor configuration enhanced by radiative detachment showed large reductions in both steady-state and ELM heat fluxes (ELMing peak values down from 19 MW m −2 to less than 1.5 MW m −2 ). Toroidal asymmetry of heat deposition was observed during ELMs or by 3D fields. The heating power required for accessing H-mode decreased by 30% as the triangularity was decreased by moving the X-point to larger radius, consistent with calculations of the dependence of E × B shear in the edge region on ion heat flux and X-point radius. Co-axial helicity injection reduced the induct...

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High non-inductive fraction H-mode discharges generated by high-harmonic fast wave heating and current drive in the National Spherical Torus Experiment

Cited by 24 publications

References 25 publications

Fusion nuclear science facilities and pilot plants based on the spherical tokamak

Fusion nuclear science facilities and pilot plants based on the spherical tokamak

Overview of the physics and engineering design of NSTX upgrade

Overview of physics results from the conclusive operation of the National Spherical Torus Experiment

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