A Neutral Beam Injector Upgrade for NSTX

Stevenson, T.; McCormack, B.; Loesser, G. D.; Kalish, M.; Ramakrishnan, S.; Grisham, L.; Cropper, M.; Rossi, G.; Halle, A. von; Williams, M. D.

doi:10.2172/793023

Cited by 10 publications

(9 citation statements)

References 2 publications

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“…Beyond the new CS and exvessel structural enhancements, a 2nd neutral beam from TFTR has been chosen to provide the factor of 2 increase in auxiliary heating and current drive power for NSTX Upgrade, as this is presently the most mature and capable technology applicable to ST plasma parameters. The performance parameters of the NSTX neutral beams [105] are provided in figure 19 for reference. Of particular relevance is the decrease in achievable NBI pulse duration (set primarily by the thermal limits of the ion dumps) from 5 to 1.6 s as the beamline power is increased 50% from 5 to 7.5 MW.…”

Section: Nd 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: Nd Neutral Beam Injectormentioning

confidence: 99%

Overview of the physics and engineering design of NSTX upgrade

et al. 2012

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“…The major radius of typical plasmas is R 0 = 0.85 m, with aspect ratios R 0 /a = 1.35 to 1.55. The plasma currents are typically in the range 600 < I P (kA) < 1300, with toroidal fields between 0.35 and 0.55 T. Neutral beam heating with powers up to 7 MW, oriented parallel to the plasma current, is used for most highperformance plasmas, [97] and can be used for β N control [98]. High-harmonic fast wave (HHFW) heating at 30 MHz with powers up to ∼6 MW is also available [99].…”

Section: Introductionmentioning

confidence: 99%

Disruptions, disruptivity and safer operating windows in the high-βspherical torus NSTX

et al. 2013

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This paper discusses disruption rates, disruption causes and disruptivity statistics in the high-βN National Spherical Torus Experiment (NSTX) (Ono et al 2000 Nucl. Fusion 40 557). While the overall disruption rate is rather high, configurations with high βN, moderate q*, strong boundary shaping, sufficient rotation and broad pressure and current profiles are found to have the lowest disruptivity; active n = 1 control further reduces the disruptivity. The disruptivity increases rapidly for q* < 2.7, which is substantially above the ideal MHD current limit. Under quiescent conditions, qmin > 1.25 is generally acceptable for avoiding the onset of core rotating n = 1 kink/tearing modes; when EPM and ELM disturbances are present, the required qmin for avoiding those modes is raised to ∼1.5. The current ramp and early flat-top phase of the discharges are prone to n = 1 core rotating modes locking to the wall, leading to a disruption. Small changes to the discharge fuelling during this phase can often mitigate the rotation damping associated with these modes and eliminate the disruption. The largest stored-energy disruptions are those that occur at high current when a plasma current ramp-down is initiated incorrectly.

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“…The major radius of typical plasmas is R 0 =0.85 m, with aspect ratios R 0 /a=1.35 to 1.55. The plasma currents are typically in the range 600<I P [kA]<1300, with toroidal fields between 0.35 T and 0.55 T. Neutral beam heating with powers up to 7 MW, oriented parallel to the plasma current, is used for most high-performance plasmas, [97] and can be used for β N control [98]. High-harmonic fast wave heating at 30 MHz with powers up to ~6 MW is also available [99].…”

Section: : Introductionmentioning

confidence: 99%