The Spherical Torus Approach to Magnetic Fusion Development

Stambaugh, R.D.; Chan, V. S.; Anderson, P.A.; Baxi, C.B.; Callis, R.W.; Chiu, H.K.; Forest, C.B.; Rao, Hong; Jensen, Trond; Lao, L. L.; Leuer, J.A.; Mahdavi, M. A.; Miller, R.L.; Nerem, A.; Prater, R.; Politzer, Peter; Schaffer, M. J.; Sevier, D.L.; Taylor, T. S.; Turnbull, A. D.; Wong, C.P.C.

doi:10.13182/fst96-a11963141

Cited by 11 publications

(5 citation statements)

References 13 publications

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“…The spherical tokamak in comparison with the conventional tokamak can operate at high beta (β ≈ 1) while its aspect ratio is also smaller [17,18]. By evacuating all non-essential components such as inboard blanket or shield, OH heating solenoid and inboard poloidal coil from the inner side of the plasma, the size of a tokamak power core can be minimized [19,20]. Research on spherical tokamak has shown the importance of shape and elongation as critical issues.…”

Section: Introductionmentioning

confidence: 99%

“…The plausibility of STs fueled with D-3 He has been specified [19,34] hence, we explore an identified ST's prospects in details with the consideration of side D-D reactions. In this paper in section 2, we have utilized the scaling laws of density, beta ratio, energy confinement time IPB98y2 and with the use of power balance equation, obtained the fusion gain and power of D-3 He fuel in ST.…”

Section: Introductionmentioning

confidence: 99%

See 1 more Smart Citation

Study of D-³He fusion fuel parameters sensitivity in spherical tokamak

Sharifi¹,

Motevalli²,

Fadaei³

2021

Phys. Scr.

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Spherical tokamak assesses the potential of great performance in high beta and is capable of steadystate operation. Controlling plasma parameters and profile could lead to a high beta for spherical tokamaks. In this paper, we used the scaling laws of density, beta ratio, and energy confinement time with D-3He fuel. We investigated the dependency of Q on confinement enhancement factor and fuel density ratio of D-3He by plasma power balance equation in spherical tokamak (ST) which H y2 ≈ 1.9 and f D3 = 0.84 would lead to Q = 5.5 and high power production about 1.6 GW in the 65 keV. Hot ion mode as an imperative circumstance in ST has been investigated and we have illustrated that the convenient ion temperature is around 60–70 keV and γ ≈ 0.15–0.3 in order to enhance the operation of ST and restrict radiation loss.

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Section: Introductionmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

Study of D-³He fusion fuel parameters sensitivity in spherical tokamak

Sharifi¹,

Motevalli²,

Fadaei³

2021

Phys. Scr.

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“…In this limit the plasma configuration deviates strongly from that of a conventional tokamak. Experimental studies of near-unity aspect ratio plasmas have been called for in several discussions [1][2][3] due to the projection of very high stable β t (≡2µ 0 nkT /B 2 t0,vac ) in the Troyon limit [4] and the desire to explore the limits of ST operational space.…”

Section: Introductionmentioning

confidence: 99%

“…The Pegasus Toroidal Experiment [5] is an extremely low aspect ratio toroidal facility developed to explore quasi- 3 Author to whom any correspondence should be addressed. spherical high-pressure plasmas by minimizing the central column while maintaining good confinement and stability.…”

Section: Introductionmentioning

confidence: 99%

The upgraded Pegasus Toroidal Experiment

et al. 2006

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The Pegasus Toroidal Experiment was developed to explore the physics limits of plasma operation as the aspect ratio (A) approaches unity. Initial experiments on the device found that access to high normalized current and toroidal beta was limited by the presence of large-scale tearing modes. Major upgrades have been conducted of the facility to provide the control tools necessary to mitigate these resistive modes. The upgrades include new programmable power supplies, new poloidal field coils and increased, time-variable toroidal field. First ohmic operations with the upgraded system demonstrated position and current ramp-rate control, as well as improvement in ohmic flux consumption from 2.9 MA Wb −1 to 4.2 MA Wb −1 . The upgraded experiment will be used to address three areas of physics interest. First, the kink and ballooning stability boundaries at low A and high normalized current will be investigated. Second, clean, high-current plasma sources will be studied as a helicity injection tool. Experiments with two such sources have produced toroidal currents three times greater than predicted by geometric field line following. Finally, the use of electron Bernstein waves to heat and drive current locally will be studied at the 1 MW level; initial modelling indicates that these experiments are feasible at a frequency of 2.45 GHz.

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“…The spherical tokamak concept proposed by Dr Peng has been conceived as a one of the candidates of an economical fusion reactor (Peng and Stricker 1986). Since successful operation of the START experiment (Sykes et al 1992, Robinson 1995, Gryaznevich et al 1998 reactor studies on a spherical tokamak (Peng et al 1996, Robinson 1995, Stambaugh et al 1996, Miller 1998) with a small aspect ratio have been actively conducted. Recently, the second generation of medium-sized spherical tokamak experiments have just begun or are under construction such as MAST (Robinson et al 1996), NSTX (Spitzer et al 1996), PEGASUS (Thorson et al 1998), GLOBUS-M (Kavin et al 1997) and TST-M (Toyama et al 1998).…”

Section: Introductionmentioning

confidence: 99%

Inductive plasma current start-up by the outer vertical field coil in a spherical tokamak

Mitarai

1999

Plasma Phys. Control. Fusion

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Plasma current-start up induced by an outer vertical field coil is studied during the ignition access phase in a spherical tokamak reactor. We have illustrated the concept that the plasma current of ∼50 MA could be induced by the outer vertical field coil in the proposed spherical tokamak with the help of the small central solenoid flux of ±5 V s and the strong heating power less than 100 MW for the internal inductance of i ∼ 0.4-0.8 without the help of bootstrap current and non-inductive current drive power. The required condition to achieve this operation scenario is that the flux produced by the equilibrium vertical field is larger than the inductive flux. Current start-up operation is achieved by adding the small ohmic heating solenoid flux for the flux waveform adjustment because the flux from the outer vertical field coil cannot solely induce the desired plasma current waveform in the case of the preprogramming of the heating power.

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The Spherical Torus Approach to Magnetic Fusion Development

Cited by 11 publications

References 13 publications

Study of D-³He fusion fuel parameters sensitivity in spherical tokamak

Study of D-³He fusion fuel parameters sensitivity in spherical tokamak

The upgraded Pegasus Toroidal Experiment

Inductive plasma current start-up by the outer vertical field coil in a spherical tokamak

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The Spherical Torus Approach to Magnetic Fusion Development

Cited by 11 publications

References 13 publications

Study of D-3He fusion fuel parameters sensitivity in spherical tokamak

Study of D-3He fusion fuel parameters sensitivity in spherical tokamak

The upgraded Pegasus Toroidal Experiment

Inductive plasma current start-up by the outer vertical field coil in a spherical tokamak

Contact Info

Product

Resources

About

Study of D-³He fusion fuel parameters sensitivity in spherical tokamak

Study of D-³He fusion fuel parameters sensitivity in spherical tokamak