Noninductive plasma initiation and startup in the DIII-D tokamak

Jackson, G.L.; Humphreys, D.A.; Hyatt, A.W.; Lohr, J.; Luce, T. C.; Yu, J. H.

doi:10.1088/0029-5515/51/8/083015

Cited by 26 publications

(39 citation statements)

References 21 publications

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“…We need quite a large power ≈ 140 kW to keep this small CFP current in the present case since the loss via the pitch angle diffusion is very fast due to the large hole of lost area even in the mid-plane segment, much larger loss in off mid-plane segments. This power is ≈ 10% of the power injected in the experiments [11,12] and seems to be rather reasonable when we take into account the same reasons as those mentioned in the previous section.…”

Section: Case In a Large Tokamaksupporting

confidence: 54%

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Open field equilibrium current and cross-field passing electrons as an initiator of a closed flux surface in EC-heated toroidal plasmas

Maekawa¹,

Yoshinaga²,

Uchida³

et al. 2012

Nucl. Fusion

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A modeling for the non-inductive initiation of a closed flux surface observed in electron cyclotron (EC) heated toroidal plasmas is presented. First, a pressure driven equilibrium toroidal current develops under a weak external vertical field so as to counter balance the pressure-ballooning and current-hoop forces. When the self field from the current almost cancels out the external vertical field, a forward energetic part of electrons in the velocity space begin to make cross field passing (CFP) orbits. The CFP electrons are generated by the EC heating of bulk electrons and subsequent pitch angle scattering, which is analyzed using the Fokker Planck equation. They provide an additional current that closes the filed lines. The model is examined for experiments in the small low aspect ratio device of LATE and in the large conventional device of JT-60U with a search for appropriate modes of EC heating. Simultaneous coincidence of the model with these two experiments is obtained in terms of microwave power and driven current. The results predict that initiation of closed flux surface requires more and more EC power as the plasma major radius increases. Especially, careful injection of high N ∥ EC waves are needed for large devices, both for initiation of a closed flux surface and for subsequent enlargement of the flux surface by usual EC current drive (ECCD) onto the closed flux area.

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Section: Case In a Large Tokamaksupporting

confidence: 54%

“…We examine the model described in sections 2-4 for experiments in large tokamaks of DIII-D and JT-60U [11,12].…”

Section: Case In a Large Tokamakmentioning

confidence: 99%

Open field equilibrium current and cross-field passing electrons as an initiator of a closed flux surface in EC-heated toroidal plasmas

Maekawa¹,

Yoshinaga²,

Uchida³

et al. 2012

Nucl. Fusion

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“…These features can be readily integrated into the development of low-loss 2212 conductors for fusion applications, such as the CS coils for fusion nuclear science facility (FNSF) and sustained high power density (SHPD) plasma startup that PPPL is developing. The outer diameter of the CS in a compact ST such as SHPD can be approximately 0.4 meter, due to the higher fraction of bootstrap currents in a compact Tokamak reactor design [24][25][26].…”

Section: Bi-2223 and Bi-2212mentioning

confidence: 99%

Low cost, simpler HTS cable conductors for fusion energy systems

Zhai

Otto

Zarnstorff

2022

IOP Conf. Ser.: Mater. Sci. Eng.

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The cost and complexity of large, high-field superconducting magnet modules and related subsystems comprise 30% to 60% of the fusion reactor core capital cost. The strategic plan for the U.S. burning plasma research, the Fusion Energy Sciences Committee Report (FESAC) “Power the Future: Fusion and Plasmas’', and 2021 NASEM report “Key Goals and Innovations needed for a U.S. Fusion Pilot Plant” recommends that the U.S. pursue innovative science and technology to enable construction of a Fusion Pilot Plant (FPP) that produces net electricity from fusion at reduced capital cost. To achieve this, a novel combination of lower-cost high temperature superconductors (HTS) in cable configurations with co-wound reinforcement for higher current density are being investigated using a simplified construction strategy to produce compact stable coils. They would be capable of generating 20 T at up to 10-20 K. Small-scale, inexpensive test coils and prototypes will help develop each feature and validate cabled conductor design models. The near term goal is to validate engineering approaches, scientific models and fabrication capabilities applicable to fusion reactor development such as U.S. fusion nuclear science facility (FNSF), sustained high-power density tokamak facility (SHPD) and FPP designs. The design options include lower-cost, high-strength, quench resistant REBCO or Bi-2212 cables in an all metal coil design that simplifies HTS coil construction and quench protection system, with co-wound reinforcements that integrate stress management in HTS cable design and provides thermal mass to help prevent quench damage.

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“…The issue of non-inductive start -up of conventional and spherical tokamaks has long history, for example using LH [8][9][10] and RF in the range of frequency of EC/EBW [11][12][13][14][15][16][17][18][19][20]. The method of coaxial helicity injection (CHI) relies on electrostatic helicity injection for initiating the plasma discharge [21].…”

Section: Introductionmentioning

confidence: 99%

Simulations towards the achievement of non-inductive current ramp-up and sustainment in the National Spherical Torus Experiment Upgrade

et al. 2015

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One of the goals of the National Spherical Torus Experiment Upgrade (NSTX-U) (Menard et al 2012 Nucl. Fusion 52 083015) is the demonstration of fully non-inductive start-up, current ramp-up and sustainment. This work discusses predictive simulations where the available heating and current drive systems are combined to maximize the non-inductive current and minimize the solenoidal contribution. Radio-frequency waves at harmonics higher than the ion cyclotron resonance (high-harmonic fast waves (HHFW)) and neutral beam injection are used to ramp the plasma current non-inductively starting from an initial Ohmic plasma. An interesting synergy is observed in the simulations between the HHFW and electron cyclotron (EC) wave heating. Time-dependent simulations indicate that, depending on the phasing of the HHFW antenna, EC wave heating can significantly increase the effectiveness of the radiofrequency power, by heating the electrons and increasing the current drive efficiency, thus relaxing the requirements on the level of HHFW power that needs to be absorbed in the core plasma to drive the same amount of fast-wave current.

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Noninductive plasma initiation and startup in the DIII-D tokamak

Cited by 26 publications

References 21 publications

Open field equilibrium current and cross-field passing electrons as an initiator of a closed flux surface in EC-heated toroidal plasmas

Open field equilibrium current and cross-field passing electrons as an initiator of a closed flux surface in EC-heated toroidal plasmas

Low cost, simpler HTS cable conductors for fusion energy systems

Simulations towards the achievement of non-inductive current ramp-up and sustainment in the National Spherical Torus Experiment Upgrade

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