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

Mitarai, O.

doi:10.1088/0741-3335/41/12/304

Cited by 18 publications

(19 citation statements)

References 27 publications

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“…When the vertical field drives the plasma current, additional heating power is necessary to prevent the plasma current from shifting inward [2]. However, in this QUEST experiment, no additional heating power was applied during the vertical field current ramp-up phase.…”

Section: Smaller Ohmic Coil Bias Current Of I Cs = 5 Kamentioning

confidence: 99%

“…A new concept has been proposed for ramping up the plasma current using vertical field and heating power [2,3]; it was successfully demonstrated on the JT-60U tokamak [4][5][6] and the TST-2 [7] and MAST [8] STs. However, as steady state was not maintained in these experiments, an experimental demonstration of steady-state operation in an ST is necessary.…”

Section: Introductionmentioning

confidence: 99%

“…Experimental operating scenarios can be categorized as (1) pure ohmic discharge, (2) ohmic and RF current drive, (3) pure RF start-up to the steady state, and (4) vertical-field-assisted plasma current start-up and RF current drive to the steady state.…”

Section: Introductionmentioning

confidence: 99%

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First Ohmic Discharge Assisted with RF Power in QUEST Spherical Tokamak

Mitarai

Nakamura

Tashima

et al. 2011

Plasma and Fusion Research

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Ohmic plasma currents of up to 17 kA with a discharge duration of 0.32 s have been obtained in the Kyushu University Experiment with Steady-State Spherical Tokamak (QUEST) with the help of electron cyclotron wave (ECW) and cancellation coils (CCs). The CCs, originally installed to create a field null in the plasma breakdown phase, are essential for producing plasma current in QUEST. Although the ohmic coil current is initially biased and then reduced completely to zero to induce the plasma current in 15-20 ms, we demonstrate that the flat top of the plasma current exceeding 20 ms is maintained by the vertical field after the ohmic current is switched off. This type of operation is quite favorable for extending pulsed operation to the steady state by electron Bernstein wave current drive (EBCD).

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Section: Smaller Ohmic Coil Bias Current Of I Cs = 5 Kamentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

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First Ohmic Discharge Assisted with RF Power in QUEST Spherical Tokamak

Mitarai

Nakamura

Tashima

et al. 2011

Plasma and Fusion Research

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“…Такое приближение обычно используется и для расчёта подъёма тока, когда параметры плазмы меняются сильно. С помощью упрощённого нульмерного подхода пока-зано [3,4], что при достаточно большой величине  p (отношение тепловой энергии в плазме к энергии полоидального магнитного поля в плазме) может реализоваться режим индуктивного подъёма тока за счёт изменения вертикальных управляющих полей. При выключенном индукторе такой режим (режим «овердрайв») может реализоваться при условии, когда изменение потока от управляющих катушек Δψ Bv превышает изменение потока от плазмы плюс резистивные потери ψ res : Δψ Bv > L p ΔI pl + ψ res , где L p -полная индуктивность плазмы, а ΔI pl -изменение полного тока плазмы.…”

Section: Introductionunclassified

Inductive Current Ramp Up Analysis in NSTX and Tin-St Tokamaks

Dnestrovskij¹,

Lukash²,

Khayrutdinov³

2012

PAST-TF

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Ввод тока в плазму термоядерного источника нейтронов на основе сферического токамака (ТИН-СТ) [1], имеющего на стадии стационарного режима работы аспектное отношение 1,7, вытянутость 2,7 и треугольность 0,4, предполагает индукционный процесс за счёт использования магнитного потока, запасённого в полоидальной магнитной системе. В предлагаемой работе приводятся результаты анализа процесса ввода тока в плазму ТИН-СТ c помощью кода ДИНА [2] с использованием моделей равновесия с фиксированной и свободной границами плазмы. Используемая в коде ДИНА физическая модель тестируется по результатам экспериментов на токамаке NSTX. В работе на основании результатов моделирования сделан ряд выводов, касаю-щихся расхода полоидального магнитного потока на стадии ввода тока в плазму сферического токамака. Главным из этих вы-водов является утверждение о том, что источником полоидального магнитного потока на стадии ввода тока, кроме центрально-го соленоида, являются катушки полоидальной магнитной системы, отвечающие за изменение формы плазмы (Bv), которые обеспечивают до 50% потока на этой стадии.Ключевые слова: токамак, источник нейтронов, сценарий подъёма тока, равновесие плазмы со свободной границей. INDUCTIVE CURRENT RAMP UP ANALYSIS IN NSTX AND TIN-ST TOKAMAKS A.Yu. Dnestrovskij, V.E. Lukash., R.R. Khayrutdinov RRC «Kurchatov Institute» Moscow RussiaPlasma current ramp up in the spherical tokamak fusion neutron source (TIN-ST) with the aspect ratio 1.7, the plasma elongation 2.7 and the plasma triangularity 0.4 at the steady state suggests the usage of the magnetic flux supplied by the poloidal magnetic coils including the central solenoid. The results of inductive current ramp up analysis in TIN-ST tokamak [1] by free and fixed boundary equilibrium models using the DINA code [2] of are presented. The used model is tested by the modeling of the NSTX experiment. One of the main conclusion from the analysis is that the poloidal magnetic flux source from the control coils (Bv) can reach up to 50% of the total magnetic flux source Key words: tokamak, neutron source, current ramp up scenario, free boundary plasma equilibrium. ВВЕДЕНИЕВвод в плазму полоидального магнитного потока от центрального соленоида (CS) и от катушек управ-ления вертикальным магнитным полем обеспечивает индуктивный подъём тока плазмы. Напряжение на обходе плазменного шнура, обычно измеряемое в эксперименте, определяется изменением полного маг-нитного потока, включающего наряду с потоком от внешних катушек поток от плазменного тока. В то же время это напряжение задаёт граничные условия для решения уравнения проникновения полоидального магнитного поля в плазму, имеющего диффузионный вид. Кроме того, вертикальное магнитное поле, обеспечивающее равновесие плазменного шнура, в свою очередь, зависит от параметров плазмы, таких, как давление и ток. Таким образом, задача индуктивного подъёма тока в плазме должна самосогласованно включать решение задачи равновесия во внешних полях, определение напряжения на границе плазмы по изменению полного потока и расчёт проникновен...

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“…A new concept has been proposed for ramping up the plasma current using the vertical field and heating power [2,3], and was successfully demonstrated on the JT-60U tokamak [4][5][6]. However, in the JT-60U experiment, the inboard turns of the triangularity control coil acted as a part of the CS and contributed 20% of the total flux.…”

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confidence: 99%

Plasma Current Start-up by ECW and Vertical Field in the TST-2 Spherical Tokamak

Mitarai

Takase

Ejiri

et al. 2004

J. Plasma Fusion Res.

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Plasma current start-up and ramp-up to 10 kA have been demonstrated in the TST-2 spherical tokamak without the use of the central solenoid. Only the electron cyclotron wave (ECW) and the outer equilibrium field coils are used. The plasma current evolution depends on the poloidal coil arrangement. It is also demonstrated that the plasma current start-up can take place without the field null. Keywords:spherical tokamak, current start-up, central solenoid, ECW, vertical field Plasma current ramp-up has been recognized as a major difficulty in low aspect ratio spherical tokamaks [1] due to insufficient room for the central solenoid (CS) on the inboard side of the torus. A new concept has been proposed for ramping up the plasma current using the vertical field and heating power [2,3], and was successfully demonstrated on the JT-60U tokamak [4][5][6]. However, in the JT-60U experiment, the inboard turns of the triangularity control coil acted as a part of the CS and contributed 20% of the total flux. While the vertical field effect on plasma current rampup has also been demonstrated in a recent MAST experiment [7], such a result was obtained by a rapid ramp of special induction coils located inside the vacuum chamber (not by a slow ramp of the usual vertical field coils). In this paper we report the first experimental demonstration of plasma current start-up by the outer vertical field coils alone with ionization provided by ECW in the TST-2 spherical tokamak without the use of the central solenoid.In the present experiment, PF2, PF3, and PF5 coils (shown in Fig. 1) were used. Two sets of poloidal field coil arrangements were used: (1) PF2 and PF5 connected in series with a uni-polar power source and PF3 coil with a bi-polar power source, and (2) PF2 with a bi-polar power source and PF3 with a uni-polar power source.The launcher for ECW, designed for electron Bernstein wave (EBW) heating, at a frequency of 8.2 GHz is installed on the outboard side of the torus, below the midplane. The cut-off density for the O-mode is 0.83 × 10 18 m -3 . In TST-2, the major radius is R ≤ 0.38 m, the minor radius is a ≤ 0.25 m, the elongation can be κ ≤ 1.8, and the toroidal field is B t ≤ 0.3 T. The diagnostics available for this experiment were quite limited.The plasma current evolution (#302405) using the first coil arrangement is shown in Fig. 2. The initially positive PF3 coil current (creating an opposite vertical field) is reduced at 0.138 s and increases in the negative direction to provide the flux and the equilibrium field. The current through PF2 and PF5 coils starts at 0.142 s when the PF3 current has reached a plateau level. The plasma current starts at t = 0.139 s around the time I PF3 crosses zero, reaches 10 kA at 0.143 s, and then author's e-mail: omitarai@ktmail.ktokai-u.ac.jp Fig. 1 Schematic of the TST-2 plasma cross section and poloidal field coils.

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

Cited by 18 publications

References 27 publications

First Ohmic Discharge Assisted with RF Power in QUEST Spherical Tokamak

First Ohmic Discharge Assisted with RF Power in QUEST Spherical Tokamak

Inductive Current Ramp Up Analysis in NSTX and Tin-St Tokamaks

Plasma Current Start-up by ECW and Vertical Field in the TST-2 Spherical Tokamak

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