Plasma current and position feedback control in ADITYA Tokamak

Balakrishnan, V.; Gupta, C.N.; Sinha, R.K.

doi:10.1016/s0920-3796(03)00323-5

Cited by 9 publications

(5 citation statements)

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“…In a recent experimental schedule, repeatable plasma discharges with a maximum plasma current of ~160 kA with a duration of a quarter of a second with a plasma current flattop duration of ~140 ms have been obtained for the first time in the first Indian tokamak ADITYA. Discharge reproducibility has been improved considerably with lithium wall conditioning [1] and by precisely controlling the position of the plasma [2]. Improved discharges are attempted over a wider parameter range to carry out various confinement scaling experiments [3][4][5].…”

Section: Nuclear Fusionmentioning

confidence: 99%

“…2 e e i i p l rad (2) where I p is the plasma current, V l is the loop voltage, V is the plasma volume, and P rad is the total radiated power. The plasma volume is calculated using the plasma radius measured with the position probes.…”

Section: Energy Confinement Time Enhancementmentioning

confidence: 99%

“…The existing ADITYA with a limiter configuration is being upgraded into a divertor tokamak. Before upgrading ADITYA, several experiments, which are very much relevant even for large size tokamaks, including ITER, were carried out [1,2,6,7,16]. The outcomes of these innovative experiments…”

Section: Introductionmentioning

confidence: 99%

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Overview of recent experimental results from the Aditya tokamak

Tanna¹,

Ghosh²,

Chattopadhyay³

et al. 2017

Nucl. Fusion

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Several experiments, related to controlled thermonuclear fusion research and highly relevant for large size tokamaks, including ITER, have been carried out in ADITYA, an ohmically heated circular limiter tokamak. Repeatable plasma discharges of a maximum plasma current of ~160 kA and discharge duration beyond ~250 ms with a plasma current flattop duration of ~140 ms have been obtained for the first time in ADITYA. The reproducibility of the discharge reproducibility has been improved considerably with lithium wall conditioning, and improved plasma discharges are obtained by precisely controlling the position of the plasma. In these discharges, chord-averaged electron density ~3.0–4.0 × 1019 m−3 using multiple hydrogen gas puffs, with a temperature of the order of ~500–700 eV, have been achieved. Novel experiments related to disruption control are carried out and disruptions, induced by hydrogen gas puffing, are successfully mitigated using the biased electrode and ion cyclotron resonance pulse techniques. Runaway electrons are successfully mitigated by applying a short local vertical field (LVF) pulse. A thorough disruption database has been generated by identifying the different categories of disruption. Detailed analysis of several hundred disrupted discharges showed that the current quench time is inversely proportional to the q edge. Apart from this, for volt–sec recovery during the plasma formation phase, low loop voltage start-up and current ramp-up experiments have been carried out using electron cyclotron resonance heating (ECRH). Successful recovery of volt–sec leads to the achievement of longer plasma discharge durations. In addition, the neon gas puff assisted radiative improved confinement mode has also been achieved in ADITYA. All of the above mentioned experiments will be discussed in this paper.

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Section: Nuclear Fusionmentioning

confidence: 99%

Section: Energy Confinement Time Enhancementmentioning

confidence: 99%

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Overview of recent experimental results from the Aditya tokamak

Tanna¹,

Ghosh²,

Chattopadhyay³

et al. 2017

Nucl. Fusion

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“…In 1986, Pulse Power supplies were installed and commissioned by AEG Germany for the copper coils. These power supplies are the Ohmic Transformer power supply (OTPS), toroidal field power supply (TFPS), and vertical field power supply (VFPS) [4]. These were designed for the operating parameters of ADITYA, but when SST-1 was commissioned, it was decided to use the same OTPS and VFPS for SST-1's copper coils [5].…”

Section: Introductionmentioning

confidence: 99%

An approach for control of equilibrium field profile through the real-time plasma current in ADITYA-U and SST-1 tokamaks

Gupta,

Shah,

Makwana

et al. 2024

Preprint

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Pairs of external magnetic field coils placed outside the vessel maintain plasma equilibrium in tokamaks. The shape and position of the plasma column are determined and controlled by the temporal current profile in the external field coils. The temporal profile of the coil current mainly controls the horizontal movement of the plasma column, which keeps the plasma column at the vessel center. Hence, it is important to set and control the equilibrium profile according to the plasma current evolution and its profile. For this purpose, a controlled power supply is used to set the current in the coils as per the given reference profile. Currently, during plasma operation in the ADITYA-U and SST-1 tokamaks, a pre-programmed profile is used to generate the coil current profile, which is a passive form of control. For modern tokamak operations, however, real-time plasma control is essential. A novel technique has been devised for controlling the equilibrium field profile in ADITYA-U and SST-1 tokamaks using the real-time plasma current. To match the output current profile with the plasma current profile, a preconditioning system called Plasma Current Conditioning (PCC) is developed. To accomplish this, a MATLAB Simulink model of a 35-year-old power supply and a mathematical model of inverse control were developed, which generates the required reference profile in real time based on the current plasma profile. Using prior experimental data, the model has been exhaustively tested and validated.

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“…Similarly, application of modern controllers in new generation tokamaks like JIPP T-II [12], HT-1 [13], HL-1 [14], ,STOR-M [15], FTU [16], HAMANA-T [17], HYBTOK-II [18], EAST [19], J-TEXT [20] & ASDEX Upgrade ________________________ Email: rohit.kumar@ipr.res.in [21] which has also implemented real-time position control of plasma column. Application of the traditional PID controller was also proposed for the position control operation in ADITYA [22] tokamak and currently implemented in ADITYA-U tokamak to control the real-time horizontal position of plasma.…”

Section: Introductionmentioning

confidence: 99%