ECRH assisted plasma experiments on tokamaks SST-1 and Aditya

Shukla, B. K.; Bora, D.; Jha, Ratneshwar; Pradhan, Subrata; Ghosh, Joydeep; Gupta, C.N.; Patel, Jatin; Babu, Rajan; Patel, Harshida; Dhorajia, Pragnesh; Tanna, R.L.; Tanna, V. L.; Atrey, P.K.; Joisa, S.; Bhatt, S. B.; Purohit, D.; Chattopadhyaya, P. K.; Raju, D.; Patel, Paresh; Manchanda, R.; Gupta, Manoj Kumar

doi:10.1109/sofe.2015.7482270

Cited by 4 publications

(3 citation statements)

References 5 publications

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“…Similar experiments have been carried out in ADITYA. The substantial reduction in breakdown voltage with ECR, as low as ~7 V is achieved as compared to ~20 V, typically required without RF [16].…”

Section: Nuclear Fusionmentioning

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%

“…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%

Overview of recent experimental results from the Aditya tokamak

Tanna¹,

Ghosh²,

Chattopadhyay³

et al. 2017

Nucl. Fusion

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“…The first discharge in ADITYA-U was obtained on December 1, 2016. For several years now, nearly all the experimental work conducted on tokamaks throughout the world-including DIII-D [4] in the USA, JET [5,6], ASDEX [7] and FTU [8] in the EU, JT60 [9,10] in Japan, KSTAR [11] and VEST [12] in Korea, ADITYA and SST-1 [13,14] in India, HL-2A [15], EAST [16] and J-TEXT [17] in China-has been to investigate the optimum operational parameters in the start-up phase. Understanding and optimization of the parameters in the start-up phase is critical not only for the future successful ITER start-up scenario [18], but also for saving the valuable voltage-second (Vs) [19], which prompts more rapid burn-through and ultimately extends the pulse length of the discharge.…”

Section: Introductionmentioning

confidence: 99%

Plasma production and preliminary results from the ADITYA Upgrade tokamak

Tanna¹,

Ghosh²,

Raj³

et al. 2018

Plasma Sci. Technol.

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The Ohmically heated circular limiter tokamak ADITYA (R 0 =75 cm, a=25 cm) has been upgraded to a tokamak named the ADITYA Upgrade (ADITYA-U) with an open divertor configuration with divertor plates. The main goal of ADITYA-U is to carry out dedicated experiments relevant for bigger fusion machines including ITER, such as the generation and control of runaway electrons, disruption prediction, and mitigation studies, along with an improvement in confinement with shaped plasma. The ADITYA tokamak was dismantled and the assembly of ADITYA-U was completed in March 2016. Integration of subsystems like data acquisition and remote operation along with plasma production and preliminary plasma characterization of ADITYA-U plasmas are presented in this paper.

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