Plasma initiation stage analysis in tokamaks with TRANSMAK code

Belyakov, V. A.; Lobanov, K.M.; Makarova, L.P.; Mineev, A. B.; Vasiliev, V. I.

doi:10.1080/1051999031000151140

Cited by 25 publications

(17 citation statements)

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“…In order to understand the key physics of start-up, a lot of simulation and analysis work (Jardin, Bell & Pomphrey 1993;Leuer & Wesley 1993;Tsutsui & Shimada 1998;Senda et al 1999;Belyakov et al 2003;Formisano et al 2017) has been done. The most successful model has been developed by Lloyd et al (1991), Lloyd, Carolan & Warrick (1996).…”

Section: Introductionmentioning

confidence: 99%

On the breakdown modes and parameter space of ohmic tokamak start-up

et al. 2018

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Tokamak start-up is strongly dependent on the state of the initial plasma formed during plasma breakdown. We have investigated through numerical simulations the effects that the pre-filling pressure and induced electric field have on pure ohmic heating during the breakdown process. Three breakdown modes during the discharge are found, as a function of different initial parameters: no breakdown mode, successful breakdown mode and runaway mode. No breakdown mode often occurs with low electric field or high pre-filling pressure, while runaway electrons are usually easy to generate at high electric field or low pre-filling pressure (${<}1.33\times 10^{-4}$ Pa). The plasma behaviours and the physical mechanisms under the three breakdown modes are discussed. We have identified the electric field and pressure values at which the different modes occur. In particular, when the electric field is $0.3~\text{V}~\text{m}^{-1}$ (the value at which ITER operates), the pressure range for possible breakdown becomes narrow, which is consistent with Lloyd’s theoretical prediction. In addition, for $0.3~\text{V}~\text{m}^{-1}$, the optimal pre-filling pressure range obtained from our simulations is $1.33\times 10^{-3}\sim 2.66\times 10^{-3}$ Pa, in good agreement with ITER’s design. Besides, we also find that the Townsend discharge model does not appropriately describe the plasma behaviour during tokamak breakdown due to the presence of a toroidal field. Furthermore, we suggest three possible operation mechanisms for general start-up scenarios which could better control the breakdown phase.

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Section: Introductionmentioning

confidence: 99%

On the breakdown modes and parameter space of ohmic tokamak start-up

et al. 2018

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“…Studies of plasma initiation in ITER are performed by simulations, which take into account eddy currents in the conducting structures and models of the power supplies (in particular, the set of resistors in the PF power supply switching networks). Several codes were used in these studies: the 2D electromagnetic code TRANSMAK [13] comprising the 0D transport code SCENPLINT and the 3D electromagnetic code BDOS [14]. The studies have shown the capability of the PF system to provide outboard or almost central plasma initiation, starting from 45% to 100% of the maximum CS magnetization.…”

Section: Plasma Initiation In Itermentioning

confidence: 99%

Chapter 8: Plasma operation and control

et al. 2007

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The ITER plasma control system has the same functional scope as the control systems in present tokamaks. These are plasma operation scenario sequencing, plasma basic control (magnetic and kinetic), plasma advanced control (control of RWMs, NTMs, ELMs, error fields, etc) and plasma fast shutdown. This chapter considers only plasma initiation and plasma basic control. This chapter describes the progress achieved in these areas in the tokamak experiments since the ITER Physics Basis (1999 Nucl. Fusion 39 2577) was written and the results of assessment of ITER to provide the plasma initiation and basic control. This assessment was done for the present ITER design (15 MA machine) at a more detailed level than it was done for the ITER design 1998 (21 MA machine) described in the ITER Physics Basis (1999 Nucl. Fusion 39 2577). The experiments on plasma initiation performed in DIII-D and JT-60U, as well as the theoretical studies performed for ITER, have demonstrated that, within specified assumptions on the plasma confinement and the impurity influx, ITER can produce plasma initiation in a low toroidal electric field (0.3 V m −1), if it is assisted by about 2 MW of ECRF heating. The plasma basic control includes control of the plasma current, position and shape-the plasma magnetic control, as well as control of other plasma global parameters or their profiles-the plasma performance control. The magnetic control is based on more reliable and simpler models of the control objects than those available at present for the plasma kinetic control. Moreover the real time diagnostics used for the magnetic control in many cases are more precise than those used for the kinetic control. Because of these reasons, the plasma magnetic control was developed for modern tokamaks and assessed for ITER better than the kinetic control. However, significant progress has been achieved in the plasma performance control during the last few years. Although the physics basis of plasma operation and control is similar in ITER and present tokamaks, there is a principal qualitative difference. To minimize its cost, ITER has been designed with small margins in many plasma and engineering parameters. These small margins result in a significantly narrower operational space compared with present tokamaks. Furthermore, ITER operation is expensive and component damage resulting from purely operational errors might lead to a high and avoidable repair cost. These factors make it judicious to use validated plasma diagnostics and employ simulators to 'pre-test' the combined ITER operation and control systems. Understanding of how to do this type of pre-test validation is now developed in present day experiments. This research push should provide us with fully functional simulators before the first ITER operation.

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“…These structures are listed below; see [1] for details. The data of the ITER poloidal field coils and their currents are presented in Table 1 [6,7]. Table 1.…”

Section: Main Elements Of the Magnetic Model Of The Tokamak Complexmentioning

confidence: 99%

“…Simulation of plasma initiation was performed with the TRANSMAK code [7]. Positions of first wall, the double-wall vacuum vessel, the poloidal field coils, the cryostat, as well as location of the BD region are shown in Figure 12.…”

Section: Simulation Of Plasma Initiation With Mmtc1mentioning

confidence: 99%

Stray magnetic field at plasma initiation produced by ferromagnetic elements of the ITER tokamak complex

Amoskov

Белов

Belyakov

et al. 2009

Plasma Devices and Operations

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The stray magnetic fields produced at plasma initiation (in particular at the gas breakdown (BD)) by the ITER tokamak complex building have been studied. The influence of global elements of the tokamak complex building is estimated. The scenario of plasma initiation was designed and simulated using virtual coils with currents producing a magnetic field in the BD region similar to that from the building.

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Plasma initiation stage analysis in tokamaks with TRANSMAK code

Cited by 25 publications

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On the breakdown modes and parameter space of ohmic tokamak start-up

On the breakdown modes and parameter space of ohmic tokamak start-up

Chapter 8: Plasma operation and control

Stray magnetic field at plasma initiation produced by ferromagnetic elements of the ITER tokamak complex

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