Plasma Radial Position Research Articles

Plasma initiation and preliminary magnetic control in the HL-2M tokamak

The first plasmas have been achieved in the HL-2M tokamak by fulfilling the usual criteria for plasma breakdown, i.e. optimal working gas pressure, field null configuration of poloidal magnetic field, reasonable toroidal electrical field and clean wall conditions. The characterization of plasma initiation, such as identification of initiation time and position, is performed by diagnostic data analysis and numerical simulation. A model-based method, which enables dedicated pick-up sensors to estimate plasma radial position in real time, is applied for the limiter plasma shape in the first plasma commissioning phase. For a circular cross-section plasma, feedback control on plasma current and radial position allowed us to extend discharge durations to ∼250 ms associated with a 120 kA plasma current in flat-top. Preliminary plasma equilibrium reconstruction by the NICE code from pick-up sensors and flux loops, as well as interferometry measurements, gives plasma boundaries, plasma profiles of toroidal current density and safety factor, as well as electron density, in an offline way.

Real-time plasma position reflectometry system development and integration on COMPASS tokamak

O-mode frequency-modulated continuous wave (FMCW) reflectometry provides an alternative to magnetic measurements in the determination of the plasma separatrix position for plasma position control. This type of measurement proves to be particularly attractive for the control of future fusion reactors where the harsh radiation environment may damage magnetic probes or induce non-compensable measurement drifts. Plasma position reflectometry (PPR), first demonstrated in ASDEX-Upgrade, is a control technique that is increasingly important to validate in diversified experimental devices and relevant plasma regimes. The COMPASS tokamak provides suitable conditions for such advanced demonstrations and regular PPR operation and development, thanks to its O-mode reflectometer and Multi-Threaded Application Real-Time executor (MARTe) based real-time control system. Herein we present the integration of a PPR system on COMPASS, both at hardware and software levels. Reflectometry swept measurements require signals to be acquired in bursts of data and streamed to the corresponding MARTe-PPR node through PCIe® fibre-optic links. The data transferred in real-time is used to reconstruct the radial density profiles from which the outer separatrix position is estimated. This estimate is then delivered to the central MARTe controller node via a dedicated Xilinx® Aurora® link at a rate matching COMPASS's 500 μs slow control cycle. The implemented system systematically met the required latency specifications, being able to deliver an estimation of the plasma radial position capable of successfully replacing the corresponding magnetic measurements in the plasma position feedback control loops.

Solenoid-free startup experiments in DIII-D

A series of DIII-D experiments was performed to investigate the potential for initiating plasma current using only poloidal field coils located outside the DIII-D central solenoid, i.e. ‘solenoid-free’. Plasma current to 166 kA was achieved using 2–3 MW of electron cyclotron (EC) heating and was limited by coil and power supply constraints. Flux conversion to plasma current was similar to standard DIII-D startup with some degradation at higher plasma current associated with stray fields and vertical stability issues. In preliminary solenoid-free experiments, neutral beam (NB) current drive (CD) levels were small and attributed to reduced CD efficiency associated with low electron temperature produced by the low current, low confinement plasma. Lack of plasma radial position control also contributed to a reduction of NBCD. Similarly, ECCD was small owing to low plasma temperature and outside EC launch which is required in the solenoid-free scenario. Synergistic experiments were carried out using standard solenoid initiated plasmas in order to study noninductive CD in limited, Lmode plasmas, typical of that generated by solenoid-free startup. While substantial noninductive current can be driven, self-sustaining levels of noninductive current have not yet been achieved with our present six-source co-injection NB system combined with EC and fast wave systems. At low plasma current and high levels of localized EC heating, substantial MHD is generated and this was seen to severely limit plasma performance. Although further optimization is possible in the limited plasma regime, full noninductive, steady-state operation may require diverted plasma with H-mode quality confinement. Discharges obtained during the solenoid-free campaign are compared with results of previous DIII-D campaigns aimed at achieving a steady state, noninductive CD solution.

Magnetic control of plasma current, position, and shape in Tokamaks: a survey or modeling and control approaches

Plasma current, position and shape control systems and modeling approaches were reviewed. The literature was divided into three categories, namely plasma radial position and current control, vertical stabilization of elongated plasmas, and plasma shape control. The plasma current and radial position control problem can be easily solved using a filament plasma model - the Shafarov lumped parameter equation and two separate PID controllers. At most, a MIMO controller can be used if decoupling is desired. Stabilization of the plasma vertical motion requires new control strategies in which the control algorithm is changed on the basis of an estimate of the measurement accuracy. For controlling the shape during the plasma formation and start-up phase, controllers based on magnetic flux control are useful when control requirements are not stringent.

Plasma modeling for position and current control in FTU

This paper describes the development and experimental validation of a simulation model for the design of FTU plasma radial position and current controllers. These controllers have been designed to be tested experimentally on the FTU tokamak. The results predicted in simulation were actually delivered during FTU operation, in two different discharges.

H-mode confinement in JET with enhanced performance by pellet peaked density profiles

The combination of two regimes of enhanced performance, the H-mode and the pellet enhanced performance (PEP) mode, has been achieved in JET. The strong enhancement of the central plasma parameters, obtained with pellet injection and subsequent auxiliary heating, is found to persist well into the H-mode phase. A characteristic of the PEP regime is that an improvement of the fusion reactivity over non-pellet discharges is obtained under the condition of nearly equal electron and ion temperatures. A maximum neutron production rate of 0.95 × 10l6 s−1 was obtained in a double-null X-point discharge with 2.5 MW of neutral beam heating and 9 MW of ion cyclotron resonance heating, with central ion and electron temperatures of about 10 keV and a central deuterium density of 8.0 × 1019 m−3. The corresponding fusion product nD(0)τETi(0) is between 7.0 and 8.6 × 1020 m−3·s·keV. The enhanced neutron production is predominantly of thermonuclear (Maxwellian) origin. The compatibility of these regimes is an important issue in the context of tokamak ignition strategies. Several technical developments on JET have played a role in the achievement of this result: (1) the use of low voltage plasma breakdown (0.15 V/m) to permit pellet injection in an X-point configuration before the formation of a q = 1 surface; (2) the elimination of ICRH specific impurities with antenna Faraday screens made of solid beryllium; (3) the use of a novel system of plasma radial position control that stabilizes the coupling resistance of the ion cyclotron heating system.

Improvement of Plasma Radial Position Control in PDX through an Automated Learning Procedure

A procedure has been developed to calculate improved reference waveforms for the radial position feedback control system of the PDX tokamak. The procedure uses the data from previous discharges in calculating the reference waveforms, correcting for position errors, and including compensation for field penetration and power supply response. The results obtained in PDX divertor discharges are presented.