Tokamak Control Research Articles

Nonlinear burn condition control in tokamaks using isotopic fuel tailoring

One of the fundamental problems in tokamak fusion reactors is how to control the plasma density and temperature in order to regulate the amount of fusion power produced by the device. Control of these parameters will be critical to the success of burning plasma experiments like ITER. The most previous burn condition control efforts use either non-model based control designs or techniques based on models linearized around particular operating points. Such strategies limit the potential operational space and must be carefully retuned or redesigned to accommodate changes in operating points or plasma parameters. In this work, a nonlinear dynamic model of the spatial averages of energy and ion species densities is used to synthesize a nonlinear feedback controller for stabilizing the burn condition. The nonlinear model-based control strategy guarantees a much larger operational space than previous linear controllers. Because it is not designed around a particular operating point, the controller can be used to move from one burn condition to another. The proposed scheme first attempts to use regulation of the auxiliary heating power to reject temperature perturbations, then, if necessary, uses isotopic fuel tailoring as a way to reduce fusion heating during positive temperature perturbations. A global model of hydrogen recycling is incorporated into the model used for design and simulation, and the proposed control scheme is tested for a range of recycling model parameters. As we find the possibility of changing the isotopic mix can be limited for certain unfavorable recycling conditions, we also consider impurity injection as a back-up method for controlling the system. A simple supervisory control strategy is proposed to switch between the primary and back-up control schemes based on stability and performance criteria. A zero-dimensional simulation study is used to study the performance of the control scheme for several scenarios and model parameters. Finally, a one-dimensional simulation is done to test the robustness of the control scheme to spatially varying parameters.

Combined magnetic and kinetic control of advanced tokamak steady state scenarios based on semi-empirical modelling

This paper shows that semi-empirical data-driven models based on a two-time-scale approximation for the magnetic and kinetic control of advanced tokamak (AT) scenarios can be advantageously identified from simulated rather than real data, and used for control design. The method is applied to the combined control of the safety factor profile, q(x), and normalized pressure parameter, βN, using DIII-D parameters and actuators (on-axis co-current neutral beam injection (NBI) power, off-axis co-current NBI power, electron cyclotron current drive power, and ohmic coil). The approximate plasma response model was identified from simulated open-loop data obtained using a rapidly converging plasma transport code, METIS, which includes an MHD equilibrium and current diffusion solver, and combines plasma transport nonlinearity with 0D scaling laws and 1.5D ordinary differential equations. The paper discusses the results of closed-loop METIS simulations, using the near-optimal ARTAEMIS control algorithm (Moreau D et al 2013 Nucl. Fusion 53 063020) for steady state AT operation. With feedforward plus feedback control, the steady state target q-profile and βN are satisfactorily tracked with a time scale of about 10 s, despite large disturbances applied to the feedforward powers and plasma parameters. The robustness of the control algorithm with respect to disturbances of the H&CD actuators and of plasma parameters such as the H-factor, plasma density and effective charge, is also shown.

A self-description data framework for Tokamak control system design

A Tokamak device consists of numerous control systems, which need to be integrated. CODAC (Control, Data Access and Communication) system requires the configuration settings of these control systems to carry out the integration smoothly. SDD (Self-description data) is designed to describe the static configuration of control systems. ITER CODAC group has released an SDD software package for control system designers to manage the static configuration, but it is specific for ITER plant control systems. Following the idea of ITER SDD, we developed a flexible and scalable SDD framework to develop SDD software for J-TEXT and other sophisticated devices.The SDD framework describes the configuration settings of various control systems, including physical and logical elements and their relation information, in SDD models which are classified into Components and Connections. The framework is composed of three layers: the MongoDB database, an open-source, dynamic schema, NoSQL (Not Only SQL) database; the SDD service, which maps SDD models to MongoDB and handles the transaction and business logic; the SDD applications, which can be used to create and maintain SDD information, and generate various kinds of output using the stored SDD information.

Data Acquisition and Automation for Plasma Rotation Diagnostic in the TCABR Tokamak

In this work we describe the implementation of a full modular system of data acquisition and processing for the plasma rotation diagnostic in the TCABR tokamak. The experimental setup uses a single monochromator and six photomultipliers (PMT), in which pair of PMTs measures the light at slightly different wavelengths. Thus, it can measure the time evolution of the Doppler shift of the impurities emission lines coming from three spatial positions (one for toroidal rotation and two for poloidal rotation). The data acquisition and preanalysis program were written with LabVIEW software and is capable of controlling the spectrometer wavelength, PMTs power supplies, data acquisition, and storage. All data are recorded in MDSplus trees that easily allow data visualization and post-processing analysis (both locally and remotely) via MATLAB, Python, Java and others programming languages. This system can run independently from other diagnostics and machine systems and can be integrated with the main tokamak control system by means of TCP/IP messages.

Three-dimensional equilibrium reconstruction on the DIII-D device

The presence of toroidal variation in diagnostic measurements indicates that the two-dimensional (2D) symmetry of tokamak equilibria can be violated when resonant magnetic perturbations (RMPs) are applied to suppress edge localized modes. While tokamak control is still possible with a 2D model, questions arise regarding the applicability of 2D equilibria when performing detailed analysis. In particular, questions regarding edge physics would benefit from equilibrium calculations which are consistent with measurements indicating toroidal variations. The ability to fit three-dimensional (3D) equilibria to diagnostic measurements has long been a challenge for stellarators, non-axisymmetric devices with an inherently 3D field structure. The STELLOPT code provides a solution to such a challenge by fitting 3D VMEC equilibria to magnetic, Thomson, motional Stark effect (MSE) and charge-exchange diagnostics. The plasma of the DIII-D tokamak with applied n = 3 RMP is reconstructed with STELLOPT (shot number 142603). The reconstruction is governed by fit to magnetic diagnostics, measured vacuum coil currents, Thomson scattering, charge-exchange spectroscopy and MSE polarimetry. The reconstructed equilibria possess features where pressure gradients become vanishingly small at low-order rationals. Such features can be associated with the applied RMP spectrum, indicating mode penetration for this shot. Boundary displacements on the order of 0.5 cm peak-to-peak were present. This suggests that while the 3D effect was small relative to the plasma minor radius, resonant mode penetration occurred, indicating the ability of 3D reconstructions to resolve small key features in the plasma.

Demonstration of sawtooth period control with EC waves in KSTAR plasma

The sawtooth period control in tokamak is important issue in recent years because the sawtooth crash can trigger TM/NTM instabilities and drive plasmas unstable. The control of sawtooth period by the modification of local current profile near the q =1 surface using ECCD has been demonstrated in a number of tokamaks [1, 2] including KSTAR. As a result, developing techniques to control the sawtooth period as a way of controlling the onset of NTM has been an important area of research in recent years [3]. In 2012 KSTAR plasma campaign, the sawtooth period control is carried out by the different deposition position of EC waves across the q =1 surface. The sawtooth period is shortened by on-axis co-ECCD (destabilization), and the stabilization of the sawtooth is also observed by off-axis co-ECCD at outside q =1 surface. In 2013 KSTAR plasma campaign, the sawtooth locking experiment with periodic forcing of 170 GHz EC wave is carried out to control the sawtooth period. The optimal target position which lengthens the sawtooth period is investigated by performing a scan of EC beam deposition position nearby q =1 surface at the toroidal magnetic field of 2.9 T and plasma current of 0.7 MA. The sawtooth locking by the modulated EC beam is successfully demonstrated as in [3-5] with the scan of modulation-frequency and duty-ratio at the low beta ( β N ~0.5) plasma. In this paper, the sawteeth behavior by the location of EC beam and the preliminary result of the sawtooth locking experiments in KSTAR will be presented.

Edge-localized-modes in tokamaks

Edge-localized-modes (ELMs) are a ubiquitous feature of H-mode in tokamaks. When gradients in the H-mode transport barrier grow to exceed the MHD stability limit the ELM instability grows explosively, rapidly transporting energy and particles onto open field lines and material surfaces. Though ELMs provide additional particle and impurity transport through the H-mode transport barrier, enabling steady operation, the resulting heat flux transients to plasma facing surfaces project to large amplitude in future low collisionality burning plasma tokamaks. Measurements of the ELM heat flux deposition onto material surfaces in the divertor and main chamber indicate significant broadening compared to inter-ELM heat flux, with a timescale for energy deposition that is consistent with sonic ion flow and numerical simulation. Comprehensive ELM simulation is highlighting the important physics processes of ELM transport including parallel transport due to magnetic reconnection and turbulence resulting from collapse of the H-mode transport barrier. Encouraging prospects for ELM control and/or suppression in future tokamaks include intrinsic modes of ELM free operation, ELM triggering with frequent small pellet injection and the application of 3D magnetic fields.

Near-infrared spectroscopy for divertor plasma diagnosis and control in DIII-D tokamak.

New near infrared (NIR) spectroscopic measurements performed in the DIII-D tokamak divertor plasma suggest new viable diagnostic applications: divertor recycling and low-Z impurity flux measurements, a spectral survey for divertor Thomson scattering (DTS) diagnostic, and Te monitoring for divertor detachment control. A commercial 0.3 m spectrometer coupled to an imaging lens via optical fiber and a InGaAs 1024 pixel array detector enabled deuterium and impurity emission measurements in the range 800-2300 nm. The first full NIR survey identified D, He, B, Li, C, N, O, Ne lines and provided plasma Te, ne estimates from deuterium Paschen and Brackett series intensity and Stark line broadening analysis. The range 1.000-1.060 mm was surveyed in high-density and neon seeded divertor plasmas for spectral background emission studies for λ = 1.064 μm laser-based DTS development. The ratio of adjacent deuterium Paschen-α and Brackett Br9 lines in recombining divertor plasmas is studied for divertor Te monitoring aimed at divertor detachment real-time feedback control.

Observation of EHO in NSTX and theoretical study of its active control using HHFW antenna

Two important topics for tokamak edge-localized modes (ELM) control, based on non-axisymmetric (3D) magnetic perturbations, are studied in NSTX and combined envisioning ELM control in the future NSTX-U operation: experimental observations of the edge harmonic oscillation (EHO) in NSTX (with lower frequency than EHOs in DIII-D and elsewhere), and theoretical study of its external drive using the high-harmonic fast wave (HHFW) antenna as a 3D field coil. EHOs were observed particularly clearly in NSTX ELM-free operation with very low n core modes. A number of diagnostics have confirmed n = 4–6 edge-localized and coherent oscillations in the 2–8 kHz frequency range. These oscillations seem to have a favoured operational window in rotational shear, similar to EHOs in DIII-D quiescent H modes. However, in NSTX, they are not observed to provide significant particle or impurity transport, possibly due to their weak amplitudes, of a few mm displacements, as measured by reflectometry. The external drive of these modes has been proposed in NSTX, by utilizing audio-frequency currents in the HHFW antenna straps. Analysis shows that the HHFW straps can be optimized to maximize n = 4–6 while minimizing n = 1–3. Also, ideal perturbed equilibrium code calculations show that the optimized configuration with only 1 kAt current can produce comparable or larger displacements than the observed internal modes. Thus it may be possible to use externally driven EHOs to relax the edge pressure gradient and control ELMs in NSTX-U and future devices. Fine and external control over the edge pressure gradient would be a very valuable tool for tokamak control.

Physics of Landau Resonances, Cyclotron Resonances, and Current Generation

Irreversible energy and momentum transfer from waves to particles for tokamak confinement, heating, and control is based on Landau and cyclotron resonances. Above a stochasticity threshold, these interactions can be viewed as a random walk in energy (action) space within the random phase approximation. We present and discuss the quasi-linear theory describing this random walk with a particular emphasis on the interplay between the dynamical picture (electromagnetic forces) and the statistical description (photons emission/absorption). Landau and cyclotron absorptions in tokamaks are thus derived, and the classical theory of current generation in tokamaks is presented in local and nonlocal regimes.

Control of resistive wall modes in tokamak plasmas

Tokamak control systems have to deal with different kinds of instabilities related to the presence of a resistive wall that surrounds the plasma. These instabilities are known as Resistive Wall Modes and are both axisymmetric and non-axisymmetric; they can occur during normal operation of the tokamak, and therefore suitable feedback controllers need to be designed and implemented. In this paper we propose a control architecture able to deal with the two main instabilities: the axisymmetric vertical instability and the non-axisymmetric kink instability. With reference to the case of the ITER tokamak, we design a controller consisting of two separate loops, one for the vertical stabilization and the other one for the stabilization of the kink instability. The two loops are designed in such a way to minimize the control effort and the interaction between them. The effectiveness of the approach is shown in simulation on an appropriate ITER configuration.

Power supplies for plasma column control in the COMPASS tokamak

The main magnetic fields in COMPASS – i.e. The Toroidal, Magnetising, Equilibrium, and Shaping Fields – are created by a set of four corresponding thyristor power supplies controlled in a 0.5ms loop.The plasma position has to be controlled both radially and vertically by two additional magnetic fields provided by two fast amplifiers (FAs) based on MOSFET technology, each supplying ±100V and up to ±5kA.Currently, an ongoing project aims at ELM triggering by fast changes of the vertical position of the plasma column, also referred to as vertical kicks. For this purpose, a new Vertical Kicks Power Supply (VKPS) capable of quick change of vertical plasma position is being constructed. This power supply should operate at up to 1.2kV with switching frequency up to 5kHz. It is designed as a H-bridge but based on IGBT transistors which can be operated at higher voltages than MOSFETs.In this paper, we focus on the FAs and VKPS engineering design and required output parameters. Both the power supplies are based on modern components with highest available ratings in their categories. Unique design of the power supplies takes advantage of the short duration of the COMPASS discharge by overloading the transistors above their maximal steady-state rating.The FA is regularly operating, so that in addition to describing its design, we also describe the achieved performance parameters. Finally, the common controller unit, communication, and error handling is described.

The ITER Fast Plant System Controller ATCA prototype Real-Time Software Architecture

IPFN is developing a prototype Fast Plant System Controller (FPSC) based in ATCA embedded technologies dedicated to ITER CODAC data acquisition and control tasks in the sub-millisecond range. The main goal is to demonstrate the usability of the ATCA standard and its enhanced specifications for the high speed, very high density parallel data acquisition needs of the most demanding ITER tokamak plasma Instrumentation and Control (I&C) systems. This effort included the in-house development of a new family of high performance ATCA I/O and timing boards.The standard ITER software system CODAC Core System (CCS) v3.1, with the control based in the EPICS system does not cover yet the real-time requirements fulfilled by this hardware, so a new set of software components was developed for this specific platform, attempting to integrate and leverage the new features in CSS, for example the Multithreaded Application Real Time executor (MARTe) software framework, the new Data Archiving Network (DAN) solution, an ATCA IEEE-1588-2008 timing interface, and the Intelligent Platform Management Interface (IPMI) for system monitoring and remote management.This paper presents the overall software architecture for the ATCA FPSC, as well a discussion on the ITER constrains and design choices and finally a detailed description of the software components already developed.

A real-time data acquisition and elaboration system for instabilities control in the FTU tokamak

A real-time data acquisition and elaboration system is being implemented to control the new ECH launcher recently installed at FTU (Frascati Tokamak Upgrade). The system is aimed at controlling different kinds of magnetohydrodynamic instabilities, in particular the deleterious 3/2 and 2/1 (neoclassical) tearing modes, (N)TM, and the saw teeth period in order to prevent the seeding of NTMs. The complete system is presented here together with preliminary offline and real-time tests. © 2001 Elsevier Science. All rights reserved.

A new remote control room for tokamak operations

This paper presents a summary of a new remote tokamak control room constructed near the offices of DIII-D's scientific staff. This integrated system combines hardware, software, data, and control of the room (R-232) into a unified package that has been designed and constructed in a generic fashion so that it can be used with any tokamak operating worldwide. The room is approximately 300ft2 and can accommodate up to 12 seated participants. Mounted on the wall facing each scientist are five 52″ LCD televisions and mounted to the wall on their right are six 24″ LCD monitors. Each seat has associated with it a 24″ monitor, network connection, and power and the scientist is either provided with a computer or they can use their own. The room has been used for operation of DIII-D, EAST, and KSTAR. Due to the long distances, data from EAST and KSTAR was brought back to local DIII-D computers in one large parallel network transfer and subsequently served to scientists in the remote control room to other US collaborators. This parallel data transfer allowed the data to be available to US participants between pulses making remote experimental participation highly effective.

A constrained control strategy for the shape control in thermonuclear fusion tokamaks

The paper deals with the application of the so-called Reference (or Command) Governor constrained control strategy to the shape control of plasmas in thermonuclear fusion reactors with the main scope of optimizing tokamak operations also in conditions very close to the operating envelope limits. A primal inner loop controlling the plasma–wall distance is first designed; the Reference Governor device is then tuned to modify, whenever necessary, the reference signals to the inner loop, on the basis of constraints due to voltage saturations on the power supply converters, limitations of currents in the active control coils, minimum clearance between the plasma surface and the vacuum chamber wall, maximum induced magnetic fields and forces on coils. As usual in model predictive paradigms, the reference signal modification is accomplished through an on-line optimization procedure which embodies plasma model forecasts computed along a finite time virtual receding horizon. The ITER (International Thermonuclear Experimental Reactor) tokamak is assumed as the case study. Numerical simulations are carried out on a finite elements nonlinear model taking into account induced currents in the passive structures. The proposed application shows how almost a hundred constraints can be managed on-line by the Reference Governor.

A real-time Grad–Shafranov PDE solver and MIMO controller

Plasma control experiments require enormous computational power to solve large problems with critical time constraints. For tokamak control, the non-linear and constrained Grad–Shafranov equation needs to be solved in real-time with a cycle time of less than 1ms. A new algorithm for the solution of this equation based on discrete sine transforms and a tridiagonal solver rather than the commonly used cyclic reduction algorithm is presented. Input signals from magnetic probes and flux loops are the constraints for the equation that must be continuously solved to calculate the magnetic equilibrium. A number of novel mathematical ideas were introduced and several generally applicable numerical strategies were developed using LabVIEW graphical dataflow programming to meet the critical timing goals. Benchmarks on CPUs are reported. Furthermore, the design of a MIMO (multiple input and output) controller to demonstrate the possibilities of tokamak position and shape control using graphical dataflow programming is discussed.

Effect of passive plates on vertical instability in the EAST tokamak

The effect of passive plates on vertical displacement control in the EAST tokamak is investigated by open loop experiments and numerical simulations based on a rigid displacement model. The experiments and simulations indicate that the vertical instability growth rate is reduced by a factor of about 2 in the presence of the passive plates, where the adjacent segments are not connected to each other. The simulations also show that the vertical instability growth rate is reduced by a factor of about 10 if all adjacent segments on each passive plate loop are connected to each other. The operational window is greatly enlarged with the passive plates.

Improved feature selection based on genetic algorithms for real time disruption prediction on JET

The early prediction of disruptions is an important aspect of the research in the field of Tokamak control. A very recent predictor, called “Advanced Predictor Of Disruptions” (APODIS), developed for the “Joint European Torus” (JET), implements the real time recognition of incoming disruptions with the best success rate achieved ever and an outstanding stability for long periods following training. In this article, a new methodology to select the set of the signals’ parameters in order to maximize the performance of the predictor is reported. The approach is based on “Genetic Algorithms” (GAs). With the feature selection derived from GAs, a new version of APODIS has been developed. The results are significantly better than the previous version not only in terms of success rates but also in extending the interval before the disruption in which reliable predictions are achieved. Correct disruption predictions with a success rate in excess of 90% have been achieved 200ms before the time of the disruption. The predictor response is compared with that of JET's Protection System (JPS) and the ADODIS predictor is shown to be far superior. Both systems have been carefully tested with a wide number of discharges to understand their relative merits and the most profitable directions of further improvements.

Quench Protection System for the KSTAR Toroidal Field Superconducting Coil

The design of the integrated quench protection (QP) system for the high current superconducting magnet (SCM) has been fabricated and tested for the toroidal field (TF) coil system of the Korea Superconducting Tokamak Advanced Research (KSTAR) device. The QP system is capable of protecting the TF SCM, which consists of 16 identical coils serially connected with a stored energy of 495 MJ at the design operation level at 35.2 kA per turn. Given that the power supply for the TF coils can only ramp up and maintain the coil current, the design of the QP system includes two features. The first is a basic fast discharge function to protect the TF SCM by a dump resistor circuit with a 7 s time constant in case of coil quench event. The second is a slow discharge function with a time constant of 360 s for a daily TF discharge or for a stop demand from the tokamak control system. The QP system has been successfully tested up to 40 kA with a short circuit and up to 34 kA with TF SCM in the second campaign of KSTAR. This paper describes the characteristics of the TF QP systems and test results of the plasma experiment of KSTAR in 2009.