Tokamak Control Research Articles

Design of the Plasma Position and Shape Control in the ITER Tokamak Using In-Vessel Coils

The International Thermonuclear Experimental Reactor (ITER) is the next step toward the realization of electricity-producing fusion power plants, and it is planned to be in operation in 2016. ITER has been designed so as to reach the plasma burning condition and to operate with high-elongated unstable plasmas. However, due to the constraints that affect the machine realization, these open-loop unstable high-performance plasmas can hardly be stabilized using the poloidal field coils placed outside the tokamak vessel. For this reason, during the ITER design review phase, it has been proposed to investigate the possibility of using in-vessel coils in order to improve the best achievable performance of the vertical stabilization (VS) system. This paper proposes a new approach for the plasma current, position, and shape control design in the presence of in-vessel coils. Two control loops are designed: a first loop that guarantees the VS driving the voltage applied to in-vessel coils, and a second loop that controls the plasma current and up to 32 geometrical shape descriptors as close as possible to the reference values. The performance of the proposed control system is shown by means of simulations of some cases of interest.

Observation of poloidal current flowed to the vessel after failure of vertical position feedback control in EAST Tokamak

Plasmas with vertically elongated cross-sections tend to be unstable to an axis-symmetric instability. This paper studies the magnetohydrodynamic equilibria in elongated plasmas after failure of vertical feedback control by using magnetic data for EAST device. Vertical forces on the vessel due to the induced polodial and toroidal currents are evaluated. The maximum force of the Fzpol in vertical displacement events for EAST designed parameters is given.

Real time magnetic field and flux measurements for tokamak control using a multi-core PCI Express system

The existing real time system for the position and shape control in ASDEX Upgrade has been extended to calculate magnetic flux surfaces in real time using a multi-core PCI Express system running LabVIEW RT. The availability of reflective memory for LabVIEW RT will allow this system to be connected to the control system and other diagnostics in a multi-platform real time network. The measured response of each magnetic probe to the individual poloidal field coil currents in the absence of plasma current is compared to the calculated value. Prior to a tokamak discharge this comparison can be used to check for failure of the magnetic probe, flux loop or integrator.

Fractional variational problems and particle in cell gyrokinetic simulations with fuzzy logic approach for tokamaks

In earlier Rastovic's papers [1] and [2], the effort was given to analyze the stochastic control of tokamaks. In this paper, the deterministic control of tokamak turbulence is investigated via fractional variational calculus, particle in cell simulations, and fuzzy logic methods. Fractional integrals can be considered as approximations of integrals on fractals. The turbulent media could be of the fractal structure and the corresponding equations should be changed to include the fractal features of the media.

Control of Tokamaks using ASTRA – Matlab Integration

The need to find new sources of power has emerged as an important issue due to the existing worldwide increase in the demand of energy. In this way, nuclear fusion has turned out to be a promising source of this power which has lead to an increasing interest in finding solutions for the different tokamak control problems. In order to find appropriate controllers for tokamaks several kinds of codes and models have been developed. In this paper one of those codes, ASTRA, is used to simulate the performance of a tokamak. The aim of this paper is to show how the ASTRA code for tokamaks can be integrated into the Matlab – Simulink tool with the purpose of making more convenient and easier the development of suitable controllers for tokamaks. In particular as a demonstrative case study, a simple PID controller has been designed for the loop voltage of a tokamak. The simulations obtained from the example presented in this manuscript show the feasibility and the goodness of the proposed ASTRA – Matlab integration. This integration composes a new and innovative issue in the tokamak control area and it provides new possibilities to the development and application of advanced control schemes to the standardized and widely extended ASTRA simulation code for tokamaks.

Development of a multichannel dispersion interferometer at TEXTOR

The design and main characteristics of 14-channel dispersion interferometer for plasma profile measurement and control in TEXTOR tokamak are presented. The diagnostic is engineered on the basis of modular concept, the 10.6 microm CO(2) laser source and all optical and mechanical elements of each module are arranged in a compact housing. A set of mirrors and retroreflectors inside the TEXTOR vacuum vessel provides full coverage of the torus cross section with 12 vertical and two diagonal lines of sight, no rigid frame for vibration isolation is required. Results of testing of the single-channel prototype diagnostic and the pilot module of the multichannel dispersion interferometer are presented.

Decoupling in the problem of tokamak plasma response to asymmetric magnetic perturbations

A general solution is proposed to the coupling problem which appears in theoretical studies of the perturbed equilibrium and the plasma stability active control in tokamaks. This is based on the separation of the plasma contribution from the total magnetic field found by solving the perturbed equilibrium problem. A numerical procedure to decouple the components of different origins was developed recently when the plasma code MARS and the electromagnetic code CARIDDI allowing a detailed 3D simulation of the outer region were integrated into the CarMa code (Albanese et al 2008 IEEE Trans. Magn. 44 1654, Portone et al 2008 Plasma Phys. Control. Fusion 50 085004). Here we show that the necessary separation can be described in the explicit form by a single equation, which gives the easiest and direct solution. The analysis is based on the first principles, and the result can be used with any model of the plasma.

Tokamak machine monitoring and control system for JT-60

The tokamak machine monitoring and control system has been developed to avoid machine damages in JT-60 safety operation. It monitors the JT-60 mechanical structures and other sub-components to maintain their conditions. The system adopts the CAMAC digitizers, the Signal Transfer Unit (STU), and the hard-wired interlocks. Over 20 year, it has never been affected by any major troubles except a water leakage in toroidal field coil (TFC). The TFC monitoring system has been developed for the TFC safety with the water leakage from a cooling water channel. It has improved the reliability and efficiency of the TFC operation.

The plasma current control in the TRIAM-1M tokamak

TRIAM-1M is a superconducting tokamak, the plasma is maintained by RF (radio frequency) wave completely without ohmic field in the steady-state operation. An active control of plasma current in fully non-inductive lower hybrid current drive (LHCD) plasmas on TRIAM-1M is demonstrated successfully for more than 1 min. A simple model to decide the PID (proportional integral and differential) control parameters based on secondary lower hybrid wave (LHW) is proposed. The PID parameters different from the prediction of the model prevent from maintaining plasma current well.

Fractional Fokker–Planck Equations and Artificial Neural Networks for Stochastic Control of Tokamak

The general form of description of Kolmogorov–Arnold–Moser (KAM) theorem in controlled plasma fusion, is obtained via the theory of artificial fuzzy neural networks. Without of the global maximum entropy principle, the complexity function is used for the Monte Carlo simulations.

Towards model-based current profile control at DIII-D

A key goal in control of an advanced tokamak (AT) discharge is to maintain safety factor ( q) and pressure profiles that are compatible with both MHD stability at high toroidal beta and a high fraction of the self-generated bootstrap current. This will enable high fusion gain and non-inductive sustainment of the plasma current for steady-state operation. In this work we report progress towards enabling model-based active control of the current profile at DIII-D. Initial results on modeling-for-control and simulation of the dynamic evolution of the poloidal flux profile during and just following the ramp-up of the plasma current are presented. The magnetic diffusion equation is combined with empirical correlations obtained at DIII-D for the density, temperature and non-inductive current to introduce a simplified dynamic model describing the evolution of the poloidal flux, and therefore the q profile, during the inductive phase of the discharge. The physical model is rewritten in a control-oriented formulation and the control challenges asocciated with the problem are discussed.

On Stochastic Control of Tokamak and Artificial Intelligence

Instead of the theory of invariant manifolds and entropy reduction, the theory of fractional Brownian motions and artificiall neural networks is used for description of advanced methods for control of tokamak plasma behaviour.

Plasma Shape Reconstruction of Spherical Tokamak using CCS Method

Plasma shape reconstruction is important for plasma control of tokamak. Cauchy-Condition Surface (CCS) method is a numerical approach to reproduce plasma shape which has good precision in conventional tokamak. In order to apply it in the plasma shape reproduction of Compact PWI experimental Device (CPD) which is a new spherical tokamak in Kyushu University, the calculation precision of CCS method in CPD is analyzed in the paper.

The collaborative tokamak control room

Magnetic fusion experiments keep growing in size and complexity resulting in a concurrent growth in collaborations between experimental sites and laboratories worldwide. In the US, the National Fusion Collaboratory Project is developing a persistent infrastructure to enable scientific collaboration for all aspects of magnetic fusion energy research by creating a robust, user-friendly collaborative environment and deploying this to the more than 1000 US fusion scientists in 40 institutions who perform magnetic fusion research. This paper reports on one aspect of the project which is the development of the collaborative tokamak control room to enhance both collocated and remote scientific participation in experimental operations. This work includes secured computational services that can be scheduled as required, the ability to rapidly compare experimental data with simulation results, a means to easily share individual results with the group by moving application windows to a shared display, and the ability for remote scientists to be fully engaged in experimental operations through shared audio, video, and applications. The project is funded by the USDOE Office of Science, Scientific Discovery through Advanced Computing (SciDAC) Program and unites fusion and computer science researchers to directly address these challenges.

Model-based dynamic resistive wall mode identification and feedback control in the DIII-D tokamak

A new model-based dynamic resistive wall mode (RWM) identification and feedback control algorithm has been developed. While the overall RWM structure can be detected by a model-based matched filter in a similar manner to a conventional sensor-based scheme, it is significantly influenced by edge-localized-modes (ELMs). A recent study suggested that such ELM noise might cause the RWM control system to respond in an undesirable way. Thus, an advanced algorithm to discriminate ELMs from RWM has been incorporated into this model-based control scheme, dynamic Kalman filter. Specifically, the DIII-D [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] resistive vessel wall was modeled in two ways: picture frame model or eigenmode treatment. Based on the picture frame model, the first real-time, closed-loop test results of the Kalman filter algorithms during DIII-D experimental operation are presented. The Kalman filtering scheme was experimentally confirmed to be effective in discriminating ELMs from RWM. As a result, the actuator coils (I-coils) were rarely excited during ELMs, while retaining the sensitivity to RWM. However, finding an optimized set of operating parameters for the control algorithm requires further analysis and design. Meanwhile, a more advanced Kalman filter based on a more accurate eigenmode model has been developed. According to this eigenmode approach, significant improvement in terms of control performance has been predicted, while maintaining good ELM discrimination.

Advanced tools for enhancing control room collaborations

The US National Fusion Collaboratory (NFC) project has been exploring a variety of computer and network technologies to develop a persistent, efficient, reliable and convenient collaborative environment for magnetic fusion research. One goal is to enhance remote and collocated team collaboration by integrating collaboration software tools into control room operations as well as with data analysis tools. To achieve this goal, the NFC recently introduced two new collaboration technologies into the DIII-D tokamak control room. The first technology is a high-resolution, large format Shared Display Wall (SDW). By creating a shared public display space and providing real time visual information about the multiple aspects of complex experiment activity, the large SDW plays an important role in increasing the rate of information dissemination and promoting interaction among team members. The second technology being implemented is the “tokamak control room aware” Instant Messaging (IM) service. In addition to providing text-chat capabilities for research scientists, it enables them to automatically receive information about experiment operations and data analysis processes to remotely monitor the status of ongoing tokamak experiment. As a result, the IM service has become a unified portal interface for team collaboration and remote participation.

CORBA-based solution for remote participation in SST-1 tokamak control and operation

The steady state superconducting tokamak (SST-1) central control system is a distributed heterogeneous process communication system built on socket programming. It consists of machine, experiment and discharge control plus timing and a database. The software controls and monitors SST-1 subsystems: water-cooling, power supplies, cryogenics and vacuum over a local area network (LAN). The SST-1 control room is the place where all the activities like session announcement, machine control, experiment control, discharge control and monitoring are performed. We have realized that, instead of having a single monitoring place, we should have multiple monitoring points and it should be made possible to control the experiment from any PC over the LAN. In order to meet such requirements for remote participation in tokamak operation, we are upgrading the existing software. The upgraded software is based on Common Object Request Broker Architecture (CORBA) technology. The software is utilizing CORBA-services such as event service, naming services, interface repository and security services. The inherent features of CORBA make the software, platform and language independent. The software supports a variety of communication paradigms including publish–subscribe, peer-to-peer, and request–reply. Based on this software, one can participate in SST-1 tokamak operation from the LAN, or a wide area network (WAN) connection anywhere on the Internet. Each user can customize plasma parameters and diagnostics data that he wants to monitor, at any time without any change in the software and a copy of these parameters will be available to him. This paper focuses on the publish–subscribe communication paradigm and its application for a machine monitoring system.

Heating, Current Drive, and Advanced Plasma Control in JFT-2M

We briefly present the experimental results obtained using the various heating and current drive systems, such as electron cyclotron heating, fast wave, lower hybrid wave, ion cyclotron range of frequency, neutral beam injection heating, or the external coil systems or the divertor bias system developed in JFT-2M from the viewpoint of the active control of the tokamak plasma. The features and design concepts of each system are described briefly, as we had to develop such systems to carry out the physical research. The experimental findings in plasma heating, current drive, and advanced plasma control in the JFT-2M tokamak contributed to the design and experiments in the larger tokamaks and to ITER.

Optimal control of tokamak and stellarator plasma behaviour

The control of plasma transport, laminar and turbulent, is investigated, using the methods of scaling, optimal control and adaptive Monte Carlo simulations. For this purpose, the asymptotic behaviour of kinetic equation is considered in order to obtain finite-dimensional invariant manifolds, and in this way the finite-dimensional theory of control can be applied. We imagine the labyrinth of open doors and after applying self-similarity, the motion moved through all the desired doors in repeatable ways as Brownian motions. We take local actions for each piece of contractive ergodic motion, and, after self-organization in adaptive invariant measures, the optimum movement of particles is obtained according to the principle of maximum entropy. This is true for deterministic and stochastic cases that serve as models for plasma dynamics.

Variation of particle exhaust with changes in divertor magnetic balance

Recent experiments on DIII-D point to the importance of two factors in determining how effectively the deuterium particle inventory in a tokamak plasma can be controlled through pumping at the divertor target(s): (1) the divertor magnetic balance, i.e. the degree to which the divertor topology is single-null or double-null (DN) and (2) the direction of the of B × ∇B ion drift with respect to the X-point(s). Changes in divertor magnetic balance near the DN shape have a much stronger effect on the particle exhaust rate at the inner divertor target(s) than on the particle exhaust rate at the outer divertor target(s). The particle exhaust rate for the DN shape is strongest at the outer strike point opposite the B × ∇B ion particle drift direction. Our data suggests that the presence of B × ∇B and E × B ion particle drifts in the scrape-off layer and divertor(s) play an important role in the particle exhaust rates of DN and near-DN plasmas. Particle exhaust rates are shown to depend strongly on the edge (pedestal) density. These results have implications for particle control in ITER and other future tokamaks.