High current and low q95 scenario studies for FAST in the view of ITER and DEMO

The Fusion Advanced Study Torus (FAST) has been proposed as a possible European satellite, in view of ITER and DEMO, in order to: a) explore plasma wall interaction in reactor relevant conditions b) test tools and scenarios for safe and reliable tokamak operation up to the border of stability c) address fusion plasmas with a significant population of fast particles. A new FAST scenario has been designed focusing on low-q operation, at plasma current IP=10 MA, toroidal field BT=8.5T, with a q95=2.3 that would correspond to IP=20 MA in ITER. The flat-top of the discharge can last a couple of seconds (i.e. half the diffusive resistive time and twice the energy confinement time), and is limited by the heating of the toroidal field coils. A preliminary evaluation of the end-of-pulse temperatures and of the electromagnetic forces acting on the central solenoid pack and poloidal field coils has been performed. Moreover, a VDE plasma disruption has been simulated and the maximum total vertical force applied on the vacuum vessel has been estimated

Improving tokamak vertical position control in the presence of power supply voltage saturation

The control of the current, position and shape of an elongated cross-section tokamak plasma is complicated by the so-called instability of the current vertical position. Linearized models all share the feature of a single unstable eigenmode, attributable to this vertical instability of the plasma equilibrium movement, and a large number of stable or marginally stable eigenmodes, attributable to zero or positive resistance in all other model circuit equations. Due to the size and therefore cost of the ITER tokamak, there will naturally be smaller margins in the poloidal field coil power supplies, implying that the feedback control will experience actuator saturation during large transients due to a variety of plasma disturbances. Current saturation is relatively benign, due to the integrating nature of the tokamak, resulting in a reasonable time horizon for strategically handling the approach to saturation which leads to the loss of one degree of freedom in the feedback control for each saturated coil. On the other hand, voltage saturation is produced. by the feedback controller itself, with no intrinsic delay. This paper presents a feedback controller design approach which explicitly takes saturation of the power supply voltage into account when producing the power supply demand signals. We consider the vertically stabilizing part of the ITER controller (fast controller) with one power supply and therefore a single saturated input. We. consider an existing ITER controller and enlarge its region of attraction to the full null controllable region by adding a continuous nonlinearity into the control. In a system with a single unstable eigenmode and a single stable eigenmode we have already provided a proof of the asymptotical stability of the closed loop system, and we have examined the performance of this new continuous nonlinear controller. We have subsequently extended this analysis to a system with a single eigenmode and multiple stable eigenmodes. The method requires state feedback control, and therefore a reconstruction of the states is indispensable. We discuss the feasibility of extracting these states from the available diagnostic information as well as other implementation details. As a complement to our ITER simulations we confirm the enlargement of the region of attraction by the new controller by a JET simulation

Modelling for JET Vertical Stabilization System

Nuclear fusion is, in a sense, the opposite of nuclear fission. Fission, which is a mature technology, produces energy through the splitting of heavy atoms like uranium in controlled chain reactions. Unfortunately, the by-products of fission are highly radioactive and long lasting. On the other hand, fusion is the process by which the nuclei of two light atoms such as hydrogen are fused together to form a heavier (helium) nucleus, with energy produced as a by-product. Although controlled fusion is extremely technologically challenging, a fusion-power reactor would offer significant advantages over existing energy sources. This thesis is devoted to the control of tokamaks, magnetic confinement devices constructed in the shape of a torus (or doughnut). Tokamaks are the most promising of several proposed magnetic confinement devices. The need to improve the performance of modern tokamak operations has led to a further development of the plasma shape and position control systems. In particular, extremely elongated plasmas, with high vertical-instability growth rate, are envisaged to reach the required performance for ignition. This request for better performance from the experimentalists’ side has motivated the development of the new vertical-stabilization (VS) system at the JET tokamak, which has been proposed within the Plasma Control Upgrade project. This thesis presents the activity carried out to increase the capability of the VS system and to understand the operational limits in order to assess what can be done to improve the overall performance with the existing hardware and control system so as to minimize the impact on JET operation. The first objective of this work is the analysis of the new diagnostic system and the influence of the mechanical structure on the magnetic measurements used as diagnostics by the VS controller; the main focus is on the influence on the controller performance in the presence of large perturbations. The second objective is to design a new controlled variable to increase the performance of the VS system. The third objective is to provide an equivalent model of an ELM (Edge Localized Mode), in terms of internal plasma profile parameters via best fit of the vertical velocity estimation. The last objective is to obtain a reliable and accurate model of the overall system, based on the new platform MARTe, developed at JET and useful also for other devices

A flexible architecture for plasma magnetic control in tokamak reactors

Plasma magnetic control is one of the core engineering issues to be tackled in a fusion device. Over the last years, model based approaches have been proposed to face this issue, proving their effectiveness and allowing to reduce the time span needed for control testing and validation. The first part of this work is intended to give an overview of the subject, from the historical milestones to the underlying physics; the most common techniques for tokamak plasmas electromagnetic modeling and control are also introduced and discussed. After this introduction, a general architecture for plasma magnetic control in tokamaks is proposed. Finally, the proposed solution is applied to the Experimental Advanced Superconducting Tokamak (EAST) tokamak, where a new plasma magnetic control architecture was developed and implemented during the 2016-2018 experimental campaigns, and to the Japan Torus-60 Super Advanced (JT-60SA) device, which is currently under construction in Japan

A flexible architecture for plasma magnetic control in tokamak reactors

Plasma magnetic control is one of the core engineering issues to be tackled in a fusion device. Over the last years, model based approaches have been proposed to face this issue, proving their effectiveness and allowing to reduce the time span needed for control testing and validation. The first part of this work is intended to give an overview of the subject, from the historical milestones to the underlying physics; the most common techniques for tokamak plasmas electromagnetic modeling and control are also introduced and discussed. After this introduction, a general architecture for plasma magnetic control in tokamaks is proposed. Finally, the proposed solution is applied to the Experimental Advanced Superconducting Tokamak (EAST) tokamak, where a new plasma magnetic control architecture was developed and implemented during the 2016-2018 experimental campaigns, and to the Japan Torus-60 Super Advanced (JT-60SA) device, which is currently under construction in Japan

Enhancing the control of tokamaks via a continuous nonlinear control law

The control of the current, position and shape of an elongated cross-section tokamak plasma is complicated by the instability of the plasma vertical position. In this case the control becomes a significant problem when saturation of the power supplies is considered. Current saturation is relatively benign due to the integrating nature of the tokamak, resulting in a reasonable time horizon for strategically handling this problem. On the other hand, voltage saturation is produced by the feedback controller itself, with no intrinsic delay. In practice, during large plasma disturbances, such as sawteeth, ELMs and minor disruptions, voltage saturation of the power supply can occur and as a consequence the vertical position control can be lost. If such a loss of control happens the plasma displaces vertically and hits the wall of the vessel, which can cause damage to the tokamak. The consideration and study of voltage saturation is especially important for ITER. Due to the size and therefore the cost of ITER, there will naturally be smaller margins in the Poloidal Field coil power supplies implying that the feedback will experience actuator saturation during large transients due to a variety of plasma disturbances. The next generation of tokamaks under construction will require vertical position and active shape control and will be fully superconducting. When the magnetic transverse field in superconducting magnets changes, the magnet generates two types of heat loss, the so-called coupling loss and the so-called hysteresis loss, grouped together as AC losses. Superconducting coils possess superconducting properties only below a critical temperature around a few K. AC losses are detrimental since they heat up the superconducting material. Thus, if AC losses are too large, the cryogenic plant can no longer hold the required temperature to maintain the superconductivity properties. Once the superconductivity is lost, the electric currents in the coils produce an enormous heat loss due to the ohmic resistivity, which can lead to a possible damage to the coils. In general, the coils are designed with enough margin to absorb all likely losses. A possible loss reduction could allow us to downsize the superconducting cross section in the cables, reducing the overall cost, or simply increase the operational cooling margin for given coils. In this thesis we have tried to take into consideration these two major problems. The thesis is therefore focused on the following main objectives: i) the stability analysis of the tokamak considering voltage saturation of the power supplies and ii) the proposition of a new controller which enhances the stability properties of the tokamak under voltage saturation and iii) the proposition of a controller which takes into consideration the problem of reducing the AC losses. The subject of the thesis is therefore situated in an interdisciplinary framework and as a result the thesis is subdivided into two principal parts. The first part is devoted to tokamak physics and engineering, while the second part focuses on control theory. In the tokamak physics and engineering part we present the linear tokamak models and the nonlinear tokamak code used for the controller design and the validation of the new proposed controller. The discussion is especially focused on the presence of a single unstable pole when the vertical plasma position is unstable since this characteristic is essential for the work presented in the control theory part. In order to determine the enhancement of the stability properties we have to bring the new proposed controller to its stability limits by means of large disturbances. Validation by means of simulations with either linear or nonlinear tokamak models are imperatively required before considering the implementation of the new controller on a tokamak in operation. A linear tokamak model will probably be inadequate since large disturbances can move its state outside its validity regions. A full nonlinear tokamak evolution code like DINA is indispensable for this purpose. We give a detailed description of the principal plasma physics implemented in the DINA code. Additionally, validation of DINA is provided by comparing TCV experimental VDE responses with DINA code simulations. To allow a study of the AC losses reduction, the nature of the AC losses has to be reduced to a simplified form. We analyse to what extent the accumulated AC losses in ITER could be reduced by taking into account the losses themselves when designing the feedback control loops. In order to be able to carry out this investigation a simple and fast AC loss model, referred to as "AC-CRPP" model, is proposed. In the control theory part we study the stability region in state space, referred to as the region of attraction, for linear tokamak-like systems with input saturation (voltage saturation) and a linear state feedback. Only linear systems with a single unstable pole (mode) and a single saturated input are considered. We demonstrate that the characterisation of the region of attraction is possible for a second order linear system with one unstable and one stable pole. For such systems the region of attraction possesses a topological bifurcation and we provide an analytical condition under which this bifurcation occurs. Since the analysis relies on methodologies like Poincaré and Bendixson's theorems which are unfortunately only valid for second order systems it is evident that there is no way to apply the results for second order systems to higher order systems. It turned out that the search for characterising the region of attraction for higher order systems was illusory and thus this research direction had to be abandoned. We therefore focused on controllers for which the region of attraction is the maximal region of attraction that can be achieved under input saturation. This region is referred to as the null controllable region and its characterisation is simple for any arbitrary high order system possessing a single unstable pole. We present a new globally stabilising controller for which its region of attraction is equal to the null controllable region. This result is obtained by incorporating a simple continuous nonlinear function into a linear state feedback controller. There are several advantages linked to this new controller: i) the stability properties are enhanced, ii) the performance, AC loss reduction and fast disturbance rejection, can be taken into account, iii) the controller can be applied to any arbitrary high order system and iv) the controller possesses a simple structure which simplifies the design procedure. We close the control theory part by focusing on the application of the proposed new controller to tokamaks. Since this controller is a state feedback controller one of the major problems is linked to the state reconstruction. Other pertinent topics are: i) the study of the effect of the disturbances on the closed-loop system stability, ii) the problem inherent to the nature of a state feedback controller when we want an output of the system to track a reference signal and iii) the discussion of the detrimental effects on stability if a pure time delay or a limited bandwidth are added to the closed-loop system, as is the case in reality. The validation of the proposed controller is carried out by means of simulations. We present results for ITER-FEAT and JET using the linear tokamak model CREATE-L. Finally, we present a validation for the case of TCV using the nonlinear DINA-CH code

DTT - Divertor Tokamak Test facility - Interim Design Report

The “Divertor Tokamak Test facility, DTT” is a milestone along the international program aimed at demonstrating – in the second half of this century – the feasibility of obtaining to commercial electricity from controlled thermonuclear fusion. DTT is a Tokamak conceived and designed in Italy with a broad international vision. The construction will be carried out in the ENEA Frascati site, mainly supported by national funds, complemented by EUROfusion and European incentive schemes for innovative investments. The project team includes more than 180 high-standard researchers from ENEA, CREATE, CNR, INFN, RFX and various universities. The volume, entitled DTT Interim Design Report (“Green Book” from the colour of the cover), briefly describes the status of the project, the planning of the design future activities and its organizational structure. The publication of the Green Book also provides an occasion for thorough discussions in the fusion community and a broad international collaboration on the DTT challenge

Exploration of candidate fusion reactor regimes by real time control of tokamak plasma shape

The potential of nuclear fusion to provide a practically inexhaustible source of energy has motivated scientists to work towards developing nuclear fusion tokamak power plants. Stable operation of a tokamak at high performance requires simultaneous treatment of several plasma control problems. Moreover, the complex physics which governs the tokamak plasma evolution must be studied and understood to make correct choices in controller design. This mutual inter-dependence has informed this thesis, using control solutions as an experimental tool for physics studies, and using physics knowledge for developing new advanced control solutions. The TCV tokamak at SPC-EPFL is ideally placed to explore issues at the interface between plasma physics and plasma control, by combining a state-of-the-art digital real time control system with a flexible and diverse set of actuators including a full set of independently powered shaping coils. The recent deployment of the real time version of the Grad-Shafranov equilibrium reconstruction code LIUQE, with a sub-ms cycle time in the digital control system, has facilitated the design of a new generalised plasma position and shape controller, based on the information on poloidal flux and magnetic field provided by the real-time Grad-Shafranov solver. The first issue addressed in the thesis is the development and experimental testing of a new real time control strategy to construct a generalised control algorithm for not only controlling the position of the plasma but also to aid in the precise control of higher order shape moments, X-points and strike points, particularly in advanced plasma configurations such as negative-triangularity plasmas, snowflake and super-X divertors, and doublets. A controller formulation ensuring flexibility through an ordering of controlled variables from the most easily to the least easily controlled, while respecting the hardware limits on the poloidal field coil currents, is developed. The successful experimental implementation of the control algorithm has been demonstrated for both fixed and time varying plasma position and shape for limiter and divertor plasma discharges. In addition, the controller has provided satisfactory performance with respect to plasma scenarios involving complex changes in the plasma shape and position. The second issue addressed in the thesis is the application of the generalised plasma position and shape controller to a snowflake plasma configuration. A comparison between the optimised generalised plasma position and shape controller with the performance of the TCV hybrid controller for a given reference snowflake plasma discharge showed a marked improvement in various geometrical properties of the snowflake plasma configuration in the vicinity of the null point. However, strong control of the poloidal magnetic field at the two X-points resulted in a tradeoff on the upper part of the plasma boundary, where the overall precision was comparable to that of the legacy controller. In the experimental time available, the snowflake shots developed exhibited a boundary that was too close to the inner wall of the vessel, modifying the edge plasma behaviour (studied with infrared cameras and Langmuir probes) and making it difficult to study the physics properties of the exact snowflake. However, further optimisation should well be possible

THREE-DIMENSIONAL EFFECTS OF ELECTROMAGNETIC FIELDS IN TOKAMAK PLASMAS

The problem of the energy harvesting to face the more and more increasing energy demand is currently challenging. The higher part of our electrical energy (about 80%) is produced by thermoelectrical power plants, which exploit the so-called Non-renewable energy resources (e.g. oil and gas), whose re-growth rate lasts millions of years and are so to be considered as in a fixed amount. On the other hand, the Renewable energy resources are not reduced by their exploitation. For instance, solar and wind energy are obviously both permanent renewable resources, because the energy flow is lower than the energy storage, contrary to the oil resource, where the flow exceeds its natural re-growth rate. Recalling that the renewable energy resources are not able to cover the energy needs (they are often used for the Peak Shaving and not to cover the basis energy demand), it is clear that a new energy resource is necessary to meet the increased energy demand. Moreover, it has to be non-polluting, renewable and continuously available with no interruptions (unlike solar and wind energy, which are affected by the presence of sunlight and wind). This new energy source can be the Nuclear Fusion Energy, a new kind of energy resource that exploits the energy released by the collision and the fusion of two light atoms (such as hydrogen or its isotopes), according to Einstein equation and the mass-energy balance. Although controlled fusion is extremely technologically challenging, a fusion power plant would offer significant advantages over the existing renewable and non-renewable energy sources, such as the practically infinite fuel supply, the absence or air pollution or greenhouses gas during normal operations and the absence of the risk of a nuclear meltdown. The collision of two nuclei can occur if and only if their kinetic energy is high enough to overcome the energy barrier opposing the fusion reaction, due to the long-range Coulomb repulsion. Therefore, the hydrogen gas is heated up to very high temperatures (one hundred million degrees and even more), reaching the Plasma state. Because of this temperature range, the plasma must be confined and must not touch any structure, in order to avoid yielding heat loads as well as mechanical loads. The Tokamak is a fusion machine aimed at the plasma confinement by means of a magnetic field generated by a set of coils surrounding the plasma itself. In principle, the plasma is supposed to be toroidal shaped during normal operations, but this symmetrical condition is ideal, because of many effects which may lead to a non-axisymmetric perturbation of the plasma column. For these reasons, this PhD thesis is devoted to the analysis of some non-axisymmetric plasma perturbations, their effects during the plasma operations and their modelling. The PhD thesis is divided as follows: 1. The first chapter is a brief overview of the main principles the controlled thermonuclear fusion is based on, focusing on the plasma confinement inside a tokamak, the additional heating and the roadmap towards the fusion energy. 2. The second chapter describes the diamagnetic flux evaluation in ITER tokamak for the estimation of the poloidal beta in the presence of non-axisymmetric effects. In particular, the COMPFLUX procedure used for the analysis is presented, then the effects of the main three-dimensional effects are evaluated and the performance of the compensation system is assessed. 3. The third chapter shows the electromechanical effects due to non-axisymmetric halo currents in ITER tokamak. After discussing the mathematical model, the mechanical effects in terms of forces and torques on the structures surrounding the plasma are evaluated. 4. The fourth chapter is devoted to the flux-density field lines tracing and to the identification of non-axisymmetric plasmas. The mathematical model and the procedures developed for the analysis are presented. Afterwards, the standard and geometrical integrators are compared with reference to test cases for which analytical solutions based on the use of Clebsch potentials are available. Finally, the field line tracing technique is used for the non-axisymmetric plasma boundary reconstruction and a novel technique for the 3-D plasma identification is presented and validated. 5. The fifth chapter reports the main conclusions regarding all the topics dealt with this PhD thesis